ML20086S327

Response Opposing Ucs 2.206 Petition for Show Cause Re Emergency Feedwater Sys.Certificate of Svc Encl

ML20086S327
Person / Time
Site:	Three Mile Island
Issue date:	02/24/1984
From:	Baxter T GENERAL PUBLIC UTILITIES CORP.
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20086S331	List: ML20086S327 ML20087P865 ML20087P868
References
NUDOCS 8403010509
Download: ML20086S327 (45)

Text

I

, i 6

00CKETED U?tpc February 24, 1984

'84 APR -9 p!! :13 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ,

BEFORE THE DIRECTOR OF NUCLEAR REACTOR REGULATION In the Matter of )

)

GPU NUCLEAR CORPORATION ) Docket No. 50-289

) (10 CFR 2.206)

(Three Mile Island Nuclear )

Station, Unit No. 1) )

LICENSEE'S RESPONSE TO UNION OF CONCERNED SCIENTISTS' PETITION FOR SHOW CAUSE CONCERNING TMI-1 EMERGENCY FEEDWATER SYSTEM I. Introduction The Union of Concerned Scientists (UCS) filed with the Commission a ". . . Petition for Show Cause Concerning TMI-1 Emergency Feedwater System," dated January 20, 1984, which seeks an order suspending the operating license for TMI-1 "un-less and until the plant's Emergency Feedwater ('EFW') System complies with the NRC rules applicable to systems important to safety (including safety-grade, safety-related, and engineered safety feature systems)."

7A Copy Has Been Senf to PDR 030 OWs 3_

The Commission has referred the UCS Petition to the office of Nuclear Reactor Regulation for treatment as a request for action pursuant to 10 C.F.R. 9 2.206.1/ The Director of Nucle-ar Reactor Regulation has stated that he will issue a decision on the UCS Petition within a reasonable time, and requested that Licensee submit a response, in writing under oath or af-

~ '

firmation pursuant to 10 C.F.R. S 50.54(f), that addresses each of the issues identified by the petition.2/ Attached hereto, and under oath pursuant to section 50.54(f), is the "GPU Nucle-ar Technical Response to Union of Concerned Scientists' Peti-tion for Show Cause Concerning TMI-1 Emergency Feedwater Sys-tem."

Here, Licensee addresses the standards which govern the Director's decision to initiate or not to initiate a show cause proceeding, and the standards for taking immediately effective action if such a proceeding is initiated. The Response then evaluates the relationship between the UCS Petition and the TMI-1 Restart proceeding, and summarizes the facts which show that, under the applicable standards, no basis exists for the action requested.

1/ 49 Fed. Reg. 5005 (February 9, 1984); letter, January 27, 1984, Harold R. Denton (NRC) to Ellyn R. Weiss (UCS).

2/ Id.; Letter, January 27, 1984, Harold R. Denton (NRC) to Henry.D. Hukill (GPU Nuclear).

II. Standards for Deciding Whether a Show Cause Proceeding Should be Initiated Section 2.206 of the Commission's regulations provides a mechanism whereby members of the public may request initiation of an enforcement action to modify, suspend or revoke a li-cense, or for such other action as may be proper. It also vests authority in the director of the appropriate NRC office to decide whether to institute an enforcement action in the form of a show cause proceeding. The only criterion set forth in the rule itself for judging the sufficiency of a petition is the requirement that "[t]he requests shall specify the action requested and set forth the facts that constitute the basip.for the request." See 10 C.F.R. 5 2.206(a).

The apparent reason for the absence of a more specific standard in the regulation is that the decision to institute an enforcement action is not an adjudicative one, but rather is a matter of " prosecutorial" discretion.3/ See Consolidated Edison Company of New York, Inc. (Indian Point, Units 1, 2 and 3), CLI-75-8, 2 N.R.C. 173, 175 (1975). Some guidance on the standards to be applied is available, however, from previous decisions.

3/ In the same vein, a petitioner has no right to Commission review of a director's decision. 10 C.F.R. 5 2.206(c)(2).

In affirming a Director's decision denying a 2.206 peti-tion (while at the same time referring certain issues to an al-ready empaneled Appeal Board), the Commission stated that "[the Director] correctly understood that a show cause order would have been required had he reached the conclusion that substan-tial health or safety issues had been raised," and that ". . .

a mere dispute over factual issues does not suffice" as a basis for issuance of such an order. Indian Point, supra, CLI-75-8, 2 N.R.C. at l'/6 (1975).4/

The Commission reiterated the " substantial health and safety issue" standard in Northern Indiana Public Service Comoany (Bailly Generating Station, Nuclear-1), CLI-78-7, 7 N.R.C. 429, 433 (1978), aff'd, Porter County Chapter v. NRC, 606 F.2d 1363 (D.C. Cir. 1979), and rejected a claim that the Director erred in failing to permit petitioner to comment on, respond to, or cross-examine the views of the NRC staff:

[The Director) is not required to accord -

presumptive validity to every assertion of fact, irrespective of its degree of substantiation, or to convene an adjudica-tory proceeding in order to determine whether an adjudicatory proceeding is war-ranted. Rather, his role at this prelimi-nary stage is to obtain and assess the in-formation he believes necessary to make that determination. Provided he does not abuse his discretion, he is free to rely on 4/ This standard has been acknowle;. $ in dicta by the D.C.

and Seventh Circuits. Lorion v. NRC, F.2d 1472, 1475 (D.C.

Cir. 1983); Rockford League of Women Vorsrs v. NRC, 679 F.2d 1218, 1222 (7th Cir. 1982).

l l

O a variety of sources of information, including staff analyses of generic issues, documents issued by other agencies, and the comments of licensees on the factual alle-gations.

Id. at 432-33.

The Director in the past has denied 2.206 petitions on grounds relevant here. First, the Director will not readily initiate an enforcement proceeding to consider actions that are inconsistent with Commission policy. See, e.g., Rochester Gas and Electric Corporation (R.E. Ginna Nuclear Power Plant),

DD-82-3, 15 N.R.C. 1348 (1982). In that case, the Director found that current Commission policy does not require that the FORV and its solenoid operated air valve be designated safety grade, and that the issue was under generic study. According-ly, the Director rejected a request that a determination spe-cific to Ginna be made. Id. at 1351.

Second, the Director will not initiate an enforcement pro-ceeding where problems are being expeditiously remedied and there 's no significant interim hazard. Southern California Edison Company (San Onofre Nuclear Generating Station, Unit 1),

DD-81-19, 14 N.R.C. 1041 (1981). In San Onofre, the Director denied 2.206 petitions which were based on new seismic informa-tion because (1) seismic improvements had been scheduled, and (2) in the short term there was a low probability of ground mo-tion in excess of that against which the structures had been designed. Id. at 1042-46.5/

l 1

5/ In addition, the Director will not initiate an enforcement proceeding on issues that are being addressed by the Commission  !

! (Continued Next Page)_

l

(

l III. Standards for Taking Immediately Effective Action if a Show Cause Proceeding is Initiated If the Director determines that a substantial health or safety issue exists,6/ he should issue a show cause order pur-suant to 10 C.F.R. 5 2.202. Licensee would then have the op-l portunity to respond to the show cause order by written answer under oath or affirmation, and to request a hearing. 10 C.F.R. 5 2.202(b). If a hearing were requested, any order to modify, suspend or revoke the license would issue after, and be based upon the record of, that hearing.

The norm for administrative action modifying outstanding licenses embraces a prior opportunity to be heard. Consumers Power Company (Midland Plant, Units 1 and 2), CLI-73-38, 6 A.E.C. 1082, 1083 (1973). The Administrative Procedure Act provides a licensee a right to notice and an opportunity to achieve compliance prior to license suspension. An order to show cause may provide,.however, for stated reasons, that the (Continued) on a generic basis through rulemaking. See, e.g., _ Vermont Yankee Nuclear Power Corporation (Vermont Yankee Nuclear Power Station), DD-80-20, 11 N.R.C. 913, 914 (1980); Public Service Electric and Gas Company (Salem Nuclear Generating Station, Units 1 and 2), DD-80-19, 11 N.R.C. 625, 627-28 (1980) (spent fuel disposal issue precluded by waste confidence rulemaking).

6/ Licensee demonstrates, in part V, infra, that such an issue does not exist.

i

proposed action be temporarily effective pending further order if the public health, safety or interest so requires. 5 U.S.C.

9 558(c); 10 C.F.R. 5 2.202(f). The Commission has stressed i that summary administrative action is a " drastic procedure."

Midland, supra, CLI-73-38, 6 A.E.C. at 1083. "Such action, l

unless warranted by comoelling safety considerations, can have i

serious consequences." Id. (emphasis added).

