ML19327B693

State of VT Suppl to Petition to Intervene.* Contentions Include State of VT Unwillingness to Accept Ownership & Liability for Low Level Radwaste for Proposed License Extension.W/Certificate of Svc

ML19327B693
Person / Time
Site:	Vermont Yankee
Issue date:	10/30/1989
From:	Roisman A COHEN, MILSTEIN & HAUSFELD, VERMONT, STATE OF
To:	NRC COMMISSION (OCM)
References
CON-#489-9389 OLA-4, NUDOCS 8911060325
Download: ML19327B693 (64)

Text

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f 92V4 1 UNITED STATES OF AMERICA before the l.

GFOu m u A1 NUCLEAR REGULATORY . COMMISSION DUChEl% A iMi B N ant,"

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In the Mattersof )

Docket No. 50-271-OLA '

VERMONT' YANKEE NUCLEAR- )

) (Operating POWER. CORPORATION

)

Extension), License '

)

(Vermont Yankee Nuclear .,

Power Station )

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STATE OF VERMONT SUPPLEMENT TO PETITION TO INTERVENE This document supplements the Petition of the State of ,

Vermont for Leave to Intervene and Request for an Evidentiary Hearing (" Petition") filed on August 22, 1989. It is filed in  ;

accordance with 10 CFR 52.714(b) and, to the extent requireo, expands the statement of the aspects of the proceeding as to which the. State of Vermont (" Vermont") wishes to intervene.

Because this proceeding was initiated prior to September 11, 1989,10 CFR 52.714 (b) , as recently amended by the Commission, is not the law which binds this proceeding. Statement of

' Vermont objects to the characterization ofThat this proceeding as a " Construction Permit Recapture." characterization Therefore, Vermont in pres of such a process,.which Vermont rejects.

its pleadings will use the neutralWe andrespectfully more accurate description request that the  ;

" Operating License Extension." Board demonstrate its neutrality by similarly de issuances.

1 8911060325 093030 PDR ADOCK 05000272 G- PDR g

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i consideration accompanying amendments to 10 CFR Part 2, 54 Fed.

- Reg. 33168, 33179 (August 11, 1989).

Vermont believes that this proceeding is premature and as such that it severely prejudices the rights of Vermont and does not serve the public interest. At an appropriate time Vermont 1

ts"- will request suspension of this hearing and postponement of any resolution of the application in order to allow the decisions involved here to be made at a time closer to when the decision must be made,'thus allowing time for numerous matters which are now uncertain to be clarified. Among these matters are the need L for power in' Vermont for the time period from 2007 to 2012, the j

options available to meet those needs other than extending the l' life of an aging nuclear power plant and in particular completion of the Comprehensive Energy Plan recently ordered to be prepared by the Governor of Vermont, the benefit of experience learned

" from this plant and others regarding the effects of aging on the ,

L operation of large nuclear power plants, the completion of the Individual Plant Examination of this plant and the development of l.

I final and lawful regulations by the NRC to govern applications l:

i for license renewal, which is what Vermont Yankee is seeking in L

this proceeding.

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j I. .

o CONTENTIONS AND BASES If Section 5(d) (2) (C) of the Low-Level Radioactive Wasta 1

is lawful, and thus if l Policy Amendments Act of 1985 ("LLRWPAA")

the State of Vermont can be corrt>. led to accept possession of, 2

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,y

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title to and liability.for low-level radioactive waste generated  :

by Vermont Yankee, it would be illegal to compel the State of Vermont'to' accept the generation of any low-level radioactive waste from the operation of the Vermont Yankee facility beyond thL date originally authorized in its operating license permit.

The " Low-Level Radioactive Waste Policy Amendments Act

a. ,

of 1985" (LLRWPAA) (Public Law 99-240) establishes, in Section 5 (d) (2 )'(C) :

"If a State-(or, where applicable, a compact region) in which low-level radioactive waste is generated is 1 unable to provide for disposal of all such waste generated within such state or compact region by January 1, 1996, each State, upon request of the generator or owner of the waste, shall take title to the waste, be obligated to take possession of the waste, and shall be liable for all damages directly -

or indirectly incurred by such generator or owner as a consequence of the future of the State to take possession of the waste as soon after January 1, 1996, as the generator or owner notifies the State that the waste is available for shipment."

The State of Vermont has not been successful in negotiating a compact agreement, nor has the State of Vermont enacted low-level radioactive waste disposal siting legislation.

b. At the time of passage of the LLRWPAA, Vermont Yankee's operating license terminated at December 11, 2007.

An operating license beyond that date was not requested as part of the original licensing proceeding for Vermont Yankee, nor at any time prior to passage of LLRWPAA. If Section 5 (d) (2) (c) is 10wful, the proposed license extension would put ownership and liability for approximately 70,620 cubic feet (2000 cubic meters) of low-level radioactive waste upon the 3

e f I b

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' State of Vermont. Vermont is unwilling to accept this f ownership and' liability,'and therefore, the license extension request must be denied,

c. If the NRC approves the application to extend the operating license of the Vermont Yankee plant, and if Section 5 (d) (2) (C) of LLRWPAA is lawful, then it will be forcing  !

Vermont to choose be';een accepting title to, possession of and liability for additional low-level radioactive waste which

.it.does not want or having the State Legislature and the Governor approve legislation to allow siting for greater capacity for a low-level radioactive waste facility in Vermont i or approve legislation, in conjunction with other jurisdictions, to enter into a Regional Compact for greater capacity. This would have the effect of compelling the State '

. Legislature and the Governor to cast votes and otherwise l

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approve legislation in order to provide a remedy for the low-level radioactive waste problem which they did not want in the first instance. Such coercion to act would violate their L

Freedom of Speech rights under the Constitutions of Vermont and the United States. '

d. By approving the application to extend the operating license of Vermont Yankee, if Section 5(d)(2)(c) is lawful, the NRC would be compelling Vermont to take title to, r

possession of and liability for low-level radioactive wastes which Vermont has stated it does not want and which wastes are not yet authorized to be generated. Such conduct by the NRC 4

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would infringe on the rights of sovereignty of the State in )

violation of the Tenth Amendment to the United States j

Constit.ution.

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TI. I t

Granting'an amendment to extend the operating life of  !

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o Vermont Yankee such that it would either be authorized to .

operate for more than 40 years from the date of issuance of 1 its construction permit or for a period longer than requested  !

in its application to operate violates the provisions of 10 CTR 550.51 which require that operation of any plant for I longer than the term originally requested in the application or longer than 40 years can only be accomplished by filing a  ;

request for renewal of an operating license. Such a renewal [

application must at a minimum meet all of the requirements l applicable to filing an initial application to operate a facility and the filing by Vermont Yankee does not meet those reetitenants.

"the Commission will

a. 10 CFR 50.51 provides that 1 issue the license for the term reauested by the annlicant

" Section 50.33, Contents of applications; general information, identifies that the " term requested by the applicant" must be identified in the license application:

50.33 (e) "(Each application shall states) The class of license applied for, the use to which the facility will be put, tha_neriod of time for which the license is sought ....

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Since it is clear from the record that tho' original Vermont Yankee application did not request operation beyond 2007, NRC ,

I cannot approve an amendment to the operating license permitting extended operation. The NRC staff is currently l

working on and has been working on for some time detailed f I

procedurew to be used by applicants seeking license renewals, f i

These proposals involve substantially greater detail and cover l substantially more issues than the summary document submitted l l

by applicant in support of its proposed amendment. See for example, NRC letter, September 12, 1989, Docket 50-029, f Meeting to Present Plant Life Extension and Individual Plant  :

Examination (IPE) Study by Licensee; and NUREG-1317, Regulatory options for Nuclear Plant License Renewal. i

b. The Vermont Yankee operating license application, 7 i

dated December 31, 1969, made reference to, and was Amendment 10 to, the construction permit application, dated November 30, l l

f 1966. The construction permit application requested:

"A Class 104 (utilization facility) license and construction permit pursuant to Section 104(b) of the ,

Atomic Enerqy Act and Part 50 of the Regulations of the l l

j Commission thereunder for a torn of 40 years."

4 As the facts below indicate the life of the plant was expected ,

to be 30 years, and the request was for 40 years from the  :

construction permit date.  ;

I c. There is no provision made in the rules, or in notes of consideration of the rules, which allows the applicant to 6

I n l

.u j adjust the terminus of its application after issue of the  !

operation license. Rather, 10 CFR 550.51 providest

" Licenses may be renewed by the commission upon the j expiration of the period."  ;

1 Thus, the proposed application must be considered a renewal application. To the extent the NRC or the Staff have  ;

established a policy allowing applicants to adjust the terminus of an operating license by amendment, the pol , is j i

in contradiction of the rules and is not binding here.

d. The initial decision for granting the operating license considered only the period to Dacamber, 2007. Vermont j Yankee Nuclear Power Corcoration, Docket No. 50-271 (LBP 8). Thus the decisions rendered in the past are of no evidentiary value to the questions now before this Board and ,

l should not be accorded any legal weight. 'No previous  ;

licensing board has ever been asked to consider or acted upon j any safety or environmental matter regarding the operation of the Vermont Yankee plant for any period after 2007. ,

1. The record of the operating license hearing board is clear in its initial decision, that its consideration and ruling on issues regarded the license terminus as December, 2007: 3

" Wherefore, it is ordered, in accordance with the Atomic Energy Act of 1954 ... that the Director of Regulation is authorized, in  ;

accordance with this Initial Decision, to issue '

to Vermont Yankee Nuclear Power Corporation, an operating license, for a term of forty (40) years, in the form of Appendix "A" attached hereto..." LBP-73-8, 6 AEC 130, at 146 1 7 1

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"IV. This license is effective as of the date of issuance and shall expire at midnight LBP-73-8, on 6 December 11, 2007." Appendix "A", l AEC 130, at 151. l

2. The original safety analyses did not evaluate operation beyond 200*/. Tns Final Safety Analysis Report l 1

(FSAR) and Safety Evaluation Report (SER) are silent l l

with regard to the tern of plant life. r i

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3. The report of the Advisory Committee on Reactor i 1971, is l Safeguards (ACRS) (10 CFR 550.58(a)), March 9,  !

also silent regarding the term of plant life.  ;

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4. Since the finding required pursuant to the i

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standards of 10 CFR 550.40(a) cannot have been made for any period after 2007 the present application cannot be  !

granted without reconsideration of all issues and subjects presented in the FSAR, SER, and ACRS report f with respect to the proposed extended period of i

operation. i

5. The original environmental evaluations did not Section 10.4 (p. 10-evaluate operation buyond 2007.

12), Conclusion to the Cost-Benefit Analysis for Vermont ,

states Yankee, supplement to Environmental Report, t (emphasis added):

"The annual cost advantage of the nuclear unit f l

over the oil-fired over alternative its so-vaar is over life, $41the million a year.

total cost advantage, even on the basis of ,

conservative values . . . ."

