Text

Util Answer to Intervenor Motion to Accept Addl Factual Bases in Support of Admitted Contention.* Concludes That Intervenor Motion Should Be Denied.W/Certificate of Svc & Svc List
	ML20071M155
Person / Time
Site:	Vogtle
Issue date:	07/21/1994
From:	Blake E; GEORGIA POWER CO., SHAW, PITTMAN, POTTS & TROWBRIDGE, TROUTMANSANDERS (FORMERLY TROUTMAN, SANDERS, LOCKERMA
To:	; Atomic Safety and Licensing Board Panel
References
	CON-#394-15470 93-671-01-OLA-3, 93-671-1-OLA-3, OLA-3, NUDOCS 9408040079
	Download: ML20071M155 (95)
	v • d • e

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IsVP 00CKETED USHRC July 21, 1994 94 JJL 25 P4 :04 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF SECRETARY Before the Atomic Safety and Licensing 1900K[71gG & SERVICE BRANCH l

)

In the Matter of

)

Docket Nos. 50-424-OLA-3

)

50-425-OLA-3 GEORGIA POWER COMPANY,

)

et al.

)

Re: License Amendment j

)

(Transfer to Southern (Vogtle Electric Generating

)

Auclear)

Plant, Units 1 and 2)

)

.l

)

ASLBP No. 93-671-01-OLA-3 l

GEORGIA POWER COMPANY'S ANSWER TO INTERVENOR'S MOTION TO ACCEPT ADDITIONAL FACTUAL BASES IN SUPPORT OF THE ADMITTED CONTENTION I.

Introduction i

Georgia Power Company ("GPC") hereby answers and opposes Intervenor's Motion to Accept Additional Factual Basis in Support l

of the Admitted Contention (July 6, 1994) (hereinafter "Interve-nor's Motion").

The additional basis (in effect a new conten-tion) being advanced by Intervenor is untimely and unsupported, and it should be rejected.

l At the outset, GPC observes that Intervonor has known of the additional basis for over four years.

Intervenor could have raised this issue in December 1992, when he filed his original contention, but he chose not to.

Now, as a very lengthy period of discovery is finally drawing to a close, Intervenor files this last-minute allegation which, if admitted for litigation, would necessitate a whole new round of discovery and result in a 9400040079 940721

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substantial delay.

Such a delay would be extremely prejudicial to GPC, and In'tervenor's t'actic of delay should not be tolerated.

Intervenor's additional basis also fails to satisfy the threshold for admissibility.

Intervenor inaccurately character-izes the Vogtle Technical Specifications, fails.to discuss or address NRC's interpretation of the Technical Specifications, and provides no meaningful support for any of his assertions.

The two-page transcript, which is attached to Intervenor's motion and appears to be the only information on which Intervenor is relying, does not establish any wrongdoing.

II.

The Expanded Basis Does Not Satisfy Pleading Standards The Commission's Rules of Practice, at 10 C.F.R. 5 2.714, set forth the requirements for the admission of contentions.

A contention must consist of a specific statement of law or fact to be raised or controverted.

It must be supported by a statement of the alleged facts or expert opinions on which Intervenor intends to rely in proving the contention at hearing, together with references to the specific sources and documents of which Intervenor is aware and on which he intends to rely to establish those facts or expert opinions.

The supporting information must be sufficient to establish the existence of a " genuine" dispute

[

on a " material" issue of law of fact.

10 C.F.R.

S 2. 714 (b),

(d).1/

1 The 1989 amendments to the NRC's Rules of Practice, which promulgated the current pleading standards, were intended to raise the threshold for the admission of contentions.

54 Fed.

Reg. 33,168 (1989).

See Arizona Public Service Co. (Palo Verde Nuclear Generating Station, Units 1, 2,

and 3), CLI-91-12, 34 N.R.C.

149, 155-56 (1991); Long Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1), LBP-91-35, 34 N.R.C.

163, 167 (1991).

These standards are to be enforced rigorously.

A Board 1

should not overlook any deficiencies in a contention or assume the existence of missing information.

Palo Verde, CLI-91-12, 34 N.R.C.

at 155-56; Long Island Lighting Co. (Shoreham Nuclear Power Station, Unit 1), LBP-91-39, 34 N.R.C.

273, 279 (1991).

As explained when these current pleading standards were pro-mulgated, a contention should not be admitted "where an interve-nor has no facts to support its position and where the intervenor contemplates using discovery or cross-examination as a fishing expedition which might produce relevant supporting facts."