The UCS Petition does not state that temporarily effec-i

tive, immediate interim relief is sought. However, UCS seeks to accomplish the same result by asking the Commission to keep in effect the present suspension of the TMI-1 license pending resolution of a new show cause proceeding

. . . the Commission now has under consid-eration action which would allow TMI-1 to operate by lifting the "immediate effec-tiveness" of its orders of July and August, 1979. Before allowing TMI-1 to operate, it is vital that the Commissioners address and resolve these safety issues. . . .

In the face of the information pres-ented herein, authorization of restart by the Commission would violate its statutory mandate to ensure the public health and safety.

UCS Petition at 2-3. Since the Commission hopes to decide in June, 1984, whether or not to lift the immediate effectiveness of its suspension orders,7/ any show cause proceed'ing 7/ Commission Memorandum to the Parties to the TMI-l Restart Proceeding, Tentative Commission Views and Plan for Resolution of Management Integrity Issues Prior to Restart, January'27, 1984.

instituted in response to this UCS petition is unlikely to be completed prior to a Commission decision on restart.

It is fundamental and imperative, however, as well as re-quired under the law, that this UCS petition -- which seeks suspension of the TMI-1 operating license on grounds different than those which underlie the Commission's 1979 suspension or-ders -- be decided on its own merits and without regard to the fact that the license already is temporarily suspended. This applies to the Director's decision on whether or not to initi-ate a show cause proceeding, as well as any decision on immedi-ately effective relief if a proceeding is initiated. UCS is not automatically entitled to continued suspension simply be-cause the license has been temporarily suspended on other grounds and may be reinstated in the near future.8/ Any sus-pension of the license must flow from the merits of the peti-tion itself. In the same vein, Licensee has all the rights and privileges of its license, under the Administrative Procedure Act, to the extent they are outside the scope of the existing suspension orders and the bases upon which they were issued.

8/ Neither is UCS entitled to expedition in the treatment of its petition simply because of the schedule of the Restart pro-ceedings. In this regard, UCS makes no presentation as to why its petition could not have been filed considerably earlier.

Staff Safety Evaluations on environmental qualification of safety-related electrical equipment at TMI-1, and on seismic qualification of the TMI-l emergency feedwater system were is-sued in December, 1982, and in August, 1983, respectively.

IV. Relationship of the Petition to the Restart Proceeding i

Steps to improve the qualification and reliability -f the EFW system were identified as issues for the Restart proceeding in the Commission's Order and Notice of Hearing, CLI-79-8, 10 N.R.C. 141, 144-45 (1979), and were addressed extensively by the Licensing and Appeal Boards in their decisions on design issues. See LEP-81-59, 14 N.R.C. 1211, 1353-75 (1981);

ALAB-729, 17 N.R.C. 814, 831-36 (1983). On January 27, 1984, the Commission issued an Order in the Restart proceeding iden-tifying the issues in ALAB-729 on which it had decided to take review. Briefs have not yet been filed, but it is clear that the EFW system will be addressed under the issues being re-viewed.

a The pendency of the Restart adjudicatory proceeding has clear legal implications for portions of the UCS Petition. In Indian Point, suora, the Commission agreed with the licensee's argument that it should not be forced to argue its case in two separate forums (2.206 and an operating license proceeding) si-multaneously:

[P]arties must be prevented from using 10 j C.F.R. 2.206 procedures as a vehicle for reconsideration of issues previously decid-l ed, or for avoiding an existing forum in

, which they more logically should be pres-

, ented.

CLI-75-8, 2 N.R.C. 173, 177 (1975).9/

9/ The Commission recognized that the Indian Point 3 licens-ing proceeding was no longer before the Licensing Board, but (Continued Next Page)

The Commission reaffirmed the Indian Point rule in Bailly, supra, CLI-78-7, 7 N.R.C. 429 (1978), and in Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2 ) , CLI-81-6, 13 N.R.C. 443 (1981); see also Rockford League of Women Voters v. NRC, 679 F.2d 1218, 1222 (7th Cir.

1982) (acknowledging the rule on avoidance of duplicative fo-rums).

Consequently, where an issue has already been decided in the Restart proceeding, a 2.206 petition should not be enter-tained to the extent it seeks reconsideration of those deci-sions. Further, where issues in the Restart proceeding are be-fore the Commission under its January 27, 1984 Order, a 2.206 petition is duplicative if it seeks to raise issues on which UCS may yet prevail in the adjudicatory forum.

Licensee has identified, in the attached Technical Re-sponse, where issues raised in the UCS Petition have been con-sidered in the Restart proceeding. These issues include: re-lationship of the Integrated Control System to the EFW system (UCS Petition at 20, 21); condensate storate tank level instru-mentation (UCS Petition at 5, 10, 13); EFW flow instrumentation (UCS Petition at 21-25); and the Main Steam Line Rupture (Continued) was pending before the Appeal Board; however, the Commission noted that further factual inquiry was not precluded.

CLI-75-8, 2 N.R.C. at 177.

f 1

Detection System insofar as it might cause the inadvertent isolation of feedwater (UCS Petition at 25-29).10/ In addi-tion, the Commission is considering whether the entire subject of environmental qualification of electrical equipment was properly excluded from the Restart proceeding. Order (unpublished) at 1-2 (January 27, 1984).

2 While Licensee has addressed, in the attached technical i response, the merits of all of the issues raised in the UCS Pe-i tition, the Director would be on firm ground in denying the pe-t

! tition summarily where it overlaps with the adjudicatory Restart proceeding.

V. The UCS Petition Does Not Provide an Adequate Basis for the Action Requested When UCS summarizes the action requested by its petition, it is that the TMI-1 license be suspended unless and until the EEW system complies with the NRC rules applicable to systems important to safety (including safety-grade, safety-related, and engineered safety feature systems).11/ UCS Petition at 1, 10/ The Commission's Order of January 27, 1984, is not clear on the scope of the MSLRDS issue. It is Licensee's position,

, however, that the potential for containment overpressurization j due to a failure to isolate.feedwater during.a main steam line l break event is not encompassed by the Restart proceeding.

l 11/ In the TMI-1 Restart proceeding, both the Licensing Board and the Appeal Board have rejected UCS's views on the proper 1 interpretation of the General Design Criteria (e.g., the term' i '

"important to safety") and the situations in which structures, systems and components should be designed to meet safety-grade

~

(Continued Next Page) l l 1

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

, 29. To the extent that this request is broader than the scope of the five more specific sections of the " FACTS" portion of the petition (i.e., UCS Petition at 5-29), the action requested should be denie'd because the petition does not "

set forth the l facts that constitute the basis for the request." See 10 C.F.R. 9 2.206(a).

A. The Backfit Process Applied to the TMI-1 EFW System Relevant to the UCS request to suspend this license, but ignored in the petition, is the applicability of current NRC requirements, used now in the licensing review of applications, to the EFW system for TMI-1. As noted in the Staff's 1973 Safety Evaluation on the TMI-1 operating license application, construction of the plant was 60 percent complete and the Final Safety Analysis Report had been filed with the AEC before pub-lication of the General Design Criteria in 1971.12/ See At-tachment at 2.

, The EFW system was not classified as a safety system at the time TMI-1 was designed and constructed.13/ The system is (Continued) criteria. See LBP-81-59, 14 N.R.C. 1211, 1342-49 (1981);

ALAB-729, 17 N.R.C. 814, 872-78 (1983).

12/ The proposed criteria were published in 1967, and(TMI-1 _

was constructed and designed to meet the intent of the s proposed

, criteria. Attachment at'2. '

13/ One member of the Staff has stated that full recognition of the EFW system as an engineered safety feature at pressur-(Continued Next Page) ,

, ,c ,

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

i undergoing extensive medifications, however, to upgrade it to full safety-grade status. See ALAB-729, 17 N.R.C. 814, 924-25 i

(1983). These modifications have been undertaken and are being

~

undertaken consistent with various NRC requirements, instruc-tions and guidance -- including but not limited to the adjudi-catory decisions in the TMI-1 Restart proceeding and under the general umbrella of the agency's TMI Action Plan.14/

, The gravamen of the UCS complaint is not with the improve-ments yet to be made to the TMI-1 EFW system, but with the schedule for implementing those modifications. It appears to

]

be the UCS position here, as it was in the TMI-1 Restart pro-

! ceeding, that whenever a safety improvement is endorsed as worthwhile, the plant by definition is not safe to operate until the improvement is implemented.15/ In short, UCS rejects i

s (Continued) ized water reactors did'not occur until the publication of the 1975 Standard Review Plan. Memorandum for Darrell G. Eisenhut.

from Roger J. Mattson, Analysis and Recommendations Related to l Plants Without Seismically Qualified Auxiliary Feedwater Sys-l tems, August 8, 1980. The Staff also stated in 1979 that j

recent design reviews treat the EFW system as a safety system in PWRs. NUREG-0578, TMI-2 Lessons Learned Task Force Status ,

Report and Short-Term Recommendations, at A-30.