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I This illustrates 1) that contemplated plant life is 30 i

years, and 2) that alternatives were not avaluated past 30 years of plant life. ,

6. Section VII (p. VII-1), Unavoidable Adverse 3 l

Effects, Final Environmental Statement related to the

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operation of Vermont Yankee Nuclear Power Station, July [

1972, states (emphasis added): (

"The estimated life of a nuclear power plant is 2.0 years." [

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Section VIII (p. VIII-2), Short-term Uses and Long-term Productivity, Final Environmental Statement related to  ;

i the operation of Vermont Yankee Nuclear Power Station, July 1972, states (emphasis added):

f "The resource which vill have been dedicated  !

exclusively to the production of electric power '

during the 30 vaars anticinated life sean of l

the Station will be the land itself." l These references again indicate that contemplated plant l i

life was 30 years and that assessments were made based f

on 30 years. Thus the standards of determination l required by 10 CFR 50.40(d) have not been made for any i

period after 2007. In order for these determinations to be made for the period of proposed extended operation l applicants must now submit an amendment of the r environmental report, extension of the consideration of l j

alternatives and environmental cost / benefits to the new license period, and preparation or amendment of the i environmental impact statement by the NRC Staff.

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7. Prior to 1982, and at the time of the Vermont c

Yankee operating license request, it was taken without 1 f

question that the 40 year period of 10 CFR 50.51 began l at the issuance date of the construction permit:

I "From the earliest issuance of construction l permits and operating licenses, the Commission has tied the expiration date of licenses to the (

issuance date of construction permits." NRC Memorandum of August 16, 1982, Dircks to l Commissioners. ,

i While NRC has now established a policy of issuing j 8.

licenses for 40 years from the date Of the operating r

license (Dircks memorandum to Commissioners, August 16, i 1982), this policy does not authorize the issuance of  !

extensions of licenses already issued for periods not f previously authorized. In fact this policy statement if h

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it were applicable to amendment requests would require substantially more detail regarding the safety and  !

See environmental factors than applicant offers here.

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Memorandum, supra. p. 3. In addition the policy statement cannot be used to eliminate consideration of i issues required to be considered by NRC regulations or the statute.

Limerick Ecoloav Action. Inc. v. HRc, 869 L

F.2d 719 (3d Cir. 1989)).

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9. The only way for the Commission to operate within its rules here is to consider the proposed amendment a ,

t renewal application.

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s. As demonstrated in paragraphs VI b. - end, below, l f

the plant components begin to age long before the commencement i of the operation of the plant or issuance.of the operating license. Thus, issuance of the initial license for forty l years from the date of construction represented an appropriate l i

-issuance of the maximum permissible license under the Atomic  :

Energy Act and the Commission regulations and any extension of l such a license can only be accomplished by a renewal ,

application and not by an amendment.

III. t The proposci to extend the operating life of the Vermont l

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Yankee plant for an additional four years and three months is a " major Federal action" within the meaning of 42 U.S.C. ,

i 54332(2)(c)) for which an environmental report is required from the applicant and an environmental impact statement is f

required from the NRC and for which a thorough assessment of I

alternatives must be conducted. Applicants have not met the requirements of 10 CFR 551.45 in that there is not an adequate discussion of the environmental impacts associated with the proposed operation or alternatives to the proposed operation. ,

a. See discussion following Contention II. for an analysis of Vermont's claim that the original construction permit and operating license proceedings did'not include any consideration of the environmental impacts of or alternatives [

to operation of Vermont Yankee after 2007.

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b. The'NRC regulations, 10 CFR 651.45, clearly require i that'a comprehensive analysis of the same factors applicable l f

E.

to the NRC in conducting an Environmental Impact Statement f

("EIS") must be developed by the applicant in its f

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anvironmental report. An examinaticn of the application m

submitted here demonstrates that no such analysis has been  !

V i conducted. ,

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1. Applicant does not attempt to quantify the l l

environmental impact of the disposal of additional low- l

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j and high-level radioactive waste generated by the '

extended operation of the plant, particularly the now  ;

I o 3 known uncertainty about the date on which high-level l j

nuclear waste will be able to be shipped off-site t (Consideration of Environmental Impacts of Temporary  !

l Storage of Spent Fuel After Cessation of Reactor 5 l

operation, 54 Fed. Reg. 39765 (September 28, 1989)) -a f l I date which the original application assumen would be i L

within five years of the date of its generation.

i Similar uncertainty now exists regarding the (

2. t j

disposal of low-level radioactive waste which, in the  !

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case of Vermont Yankee, could be left with no disposal j site and, if Section 5(d) (2)(c) of the LLRWPAA is lawful, in the possession of the State of Vermont. See l Contention I and discussion following it. .

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l Applicants must evaluate the potential environmental  !

3. ,

consequences associated with these indefinite periods, {

during which nuclear waste will be in limbo, with respect'to the nuclear wastes to be generated during the ,

extension period. No such analysis has been conducted 1

.by the applicant and existing NRC conclus ions regarding such impacts as presented in Tables S-3 and S-4 are ,

I inapplicable since they are based on assumptions  !

i regarding what will happen to the nuclear waste which are now undeniably in error. Those erroneous assumptions are that solutions to the problem of nuclear  :

I waste would be implemented by 2009 and that all low-level nuclear waste would have a safe and approved i disposal site. Also erroneous is any implication l concerning the weighting of impacts from land j l

permanently and irretrievably committed.  !

4. Applicants agree effort to consider alternatives to r the operation of the Vermont Yankee plant during the l i

If as applicants  :

l period 2007-2012 is fatally flawed. I assert the operation of a plant the size of Vermont l Yankee will be needed by 2007, then operation of this  !

plant for another four years will not obviate the need to build such a facility but only postpone by four years i

the date on which such new capacity will be needed. I Depending on the cost of construction, the cost of money and similar factors it may be cheaper to build the 13

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1 additional capacity to come on line in 2007 rath3r than

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l'X wait until 2012. In any event, the most that the  ;

limited extension of operation of Vermont Yankee will ,

j provide as a benefit is the difference in fuel, operating and maintenance costs per KWH generated, if l i

any, between Vermont Yankee and the selected long term j solution and the savings, if any, associated with the f postponed investment in the new plant.8 In addition l i

applicant does not quantify the cost of its cure-all l

f'  :

maintenance and replacement program in the waning years i of Vermont Yankee. Presumably, the older the facility  !

the more maintenance and replacement will be required and thus the last years are likely to be the most In comparison of alternatives in "The [

expensive.  !

Assessment" (Section 2.3), applicant has not adequately  ;

assessed the costs in the extended period (for example, i

those required to remedy aspects described in the {

discussion of Contentions I, V, VI, VII, VIII, and IX). l Applicant relies on replacing equipment in "The t Assessment," sections 3.2.2.2, 3.4.2, and 3.4.3. '

t l- [

3 of course, it may be that what Vermont Yankee really has  ;

in mind is an attempt to renewUntil the Vermont Vermont Yankee has the license opportunity for a to t much longer period of time.

conduct discovery such an intent cannot be confidently asserted r here. However, if that is the case then this case presents a classic example of the illegal attempt to segment a single proposed action into several small essentially meaningless parts i in order to attempt to avoid the " major federal action"Such segmentatio provisions of NEPA.

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case law under NEPA.  :

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Applicant acknowledges that backfits will be required in f i

Section 3.4.4.1. Applicant has not correctly included '

costs for the magnitude of replacements and backfits I

which will be required. Also, in Section 3.4.4.1,

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applicant identifies procedural changes as necessary. l f

Applicant also has not correctly included increased costs for maintenance. Applicant has not adequately i assessed the capacity factor, considering unplanned and extended refueling outages caused by aging equipment. l i

Applicant has not included adequate costs for indefinite l temporary storage and/or disposal of high- and low-  ;

level radioactive vaste, if indeed such disposal were f The correct assessment of f ever to come into existence.  !

these costs will exert a great effect on the  :

cost / benefit balance.

I

5. The analysis of alternatives excludes numerous  ;

1. viable options, which are included in the discussion of ,

j the discussion of the following contention (Contention IV).

In addition the Governor of Vermont has just  :

announced the preparation of a comprehensive Energy Plan J

- to be completed by January 1, 1991. Executive Order No.

i 79, October 23, 1989. No responsible agency would i

1 proceed to act on a proposal which does not include I

consideration of the results of such an important and  !

official study of one of the critical issues involved in deciding the proposal.

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Also missing in the applicant's environmental j

'L 6.

analysis is any discussion of the irretrievable f i

commitment of land in Vermont which will be required in  ;

order to provide a final resting place for the low-  !

l 1evel nuclear waste generated during the period 2007- i i

2012. While the total land so committed may not seem l t

significant, the fact that the commitment is forever is, See conservation society of Southern vermont. Inc. v.

[

Voice, 343 F. Supp. 761, 767-8 (D.Vt., 1972) reviewed .

and altered in numerous ways on appeal on other grounds i but still authority for the proposition that even the i f

smallest piece of land in Vermont is precious -

" Wilderness may not be shipped; it stays where it is, f broken only by the intrusion of man." Pursuant to 10  !