54 Fed. Reg. at 33,171.

Admission of a contention may be refused if its appears unlikely that the Intervenor can prove a set of facts in support of its contention.

Id. at 33,168.

1/

These requirements were summarized in the Federal Register notice commencing this proceeding.

57 Fed. Reg. 47,127 (1992). -

The additional basis proposed by Intervenor does not satisfy these pleading standards.

Intervenor makes only conclusory alle-gations and provides no real support to demonstrate that a genu-ine dispute exists.

Intervenor's proposed additional basis can be broken down into several allegations, not one of which is properly supported.

These allegations are that (1) the opening of the containment hatch on the evening of March 20, 1990, without operable diesel generators, violated Technical Specification 3.9.8.2 and 3.8.1.2; (2) the opening of the containment hatch on the evening of March 20, 1990, without operable diesel generators, breached a commit-ment to the NRC Staff; (3) these alleged violations were commit-ted knowingly and intentionally by line management up through and including the Executive Vice President and placed the plant in a less safe condition in order minimize outage time; and (4) a sub-sequent waiver of a technical specification was intended to cover up the violations.

See Intervenor's Motion at 1-4.

None of these issues is shown by Intervenor to be genuine.

1 A.

Alleged Violation of Technical Specifications Intervenor does not identify any evidence supporting the allegation that opening the containment hatch violated technical specifications.

Intervenor identifies no facts, expert opinion, documents or other sources that would support this claim.

He 1

1 1 )

i l

'1 does not meaningfully discuss TS 3.8.1.2 and 3.9.8.2 -- the two technical specifications which he states were violated.

TS 3.8.1.2 established A.C. electrical power systems limit-ing conditions for operation applicable to the refueling and cold shutdown acdes.

It required one off-site power source and one diesel generator to be operable in these modes.

With less than these minimum requirements, certain operations (such as those involving core alterations) were prohibited, and certain correc-tive actions were required.

TS 3.8.1.2, however, imposed no action regarding the containment hatch.

See Enclosure 1 to the attached Exhibit A, at 3/4 8-10.

TS 3.9.8.2 established a limiting condition for operation applicable to the refueling mode when water level was less than 23 feet above the top of the reactor vessel flange.

It required two independent residual heat removal (RHR) trains to be operable and at least one RHR train to be in operation.

With less than the required RHR trains operable, actions were required as soon as possible to restore the operability of the RHR trains or establish reactor vessel water level at least 23 feet above the vessel flange.

With no RHR "in operation," certain additional actions were required, including closing all containment penetra-tions providing direct access from the containment atmosphere to l

the outside atmosphere within four hours.

See id. at 3/4 9-9.

j

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1 i

i Intervenor states that the inoperability of the two diesel generators on the evening of March 20 made both trains of the RHR inoperable.

Intervenor's Motion at 2.

This statement is inaccu-rate and unsupported.

The definition of operability in the tech-i nical specifications did not require emergency power to be available.2/

As discussed in an August 16, 1991 Memorandum from C.

Rossi to W. Russell, "RHR Operability Requirements During Shutdown" (Exhibit A hereto), an NRC letter dated June 11, 1980, l

instructing all PWRs to amend plant technical specifications i

i regarding decay heat removal, had included a model technical I

specification stating, "The normal or emergency power source may i

I be inoperable for each RHR loop."

Further, in 1981, the NRC had j

revised the standard technical specifications to eliminate the I

reference to emergency electrical power from the definition of operability.

See Exhibit A at 2.

This led to an NRC interpre-i tation that both trains of RHR,would be operable if they received power from their respective safety electrical buses.2/

Id.

2/

Definition 1.20 stated, "A system, subsystem, train, compo-nent or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, j

cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s)."

Exhibit A,,

at 1-4.

2/

This interpretation was not uniformly accepted by all offices in the NRC and led to a differing professional opinion with regard to whether Vogtle met the RHR operability Footnote continued on next page. -

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i Based on this interpretation, both trains of RHR were opera-

~

ble when the containment h'atch was open.

The reserve auxiliary transformer (RAT) supplying offsite power to one safety electri-cal bus was available at approximately 11:30 a.m. after the event, and a second RAT supplying offsite power (independent from the other source) was restored at approximately 6:40 p.m. on the same day.

Therefore, at approximately 6:40 p.m. on March 20, 1990, each train of RHR was receiving power from its respective, independent safety electrical buses and was thus operable.