14/ Also relevant to the UCS petition are the generic NRC pro-j grams for environmental qualification of safety-related elec-trical equipment and for improving the seismic qualification of EFW systems.

I 15/ Throughout its petition, UCS consistently equates (and ,

l misleadingly attributes its own view t'o Lic~ensee and others) a l planned' modification at TMI-1 during the Cycle 6 refueling out -

age with evidence that the plant now is unsafe for. operation under applicable NRC requirements.

l 13-4 s.-, w~ pv.--m ,

e y., w - m , o, - - p

the concept, endorsed by the NRC and reviewing courts, that

backfitting safety improvements to operating plants involves the exercise of judgment and may be accomplished in a phased manner over time.

The Commission's backfit regulation, 10 C.F.R. 5 50 109(a), provides that "[t]he Commission may . . . require the backfitting of a facility if it finds that such action will i provide substantial, additional protection which is required for the public health and safety or the common defense and se-a

! curity." In promulgating that regulation, the Commission stat-i ed that: "The rapid changes in technology in the field of atomic energy result in the continual development of new or im-I proved features designed to improve the safety of production and utilization facilities." 35 Fed. Reg. 5317 (1970). Taking steps to improve safety does not mean, however, that a facility is unsafe without the improvements. In applying the backfit rule, the Diractor previously has held that a decision to ret-refit an existing facility does not necessarily imply that it is unsafe, but rather that substantial benefit to the public health and safety can be attained. In the Matter of Petition Recuesting Seismic Reanalysis, DD-80-1, 11 N.R.C. 153, 166 (1980). -

The UCS concept of the appropriate standard for deciding whether and when to require =- modifications at operating plants was rejected by both the Licensing and Appeal Boards in the i

l l

TMI-1 Restart proceeding. The Appeal Board's discussion applies here with equal force:

In our view, UCS' fundamental dis-agreement with the Board is over the issue of relative degree of safety assurance that should be required before TMI-l is autho-rized to resume operation. UCS would seem to have us conclude that actions should be deemed "necessary" any time they might po-tentially produce substantial and addition-al protection to the public health and safety. Under such definition, actions would be "necessary" every time a potential safety improvement appears on the horizon, even if the technology to produce the im-provement is only in the developmental stages. Implicit in UCS' argument is the notion that nuclear power plant operation cannot be considered reasonably safe as long as scientific efforts are under way to develop new and better safety features. As we read the Licensing Board's decision, it concluded simply that safety should proper-ly be assessed on the basis of whether present systems can assure reasonable pro-tection of the public health and safety.

Such an approach is generally consonant with the requirements of the A' comic Energy Act. See Citizens for Safe Power v. NRC, 524 F.2d 1291, 1297 (D.C. Cir. 1975); Nader

v. NRC, 513 F.2d 1045, 1052-54 (D.C. Cir.

1975). It is also in accord with long standing agency practice. See, e.g.,

Petition for Shutdown of Certain Reactors, CLI-73-31, 6 AEC 1069, 1070-71 (1973);

Metrocolitan Edison Co. (Three Mile Island Nuclear Station, Unit No. 2), ALAB-486, 8 NRC 9, 46 (1978). Cf. ALAB-697, supra, 16 NRC at 1272 and ALAB-698, supra, 16 NRC at 1299-1301.

ALAB-729, 17 N.R.C. 814, 827-28 (1983) (footnotes omitted).

The Commission also has recognized, beyond the backfit

\ -=-

rule itself, the appropriateness of prioritizing improvements l

and implementing them in a deliberate manner over time. Ad-dressing the TMI Action Plan, the Commission has stated:

. . . [M]any actions were taken to improve safety immediately or soon after the acci-dent. These actions were generally consid-ered to be interim improvements. In scheduling the remaining improvements, the availability of both NRC and industry re-sources was considered, as well as the safety significance of the actions. Thus, the Action Plan approved by the Commission presents a sequence of actions that will result in a gradually increasing improve-ment in safety as individual actions are completed and the initial immediate actions are replaced or supplemented by longer term improvements.15/

Revised Statement of Policy on Further Commission Guidance for Power Reactor Operating Licensees, 45 Fed. Reg. 85236, 37 (1980).

B. The UCS Allegations The question raised by the UCS Petition is whether all contemplated improvements to the TMI-1 EFW system must be im-plemented prior to Cycle 5 operation. That question must be answered in the negative. As the attached Technical Response demonstrates, there is reasonable assurance that the ED1 sys-tem, based upon the improvements already made to both the equipment and plant procedures, will perform its function if

, .. 1s/ Other operating reactors have been allowed to operate while implementing the longer-term improvements called for by the TMI Action Plan, as well as by other generic NRC programs.

, l l

called upon to do so. See San Onofre, supra, DD-81-19, 14 I N.R.C. 1041 (1981) (no significant interim hazard).

The environmental qualification issue raised in the UCS Petition has been addressed by the Commission on a generic basis and through the promulgation of a rule -- 10 C.F.R. 6 50.49. The compliance date for TMI-1, under that regulation, is March 31, 1985. The Staff and its contractor Franklin Re-search Center have evaluated in detail the environmental quali-fication of TMI-1 safety-related electrical equipment for harsh environments created by postulated accidents. Licensee has responded to the outstanding concerns raised in that review.

The environmental qualification of the TMI-l EFW system, including the replacement of the EFW flow control valves' E/P converters with qualified I/P converters, will be completed by June, 1984 -- in time for the Commission's decision on plant restart. Attachment at 2-5.

The backfit of seismic requirements to the EFW systems of operating pressurized water reactors has been the subject of an extensive and generic NRC program. The Staff has stated that l

l

it has not decided to require EFW systems to be modified to meet facility design seismic requirements, but that its i ". . . plan is to increase the seismic resistance, where neces-

! sary, to ultima ~tely provide reasonable assurance that the sys-tem will function after the occurrence of earthquakes up to and including the SSE."13/

The Staff's contractor, Lawrence Livermore National Labo-t ratory, and the Staff itself have conducted an extensive review

of the seismic qualification of the TMI-1 EFW system. The Staff has concluded that at restart there is reasonable assur-ance that the system will be able to withstand a SSE and per-form its safety function. The basic UCS quarrel is with the -

fact that certain hardware improvements will not be made until the Cycle 6 refueling outage.19/ UCS Petition at 9-18. Li-censee has implemented, howe'ver, a plan of procedural actions which will enable the EFW system to perform its system function in the unlikely event it is called upon to do so following a

, seismic event during Cycle 5 operation. Attachment at 5-12.

The UCS position that only equipment can provide reasonable as-surance of safety is without merit. See, e.g., TMI-1 Restart, ALAB-729, 17 N.R.C. 814, 869-71 (1983) (reliance on reactor

! operators versus automatic safety systems).

Ig/ NRC Generic Letter No. 81-14, Seismic Qualification of Auxiliary Feedwater Systems, February 10, 1981.

l 19/ UCS cr.iticisms of the qualification-analyses are also ad-i dressed in the attached Technical Response.

i i

l

. l l

UCS next alleges that the EEW system does not meet the single failure criterion because there is only one flow control valve in the line to each steam generator, and because of the control of those valves by the non-safety-grade Integrated Con-trol System. UCS Petition at 19-21. The Director need not in- '

stitute a show cause proceeding to determine that the EFW sys-tem will not be fully safety-grade until after the Cycle 6 refueling outage. This is known to be the case. UCS simply assumes, however, that this criterion must be applied and must be met now. In fact, the EFW system was not designed to be ,

safety-grade and an upgrade program has been implemented by Li-censee and the NRC. In the Restart proceeding, the Licensing ,

and Appeal Boards have examined the EFW system and have ap-proved Cycle 5 operation, considering main feedwater transients and small-break loss-of-coolant accidents. Here, UCS raises main steam line breaks and steam generator tube ruptures.20/

As explained in the attached Technical Response, adequate pro-cedures are in place to provide for manual control of the EFW control valves during Cycle 5 operation. Attachment at 12-13.

The fourth subject of the UCS petition is the accuracy of the newly installed EEW flow instrumentation. UCS Petition at 20/ The interim steps taken to provide for-EEW control inde-pendent of the ICS were thoroughly considered and approved in the Restart proceeding. LBP-81-59, 14 N.R.C.'1211, 1285-86, 1362 (1981); ALAB-729, 17 N.R.C. 841, 833-34 (1983).

l

i . .

21-25. This requirement aror,e under the Restart proceeding, and the dispute now raised by UCS is over whether the indica-tors must be accurate to plus or minus ten percent at low EFW flow conditions. The accuracy criterion cited is not, as UCS suggests, a " lessons learned requirement." See Attachment at 14-15; NUREG-0737, Clarification of TMI Action Plan Require-ments, Item II.E.1.2. In any event, recent test data shows that the flow oscillations caused by cavitation of EFW control valves at low flow are within plus or minus ten percent at flows of 120 gpm and above. Since the operators do not rely on the EFW flow instrumentation at such low flow rates (i.e.,

below 225 gpm), the UCS Petition does not raise a safety issue.