CFR 551.51 the information contained in Table S-3 can be >

supplemented by a discussion of environmental significance. Given the unique importance of land in the state of Vermont such a discussion is clearly .

i warranted in this case particularly in light of the j

strong public resistance to any nuclear waste disposal in the state as evidenced by the refusal of the State r Legislature to adopt waste siting legislation. ,

l c

7. The substantial environmental impacts which will occur from this proposal include the environmental i impacts associated with the potential failure to find any acceptable site to dispose of high- or low-level 16

waste, the commitment, forever, of lond in vormont for j

disposal of low-level waste, the potential impact from i l

accidents caused by the failure to properly address the increased risks created by problems of aging, inadequate f maintenance, inadequate compliance with ASME codes and

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quality assurance criteria, the absence of an t

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individualized Probabilistic Risk Assessment for Vermont Yankee and the overall additional environmental impact ~

caused by the operation of a nuclear power plant for a period of more than 12% of its originally proposed i

operating life.  !

i

8. The NRC has consistently considered even short  ;

operation of a nuclear facility as a major action as i evidenced by the Temporary operating License provisions, j Section 191 of the Atomic Energy Act of 1954, and the NRC's concern with even a time period of two years of l

operation without the use of the most up to date codes.

Statement of Consideration accompanying adoption of 10 CFR 550.55a, 34 Fed. Reg. 18822.

9. Pursuant to 10 CFR 51.53 applicant must file a it is complete environmental report where, as here, seeking a renewal of its operating license.

l IV. ,

Even if the proposal to extend the operating life of the Vermont Yankee plant for an additional four years and three i

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f months is not a major fcd3rol ccticn, it ntnstholoco involvos i

" unresolved conflicts concerning alternative ases of available i l

resources" for which NRC must " study, develop and describe i alternatives" within the meaning of 42 U.S.C. 54332(2)(E) and l i

for which the applicant must submit such a study as part of i i

1 a its environmental report pursuant to 10 CFR I 51.45. ,

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a. The alternative uses of available resources which  !

are in conflict involve the commitment of land resources in j

Vermont to dispose of low-level radioactive wastes, additional i land elsewhere for high-level radioactive wastes and the  ;

release into the environment of radiation from the operation For a fuller j of the plant for an additional period of time. i discussion concerning the conflict regarding the use of ,

Vermont resources to receive or absorb radioactivity, see the j f

discussion in Contention I. The land could be left as, or r restored to, wilderness and the environment could be preserved 1 i

As i

from having to absorb any additional man-made radiation. l low as reasonable achievable is zero if Vermont Yankee ends  :

p its operation in 2007.

b. A natural gas power plant, cogeneration and other qualifying facilities, an advanced nuclear facility, and conservation are feasible and realistic alternatives to the {

proposed action. This fact is demonstrated by numerous I

documents, not limited tot

1. Assassina NEPOOL's Resource Adeauncy and Potential l

l Resources (1991 throuah 2004), Technical Supplement, 18 4 i

1 s

June 1989, which, und0r cOnting;ncy plcns, id0ntifics, for the NEPOOL region in the year 2004, the availability of 2305 MW of Non-utility generation (cogeneration and l other qualifying facilities), 3925 MW of utility generation (major facilities such as a natural gas  ;

plant), 10,396 MW available for purchase from Canada, j and 1025 MW available from demand side management I

This demonstrates that any of (conservation) - Table 4.

these sources, singly or in combination, have sufficient  :

power to Vermont Yankee. This document also identifies f j

the ocean State Power Units 1 and 2, a 500'MW, two unit natural gas plant under construction in Rhode Island, and estimated for unit 1 in service in 1991 and unit 2 in service in 1994 - pp 33-34. This demonstrates the feasibility of a 500 MW natural gas plant as an j i

l alternative to Vermont Yankee.  !

2. NEPOOL Forecast Report of Canacity. Enerav. Loads and Transmission. 1989-2004, April 1, 1989, which ,

l j

identifies Ocean States Power Units 1 and 2, a 500 MW,  ;

two-unit natural gas plant, scheduled to be inservice in  ;

f 1990 and 1991 - p. 33. This document also identifies proposed combined-cycle, natural gas and fuel oil or kerosens units, proposed to be added by NEP00L utilities i after 1994, amounting to 1307 MW, thus demonstrating  !

feasibility - p. 33. Also identified are 1395 MW of ,

I uncommitted non-utility generated power in 2004, ,

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, - r demonstrating the feasibility of cogeneration and other qualifying facilities - p. 34,

3. Symmary of Generation Task Force Lena-Rence Study Assumotion, NEPLAN and Generation Task Force, April 1988, which, in Section I.B.2, Tuture Resources, states, "In addition to NEPOOL generating units future resources will include options such as DSM, NUG, life extension programs, and pool imports." This document also includes economic information related to the identified alternatives.

4. coaeneration. Small-Power and Indeoendent Power Facilities in New Enaland, The New England Governors' Conference, Inc., December 15, 1988, and May, 1989, update, which identifies, in New England, 2076 MW of I

operating NUG capacity, 915 MW of under-construction NUG capacity, and 2003 MW of planned NUG capacity, thus demonstrating the feasibility of NUG (cogeneration and 3.

other qualifying facilities) as an alternative - p.

5. Green Mountain Power corooration Intearated Reenurce Plan, February 1989, which is indicative of resource options available to New England utilities.

This document identifies long term resources of coal, combined-cycle and gas turbine plantst Hydro-Quebec purchases (a 500 MW contract); and 24 proposals for alternative power, solicited on May 31, 1988, for a 20

total of 806 MW of capacity - Section 4.2.2. This indicates the feasibility of cogeneration and other The document, in qualifying facilities alternatives.

section 6, also identifies 54.8 MW which can be attained through demand side management (conservation),

demonstrating that demand side management is real and feasible to be considered as an alternative to the proposed action, in conjunction with other sources.

6.

Central Vermont Public Service Corporation, News Release of November 23, 1988, which indicates that 28

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proposals were received from non-utility generators for a total of 660 MW of capacity, thus demonstrating the feasibility of cogeneration and other qualifying facilities.

7. Docket No. 5270, " Investigation into Least-Cost Investments, Energy Efficiency, Conservation, and Management of Domand for Energy," Public Service Board, state of Vermont, unarina of ficer's manort and Pronosal for Decision, July 13, 1989, which indicates:

"The record demonstrates a potential of far more than 120 MW in energy efficiency savings by 2000.

However, the record does not allow one to specify a specific quantity that reliably defines the outer bound of cost effective efficiency solutions...however, the record suggests considerably greater potential." - pp III-28,29 .

"I conclude that ... existing load forecasts overstate the load growth that will occur 11 utilities aggressively and effectively pursue the 21

I acquisition of all cost-effective demand-side l l

resources." - p. I-5 l "I conclude that there is a high potential for 1 acquiring cose effective efficiency resources from j a great majority of houses, businesses, farms, and i l

N. factories in Vermont. That potential is very large, but its upper limit cannot yet be l quantified. I also conclude that price signals, .

while necessary, are not sufficient to acquire i those resources." - p. I-6 l I

i "I conclude that utilities should seek demand-side efficiencies as actively as they pursue supply l resources; in other words they should try to " buy"  :

all cost-effective efficiency savings from their i customers." p. I-6 l "I conclude that utilities should consider the  ;

costs and benefits of efficiency improvements on a societal basis when deciding which energy-savinq  !

programs to pursue." p. I-6 l "I conclude that supply and demand-side options  !

must be integrated on an equal footing, and that l this has not happened historically. I also  !

conclude that this integration requires Vermont utilities to enhance their ability to acquire [

' demand-side resources. Finally, I conclude that l l supply and demand options cannot be compared fairly 1 l

unless all the costs of both options are i

! considered. These costs include transmission costs, relative risks of non-delivery, backup l j

supply needs, and environmental eff6 cts that are r

l often hard to price in monetary terms but are 1 j

' nonetheless of vital significance." pp. I-6,7  !

Docket No. 5270 demonstrates the real and feasible 7 I

nature of demand side management as an alternative, t

\

in conjunction with other sources.  !

c. The State of Vermont will provide testimony i t

concerning a comprehensive Energy Plan, which incliades alternative means of energy production and conservation.

22 [

l

i l

Governor Kunin, in Executive Crder No. 79, dated October 23, j 1989, established:  ;

i "The Department of Public service in conjunction l with the Agency of Natural Resources shall submit a  !

Comprehensive Energy Plan by January 1, 1991. This l Plan will guide the purchase, regulation and use of j all forms of. energy within Vermont. The Plan shall )

be directed toward goals of protecting the  ;

environment, increasing energy efficiencies, and  :

reducing overall energy costs." l' 1

This information will be available by January 1, 1991.

i I

V.  ;

The application should be denied because the applicant

1) has not evaluated the difforence between the Vermont Yankee i licensing basis and the current licensing basis for plants i originally licensed to operate through 2012, and 2) has not  !

demonstrated the effect on the environment and public health i and safety of each difference. One example of this is that I i the American Society of Mechanical Engineers (ASME) Codes and l

( f quality assurance requirements for reactor coolant pressure j

boundary (RCPB) pressure vessels, piping and pumps and valves  ;

i and the codes to which it has been constructed are inadequate for extended operation beyond 2007.

a. Numerous licensing requirements were developed by (then) AEC between 1967, applicant's construction permit date, and 1972, applicant's fuel load license. Each requirement was l J  !

j carefully evaluated by AEC, and adopted because it provided l the degree of safety and environmental protection deemed necessary for plants which would operate after 2007.

23 i

_ - - - _ _ _ _ - - - - ~ - . _ _ -

b. The Cxtont of th0Co diffor nc0c b;twocn liconoing bases will be further developed through discovery. For l 1

demonstration of validity of this contention, examples of ASME l Code and quality assurance requirements are elaborated below. l J

c. The ASME codes which are applicable to the proposed  !

l extension but which are not met by Vermont Yankee are Section {

III, ASME Coder USA Standard Code for Pressure Piping (USAS B31.7); and Draft ASME Code for Pumps and Valves, and the f i

requirements applicable to pumps and valves set forth in articles 1 and 8 of Section III of the ASME Codel available in f March 1972, and applicable to construction permit plants on i I

l that date. i

d. The safety significance of the codes listed above j j

was given as the basis for the Commission requirement for  !

implementation at the earliest feasible time, and specifically t for all plants with constructica permits on or after January 1, 1971, which would operate on or after January 1, 2011. The  !