More-over, one train of RHR was in operation.

Consequently, there was no violation of TS 3.9.8.2 when the containment hatch was later opened.

Because these matters were discussed when Intervenor was employed at Vogtle, Intervenor must have been aware of the NRC interpretation of TS 3.9.8.2 and the resolution of the issue.

Yet he makes no reference to August 16, 1991 memorandum and pro-vides no explanation why, in light of the determination reflected in this memorandum, Vogtle should be accused of violating techni-cal specifications.

In sum, Intervenor does not meaningfully Footnote continued from previous page.

requirements at the time of the March 20, 1990 event.

Exhibit A addressed the differing professional opinion and concluded that Vogtle met its Technical Specification requirements for operabil-ity at the time the March 20 event occurred.

At the time the event occurred, the plant had one operable offsite power source and one operable diesel generator.

4 t

address the NRC documents related to this matter and hence pro-vides no basis for his allegation.S/

B.

Alleged Breach of Commitments Intervenor also provides no meaningful support for his assertion that " commitments were made to the NRC in a meeting on or about the evening of March 20, not to open the containment hatch until the diesel and the RAT.

were operable."

See Intervenor's Motion at 2.

Intervenor does not allege that he has any first hand knowledge of such a meeting, or that any such com-mitment was memorialized.

The only purported basis for this allegation is rather con-fusing double hearsay in a two-page transcript of a conversation between Intervenor and George Frederick.

As transcribed by Intervenor, Mr. Frederick is alleged to have said, And basically, at the meeting I thought that the final dis-cussion that I got from George [Bockhold) and Skip (Kitch-ens], because they said it 4 times for clarification, I

remember it had to be said four times before everybody understood, that we wouldn't reopen the hatch until we had the diesel and the RAT.

i/

Intervenor has "an ironclad obligation to examine publicly available documentary material pertaining to the facility in question with sufficient care to enable it to uncover any infor-nation that could serve as the foundation for a specific conten-tion."

Duke Power Co. (Catawba Nuclear Station, Units 1 & 2),

ALAB-687, 16 N.R.C. 460, 468 (1982), vacated in part on other grounds, CLI-83-19, 17 N.R.C.

1041 (1983).

D Frederick, however, does not state that this was a " commitment" to the NRC.5/

I Further, even if there was a commitment, it is not clear what the precise commitment was.

It is unclear what was meant or understood by the statement that the hatch should not be reopened I

until GPC "had" the diesel and the RAT (or even whether Frederick was accurately paraphrasing earlier statements).

As discussed earlier, two RATS, providing independent offsite power supply, had been restored by approximately 6:40 p.m. on March 20.

In addition, prior to opening the containment hatch on the evening i

of March 20, Diesel Generator lA was successfully started three times.

The first start ran the diesel at nearly full load for

[

about three quarters of an hour.

At that point, the diesel was intentionally tripped and restarted two more times.

Thus, at the time the containment hatch was opened, in addition to the two independent sources of offsite power, the plant staff had also i

demonstrated that emergency diesel generator power would be available if needed.

Intervenor does not address any of this i

5/

In the time allowed to prepare this response, we have not been able to determine that any commitment was made.

The recol-lections of individuals more than four years after the event are too vague, and no document memorializing any such commitment has been identified.

This difficultly in addressing Intervenor's allegation demonstrates the prejudice that results when an Inter-venor hoards an allegation, revealing it only at the eleventh hour.

In light of this prejudice, any uncertainty should be resolved against Intervenor.

_g_

7 information, which is all readily ascertainable from documents produced in discovery.

In short, Intervenor provides insufficient information to establish that a genuine dispute exists on a material issue.

He has not demonstrated that any commitment was in fact made to the NRC, let alone established with any precision the terms of that commitment.

Even if some oral commitment relating to power sup-ply was made, it may well have been satisfied by GPC's actions to re-energize two RATS, thereby providing independent sources of power to each train of RHR, and to test the diesels, thereby showing that a source of emergency power would be available if needed.

Intervenor provides no evidence to the contrary.

Pre-sumably, if GPC had violated some oral commitment to the NRC, the NRC would have said something about it long ago.

C.

Alleged Willfulness Intervenor has not established that any violation of techni-cal specifications or commitments occurred, and therefore has equally failed to establish that there was any willful wrongdo-ing.