Attachment at 14.

Finally, UCS asserts that the Main Steam Line Rupture De-tection System (MSLRDS) ". . . is not safety grade and requires modifications so that a single failure will not prevent isolation of main feedwater to the steam generator affected by a main steam line break." UCS Petition at 29. As UCS notes, the potential for inadvertent isolation of feedwater was con-sidered in the TMI-1 Restart proceeding as a part of the emer-gency feedwater reliability issues. LBP-81-59, 14 N.R.C. 1211, 1373-74 (11 1060-64) (1981). The Appeal Board found that the operators' capability to bypass the MSLRDS and manually open the EFW flow control valves if the MSLRDS isolates feedwater inadvertently is an adequate solution for resuart. ALAB-729,

_ _ ~. _ _ _ _ _ _ _ . __

a 17 N.R.C. 814, 834, 887-88 (1983). In an Order (January 27, t

- 1984) issued in the TMI-l Restart proceeding after the UCS Pe- l j tition was filed, the Commission called for comments on the ad-i equacy of Licensee's proposed solution to the MSLRDS " problem."

The issue here is the potential failure to isolate feedwater. Again, the UCS Petition is relevant only to Cycle 5 1

j operation, until a safety-grade MSLRDS is installed. In the attached Technical Response, Licensee fully describes the con-figuration of the MSLRDS and shows that, in the interim, there

is reasonable assurance that the containment will not overpressurize due to a failure of the MSLRDS to isolate i feedwater during a main steam line break. The MSLRDS is ade-quate from a single failure standpoint -- that is, a single ac-i tive failure (such as in a pressure switch, solenoid, control relay or a 125V DC power source) will not prevent isolation of feedwater and will not result in inadvertent isolation of feedwater. The MSLRDS is seismic Class I inside containment.

Following a main steam line break in the reactor building the I

system will function to isolate feedwater from the affected steam generator since the pressure switches (for MSLRD) to be installed in June, 1984 will be suitable for the accident envi-ronment. While electrica.1 separation between the redundant f

j circuits is not maintained outside containment, since some of them run in the same trays / conduits, a main steam line break inside containment will not cause a common mode failure of

-- + , , , ,r -- , . , u-- -- , - - - -

'. I 1

l these circuits. The MSLRDS, therefore, is adequate for op- i eration until the fully safety-grade modification is installed.

Attachment at 15-17.

VI. Conclusion Licensee has shown that UCS has not raised, in its peti-tion, substantial health and safety issues which would warran*:

institution by the Director of a show cause proceeding. The petition has been shown to be replete with factual error 21/ and to be based upon erroneous legal standards. In every other case, the concern raised has been recognized by Licensee and the NRC, and a plan is in place to implement equipment modifi-cations. In the interim, procedural and other hardware changes already made provide reasonable assurance that the EFW system 21/ The UCS petition is supported by the Affidavit of Robert D. Pollard, a self-styled " nuclear safety engineer" with UCS. The record in the TMI-l Restart proceeding'shows that Mr. Pollard, with a degree in electrical engineering, served on the AEC Staff from 1969 to 1974 principally as an instrumen-tation reviewer. Then for one and one-half years Mr. Pollard was a project manager with the NRC Staff where, in his own words, he became responsible for planning and coordinating li-cense application reviews. Robert D. Pollard Qualifications, ff. Tr. 8091. Mr. Pollard has never designed an electrical or mechanical system. Tr. 6467-68'(Pollard). Since leaving the NRC Staff in February, 1976, Mr. Pollard's only identified ex-perience is "following developments in the nuclear safety field" for UCS. UCS Petition, Pollard Affidavit at 2.

t

will function and that TMI-1 will operate without endangering the health and safety.of the public:

I Respectfully submitted, George F. Trowbridge, P.C.

4 Thomas A. Baxter, P.C.

David R. Lewis SHAW, PITTMAN, POTTS & TROWBRIDGE 1800 M Street, N.W.

Washington, D.C. 20036

.( 202 ) 822-1000 Councel for Licensee Dated: February 24, 1984 -

i N

, b s

r,

• 4

+

6 m f' ,

A r 1

i

+ w=l i

I l

- 2 3 '-

I C

t

. .o

,# . , c .- . , . . + ,n_ _ _ , , , _ , , , .

(

6 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE DIRECTOR OF NUCLEAR REACTOR REGULATION In the Ma:ter of )

)

GPU NUCLEAR CORPORATION ) Docket No. 50-289

) (10 CFR 2.206)

(Three Mile Island Nuclear )

Station, Unit No. 1) )

CERTIFICATE OF SERVICE I hereby certify that copies of " Licensee's Response to Union of Concerned Scientists' Petition for Show Cause Con-cerning TMI-l Emergency Feedwater System" and "GPU Nuclear Technical Response to Union of Concerned Scientists' Petition for Show Cause Concerning TMI-l Emergency Feedwater System" were served this 24th day of February, 1984, by hand delivery to:

Lillian N. Cuoco, Esquire Office of Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. James A. Van Vliet Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 and by deposit in the U.S. mail, first class, postage prepaid, to:

Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission

- Washington, D.C. 20555 Mr. Harold R. Denton Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

/ .

Ellyn R. Weiss, Esquire Harmon & Weiss 1725 Eye Street, N.W., Suite 506 Washington, D.C. 20006

% ~*-- .

Thomas A. Baxter, P.C.

t i

Attachment REVISED (March 26, 1984)

GPU NUCLEAR TECHNICAL RESPONSE TO UNION OF CONCERNED SCIENTISTS' PETITION FOR SHOW CAUSE CONCERNING TMI-l EMERGENCY FEEDWATER SYSTEM I. Introduction The UCS Petition describes what UCS alleges to be defi-ciencies in the Emergency Feedwater (EFW) System at TMI-l as it will be configured at the time of plant restart and throughout Cycle 5 operation. Each of the alleged deficiencies is ad-dressed below. While the UCS Petition concentrates on per-ceived shortcomings in the EFW system, these allegations should not be weighed in a vacuum, but rather should be assessed with an understanding of the capabilities of the EFW system and the substantial improvements made to the qualification and reliability of that system since the accident at TMI-2. In brief, Licensee has already implemented the following modifica-tions to the EFW system:

. safety-grade automatic starting of the EFW pumps;

. control of EFW independent of the ICS;

. condensate storage tank low-level alarm;

. safety-grade steam generator level indica-i tions, independent of the ICS;

. redundant two-hour air supply in the event of a loss of all AC power;

. EFW flow control valves' failure mode modified to fail open on loss of instrument air; and

. addition of flow-limiting cavitating venturis in each EFW line.

The additional modifications which will be. undertaken dur-ing the Cycle 6 refueling outage will result in a fully safety-grade EFW system. Contrary to UCS's assertion that Li-censee admitted, in our August 23, 1983 submittal, that the "EFW system needs to be upgraded" in order to provide increased

r reliability to mitigate design basis accidents (UCS Petition at 4, emphasis added), Licensee's submittal was merely noting the

" purpose" of the additional, long-term modifications. (Ref.

2.) Licensee stands by its original position that the TMI-1 EFW system is sufficiently reliable to allow operation during Cycle 5, pending completion of the long-term modifications.

II. Environmental Qualification UCS alleges that the TMI-l EFW system is not environ-mentally qualified, and begins the discussion in its petition on this point with a reference to General Design Criterion 4 of Appendix A to 10 C.F.R. Part 50. As relevant background for this and other references in the UCS Petition to the General Design Criteria, the Staff's finding associated with the issu-ance of the TMI-1 operating license is quoted:

The Three Mile Island Unit 1 plant was de-signed and constructed to meet the intent of the AEC's General Design Criteria, as originally pro-posed in July 1967. Construction of the plant was about 60% complete and the Final Safety Analysis Report (ESAR) had been filed as Amendment 12 with the Commission before publication of the revised General Design Criteria in February 1971 and the present version of the criteria in July 1971. As a result, we did not require the applicant to reanalyze the plant on the basis of the revised criteria. However, our technical review did as-sess the plant against the General Design Design Criteria now in effect and we conclude that the plant design conforms to the intent of these newer criteria. (Ref. 1 at 3-1.)

With respect to safety-related electrical equipment, the NRC has been pursuing environmental qualification (i.e., com-pliance with GDC-4) on a generic basis first through IE Bulle-tin 79-01B, and now through its regulation on environmental qualification of electric equipment important to safety for nu-clear power plants, 10 C.F.R. $ 50.49, which first became ef-fective June 30, 1982. Pursuant to section 50.49, TMI-1 is to achieve final environmental qualification of the electric equipment within the scope of that section by March 31, 1985.