Commission indicated, in Notes of Consideration, that industry practice of delaying implementation of these new codes for 2 l I

to 5 years was unacceptable. Thus, the commission'was unwilling to allow a practice to prevail that would allow design without these codes for plants which would operate after January 1, 2011. j

e. Applicant has requested an operating license extension to March 21, 2012 (Applicant letter, BVY 89-55, June  ;

23, 1989).

24

I According to NRC (formerly AEC) policy before 1982, f

f. i operating licenses were issued for 40 years from the date of the ccnstruction permit (NRC Memorandum of August 16, 1982, l

Dircks to commissioners). t

g. In order to operate until March 21, 2012, a l l

construction permit would have to have been issued on March j l

21, 1972 pursuant to the policy in existence at the time this plant was given its operating license (NRC Memorandum of i 1

August 16, 1982, Dircks to Commissioners). I i

h. In addition at the time the NRC adopted the January 1, 1971 construction permit cutoff date it believed that all j preexisting plants would be operating for a period which would f f

not be longer than 40 years from the date of their  !

construction permits and obviously balanced the relative risk l i

j of allowing those plants to operate without meeting the codes  !

for that period of time. Had the NRC known that those plants  ;

would be seeking to operate beyond that time period they may  !

well have not been willing to provide the full " grandfather" I

clause. Thus any plant which seeks to operate beyond forty i years from the date of its construction permit should have to meet the newer code requirements or be denied the proposed J t

extension.

l

i. Reactor coolant pressure boundary (RCPB) minimum  :

l <

! quality standards for design, fabrication, erection, construction, testing and inspection for a construction permit I

on March 21, 1972, were (36 FR 11423):

l 25 ,

1

. .- - , . . - . - . - .-._- - .. . - - - .. - - -_. ~ _ _ _ - - _ _ _ _ _ . _ _ . _ . _ - _ - _ . --- . _ - _ _ _ . - - - - -

i

1. Pressure Vessels - R0guiron0nto for Cloco A or .

Class 1 Vessels,Section III, ASME Codel code and .

I addenda in effect on the date of vessel order or 18 l I

months prior to construction permit, which ever is 1 less.  ;

r

2. Piping - Requirements for USA Standard code for Pressure Piping (USAS 831.7); code and addenda in f effect on the date of piping order or 6 months  ;

prior to construction permit, which ever is less.  :

I

3. Pumps and Valves - Requirements for Class I ,

pumps and valves, Draft ASME Code for Pumps and j Valves, and the requirements applicable to pumps  :

and valves set forth in articles 1 and 8 of Section i III of the Boiler and Pressure Vessel Codel code '. l

! and addenda in effect on the date of order or 12 '

L L months prior to construction permit, which ever is less. '

i

j. Applicant's RCPS is designed to " applicable codes in f

- effect at the time of the components were ordered (Vermont  !

Yankee Safety Evaluation Report (VYSER) - p. 24)." The l reactor pressure vessel is " designed, fabricated and inspected  ;

to the Class A requirements of Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition including published l I

addenda through and including summer of 1967 (VYSER - p. 26)."

l RCPB piping is " designed, fabricated, and inspected in 26 1

l-i l

l

i i

accordance with tho USAS 831.1.0 - 1967 Pcw0r Piping Codo f (vYSER -p. 29)."

The VYSER is silent regarding codes for RCPB pumps and valves, '

k. Safety significance exists in using the codes An item 1. above when compared to the codes of item j. Noter of i consideration for 10 CFR 50.55a include the following statements:  ;

"Because of the safety significance of uniform early compliance by the nuclear industry with the requirements of these ASME and IEEE codes and ,

published code revisions, the Commission has adopted the following amendments ... which require l that certain components of water-cooled reactors ,

important to safety comply with these codes and appropriate revisions to the codes at the earliest  !

feasible time." 36 Fed. Reg. 11423.

I "The Commission believes these changes adopted will  !

... provide an equivalent increase '.n protection of the health and safety of the public." Id. j "The Commission considers that a significant l improvement in the level of quality in design,  !

fabrication, and testing of systems and components t I important to safety of water-cooled reactors will be afforded by compliance with the requirements of  ;

more recent versions of the codes." Id.  ;

1. The Commission acted to assure expeditious i L

V l application of improved codes, indicating that it did not wish l  !

to allow extension of the use of less safe code designs into f

the indefinite future.

Notes of Consideration, 37 Fed. Reg. 17021, 36 Fed. Reg. 11423 and 34 Fed. Reg. 18822.

m. The 1971 piping code (531.7) and the 1971 ASME, i

Section III, code for pumps and valves, which would have been ,

in effect for plants originally allowed to operate through 2012, established the requirement for Manufacturer's certified ,

27 l

l

l i

Mate' rials Test Reports (CMTRs) assuring th3 c3ntCnt of

, This became f material used for RCPR piping, pumps and valves.

the basic material requirement for plants originally licensed Without CMTRs, which were in 1971 and operating through 2011.

l not required by applicant's codes for RCPB piping, pumps and f

valves, the actual content of materials is in doubt, and i uncertainties exist concerning claims of material capabilities. The failure of RCPB piping, pumps or valves,  !

under normal,' design basis accident and severe accident conditions, and through aging effects, due to materials other t l l l

than specified, would affect the environment and the health l f and safety of the public, t

n. The B31.7 piping code established fatigue design Similar fatigue requirements for RCPB piping and fittings.  !

i requirements were present for RCPB pumps and valves in " Draft i j

These ASME Code for Pumps and Valves for Nuclear Power."

' f codes were in effect for plants originally allowed to operate through 2012. For example, fatigue design requirements are  !

i l.

found in 831.7, sections 1-702.3.1(b)3, 1-705.3.3, E-180, F-I

' 106, F-167, F-109. Applicant's RCPB piping code, 831.1.0, did l Applicant's RCPB not contain fatigue (design requirements.  ;

piping and fittings, pumpe and valves have not all been designed or analysed with appropriate fatigue requirements of '

B31.7. Fatigue failure is an age-related mechanism, directly P related to the proposed action.

28 1

i l

l While applicant's claims in section 3.4.1.1 of "The l l

I Assessment" to have performed a fatigue evaluation, no proper reference or, indication that NRC has reviewed this ,

i evaluation, is provided. Nor is the difference in codes j identified. Inadequate consideration of fatigue could lead to i failure, in the extended period, during normal, design basis l J

or severe accidents which would affect the health and safety of the public.

o. Adequacy of early codes is questioned in EPRI NP- l l

5181M, SWR Pilot Plant Life Extension Study at the Monticello Plant: Phase 1:

" Earlier construction codes (prior to 1971) did not i prescribe explicit fatigue evaluations for piping, l pumps and val.ves. Fatigue reevaluations may be necessary for components susceptible to fatigue degradation in order to verify extended life potential and to identify locations for enhanced inspections." (p. 5-5) i.

l p. The B31.7 piping code established more reliable nondestructive examination (NDE) requirements for RCPB for l

l plants originally licensed to operate through 2012. These NDE 3

requirements were not required in applicant's RCPB piping ,

f code, 831.1.0. The lack of this NDE leaves uncertain the f adequacy of applicant's RCPB manufacture, assembly, and l l

i t installation, potentially leading to failure, through aging processes in the extended period, of material with manufacture, assembly or installation flaws.

1 1

29

__ __._____..._._._.1

t

q. Article 8 of ASME,Section III, Summer 1969 I addendum, set the requirements >

"N-832 Quality Assurance Program - Any manufacturer of ... Class I piping of the USAS B31.7 Code for [

" Nuclear Power Piping," or class I pumps and valves I l

of the ASME Code for Pumps and Valves for Nuclear j Power shall have a quality assurance program which meets the requirements given in IX-200 of Appendix r IX of Section III."

This requirement was specifically applicabic to RCPB j piping, pumps, and valves for plants originally allowed to [

operate through 2012. However, applicant's codes contained no i

-such requirement.

Furthermore, on June 27, 1970, quality assurance requirements which have become the standard for the industry, "the 16. criteria of 10 CFR 50, Appendix B," came into  ;

existence (35 FR 10498). This revolutionized nuclear industry quality assurance practices. The date of June 27, 1970 was after many of applicant's RCPB piping, pumps and valves were manufactured. ,

t i

i

+

Therefore, applicant's RCPB piping, pumps and valves

' were not manufactured, assembled, and installed to the quality assurance standards for plants originally licensed to operate  !

through 2012. The lack of this level of quality assurance casts doubt on all statements of vendors, suppliers, i

manufacturers, assemblers and installers, regarding the adequacy of materials and processes. Deficiencies in these  :

1 30  ;

materials or processes can cause age related failurco in tho extended period which would affect public health and safety.

j

r. The quality concern was expressed by General

)

Electric (GE) in its 1975 " Nuclear Reactor Study (Reed i Report)." Under the findings of " Control of design of  ;

t purchased components," GE statest .

"NED relies almost entirely on vendor design expertise to produce components and equipment to performance and functional purchase specifications

... Because of reliance on this vendor expertise, ,

j NED ... has exercised essentially no control of '

design details of critical purchased equipment."

i (p. 44) i "The present system of rewards and penalties does '

not operate ef fectively to motivate vendors to  ;

produce purchased components to the level of  !

reliability and durability justified for nuclear plant service." (p. 45)

s. The specific relevance of these concerns, related to aging in the extended period, is demonstrated by ongoing ,

issue of Feedwater Check Valve V285 (Refer to applicant letter, BVY 89-31, Response to USNRC Request for Vermont Yankee Feedwater Check Valve V288 Flaws Evaluation, March 28, 1989; and NRC letter, Feedwater Check Valve Flaws, April 5, 1989.

A flaw of about 2.5 inches long and 0.65 inches deep was ,

visually discovered in feedwater check valve, V28B, a RCPB valve. Additional, similar flaws were found in this and otlier RCPB valves. Applicant claims these flaws have existed since initial operation of the plant. Applicant also claims these 31

. _ - _ . - . - _ . _ - _ - - _ - - . . - - . - . - - . . . - . . . _ . . ~ . - . - - - . _ - .-

l 1

j flaws to not constitute a safety probica. H w0vor, NRC otaff J

has not completed its evaluation of applicant's report.