Intervenor states that "The above actions involved the deliberate and knowing violations of tech. specs. by SRO licensed personnel including Lackey, Beasely, Kitchens and line management i

up to and including R.P. Mcdonald" (Intervenor's Motion at 4),

but there is not one whit of information to even suggest that any of these individuals acted improperly.

With respect to the 1

-Io-

3 interpretation of TS 3.9.8.2, it is clear from the NRC's analysis that differing, reasonable interpretations applied, and GPC's position was in fact in keeping with earlier NRC guidance.

Thus, even if one were to interpret the technical specification differ-ently, one could not deny that GPC's position was supportable.

In this light, Intervenor's claim that named individuals commit-ted deliberate and knowing violations of technical specification is baseless.

Intervenor allegations concerning improper motivation are similarly unsupported concoctions.

Intervenor asserts, for exam-

ple, The motivation for taking this action (opening the contain-ment hatch] stems from the fact that containment integrity (required while the plant was at mid-loop with no OPERABLE RHR Systems) was blocking critical outage progress end slow-ing down SONOPCO's planned outage scheddle.

Without regard to prior commitments made to the NRC or the precarious con-dition of having no OPERABLE emergency AC power, Plant Vogtle was intentionally placed in a lese safe condition by removing the containment equipment access hatch.

Intervenor's Motion at 2-3.

Intervenor does not identify a sin-gle fact, document, or reference to support this claim.

In con-trast, opening the hatch was important to support expeditious work to tension the reactor vessel head, fill and vent the RCS system to increase inventory, and make the steam generators available for heat removal should they be required.

As indicated in the GPC's March 22, 1990 letter following up on a waiver of compliance (Exhibit B), this activity improved the plant's margin of safety.

Intervenor ignores this legitimate, documented

)

purpose and substitutes his own theory, manufactured out of whole cloth.

Intervenor's conjecture and innuendo is clearly insuffi-cient to support his attacks on the integrity of individuals.

D.

Alleged Cover Up l

Intervenor does not allege any evidence at all supporting his assertion that the waiver of technical specifications was intended to cover up violations of the technical specifications.

First, as previously discussed, opening the containment hatch did t

not violate any technical specification.

Second, the waiver i

later obtained by GPC did not relate to the hatch.

Instead, a waiver was obtained from TS 3.0.4, which would otherwise have prohibited Vogtle from making a change from Mode 6 to Mode 5

'(i.e., changing from a refueling mode to a cold shutdown mode) without verification of diesel generator operability.

See Exhibit B.

The request for waiver did nothing to change or con-ceal the opening of the hatch on March 20, and there is nothing but Intervenor's fanciful and unsupported conjecture to the con-trary.

III. Intervenor's Late-Filing Is Unjustified As discussed above, Intervenor's new allegations are unsup-ported and fail to satisfy pleading requirements.

They are also untimely and fail to satisfy standards for late filing.

I Under the Commission's Rules of Practice, untimely conten-tions are not entertained unless the Intervenor demonstrates that admission of the late-filed issue is justified by a balancing of five factors.

The five factors are:

(i)

Good cause, if any, for failure to file on time; i

(ii)

The availability of other means whereby (Interve-l nor's) interest will be protected; (iii)

The extent to which [Intervenor's) participation may L

reasonably be expected to assist in developing a sound record; (iv)

The extent to which [Intervenor's represented by existing parties; a)ndinterest will be i

(v)

The extent to which [Intervenor's] participation will broaden the issues or delay the proceeding.

i 10 C.F.R.

S 2. 714 (a) (1).

I These five factors are not weighed equally.

Of the five, good cause is the most important.

Detroit Edison Co. (Enrico l

Fermi Atomic Power Plant, Unit 2), ALAB-707, 16 N.R.C.

1760, 1765 l

(1982).

Good cause for an amendment must be established by show-ing that the new information appears in previously unavailable documents and that the request to amend, being otherwise proper, is expeditiously presented.

Northern States Power Co.

(Monticello Nuclear Generating Station, Unit 1),

LBP-75-45, 2 N.R.C. 263, 268 (1975).

The lack of good cause for Intervenor's late filing is made evident by the very transcript that Intervenor attaches to his motion.

That transcript, dated March 30, 1990, demonstrates that.. -.

3 f

Intervenor was aware over four years ago of the issue he now Peeks to raise.

In fact, the four year old conversation is the only source of purported support for Intervenor's allegatio.a.