The EFW system has been included in the overall evaluation of TMI-1 under these generic programs.

Focusing upon a steam line break outside of containment, t

UCS states ". . . GPU recognizes that the TMI-1 EFW system is not qualified for the hostile environmental conditions result-j ing from a main steam line break." UCS Petition at 6. What

I l

l GPU in fact stated in the reference cited by UCS, which de-scribes long-term modifications to the system, is that:

Equipment which is part of the EFW system or which is required to act in support of this system and which is located in the Intermediate Building, shall either be upgraded to be qualified for the hostile environmental conditions resulting from a Main Steam Line Break (MSLB) in this building or be replaced with qualified equipment or be relocated to an environmentally acceptable loca-tion which is otherwise suitable for their safety function. (Ref. 2, Enclosure at 11.)

While UCS asserts that ". . . several pipes carrying steam or high temperature water are located in the Intermediate Building . . .", UCS Petition at 6, the qualification program has utilized two specific main steam line breaks (24 inch and 12 inch), which produce the most severe environment for elec-trical equipment. Other breaks in the feedwater lines produce a much less severe environment and are not the basis for quali-fication.

The implications for the EFW system of a high energy line break in the Intermediate Building were recognized in the orig-inal licensing of TMI-1. As a result of an analysis of the consequences of all the postulated breaks in the Intermediate Building, utilizing criteria and guidelines provided by the Staff, corrective actions were identified. These included shielding of the EFW suction line and installation of addition-al piping restraints to prevent pipe whip damage and the fail-ure of a line connected to one steam generator from causing the failure of a line connected to the other steam generator. In addition, a significantly augmented inservice inspection of critical welds was instituted for the postulated break loca-tions. The Staff's conclusion was stated as follows:

The staff has evaluated the assessment per-formed by the applicant and has concluded that the applicant has analyzed the facilities in a manner consistent with the criteria and guidelines pro-vided by the staff. The staff agrees with the ap-plicant's selection of pipe failure locations and concludes that all required accident situations have been addressed appropriately by the appli-cant. Furthermore, the staff has evaluated the locations where increased inservice inspection is proposed in lieu of plant modification'and we find this justified and acceptable. (Ref.-1 at 10-7.)

n

i The augmented inservice inspection program for the Main Steam system is incorporated in the TMI-l operating license (No. '

i DPR-50, Technical Specification 4.15).

The harsh environment in the Intermediate Building follow-ing a main steam line break is being addressed in the review for TMI-1 under IE Bulletin 79-OlB and section 50.49. UCS ar-gues that the current status is not known of EFW system compo-nents for which the Technical Evaluation Report (TER) concluded that environmental qualification had not been established, and that "it is known that many vital components in the TMI-1 EFW remain incapable of functioning properly during a steam line break." UCS Petition at 7, 8.

As UCS and the Staff are aware, the deficiencies identi-fied in the Franklin Research Center TER on TMI-1, dated November 5, 1982, were predominantly based on the uncertainty by Franklin Research Center as to whether Licensee had adequate documentation to demonstrate the qualification of the identi-fied equipment (although Franklin had not requested the docu-mentation). The purpose of the October 5, 1983 meeting with the Staff was not to achieve final resolution of the TER deficiencies, as UCS implies, but to discuss Franklin's con-cerns. (UCS also inaccurately represents the December 16, 1983 meeting. Licensee discussed 120 equipment deficiencies, not 120 types of equipment having deficiencies. The 120 deficiencies address the entire plant and not just the EFW sys-tem -- the focus of the UCS Petition.) There is no equipment at TMI-l classified by the NRC in the category II.b, " EQUIPMENT NOT QUALIFIED." (Ref. 3, TER at 4-3.) As discussed below, some equipment is classified category II.a, " EQUIPMENT QUALIFI-CATION NOT ESTABLISHED."

While UCS may not be aware of the current status of the specific components identified in its petition, Licensee docu-mented the resolution of outstanding qualification items in

• 1etters to the Staff of February 10 and 22, 1984 (Refs. 4, 24.)

• and by the Revised Technical Response. The environmental qual-

• ification of the TMI-l EFW system under 10 C.F.R. 550.49 will

• be completed by June, 1984, including replacement of the Bailey

• E/P Converters for the EFW control valves with qualified I/P

• Converters. (Licensee has continued to work on improving the

• schedule for this modification, which had been set for the

• Cycle 6 refueling outage, and has now determined that it will

• be completed by June, 1984.) Thus, the environmental qualifi- i

• cation of the TMI-1 EFW ystem poses no undue risk to the public I

• health and safety and does not provide an appropriate basis for I

• the UCS Petition.

1

I III. Seismic Qualification The seismicity analysis for the licensing of TMI-1 indi-cated that the Pennsylvania area is relatively inactive seismically, based upon 200 years of historical data and 40 years of instrumental data. The TMI site is characterized by infrequent earthquakes of low intensity. This low intensity corresponds to a ground acceleration of 0.04g. (Ref. 5, sec-tion 2.8.) The Seismic I portion of TMI-l was designed to withstand a ground acceleration of 0.12g acting horizontally for the Safe Shutdown Earthquake (SSE) condition (Ref. 5, sec-tion 5.1.2), which exceeds the 0.lg specified ground accelera-tion of Appendix A to 10 C.F.R. Part 100. Consequently, the portions of the TMI-1 EFW system that are Seismic Category I are designed to more severe criteria than NRC regulations re-quire. Mechanical portions of the EFW system that are not now Seismic Category I are designed to the requirements of ANSI B31.1, " Power Piping." Fossil power plants and conventional portions of nuclear power plants designed to this standard have exhibited significant seismic resistance. (Refs. 6, 7; Ref. 8 at 2.)

It is clear that while Staff guidance for seismic qualifi-cation of PWR auxiliary feedwater systems has been evolving over a long period of time, the evaluation to determine how to backfit seismic requirements to earlier plants has not resulted in the imposition of specific seismic requirements. (Ref. 9.)

In its information request of February 10, 1981 (Ref. 8), the Staff stated:

Although we are not at this time requesting that the AEW System be modified to be in con-formance with the facility design seismic re-quirements, we have stated that our plan is to increase the seismic resistance, where necessary, to ultimately provide reasonable assurance that the system will function after the occurrence of earthquakes up to and including the SSE.

Licensee has made numerous submittals of information to the Staff, in response to Generic Letter 81-14, on the seismic qualification of the TMI-1 EFW system. The Staff's contractor, Lawrence Livermore National Laboratory (LLNL), has reviewed these responses and issued Technical Evaluation Reports dated October 29, 1982 and July 7, 1983. While the first TER identi-fied deficiencies in Licensee's responses, LLNL concluded in its second TER that, with the actions taken and planned by Li-censee (i.e., the long-term EFW modifications detailed in Ref-erence 2), the TMI-1 EFW system will be fully qualified to Seismic Category I at the next refueling outage (prior to start o

I l

up for Cycle 6 operation). Based upon this TER and its own evaluation of Cycle 5 operation, the Staff has concluded that there is reasonable assurance that the TMI-l EFW system will be able to withstand a SSE and perform its safety function. (Ref.

10.) .

UCS challenges this conclusion, apparently, in its asser-tions that the TMI-1 EFW system is not seismically qualified and that operation of TMI-1 therefore would pose an undue risk to the health and safety of the public. As the assessment below will demonstrate, the UCS Petition is without technical merit and does not undermine the validity of the Staff's previ-

)

ous safety evaluation.

A major fault in the UCS Petition is the extensive refer-ence, in the present tense, to findings in the first TER issued by LLNL, while virtually ignoring the second TER. UCS Petition at 9-15 (especially the list of "many vital components in the TMI-1 EFW system which are not environmentally qualified," UCS Petition at 10-11).

In its final TER, LLNL concluded that the TMI-1 EFW system piping, valves, structures and power supplies possess a SSE level of seismic capability, and that the initiation / control system will possess such capability after the Cycle 6 refueling outage.

The available information, which provides reasonable as-surance that the EFW system will perform its safety function after a SSE, and that has been ignored by the UCS Petition (at 10-11), includes:

a. Recirculation lines of the EFW pumps. The TMI-l Emergency Procedure for Earthquakes (1202-30) calls for closing of the Condensate Storage Tank B isolation valve (CO-V-176) and the EFW pump recirculation isolation valves (EF-V20A/B and EF-V22) if the EFW pump recirculation lines are ruptured.

(Ref. 11, Item 1.)

b. Portions of the EFW suction piping to the condenser hotwell, for which there are no double isolation valves between the seismic Class I piping and the non-seismic Class I piping.