Even should NRC's evaluation agree with the applicant's, l

)

First, if the validity of this contention is demonstrated.

j this is not a safsty problem, why has applicant agreed to l replace the valve at the next outage? Second, applicant has l not attempted to show such flaws would be safe through the f extended period. Third, applicant has not volunteered to f perform NDE on all RCPB and safety related valves to identify f I

similar flaws. And, fourth, the fact that such flaws can i exist is a direct result of inadequate codes in the areas of A

materials certification, quality assurance, nendestructive ,

examination, and fatigue evaluations.

Applicant, while requesting to extend use of less t.

is safe codes and quality assurance practices into the future, I l

silent regarding the Commission's concern over code

( application, does not identify design codes for reactor ,

l '

pressure vessel, RCPB pumps, valves and piping, does not identify requirements for these components for plants originally licensed for operation to 2012, and does not 1

identify differences or offer justifications for extension of the older code design (Applicant letter, BVY 89-41, April 27, 1989, Attachment 2, Section 3.4.1).

u. This plant's margin of safety has fallen below the h .

minimum acceptable margin of safety for operation in the l

32

extended period due to the inadequate, inappropriate, and j L I outdated licensing basis.

VI.

The application should be denied because the applicant has failed to demonstrate that there is reasonable assurance I

, that operation of the plant beyond the date for which I

g

' operation was originally approved will provide adequate protection to the public health and safety due to the excessive aging of safety significant components and the absence of any ef fective and comprehensive program to detect the presence of such excessive aging. l

a. For the discussion of the extent to which the L

original application involved only consideration of the plant for a period through 2007 and thus that analysis by the Board I

of the aging of components did not consider the operation.of the Vermont Yankee plant for the time period or longevity requested here, see contention II above.

obsolescence is a legitimate aging factor. In b.

EPRI providing a general method of equipment life assessment, EL-5885, Generic Guidelines for the Life Extension of Plant Electrical Equipment, July 1988, states:

"The decision to either continue in service or to replace can be greatly influenced by current availability of technologically advanced designs i that offer improvement in reliability and decreased L

maintenance requirements ... many electrical equipment manufacturers are no longer inThe business supply or no longer manufacture spare parts.

33 I

l l l

.s.

~

of'cpero'porto "for cbsolcto equipnOnt io uncertain in manyfcases. - (EPRI EL-5885, p. 3-6).

4< ,

v

Obsolescence began.at the< time of manufacture, before issuance lof.the operating license.

c. Applicant's management consultant identifies Also identified is the cbsElescence:as a serious concern.  :

fact that' applicant did not have a program to monitor i equipment performance vis-a-vis obsolescence as of october,

]

1988; nor does' applicant give any indication'the a program has f t

been established since that time. t "The potential problems associated with mature plants is of current and increasing interest to the -l NRC,:and to the involved utilities as well. There is no apparent serious degradation in the safaty ,

( ,' status of Vermont Yankee However, that there woulddo necessitate appear to be '

" short~ term action.'

isolated indications of the need to address specific items before they develop into significant ,

concerns.. The comments of operators on the i reliubility of instrument recorders,;which will be i discussed in more detail in this. report, may be  ;

such an indication. ~

L While senior mEnagement appears alert to the need  :

to monitor the aging of electronic equipment and  !

instruments'and is considering concerns as they arise, it may be prudent at this time to consider l-the development and the measured implementation of l a specific program to detect early' signs of L

equipment performance degradation or reliability

[ losses, and to assure that requisite levels are ,

L-maintained." (LRS Inc., Report of October 10 - 13, L p. 8) .

See also the discussion for Contention VII, which identifies the absence of a sufficiently effective maintenance and surveillance program.

d.

Dominant. aging mechanisms for transformers are cold-l flow deformation (creep) and dielectric breakdown, which began 34 L

l,

a.-n a . - - e + w - a .-.. - -a - -

4  :

l at the time of manufacture before the issuance of the r operating license:

r

" Cold-flow deformation of the structural insulating

. parts (creep) may gradually loosen the blocking which holds the windings in place and will allow  ;

some. movement during faults." (EPRI EL-5885, p. ,

> 3.2.2)-

Additional aging mechanisms began with first operation, which occurred during preoperational tests before issuance of the P

operating license: ,

"[ Dielectric) breakdown is hastened by contamination, thermal aging, repetitive voltage '

stress, and mechanical deformation." (EPRI EL-5885, p. 3.2.2)

e. Dominant aging mechanisms for motor stator ,

insulation result from moisture penetration, heat, oil or I other surface contaminants, which began 12t the time of .

l J

I manufacture before the issuance of the operating license:

" Moisture penetration can result form prolonged exposure to high humidity, especially during idle Motors in storage may be or shutdown periods.

particularly vulnerable to moisture unless heaters y

are provided." (EPRI EL-5885, p. 3.3.2)

Numerous other aging mechanisms, such as overvoltages, corona, abrasion, mechanical wear, material fatigue and vibration, began with first operation, which occurred during preoperational tests before issuance of the operating license.

f. Dominant aging mechanisms for switchgear result from corrosion and dielectric breakdown, which began at the 35

c .

L i

' time.of manufacture before the issuance of the operating

' license:

i " Corrosion may occur whenever moisture is available i t

or dissimilar metals are in contact." (EPRI EL-5885, p. 3.4.2) i

" Dielectric failure in switchgear is cause by ...

degradation of insulation by water absorption."

(EPRI EL-5885. p. 3.4.2)

. Numerous other aging mechanisms, such as erosion of breaker arcing contacts and arc chutes, electrolytic w corrosion, mechanical wear, and internal corona, began with first operation,-which occurred during preoperational. tests  !

'I before issuance of the caerating license.

g. Dominant aging mechanisms for power and control l

1

! cables result from moisture or chemical attack of rubber or PVC insulated cable, thermal embrittlement and formation of electrical and water trees in polyethylene or cross-linked polyethylene insulation, which began at the time of l manufacture before the issuance of the operating license (EPRI i EL-5885).

Other aging mechanisms began with installation, also i before issuance of the operating license:

l

" Creep deformation occurring at points of pressure such as at duct mouths, conduit corners, tray edges, cabinet corners." (EPRI EL-5885, p. 3.6.2) 1 other aging mechanisms, such as electrolysis attack, I began with first actuation during preoperational testing before issuance of the operating license.

36 l

l

l

.i i

4

h. Protective and aux!.11ary relays degrade by moisture intrusion and thermal degradation which began at the time of manufacture. Other aging nachanisms, such as mechanical wear, L

electrolytic corrosion and dielectric breakdown began with I first actuation during preoperational testing before issuance l J

I of the operating license (EPRI EL-5885, p. 3.8.2).

1 Primary aging mechanisms of batteries are positive 1.

l plate corrosion / growth, negative plate sulfation, presence of contaminants in the electrolyte and separator degradation.

These mechanisms began with battery manufacture, and continued with addition of electrolyte and first connection to circuits l

during preoperational tests (EPRI EL-5885, p. 3.9.2). i The basic failure mechanism of a cathodic I

j. l l

l protection system is tne gradual corrosive wasting away of the l anodic material, which began at the installation of the system j I

l before issuance of the operating license (EPRI EL-5885, p.

i 3.10.2)

k. Aging mechanisms for light water reactor (LWR) structural materials (carbon steel, low alloy steel, stainless steels and Inconel) are evaluated in EPRI NP-5461, Component I Life Estimation: LWR Structural Materials Degradation Mechanisms, September 1987. The following observations are made:

I 37 I

"Th'e family of materials degradation mechanisms  ;

classified as corrosion mechanisms are characterized by:

the deterioration or destruction of a material '

reaction with its environment, or extractive metallurgy in reverse.

The mechanisms considered are: General Corrosion

, i l

and Wastage, Stress Corrosion Cracking, '

Intergranular Attack, Crevice Corrosion, Hydrogen '

Embrittlement, Erosion and' Erosion / Corrosion, Microbiologically Induced Corrosion and Pitting." '

(EPRI NP-5461, p.-C-4)

"All the listed corrosion mechanisms have been observed'in LWR's during construction ....": (EPRI NP-5461, p. C-5)

Thus, all systems and components constructed with the t

' covered materials,'such as piping, pump, valves, and vessels, i are shown>to begin aging in the construction period, before 1

' issuance of the operating license.  ;

L Additional aging mechanisms identified in EPRI NP- i 1.

f: '

5461, such as fatigue and embrittlement mechanisms, plastic deformation, dynamic mechanisms, creep, fretting and wear, t 4

began with first operation during preoperational testing before issuance of the operating license.

L .

m. EPRI NP-5181M, BWR Pilot Plant Life Extension Study at the Monticello Plant: Phase 1, May 1987, identifies aging mechanisms for BWR systems, structures and components.

General corrosion, which began from manufacture and construction, is identified as an active mechanism for the 38

o l

\1 L reactor pressure vessel-(RPV),'RPV safe ends, TPV support, reactor recirculation piping, reactor coolant piping, plant j

't )

control center, drywell metal shell, suppression chamber and l concrete structures (EPRI NP-5181M, p. 5-28).

n. Technical obsolescence, which began at manufacture, is identified as a. primary aging mechanism for the plant L

E control center and diesel generators (EPRI NP-5181M, p. 5-28).

o. Concrete aging mechanisms of creep, shrinkage, cracking and chemical attack, all of which began at construction and before issuance of the operating license, are  ;

i identified as the aging mechanisms for concrete structures (EPRI HP-5181M, p. 5-8).

p. EPRI NP-5836M, BWR Pilot Plant Life Extension Study at the Monticello Plantt Interim Phase 2, October 1988, corroborates that significant aging mechanisms begin before issuance of the operating license. For example, aging mechanisms of humidity, thermal gradients and related componen't debilitation such as cracking of concrete and corrosion of metal substrates, are identified for containment coatings, including the statement:

"The effects that debilitate the protective coating systems are related to ... construction ...." (EPRI NP-5836M, p. 3-71) 39

. 1 I

o <

p qi EPRI NP-5181M indicates at pp. 5-20,21: J i

"Results from the specified testing and ,

surveillance activities at Monticello include the following "significant" findings: 1. Drywell ,j Shell Corrosion at concrete Floor Interface ... The h visual state of the corrosion indicated that the oxidation process at the shall had initiated -

several years ago... Continued propagation of corrosion at this location without repairs / preventative measures could produce significantly reduced wall thickness within this critical section of the drywell shell."