Thus, Intervenor could have raised this issue in its December 1992 contentions and simply chose not to do so.

Intervenor attempts to explain away its lateness by arguing that the Commission's Rules of Practice are ambiguous as to whether an intervenor is required to submit all known factual bases at the time an intervenor seeks to admit a contention.

Intervenor's Motion at 5.

This argument is tantamount to an admission that Intervenor could have raised this issue and appar-ently chose not to, perhaps to gain some strategic advantage through non-disclosure.

In any event, the rule is not ambiguous.

10 C.F.R. S 2.714 (b) (2) requires an intervenor to identify the facts and expert opinion on which he intends to rely to prove his contentions, as well as the specific sources and documents of which he is aware and on which he intends to rely to establish those facts or expert opinion.

This regulation does not allow a petitioner to plead some of the facts on which he intends to rely, while omitting others of which he is perfectly aware.

To the contrary, as explained by the Commission when it promulgated the 1989 amendments, the current regulation requires disclosure of " facts or expert opinion, be it one fact or many, of which

[intervenor) is aware at that point in time which provide the basis for its contention."

54 Fed. Reg. at 33,170. _

3 Further, even if one accepts for argument's sake Interve-nor's suggestion that he was confused when he filed his December 1992 contentions, no such confusion could have existed after the Licensing Board's September 24, 1993 Memorandum and Order, LBP-93-21, 38 N.R.C.

143.

There, the Board ruled that Intervenor had voluntarily excluded from the scope of this proceeding those allegations of which Intervenor was aware and did not discuss in his petition.

Id. at 148.

Intervenor provides absolutely no explanation why it waited another ten months to reise his addi-tional allegations.

Given the scheduled closure of discovery, this delay is particularly egregious and prejudicial.

Intervenor also attempts to explain away his lateness by arguing that he did not possess enough facts to file the instant allegations until the " War Room" log became publicly available to show the date and time the containment hatch was opened.

This i

argument also lacks merit.

The transcript attached to Interve-nor's Motion shows that Intervenor was aware that the containment hatch was opened on either the evening of March 20 or the morning of March 21, 1991, and the exact time in this general time frame appears irrelevant to Intervenor's allegations.

Moreover, the particular log to which Intervenor alludes was made available to Intervenor on November 1, 1993, when it was produced by GPC in response to Intervenor's first document request.

See letter from J. Lamberski to M. Kohn (Nov.

1, 1993).

Having failed to show good cause for its late-filing, Inter-venor must make a compelling showing on the other four factors.

Fermi, supra, ALAB-707, 16 N.R.C.

at 1765.

Intervenor has not done so.

Most importantly, the fifth factor (the extent of potential delay) strongly militates aga. inst admitting Intervenor's addi-tional and untimely allegation.

Although Intervenor aseerts that

)

"[a]dmitting the new factual basis will not broaden or delay the proceeding" (Intervenor's Motion at 8), he provides no credible basis for the assertion.

Intervenor remarks that discovery could be completed witlin the allotted discovery time period if it could "immediately commence."

Id.

This remark is meaningless I

and misleading.

Presumably, Intervenor means that if discovery on this allegation had started on July 6 (the date of Interve-nor's Motion), before the scheduled depositions on diesel genera-tor issues, an extension of Intervenor's discovery might not have been necessary.

Intervenor offers no meaningful projection of the amount of discovery that would be necessary if the new alle-gation were admitted after the current depositions on diesel gen-erator issues are conducted.

Given Intervenor's insistence in this proceeding on deposing dozens of individuals on every issue 1

coupled with multiple written discovery requests, GPC believes that admission of an additional allegation might in fact delay completion of discovery by months, with an attendant or greater delay in summary disposition and hearing schedules...

l Intervenor's conclusory discussion of the other factors is equally unconvincing.

With respect to the second factor, Inter-l venor claims that no other party will have the ability or stand-ing to adjudicate this issue before a licensing board.

Interve-nor's Motion at 5-6.

The test is not whether the issue will be litigated before the Board, but whether there are other means i

whereby Intervenor's interest will be protected.

Intervenor has apparently referred this allegation to the Office of Investiga-tions, and OI and the NRC Staff are certainly able to address it.

With respect to the third factor, Intervenor argues that admitting the additional factual basis will develop a more com-plete record on the issue of character.

Intervenor's Motion at 7.

Intervenor, however, makes no showing that he has anything in particular to offer on this particular allegation.