Although TMI-1 does not have a second isolation valve between SI/SIII piping to the condenser hot well for each line, the condensate storage system is single failure proof. There are j two condensate storage tanks (CST) and Technical Specifications water inventory in either tank is sufficient for safe shutdown.

The common cross connect between the two condensate pipes (containing CO-V14A/B) has two isolation valves-(CO-V111A/B).

and closure of either valve (CO-V111A/B) will ensure integrity of one CST inventory if one of.the CO-V14A/B cannot be closed.

i l l J

'1 l

A

4 All of the valves involved (CO-V14A/B & CO-Villa /B) are Seismic I and by the end of Cycle 6 refueling outage their routing (Co-V14A/B and CO-Villa /B) and power supplies (CO-Villa /B) will also be Seismic I. In the interim, manual operator action will ensure proper operation following a seismic event.

The TMI-1 Emergency Procedure for Earthquake (1202-30) and -

relevant Alarm Response Procedures have been revised to in-struct the operator to isolate the damaged Condensate Storage Tank from the EFW system by closing valves CO-V14A/B and CO-Villa /B when tank level reaches the Tech Spec limit follow-ing EFW actuation, and following any recognizable seismic event (a seismic instrumentation alarm is available in the control room). (Ref. 12, TER Item 2.)

c. EFW pumps' minimum flow valves (recirculation valves) and their controlling flow switches and associated circuitry.

The EFW pumps' minimum flow valves (EF-V8A/B/C) are seismically qualified. (Ref. 25.) The fact that their controlling flow switches and circuitry are not seismically qualified has been resolved by locking open EF-V8A/B/C. This will prevent the possibility of dead heading the EFW pumps, and sufficient flow will still be available to the steam generators. (Refs. 18, 19.)

d. Electro-pneumatic converters for the EFW flow control

• valves, EF-V-30A and EF-V-30B.

The E/F Converters will be re-placed by June, 1984 with seismically qualified I/P Converters.

A seismic event will not result in a failure of the converters for the EFW flow control valves and thus sufficient flow will be established for the EFW system to perform its safety func-

• tion.

e. Condensate storage tank low level alarms. The ac-tions described above in "a, b and c" will ensure sufficient inventory in the Condensate Storage Tanks and a sufficient flow path to the steam generators for the EFW system to perform its safety function. (Ref. 11, Item 1.) Licensee has reviewed the failure modes in a seismic event for the condensate tank level instrumentation, (Ref. 11, Item 3.), and concluded that only in the event of a transmitter sensing line crimp (due to the transmitter falling) would the transmitter continue to read a static level. However the operator would note that no drawdown l is indicated and investigate the problem. It is incredible to l assume that both transmitters would fail in this manner.

Therefore, at least one transmitter is expected to be avail-able.

In the Restart proceeding, the Licensing Board recognized l and explicitly endorsed for Cycle 5 operation the non-safety-grade CST low-low level alarms as adequate pending the j I

installat ion of safety-grade alarms during the Cycle 6 refueling outage. LBP-81-59, 14 N.R.C. 1211, 1363-64, 1373 (11 1033, 1037, 1059). These low-low alarms use the same transmitter as the low level alarms.

f. Circuitry for main steam dump isolation valves MS-V2A, MS-V2B, MS-V8A and MS-V8B. Since the EFW system safety function can be achieved with the motor driven EFW pumps with-out relying on the turbine driven pump, the circuitry for these valves is not essential and need not be seismically qualified.

(Ref. 10, TER at 5; Ref. 12, Item 7.)

g. Circuitry for condensate storage tank isolation valves CO-V10A, CO-V10B, CO-V14A and CO-V14B. The only non-seismic parts of the circuitry for valves Co-V10A/B are the cable routing through the turbine building and the electric power supplies. CO-V10A/B are normally open and are not re-quired to change position for the system to become operational.

Valves CO-V10A/B are locked open now and there is no need to seismically qualify the circuitry for these valves. The only non-seismic part of the controls for valves CO-V14A/B is the cable routing through the turbine building. CO-V14A/B are nor-mally open and are required to change position for the system to become operational if a pipe break occurs in the hotwell makeup piping. (Ref. 19.) Manual closing of CO-V14A/B is pro-vided as discussed above in "b".

h. Circuitry for condensate storage tank cross connect valves CO-Villa and CO-V111B. The non-seismic parts of the circuitry for valves CO-V111B are the cable routing through the turbine building and the electric power supplies. CO-V111A/B are not required to change position for the system to become operational. (Ref. 19.) (See "b" above.)

1. Control systems for the atmospheric relief valves MS-V4A and MS-V4B. These valves are within the seismic bound-ary and will maintain their structural integrity during a seismic event. However, the control of these valves is not es-sential for safe hot shutdown and, therefore, the control sys-tem need not be seismically qualified. These valves will re-main closed on loss of instrument air or loss of electrical signal. The MSV-4A/B can be manually operated.

j. Vent stacks for both the main steam relief and atmospheric dump valves. UCS argues that "it is very likely that the operator will not be able to enter the Intermediate Building to isolate the leak following an earthquake because of steam released to the building by failure of equipment which is not seismically qualified" -- the vent stacks for MS-V-22A/B and MS-V-4A/B valves. UCS Petition at 13.

I l

1 l

l t A

• l l

l The pressure control valve (MS-V6) upstream of valves MS-V22A/B was modified to limit its travel at 65% of stroke to protect the EFW pump turbine from overpressurization due to the failure of any steam supply valve. This reduces the potential for opening of valves MS-V22A/B. In addition, these valves will not lift simply because a vent stack fails or the EFW tur-bine driven pump is started.

Licensee previously had evaluated the design of the vent stacks for these valves and found that these vent stacks were classified non-seismic and were designed for dead weight and discharge loads only. However, the supporting scheme for the MS-V22's stacks was judged by inspection to be seismically ac-ceptable. (Ref. 14, Question 1 of Enclosure 1; Ref. 15.)

! Also, as noted in item "i" above, operation of MSV-4A/B is not required for safe hot shutdown and the failure mode of these valves is closed. Consequently, there is a low probability of release of steam to the Intermediate Building from these vent stacks, and there is reasonable assurance, during Cycle 5 op-eration, that the operator will be able to function in the In-termediate Building.

k. Main steam isolation valve circuitry. Circuitry for these valves (MSV-1A, B, C, D) is not essential for plant shut-down (since the EFW turbine driven pump is not needed) and need not be seismically qualified. (Ref. 10, TER tat 5; Ref. 12, i Item 9; Ref. 11, Item 9.)

Following the dated list which is evaluated above, the UCS Petition proceeds to criticize use of a " static analysis" to establish the seismic qualification of valves. UCS Petition at

11. The very Standard Review Plan passage quoted by UCS belies its claim that static analysis has been rejected by the NRC:

" Analysis without testing is acceptable if structural integrity

' alone can assure the intended function." UCS Petition at 12.

Further, the seismic analyses for the 47 EFW valves utilized as inputs accelerations which were determined from a dynamic anal-ysis of the EFW piping system -- using the response spectrum approach specified in the Standard Review Plan. The valves and their characteristics (i.e., center of gravity, weights and ge-ometry) were realistically included in the dynamic model of the piping system. The piping was analyzed considering the Op-erating Basis Earthquake, and the acceleration results were then doubled to account for the SSE pursuant to the TMI-1 FSAR.

This approach is conservative since the increase in damping of the piping system during the SSE was not considered.

The accelerations used to analyze the valves were gener-ated using a fully qualified, realistic, " state of the art" dy-namic analysis of the EFW piping system. The dynamic model has been checked during the TMI-1 review in response to IE Bulle- l tins 79-02 and 79-14, which showed that-the pipe routing l 1

=

- .. . - . . - . - . - .- N

support locations and pipe support construction are consistent with the analysis.

The analyses applied the dynamic acceleration from the piping analysis to the valve internals, pressure boundaries and actuators in a static manner, along with other consequential loads. This approach is justified because the valve internals are sufficiently stiff to preclude dynamic amplification within the valve itself.

Here, stress analysis of the valves, considering accelera-tions derived from a dynamic analysis of the EFW piping system, reveals that the highest stress in the valves -- considering

, c'nsequent loads due to the SSE, internal pressure and dead-weight -- ranges from 3 to 91 percent of the ASME Code allow-able stress values. (These ASME allowable stresses are based on a safety factor of at least four, considering the ultimate strength of the materials.) This means that both the structur-al integrity and operability of the valves are assured because the materials experience stresses and strains within their elastic limits. Consequently, deformations are small and tem-porary, such that the moving parts inside the valves and actuators are not affected. For all of these reasons, the valve analyses are valid.