This aging mechanism began at the time of construction ,

~and became apparent only after years of cperation.- Applicant,  ;

in "The Assessment", Section 3.4.4.1, Primary Containment, fails to discuss this, or any other aging mechanism of primary containment.

V

r. Th.s construction period is indeed significant in f

lh aging and life considerations:

" Incidents which occurred during the construction ,

period may have an innact on component life."

(EPRI NP-5181, p. 5-13)

Some items required to assess the construction period effects are a review of nonconformances and special L construction procedures, dispositions of significant process deviations (welding, heat treatment, etc.) and out-of-tolerance dimensions, number and location repair welds, preopera'tional history, nonconformances of preoperational tests, storage environments and methods, and scope and results of nondestructive examinations (EPRI NP-5181M, p. 5-13).

Applicant provides no indication in "The Assessment" that this review was completed or that it is even possible.

40

. . . . - , , -,,, - - . - - , . - .... ~ _ - ... - - -. _

s. Applicant's environmental'qualificction  ;

documentation in accordance with 10 CFR 50.49 makes clear the f

. fact;th'et aging begins during the construction period, and that the effects are significant. For example, from QDR 6.1: ,

i "When a motor is stored, the effects of the

- environment on the components of the motor become more pronounced as time in storage increases... maintenance during storage is ,

vital...obviously, if no maintenance program is c .followed during storage, costly damage to the motor may go unnoticed or unprevented." (Westinghouse UHD Newsletter #16, 6730806)

t. Further, applicant's environmental qualification i documentation does not meet the requirements of 10 CFR 50.49. [

Applicant's program evaluates a 40-year equipment life. 9 However, applicant's program assumes this 40-year period begins with initial operation, which it is shown above that j

I aging life begins earlier, during the construction period.

L The application cannot be granted until applicant's program ,

demonstrates qualified lives up to 44 years and 3 months.

l-

! u. Above it has been shown that each aspect of the-facility, its systens, structures and components, begins the aging process before the issuance of the operating license.

- This aging process causes failures of equipment (such as those identified in paragraph IX - e), and these failures extend into, and are unevaluated for, the proposed extension period.

v. One potential effect of improper aging control is failure of system to control the effects of postulated design _

basis accident, thus resulting in a failure to meet ECCS criteria.

41

(

w. This plant's margin of' safety has fallen below the minimum' acceptable margin of safety for operation in the extended period due to aging of safety significant components.

VII.

The application should be denied because the applicant

'has' failed to demonstrate thct there is reasonable assurance that operation of the plant'beyond the date for which operation was originally approved-will provide adequate

.i protection to the public health and safety due to the absence of a sufficiently effective and comprehensive program to maintain and/or determine and replace all components found to have aged to a point where they no longer meet the safety standards applicable to this plant and upon which this plant was originally granted its operating license.

a. Applicant relies extensively on the adequacy of its maintenance and surveillance program to guarantee that failure i 1

of' aging systems, structures and components will not cause l safety problems. Refer to statements in "The Assessment" Sections 3.2.3, 3.4.1.3, 3.4.2, 3.4.3, 3.4.4.1, 3.4.4.2.

b. Applicant's maintenance program has been evaluated 4 as only generally effective by the NRC staff, with significant  !

weaknesses noted. This is the current state of the record concerning applicant's program, and no closure, showing 42

+k o l

improvement of weakness has been accomplished. Specific I x .y L weaknesses' identified are < 1 al. Lack of comprehensive and formally documented l maintenance plan and policies, i

2. Lack of comprehensive and structured review for  :

s adequacy and applicability of the plant's 1 maintenance requirements.

3. Review for the appropriateness and technical '

adequacy of completed maintenance activities were not being performed in a timely manner. ,

s-  :

4. Lack of effective policy and procedures for controlling and updating manufacturer technical manuals.

5. PRA concept not incorporated into Vermont Yankee maintenance program. -

6. No integrated Master Equipment List for Vermont  !

Yankee.

7. Post-maintenance testing requirements were not

• proceduralized in the administrative procedure." '

(NRC Region I Inspection Report No.. 50-271/89-80 (IR 89-80), June 2, 1989, Appendix 3)

I c. From IR 89-80, p. 4, NRC states:

" Currently, the licensee's maintenance program is based more on the stability of maintenance staff, their skill in their professions, and their knowledge of plant system characteristics that come with long-term experience, than on formally and clearly established management controls. This area L is considered a weakness in the licensee's '

H maintenance program."

As the plant ages, experienced maintenance personnel will retire. Thore is no assurance that qualified replacement L personnel can be obtained. See for example, " Outlook on Skilled Personnel", Inside N.R.C., October 9, 1989, which states:

43

D w '

1 h  !

j L

"The-U.S.'is losing its core of nuclear expertise and doing very-little to get it back.- By the time 4 U.S. policy makers see the:need for the nuclear -j I

option, and U.S. utilities find alternatives.

eclipsed by environmental needs, the option may be i i

gone because the people are...

i The need gets more critical as plants age, but ... 1 attracting young people to become experts in  ;

systems in a: plant that might close in five, 10 or ,

15 years can be ,tifficult." i A maintenance program which relies on experienced j workers rather than on " clearly established management  ;

controls" cannot be judged' satisfactory for the extended a

period.

d. From IR 89-80, p. 4, NRC states:

"The licensee has a program to review the appropriateness and technical adequacy of completed i maintenance-activities. However, these reviews are .

not being completed in a timely fashion. There is  !

a backlog of completed Maintenance Raquests awaiting such review."  ;

The fact that' Maintenance Request review is not completed in a timely fashion is indicative that an age-related problem might be allowed to continue, resulting in ,

failure before correction.

e. From IR 89-80, p. 4, NkC states:

" Lack of effective policy and procedures for controlling and updating manufacturer technical '

manuals is a weakness in the current maintenance program as implemented."

The lack of timely updating of vendor manuals is significant for the proposed extended period since information ,

44

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on aging equipment from the vendors will be critical to

, r prevent failure of safety' equipment, t 4

f. From IR 89-80, p. 4., NRC states  ;

" Licensee management has not established a program, goals, or the necessary training for the use of PRA concepts in its maintenance activities." ,

Since certain age related failures can only be prevented through reliability studies and other PRA techniques (See applicant's reliance on this in "The Assessment", Section ,

3.4.3), this weakness in the maintenance program exposes its unsatisfactory nature for the proposed action. See also the discussion for Contention IX.

t

g. From IR 89-80, p. 12, NRC states (emphasis added): t "This [ performance data) is not formally documented I and there is not an established mechanism...to disseminate the information to higher management or to pursue an issue, as in the case of an adverse trend ... there was no formal plan available +

indicating the existence of a long term program to reverse the trend. Also, there was no documented I

evidence that these analyses had becn communicated l to plant or corporate higher management for their l review and evaluation."

1' This demonstrates as false and unreliable, applicant's statement from "The Assessment," p. 28, " Equipment is replaced when required or when a trend analysis indicates that equipment reliability or expected life has decreased."

45

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j i

h. Applicant does.not have in place, and has made'no s commitment for a Reliability-Centered Maintenance (RCM)

I program (EPRI NP-6152, January 1989), despite its demonstrated 1

effectiveness,'specifically for life-extension:

"These (RCM) characteristics also facilitate RCM's interaction with licensing, plant life extension, and reliability programs." (EPRI NP-6152, p. 1-8)

h. Maintenance has an effective on safety. In a proposed rule, 53 FR 47823, NRC has stated:

"A limited NRC examination of nuclear. power plant maintenance has found a wide variation in the effectiveness of these programs. At some plants, maintenance has been a significant contributor to plant reliability problems and hence, is of safety concern."-

Weakness in the maintenance programs, such as those j identified by IR'89-80, are of specific risk, according to NRC Draft Regulatory Analysis for Proposed Rulemaking on Maint$ nance of Nuclear Power Plants, November 1988:

! " Risks to public health and safety will be reduced l substantially, as the performance of plants with weak maintenance is improved." (p. xi)

This document (Draft Regulatory Analysis ...) provides a summary and identification of aspects of maintenance problems relevant to the proposed action.

E l

l

i. Applicant has not incorporated the lessons learned from NUREG-1333, Maintenance Approaches and Practices in Selected Foreign Nuclear Power Programs and Other U.S.

Industries: Review and Lessons Learned, Draft for comment, 46

I

. November 1988. Applicant has not demonstrated 1) a proactive 1 maintenance program as opposed to reactive maintenance; 2) use

?

of reliability-centered approach, 3) use of an integrated i information system for collecting date and monitoring the  ;

effectiveness, and 4) derivation of planning and scheduling ,

i from overall program objectives. (NUREG-1333, p. ix)  !

Applicant does not use performance indicators. $

This is indicative that applicant's maintenance program ,

'is not acceptable for the reliance necessary for the proposed life extension.

j. The weakness of applicant's maintenance program  ;

continued to be present at the NRC inspection (IR 89-80),

despite previous identification by applicant's management consultant, LRS Incorporated. See excerpts :

"VY plant management should review the industry practice in maintenance planning to determine if the industry trend-in establishing formal maintenance planning groups would aid in increasing  ;

productivity in the maintenance area." (LRS Report, #3-88, p. 10)

"The backlog of maintenance requests (MR's) is large and there is no visible push to rectify the I situation. Better computerization of the MR system would be beneficial. Also, better utilization of the Assistant to the Operations Supervisor in handling corrective MR's and preventive maintenance l (PM) activities through MR's, to reduce or I eliminate Shift Supervisor (SS) review time and/or l' sign-offs before and after PM's are conducted, could significantly reduce the administrative burden on the SS's." (LRS Report, #3-88, p. 7) l

" Generally, the operators do not feel that Vermont Yankee pays sufficient heed to their desires for hardware repair and replacement." (LRS Report, #3-88, p. 7) 47

~ , , , . - , - , . , ~

i "One weakness was noted in the maintenance training program and that was its training skills ,

certification ... Vermont Yankee has parts of this ultimate program, but much more work has to be done in the certification area." (LRS Report, #1-89, p.