As discussed above, Intervenor has mischaracterized the particular technical specification he claims was violated, and appears to have no first hand knowledge of the commitment he claims was breached.

He has offered no reliable evidence of any wrongdoing, and abso-lutely nothing to its claim of a " cover-up involving the entire chain of management up to and including the Executive Vice Presi-dent" (see id.).

In addressing the extent to which he can assist in develop-ing a sound record, Intervenor should set forth with as much par-l ticularity as possible the precise issues he plans to cover, I

identify the prospective witnesses, and summarize their proposed testimony.

Vague assertions regarding an intervenor's ability or resources are insufficient.

Fermi, supra, ALAB-707, 16 N.R.C.

at 1766; Mississippi Power & Light Co. (Grand Gulf Nuclear Station, Units 1 and 2), ALAB-704, 16 N.R.C.

1725, 1730 (1982).

Here, Intervenor has not shown even the slightest ability to support its reckless allegations.

With respect to the fourth factor, Intervenor states that NRC Staff has already indicated that it is not interested in pur-suing this issue.

Intervenor's Motion at 8.

The NRC Staff has not communicated such a position, and GPC is not certain how Intervenor may be privy to the Staff's internal deliberations.

Nevertheless, even if accepted, Intervenor's assertion merely indicates that the Staff has indeed considered the allegation and found it lacking.

If the Staff, which is the party with the most expertise in interpreting technical specifications and the great-est understanding of the commitments it expected to be honored, has concluded that this additional allegation is not worthy of any further consideration, it is very unlikely that the allega-tion is significant enough to warrant expansion and substantial delay of this proceeding.

. i 1

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7 III.

Conclusion In summary, Intervenor's late-filed allegations are unsup-ported and unjustified.

For all of the reasons stated above, Intervenor's motion should be denied.

This proceeding has already consumed ah inordinate amount of time and resources, and Intervenor's attempt to further expand and delay it is inappropriate.

Respectfully submitted, Ernest L.

Blake, Jr.

David R. Lewis SHAW PITTMAN POTTS & TROWBRIDGE 2300 N Street, N.W.

t Washington, D.C.

20037 (202) 663 8000 i

John Lamberski l

TROUTMAN SANDERS

}

600 Peachtree Street, NE Suite 5200 Atlanta, GA 30308-2216 (404) 885 3360 Counsel for Georgia Power Company Dated:

July 21, 1994 i 4

s UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board

)

In the Matter of

)

Docket Nos. 50-424-OLA-3

)

50-425-OLA-3 GEORGIA POWER COMPANY,

)

et al.

)

Re: License Amendment

)

(Transfer to Southern (Vogtle Electric Generating

)

Nuclear)

Plant, Units 1 and 2)

)

ASLBP No. 93-671-01-OLA-3 1

CERTIFICATE OF SERVICE I hereby certify that copies of " Georgia Power Company's Answer to Intervenor's Motion to Accept Additional Factual Basis in Support of the Admitted Contention," dated July 21, 1994, were served by deposit in the U.S. Mail, first class, postage prepaid, upon the persons listed on the attached service list, this 21st day of July, 1994.

i l

N i

'=~__

David Lewis i

Dated: July 21, 1994 i

l

3 00CKETED USHRC UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 94 JUL 25 P4.04 Before the Atomic Safety and Licensing Board 0FFICE CF SCCRETARY DOCKETING & SERVICE

)

BRANCH In the Matter of

)

Docket Nos. 50-424-OLA-3

)

50-425-OLA-3

' GEORGIA POWER COMPANY,

)

et al.

)

Re: License Amendment

)

(Transfer to Southern (Vogtle Electric Generating

)

Nuclear)

Plant, Units 1 and 2)

)

ASLBP No. 93-671-01-OLA-3 SERVICE LIST Administrative Judge Carolyn F.

Evans, Esq.

Peter B. Bloch, Chairman U.S. Nuclear Regulatory Commission Atomic Safety & Licensing Board 101 Marietta Street, N.W.,

Suite 2900 i

U.S. Nuclear Regulatory Commission Atlanta, Georgia 30323-0199 Washington, D.C.

20555 office of the Secretary Administrative Judge U.S. Nuclear Regulatory Commission James H. Carpenter Washington, D.C.

20555 Atomic Safety & Licensing Board ATTNr Docketing and Services Branch 933 Green Point Drive Oyster Point Mitzi A.

Young, Esq.

Sunset Beach, N.C.