As shown above, the TMI-1 EFW system has the capability to perform its safety function following a seismic event, coin-cident with loss of offsite power with a single failure of any active component. Even if the inventory from either one or both Condensate Storage Tanks is depleted due to the single failure of isolation valve CO-V14A or B, a secondary backup supply of river water is available from the reactor building emergency cooling pumps -- an entirely seismic Class I supply, although establishment of this supply may require operator ac-tion in the Intermediate Building. (Ref. 14, Question 1 of En-closure 1, Enclosure 2 at 5.)

UCS states that GPU apparently performed no evaluation of the potential effects of flooding the Intermediate Building from failure of the EFW system, and concludes that this is a "significant omission." UCS Petition at 14. It might be if it were true, but it is not. Licensee has evaluated the conden-sate piping from valves CO-V14A/B to the turbine building wall to determine if this piping will stay intact during an earth-quake. Seismic stress analysis of the condensate piping has included the restraining capability of the supports in the non-seismic piping from the valves CO-V14A/B to the Turbine Building wall and into a portion of the piping that extends into the Turbine Building. These supports, which have a com-l bined restraining capability in three directions, will result l

in low seismic stresses in the non-seismic part of the system.

( If a pipe rupture is postulated beyond these supports, the

.I

break would be isolated and will not cause flooding in the In-termediate Building. Furthermore, there are no components vital to the EFW system which can be adversely affected by spray from a broken EFW pump recirculation line. (Ref. 11, Item 1.) Finally, the procedur,al action (discussed.above) to isolate the recirculation line will limit the leakage rate through this small line and avoid a flooding problem.

With respect to a main feedwater line break, the time re-quired to jeopardize EFW equipment is 5.5 minutes, not 86 sec-onds -- UCS Petition at 15, n. 40. (Ref. 16.) In addition, evaluation of the stress analysis for the main feedwater lines from containment penetration to the turbine building indicates that the maximum stress levels from combined operating and seismic conditions are at most 51 percent of the limits desig-nated as the potential pipe rupture stress level. (Ref. 5, Section 3.1 of Appendix 14A.) The results of these stress analyses show that the non-seismic portion of the main feedwater lines inside the Intermediate Building has seismic resistance. Consequently, there is a low probability that a main feedwater line break would cause flooding in the Interme-diate Building following a seismic event.

Finally, Licensee notes : hat UCS repeatedly cites to the plans for further hardware modifications to the EFW system (Ref. 2) as support for the proposition that the system is not seismically qualified, and asserts that GPU has concluded that at restart the TMI-1 EFW system cannot withstand a Safe Shut-down Earthquake. UCS Petition at 16. In contrast, it is,Li-censee's position that the TMI-l EFW system at restart, consid-ering accomplished modifications and with the implementation of the plan of procedural actions described above, will be able to perform its system function, in the unlikely event it should be called upon to do so following a design basis seismic event during Cycle 5 operation.

IV. Single Component Failure UCS states that "[t]he TMI-l EFW system does not meet'the single failure criterion because there is~only a single flow control valve in the pipe used to deliver EFW to each steam generator." UCS Petition at 19, 20. UCS does not address, however, the design modifications already accomplished which improve the reliability of the system.

The Main Steam Line Rupture Detection System (MSLRDS) sig-nals to the EFW control valves, EF-V30A/B, have been deleted to prevent unnecessary isolation of emergency feedwater under sin-gle failure conditions. In addition, a cavitating venturi in-stalled for each EFW line will-limit flow to a ruptured steam generator to prevent containment overpressurization (or steam generator overfill condition), and will also ensure sufficient EFW flow to the intact steam generator. (Ref. 17.)

• At restart, the arrangement of the EF-V30A/B controls will

• result in the valves failing open on either loss of instrument

• air or loss of control signal. Additionally, the EFW control valves are equipped with a handwheel which permits manual oper-ator action to establish ficw to the intact steam generator.

When there is an initiation of the EFW system or failure of an EFW control valve, an auxiliary operator will be stationed at the control valves. (See TMI-1 Abnormal Transient Procedure 1210-10.) The auxiliary operator will establish communications with the control room and will control the valves if EFW flow cannot be established from the control room.

Isolation of EFW flow, if required, to a ruptured steam generator can be achieved either by closing the affected EFW control valve or by closing the discharge header sectionalizing valves (EF-V2A/B), and then tripping the respective EFW pump.

UCS next states that "[a]nother way in which the EFW sys-tem does not meet the single failure criterion is that the EFW flow control valves are presently controlled by the Integrated Control System (ICS) which is not safety grade." UCS Petition at 20. The relationship between the EFW system and the ICS was considered extensively in the TMI-1 Restart proceeding. Pursu-ant to Short-term action 1(b) of the Commission's August 9, 1979 Order and Notice of Hearing in that proceeding, Licensee has implemented automatic initiation of the EFW pumps indepen-dent of the ICS and, further, has provided separate manual EFW flow control capability in the control room, which will allow the operators to manually control EFW flow to the steam genera-tors in the event of an ICS malfunction. The Licensing Board examined this issue and required no further modifications, finding that the actions taken provided a significant improve-ment in safety. LBP-81-59, 14 N.R.C. 1211, 1285-86 (1 802),

1362 (1 1031) (1981). The Appeal Board also evaluated the mat-ter and considered ". . . the concerns regarding dependence on the ICS for control of emergency feedwater to be resolved."

ALAB-729, 17 N.R.C. 814, 833-34 (1983).

In addition, Licensee notes that the ICS has a reliable, uninterruptible, on-site power supply. It is normally fed from an inverter which is powered from the "A" diesel backed 480 Volt AC bus. When the 480 Volt bus is unavailable, the in-verter takes its power directly from one of the DC station bat-teries. In the unlikely event of an independent inverter fail-ure, the ICS power supply will be switched to a regulating transformer which is fed directly from the same 480 Volt AC

{

bus. The independent manual control stations described in the previous paragraph are powered from a different inverter which l

is backed up by a separate set of DC station batteries. In the event of an independent failure of this inverter, the power supply for the manual control stations automatically switches to an alternative source backed by the "B" diesel generator.

In summary, means are available during Cycle 5 operation to prevent the EFW system from being disabled by a single com-ponent failure.

V. Emergency Feedwater Flow Instrumentation UCS attacks the adequacy of the new EFW flow indicators, alleging that the replacement of the unqualified sonic flow de-vices by differential pressure (D/P) transmitters " amounts to a request for exemption from the short-term lessons learned re-quirement for safety grade EFW flow instruments." UCS Petition at 24. (UCS's complaints regarding the EFW flow indicators are currently pending before the Commission in the Restart proceed-ing by virtue of UCS filings dated December 9, 1983 and January 6, 1984.) UCS here is patently wrong; as detailed in our sub-mittal to the Staff of August 25, 1983, the EFW flow instrumen-tation meets all applicable environmental, seismic and other safety-grade criteria. (Ref. 20, Attachment at 1, 2).

UCS's complaints regarding the qualification of the EFW flow indicators rest upon its claim that this instrumentation does not " meet the 1"10% accuracy requirement in effect during the restart hearing. UCS Petition at 24. As Licensee re-ported, at low EFW flow conditions (i.e., below approximately 120 gpm), cavitation of the EFW flow control valves (EEV-30's) due to low flow against negligible backpressure resulted in in-dications rate.

of EFW flow oscillations outside 1 10% of'the flow (Ref. 21; Ref. 22, Attachment at 1). However, recently reported test data, requested by the NRC (Ref. 23), confirm that at flows of 120 gpm and above, the flow oscillations recorded are within i 10% (e.g., at 200 gpm flow rate the os-cillations were 1 7.5% (15 gpm); at 600 gpm, the oscillations were 1 4.2% (25 gpm).) (Ref. 22, Attachment at 1.) (The os-cillations reported were measured on recorder traces. The EFW flow meter face contains 25 gpm graduations and thus these small oscillations combined with meter damping are not readable on the meter itself. (Ref. 22, Attachment at 1.)) Further, as discussed in Licensee's most recent submittal, operators are directed to refer to the EFW flow indicators only in limited circumstances (i.e., upon EFW actuation with steam generator (SG) level below the SG level setpoint) and, additionally, are

. instructed not to rely on EFW flow indication for flow control at rates below 225 gpm. (Ref. 22, Attachment at 2.) Thus, it is clear that the EFW flow indicators are sufficiently accurate to perform their intended function.

6 O

With respect to UCS's reliance on the 1 10% accuracy re-quirement, Licensee would merely note that (while this criteri-on was part of an interim clarification of lessons learned re-quirements dated October 30, 1979) Item II.E.1.2 of NUREG-0737, which sets forth the latest position and clarification for EFW flow indication, contains no such set accuracy requirement.

(Moreover, the Licensing Board decision itself makes no refer-ence to this 1 10% accuracy requirement. LBP-81-59, 14 N.R.C.