4)

"The (maintenance) program lacks much of the formality found at many other plants." (LRS Report,

1. 1-89, p. 6)

k. Following the NRC Inspection (IR 89-80), LRS concluded that the program informality: ,

" leaves VY vulnerable to several things, such as attrition in the experienced maintenance organization, communications problems, incidents involving vendor data shortcomings, procedural inadequacies, or even human error." (LRS Report,

1. 2-89, pp. 2-3) ,

LRS recommends: [

"A long range project should be initiated to upgrade the formality of the maintenance activities through improved and modernized programs. High on the list of items to be pursued are the Master Equipment List and Vendor Manual updates."

Until such maintenance improvements are made and tested with satisfactory performance, the proposed action cannot be considered.

1. Improper maintenance can effect the life of plant systems, structures and components:

" Improper Maintenance. Industry data shows that improper maintenance can impact the life of valves.

In addition to switch settings, incorrect installation of replacement parts and incorrectly reassembly sequences have caused a number of valve failures in the industry." (EPRI NP-5836M, p. 3-48)

m. Applicant has a history of maintenance induced problems and incorrectly executed maintenance and surveillance 48 l

r

[

! 'l programs. This fact is apparent in the failure to maintain the Uninterruptible Power Supply (UPS) and the toxic gas ,

monitors to meet reliability standards. This is shown in INPO l reports on overall industry equipment reliability and specifically for Vermont vankee. ,

The "tip of the iceberg" of this fact is also shown in the Licensee Event Reports through the life of the plant.

Maintenance and surveillance related items for the last two ,

years are presented for examples:

l LER 89-24 Missed residual heat removal system valve leakage surveillance LER 89-23 Failure to perform daily instrument checks on the low pressure coolant -

injection system crosstie monitor i

LER 89-17 Service water check valves inoperable

• due to corrosion of internal parts LER 89-14 Reactor core isolation cooling system -

inoperable due to motor burn out on RCIC-21 valve LER 89-10 Missed diesel fuel oil sample to inadequate administrative controls l LER 88-14 Missed effluent sample l

L LER 88-13 Missed effluent sample due to i personnel error LER 88-09 Reactor scram due to suspected turbine pressure control malfunction (inadequate non-safety related maintenance caused reactor trip)

LER 88-05 Potential loss of standby gas treatment train due to extension of loop seal 49

l I

LER 88-04 Isolation of radiation monitors due to ]

personnel error LER 88-03 Missed surveillance on high water )

level in scram discharge volume trip channel and alarm

n. Further evidence of the inability of applicant's maintenance and surveillance program to provide the reliance claimed is the failure of the containment leakage monitoring program for successive tests (LER 84-11, LER 85-07, LER 87-07, LER 89-07)

o. Applicant has not retained and cannot document adequate maintenance, surveillance and storage methods during the construction period, which affects component-life. See ,

the discussion for Contention VI.

i

p. The significance of the maintenance and surveillance program is heightened by the fact that applicant proposes to opera'.e systems, structures and components beyond 40-years of life (see the discussion of contention VI).

4

q. It is shown above that both past and present maintenance and surveillance programs are judged as weak.

These weaknesses cast doubt on the applicant's ability to rely on these programs to discover and correct aging equipment failures. Thus, the application must be denied.

50

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1

r. .one adverse consequence of failure to properly.

1 maintain equipment is the failure to meet-ECCS criteria in the l case of a postulated design basis accident.

s. This plant's margin of safety has fallen below the minimum acceptable margin of safety for operation in the extended period due to the lack of a'sufficiently-effective '

and comprehensive maintenance and surveillance program.

VIII.

Applicant has not demonstrated the capability of the Mark I containment used in this plant to withstand and mitigate design basis and severe accidents during the proposed period of extended operation. The most significant factor which has not been adequately analyzed by the applicant is the impact of aging during construction and during the proposed extended operation on the Mark I containment.

a. Contrary to applicant's conclusion in Section l

I 3.4.4.1 of "The Assessment", significant unresolved items '

t

\

exist concerning the adequacy of the Mark I containment to mitigate severe accidents (such as early containment failure through liner melt-through, radioactivity release through l' systems which bypass containment mitigation provisions, and L

radioactivity released from failure of interfacing systems).

See, among others, the following documents:

51 l

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1. Generic Letter 89-16, Installation of a Hardened- ,

O. Wetwell. Vent - This letter states "the staff undertook a -

'/: program to determine if any actions should be taken, on ,

. c, a generic basis, to reduce the vulnerability of BWR. Mark ,

I containments to severe accident challenges...the staff l

' identified a number of plant modifications ...  ;

1) improved hardened.wetwell vent capability, 2) improved-reactor pressure vessel depressurization system reliability, 3)an alternate water supply to the reactor vessel and drywell sprays, and 4) updated emergency  ;

procedures and training." The letter further identifies that all but the wetwell vent will be evaluated further in Individual Plant Evaluations (IPE's). j

2. Generic Letter 88-20, Supplement 1, Initiation'of the Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54 (f) - This letter identifies specific Mark I items. IPE information is identified in NUREG-1335, Individual Plant Examination:

Submittal Guidance, August 1989; and NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, June 1989.

3. NUREG/CR-5124, Interfacing Systems LOCA: Boiling Water Reactors, February 1989, identifies failure of interfacing systems as a major unresolved problem for ,

Mark I BWRs like Vermont Yankee.

1 52

. . - . . . - . .~ . ~ - -. - -- _ -_____ -_ . -. . _ - - . - - - - - -

f

4. Draft report', Review of the Status of the Mark I ,

BWR Liner Melt-Through. Issue, Submitted to NRC November 27, 1987, comments / report by G.A.Greene, which finds p

that the question of overpressurization versus leak-before-break would be moot because melt-through would  ;

occur "within minutes of (core) debris-liner contact."

(from Inside NRC, January 4, 1988). See also NUREG-1079, Estimates of Early Containment Loads.from Core Melt Accidents, December.1985.

5. Notes and records of OECD Nuclear Energy Agency (NEA) meeting of May 1988 concerning international opinion on. filtering severe accident containment vent >

paths. It is clear from the records of this meeting i that filtering is an unresolved issue which bears upon the applicant's Mark I containment. .

All of the above severe accident concerns are related to aging, effective in the proposed extended period. Liner thinning will further hasten corium melt-through. Failure of aging. equipment will defeat systems relied upon for prevention and mitigation, and specifically on interfacing systems.

b. Aging mechanisms for light water reactor (LWR) structural materials which make up portions of the primary containment boundary (carbon steel, low alloy steel, stainless steels and Inconel) are evaluated in EPRI NP-5461, Component Life Estimation: LWR Structural Materials Degradation 53

]

t 4

LMechanisms, September 1987. The following observations are made :- j "The family of materials degradation mechanisms classified as corrosion mechanisms are characterized by:  :

the deterioration or destruction of a material reaction with its environment, or extractive metallurgy in reverse.

The mechanisms considered are: General corrosion and Wastage, Stress Corrosion Cracking, Intergranular Attack,. Crevice corrosion, Hydrogen

. Embrittlement, Erosion and Erosion / Corrosion, Microbiologically Induced Corrosion and' Pitting."

(EPRI NP-5461, p. C-4)

"All the listed corrosion mechanisms have been observed in LWR's during construction ....": (EPRI NP-5461, p. C-5) q Thus, all systems and components constructed with the covered materials, such as containment liner and containment isolation valves, are shown to begin aging in the construction period, before isruance of the operating license.

c. Additional aging mechanisms identified in EPRI NP-5461, such as fatigue and embrittlement mechanisms, plastic deformation, dynamic mechanisms, creep, fretting and wear, began with first operation during preoperational testing before issuance of the operating license.

d. EPRI NP-5181M, BWR Pilot Plant Life Extension Study at the Monticello Plant: Phase 1, May 1987, identifies aging mechanisms for BWR systems, structures and components.

General corrosion, which began from manufacture and 54 l 1

construction, is identified.as an active mechanism for the j drywell metal shell, suppression chamber and concrete structures (EPRI NP-5181M, p. 5-28). 1 I

e. Concrete aging mechanisms of creep, shrinkage, I

cracking and chemical attack, all of which began at 1

construction and before issuance of the operating license, are i

identified as the aging mechanisms for concrete structures

-l

'(EPRI NP-5181M, p. 5-8). 1

f. EPRI NP-5836M, BWR Pilot Plant Life Extension Study at the Monticello Plant: Interim Phase 2, October 1988, corroborates that significant aging mechanisms begin before issuance of the operating license. For example, aging <

mechanisms of humidity, thermal gradients and related component debilitation such as cracking of concrete and corrosion of metal substrates, are identified for containment coatings, including the statement:

"The effects that debilitate the protective coating systems are related to ... construction ...." (EPRI NP-5836M, p. 3-71)

g. EPRI NP-5181M indicates at pp. 5-20,21:

"Results from the specified testing and surveillance activities at Monticello include the following "significant" findings: 1. Drywell Shell Corrosion at Concrete Floor Interface ... The visual state of the corrosion indicated that the oxidation process at the shell had initiated several years ago... Continued propagation of corrosion at this location without repairs / preventative measures could produce significantly reduced wall thickness within this critical section of the drywell shell."

55 ,

This aging' mechanism began at the time of construction I

and become apparent only after years of operation. Applicant, ins"The Assessment", Section 3.4.4.1, Primary Containment, '

fails to' discuss this, or any other aging mechanism'of primary containment.

h.- The construction period is indeed significant in

. aging and life considerations:-

" Incidents which occurred during the construction period may have an imoact on component life."

(EPRI NP-5181, p. 5-13)

Some items required to assess the construction period effects are a review of nonconformances and special l construction procedures, dispositions of significant process deviations (welding, heat treatment, etc.) and out-of- ,

tolerance' dimensions, number and location repair welds, preoperational history, nonconformances of preoperational tests, storage environments and methods, and scope and results ,

of nondestructive examinations (EPRI NP-5181M, p. 5-13).

l Applicant provides no indication in "The Assessment" that this review was completed or that it is even possible.

1' L 1. Above it has been shown that the primary containment began the aging process before the issuance of the operating license. (Also, see the discussion for contention VI which shows aging effects of many other systems, structures and components, which support the performance of the Mark I 56

s

'b*.b:. . .

. i l

containment, ure significant, and began before the issuance of 1 the operating license.

i j.- Aging effects of the containment structure are significant encugh to require repair:

"Addi,tional maintenance items were recommended where appropriate to increase life expectancy ... .

Repair cracks, spalls and popouts following survey of critical-concrete structures." (EPRI NP-5181M,

p. 5-2.1)

k. . Containment isolation valves have age failure mechanisms which began before the issuance of the operating license and which will continue in an exacerbated manner in the period requested for extension. Such age failure

mechanisms are identified in EPRI NP-5836M, Section 3.7.

1. Applicant's maintenance program has weakness in both past and present applications - see contention VII. The adequacy of the containment for the extended period is put in doubt by past and future inadequacies in the maintenance program.

m. Applicant has failed successive containment leakage tests - see LER 84-11, 85-07, 87-07 and 89-07. This fact illustrates that oresent leakage criteria cannot be met due to improper maintenance and aging of valves. Applicant presents no evaluation in "The Assessment" of these failures, or any justification that containment criteria can be met, for the 57

nu 4

design basis accidents or severe accidents, in the present, not to mention in the extended period,

n. Applicant fails to identify in "The Assessment" that gross: age failure of the drywell paint system has and is occurring. In letter, BVY 89-69, Pelletier to NRC, applicant identifies this failure, along with 18 references indicating a long history with this problem. While applicant claims no present safety problem from paint chips and no future loss of integrity, no confidence exists in these statements, based on past performance. We are unaware of NRC review and acceptance of applicant's letter. However, the letter does not evaluate the adequacy of the coating system in the extended period.

(See Llso other industry related problems, such as torus wall thinning experienced at Nine Mile Point.)

o. One adverse consequence of the failure of coating

- systems and production of other corrosion products is the fouling of ECCS pump suctions such that ECCS criteria is not met. ECCS pump susction must also be evaluated with regard to !

~

the effects of operation and misoperation of a proposed a

, hardened containment vent. l L

p. This plant's margin of safety has fallen below the  !

I minimum acceptable margin of safety for operation in the extended period due to the aging of the Mark I containment.

58

l IX.

The applicant cannot obtain an extension of its existing operating license at least until it satisfactorily completes a l l

probabilistic risk assessment ("PRA") for this plant and determines and identifies in that PRA all modifications necessary for risk reduction of a severe accident during extended operation of the plant, commits to implementation of these modifications before the beginning of the extended period, and incorporates the cost of such modifications into economic evaluations (see Contention III).

a. Applicant must prepare an Individual Plant (probabilistic) Evaluation (IPE), specifically for Vermont Yankee, by November 23, 1991 (Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54, November 23, 1988).

b. Aging is clearly effects the considerations in'an

.IPE, specifically the failure probabilities, which must be greater for older plants which used antiquated codes and standards.

c. See Contention V, VI and VIII for a discussion of the impact of aging and inadequate design codes on plant safety.

d. For example, some specific unknowns and uncertainties which are affected by aging in the proposed extended period are:

59

" )

H ,

NUREG/CR-5124, Interfacing Systems LOCA: Boiling l

> Water Reactors,~ February 1989, identifies failure of interfacing systems as a major unresolved problem for Mark I BWRo like Vermont Yankee.

Draft report,EReview of the Status of the Mark I '

BWR Liner Melt-Through Issue, Submitted to NRC November 27, 1987, comments / report by:G.A.Greene, -

'which finds that the question of overpressurization ,

versus leak-before-break would be moot because melt-through would occur "within minutes of (core) debris-liner contact.". (from Inside NRC, January ,

4, 1988). See also NUREG-1079, Estimates of Early l Containment Loads. fret Core Melt Accidents, December 1985.

Notes and records of OECD Nuclear Energy Agency ,

(NEA)~ meeting of May 1988 concerning international  ;

opinion on filtering severe accident containment '

vent paths. It is clear from the records of this meeting that filtering is an unresolved issue which  !

bears upon the applicant's Mark I containment. .

All of the above severe accident concerns are related to aging, effective in the proposed extended period. Liner ,

thinning will further hasten corium melt-through. Failure of aging equipment will defeat systems relied upon for prevention and mitigation, and specifically on interfacing systems.  ;

e. The industry, and the applicant specifically, have not been able adequately to assess aging of plant equipment.

Despite deterministic statements that systems, structures and i

components are designed for certain qualified lives, nevertheless unplanned failures of these equipment regularly occur. See for example, applicant's LER's in this present year alone:

60

l LER 89-21 Failure of RM-16-19-1B Primary n containment High Range Radiation I Monitor l ft j LER 89-19 Inadvertent Primary Containment Isolation System Actuations Due to a Malfunction of the Reactor Building Ventilation Radiation Monitor Sensor Converter LER 89-17 Service water check valves inoperable j due to corrosion of internal parts LER 89-14 Reactor Core Isolation Cooling System j Inoperable due to Motor Burn Out on i RCIC-21 Valve

)

LER 89-07 1989 Appendix J Typp B and C Failure  ;

Due to Seat Leakage l LER 89-04 Snubber MS-35 Fluid Loss and Function f Failure Due to Worn Rod Bushing i I

LER 89-03 Inadvertent Primary Containment I Isolation System Actuations Due to l Spurious Spt.ses on the Reactor  :

Building Vei.tilation Radiation Monitor [

Some examples, from this current year, of the l l I industry's difficulty in understanding aging life issues are  !

l found in Information Notices 89-66, 89-65, 89-64, 89-36, 89- l 30, 89-20, 89-17, 89-15, 89-01. I l

f. Design errors also affect the risk evaluation for {

the extended period. Design errors are a fact in applicant's (

l l  !

f 3 Note that this LER alarmingly identifies four successive [

l failures of the Type C tests, in LERs 84-11, 85-07 and 87-07)!  !

L Followon report, "Drimary Containment Leak Rite Test Report," l

. BVY-89-64, July

• 1989, identifies, time-after-time, ace-failure  ;

o_f valves as the root cause for the failure or performance of the  ;

containment isolation system.)

61 i

l 1

L I

! plant. Based on past history, a statement that no-known (

design errors exist cannot be accepted for risk analysis. The j i

IPE must include a factor, related to the historical l t l performance of Vermont Yankee, to account for design errors j which will be detected in the future. See specifically:

LER 89-22 Reactor Building Closed Cooling Water i l' Return Motor Operated Valve 70-117 Not l i Powered from Emergency Bus as Required J by FSAR [

LER 89-09 Lack of Redundancy in Residual Heat l Removal Service Water Systems i These design errors have the potential of being  !

t exacerbated by aging in the proposed extended period, and the  !

potential of not functioning as designed when other equipment  !

fails through aging mechanisms. e f

g. The effects above must be treated probabilistically, and specifically in the IPE submittal due from applicant.

Specific factors which must be included for all PRA sequences l are: 1) age-failure factor, 2) older-plant (outdated fabrication i code and licensing basis) factor, 3) age-corroded liner factor,  ;

and 4) design error factor. f i

i l

62

,_ _- _.___m_ _ . _ - , . - - . _ _ . . _ . - . - -

E ,

' ~

h*A y_I /u; 3 4 ., y i

h.. Until a PRA is prepared, which' includes.these age J 4 .. ,

related. factors, operative for;the extended period, no, '

' 3 - , ' , .,

11 decision _concerning th'e proposed action can be made, i 3

E e

Respectfully submitted, [

,. 3

/I .

,/ s.

.' if (W

. /pl. A -vn _

Anthony Z. oidman i COHEN, MIL TEIh & HAUSFELD  !

Suite 60  :

1401 New York' Avenue, N.W.  :

Washington, DC 20005 }

.c (202) 628-3500 .i L

COUNSEL F02 THE STATE OF VERMONT  ;

r Dated: October 30, 1989  ;

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t , . - . - . . ,. , s ,, . , . , a - . - - - .

,-v-.,. --, , - .-,-n I

UNITED STATES OF AMERICA before the l NUCLEAR REGUIATORY COMMISSION ,

j

'89 OCT 31 P2:35 l

) ,

In the Matter of ) us i VERMONT YANKEE NUCLEAR ) Docket No. 50-271-01A E ,0.,,,ae,y l POWER CORPORATION ) (Operating License MW- l

) Extension) l (Vermont Yankee Nuclear ) l Power Station) )  :

)

CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing State of l Vermont Supplement To Petition to Intervene dated October 30,  !

1989, has been served upon the following persons by U.S. mail,  !

t first class, except as otherwise noted and in accordance with the requirements of 10 C.F.R. I 2.712.

• Administrative Judge

• Administrative Judge ,

Robert M. Lazo, Chairman Jerry Harbour  ;

Atomic Safety and. Licensing Board Atomic Safety and Licensing U.S. '

Nuclear Regulatory Commission Board -

Washington, DC 20555 U.S. Nuclear Regulatory i i Commission l Washington, DC 20555 ls

• Administrative Judge

• Ann P. Hodgdon, Esq. l Frederick J. Shon office of the General Counsel

Atomic Safety and Licensing Board U.S. Nuclear Regulatory L U.S. Nuclear Regulatory Commission Commission Washington, DC 20555 Washington, DC 20555 l

l *R. K. Gad, III, Esq. James Volz, Esq. j Ropes & Gray Interim Director for Pub.

l One International Place Advocacy Boston, MA 02110 120 State Street a Montpelier, VT 05602 ,

Adjudicatory File ,

Atomic Safety and Licensing Board Panel [

U.S.N.R.C. 9 Washington, DC 20555 l ~ Anthony Z(/Rdisman Dated: O

• Federal Express I

64 L ___ _ .____ -.. . _ _