1211, 1362 (1 1029) (1981).) Rather, as recognized by UCS, NUREG-0737 merely referenced IEEE Standard 279-1971 which states, in pertinent part, that the system design basis shall document the " minimum performance requirements including . . .

system accuracies." See "UCS Rebuttal to Licensee's Reply Re-garding EFW Flow Instrumentation," (January 6, 1984) at 5, quoting IEEE 279-1971, 5 3(9). Licensee contends that its doc-umentation of EFW flow indication accuracy meets this require-ment and, moreover, that the earlier 1 10% accuracy criterion is met at EFW flows of 120 gpm and above.

VI. Main Steam Line Rupture Detection System UCS asserts that the Main Steam Line Rupture Detection System (MSLRDS) ". . . is not safety grade and requires modifi-cations so that a single failure will not prevent isolation of main feedwater to the steam generator affected by a main steam line break." UCS Petition at 29. As UCS notes, the potential for inadvertent isolation of feedwater was considered in the TMI-1 Restart proceeding as a part of the emergency feedwater reliability issues. LBP-81-59, 14 N.R.C. 1211, 1373-74 (11 1060-64) (1981). The Appeal Board found that the opera-tors' capability to bypass the MSLRDS and manually open the EFW flow control valves if the MSLRDS isolates feedwater inadver-tently is an adequate solution for restart. ALAB-729, 17 N.R.C. 814, 834, 887-88 (1983). In an Order (January 27, 1984) issued in the TMI-1 Restart proceeding after the UCS Petition was filed, the Commission called for comments on the adequacy of Licensee's proposed solution to the MSLRDS " problem."

In its submission of August 2, 1982 to the Staff, Licensee described the design changes to the MSLRDS to prevent unneces-sary isolation of emergency feedwater under single failure con-ditions. (Ref. 17.) In addition to those changes, existing pressure switches inside containment for MSLRD (Static-O-Ring devices) will be replaced by June, 1984, with fully qualified pressure switches. (Ref. 4.) Therefore, in the event of a main steam line rupture in containment, the pressure switches will be capable of performing their intended function. All components of the MSLRDS located-inside containment will then be environmentally qualified. The following describes the MSLRD system configuration:

1. Each steam generator (S.G.) has two outgoing steam lines, each line has two pressure switches for MSLRD.

2. Each S.G. has a parallel combination of startup and main FW control valves, and each control valve has a motor operated block valve upstream.

3. Upon MSLRD, the FW is isolated from the af-fected S.G. by closing its control valves and the block valves. Valve isolation logic is as follows:

A. Startup and Main Control Valves (FW-V16A/B & FW-V17A/B):

(1) For isolation purposes, each valve is provided with two paths in the pneumatic control circuit; however, only one path is required to achieve isolation.

(2) Each isolation path in the pneumat-ic control circuit has two sole-noids. Each solenoid is energized by a separate pressure switch upon MSLRD. Both solenoids in either of the control paths must be energized for isolation.

(3) The solenoids in the same control path are powered from the same source but the two paths receive power from separate sources.

B. Block Valves:

(1) For Main FW Controls Valves (FW-VSA/B):

Two pressure switches associated with either of the pneumatic con-trol paths (discussed in paragraph 3.A.2) must detect MSLR to cause a closure signal for the block valves. In this case, the isolation signals from RED & GREEN sources are tied together. Also the power for both the block valves is from the same source.

J

F (2) For startup EW Control Valves (EW-V92A/B):

Separate power sources are avail-able to the. motor operators. A.

single failure will prevent block valve isolation, but the same fail-ure will not prevent control valve isolation.

4. On loss of instrument air, the control valves (FW-V16A/B and 17A/B) will fail closed which will result in FW isolation.

5. Electrical Separation. Outside containment the MSLRDS circuits are not all routed in safety-related trays and therefore separation is not maintained throughout.

In conclusion, the MSLRDS is considered to be adequate from a single failure standpoint -- that is, a single active failure (such as a pressure switch, solenoid, control relay, 125V DC power source) will not prevent isolation of feedwater and will not result in inadvertent isolation of feedwater. The MSLRDS is seismic Class I inside containment. Following a main steam line break in the reactor building the system will func-tion to isolate feedwater from the affected steam generator since qualified pressure switches (for MSLRD)_ to be installed by June, 1984 will be suitable for the accident environment.

While electrical separation between the redundant circuits is not maintained outside containment, since a few of them run in the same trays / conduits, electrical separation outside contain-ment is not required for a main steam line break inside con-tainment. The MSLRDS, therefore, is adequate for operation until the fully safety grade modification is installed.

VII. Conclusion There is reasonable assurance that the emergency feedwater system at TMI-1, as modified for restart and as augmented with

(

plant procedures, will perform its function if called upon to do so. ,___

7 w<\ \g, u Richard F. Wilson)

Vice President-Technical Functions GPU Nuclear Corporation Sworn to and subscribed before me this 76 /,' day of March, 1984.

7 i

.- 4 is

\/

Hi ~ b' l Y H f D Notary Public J ptm- u I.o':23 r;v .. , .

"~1 My commission expires t.' / C  :-A% :." D. C ,

l l

[

\ r

REFERENCES

1. Safety Evaluation by the 71 rectorate of Licensing, U.S.

Atomic Energy Commission,.in the Matter of Metropolitan Edison Company, Jersey Central Power & Light Company, Pennsylvania Electric Company, Three Mile Island Nuclear Station Unit 1, Dauphin County, Pennsylvania, Docket No.

50-289, July 11, 1973.

2. GPU Nuclear letter 5211-83-232 to NRC, Long Term EFW Mods (NUREG 0737 II.E.1.1), August 23, 1983.

3. Safety Evaluation Report by the Office of Nuclear Reactor Regulation for GPU Nuclear Corporation, TMI-1, Docket No.

50-289, Environmental Qualification of Safety-Related Electric Equipment, December 10, 1982.

4. GPU Nuclear letter 5211-84-2038 to NRC, Environmental Qualification of Electrical Equipment, February 10, 1984.

5. GPU Nuclear, Final Safety Analysis Report (Updated Ver-sion), Three Mile Island Nuclear Station Unit 1.

6. USNRC NUREG-0766, Reconnaissance Report: Effects of November 8, 1980 Earthquake on Humboldt Bay Power Plant and Eureka, California Area.

7. USNRC NUREG/CR-1665, Equipment Response at the El Centro Steam Plant During the October 15, 1979 Imperial Valley Earthquake (October 1980).

8. USNRC Generic Letter No. 81-14 to All Operating Pressur-ized Water Reactor Licensees, Seismic Qualification of Auxiliary Feedwater Systems (February 10, 1981).

9. USNRC letter to All Operating Pressurized Water Reactor Licensees, Seismic Qualification of Auxiliary Feedwater Systems (October 21, 1980).

I 10. USNRC Safety Evaluation Report, Three Mile Island Unit 1, Seismic Qualification of the Auxiliary Feedwater System, August 12, 1983.

11. Attachment 1 to GPU Nuclear letter 5211-83-040 to NRC, EFW Seismic Qualification, February 4, 1983.

12. GPU Nuclear letter 5211-82-301 to NRC, Emergency Feedwater System-Seismic, December 20, 1982.

13. GPU Nuclear letter 5211-83-133 to NRC, EFW Seismic Quali-fication, May 2, 1983.

l l

[ 4

14. GPU Nuclear letter 5211-82-150 to NRC, Emergency Feedwater System-Seismic, July 7, 1982.

15. GPU Nuclear letter 5211-82-238 to NRC, Seismic Qualifica-tion of Emergency Feedwate.r System (EFW), September 29, 1982.

16. GPU Nuclear TDR-250, Rev. 1 (January 16, 1984), Review of Intermediate Building Flooding Following a Feedwater Line Break in the Intermediate Building of TMI Unit 1.

17. GPU Nuclear letter 5211-82-153 to NRC, Main Steam Line Rupture Detection System Changes, August 2, 1982.

18. GPU Nuclear letter 5211-83-055 to NRC, EFW Seismic Quali-fication Supplement, March 22, 1983.

19. GPU Nuclear letter 5211-82-018 to NRC, EFW Seismic Qualification-Electrical, February 16, 1982.

20. GPU Nuclear letter 5211-83-231 to NRC, EFW Flow Devices --

D/P Transmitters, August 25, 1983.

21. GPU Nuclear letter 5211-83-346 to NRC, EFW Flow Devices (D/P) Testing, November 23, 1983.

22. GPU Nuclear letter 5211-84-2032 to NRC, EFW Flow Instru-mentation, February 22, 1984.

23. USNRC letter to GPU Nuclear, January 18, 1984.

24. GPU Nuclear letter 5211-84-2044 to NRC, Environmental Qualification of Electrical Equipment, Supp. 1, February 22, 1984.

25. GPU Nuclear letter 5211-82-216 to NRC, Seismic Qualifica-tion of Emergency Feedwater System, September 14, 1982.

l