28468 Charles Barth, Esq.

Office of General Counsel Administrative Judge One White Flint North Thomas D. Murphy Stop 15B18 Atomic Safety & Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear P.egulatory Commission Washington, D.C.

20555 Washington, D.C.

20555

Director, Michael D. Kohn, Esq.

Environmental Protection Division Kohn, Kohn & Colapinto, P.C.

Department of Natural Resources t

517 Florida Avenue, N.W.

205 Butler Street, S.E.

j washington, D.C.

20001 Suite 1252 Atlanta, Georgia 30334 Stewart D. Ebneter i

Regional Administrator Office of Commission Appellate

{

USNRC, Region II Adjudication 1

101 Marietta Street, NW U.S.

Nuclear Regulatory Commission Suite 2900 Washington, D.C.

20555 Atlanta, Georgia 30303 J

1

Adjudicatory File Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Bruce H. Morris (Appearing for George R. Frederick)

Finestone, Morris & Wildstein Suite 2540 Tower Place 3340 Peachtree Road, N.E.

Atlanta, Georgia 30326 t

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UNITED STATES

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j NUCLEAR REGULATORY COMMISSION

,j WASHINGTON,0. C. 20555 t

f August 16, 1991 MEMORANDUM FOR:

William T. Russell Associate Director for Inspection and Technical Assessment Office of Nuclear Reactor Regulation FROM:

Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation l

SUBJECT:

RHR OPERABILITY REQUIREMENTS DURING SHUTDOWN As you requested in your memorandum dated July 19, 1991, this memorandum provides DOEA's comments on the Differing Professional View (DPV) regarding TS requirements for RHR systems during shutdown as reflected in the conclusions of the Vogtle IIT.

DST will respond to your request separately.

The central issue of the DPV is the relationship between the operability requirements for the RHR system and diesel generators during operation with r

reduced RCS inventory.

Overall, there is a substantial inconsistency within the NRC on this issue because there are two interpretations of the TS requirements.

Vogtle has Standard Technical Specifications (STS) and the Vogtle TS (Enclosure 1) require that two trains of RHR be operable in operation with reduced RCS inventory.

Operation with reduced RCS inventory occurs when the plant is shutdown (Modes 5 and 6) with less than 23 feet of water above the reactor vessel flange.

The i

TS also explicitly require the operability of one offsite power source and one diesel generator in Modes 5 and 6.

In addition, the definition of OPERABILITY in the TS requires systems which i

support the RHR to be capable of providing the necessary support.

The definition of OPERABILITY specifically includes electrical power as a necessary support.

The interpretation of the words

" electrical power" in this definition is the source of the inconsistency.

The DPV is based on the interpretation that

" electrical power" means both normal and emergency electrical power.

This interpretation makes the operability of an RHR train contingent on the operability of (1) the safety electrical bus supplying power to the RHR train and (2) the diesel generator aligned to that bus.

Using this interpretation, the DPV concludes that Vogtle was not in compliance with the plant TS requirements for operability of the RHR when the event occurred on March 20, 1990.

e

/,0G 1 6 '.SSI

J In a letter from D.

G.

Eisenhut dated April 10, 1980 (Enclosure 2), the NRC instructed all power reactor licensees to incorporate the definition of OPERABILITY into plant TS.

That definition was incorporated into the STS and included the words " normal and emergency electrical power."

LCO 3.0.5 was implemented along with the definition.

LCO 3.0.5 elearly stated that a system did not have to be declared inocerable solelv because its emerarngy cover source (usually, the associated diesel cenerator) was inocerable orovided certain conditions were met.

The normal power source had to remain operable and the redundant system had to remain operable.

LCO 3.0.5 further stated that it was not applicable in Cold Shutdown or Refueling.

LCO 3.0.5 was meaningless in those Modes because only one offsite A.C.

source and one onsite A.C.

source were required in those Modes.

Later, in a letter from D.

G.

Eisenhut dated June 11, 1980 (Enclosure 3), the NRC instructed all PWRs to amend plant TS regarding decay heat removal capability.

This generic letter included model STS pages with this footnote on the RHR LCO for operation with reduced RCS inventory "The normal or emergency power source may be inoperable for a ach RHR loop."

Again, the NRC guidance said that a system is not inoperable solely because its associated diesel generator is inoperable.

In 1981, the STS were revised (1) to eliminate the words " normal and emergency" from the definition of OPERABILITY, (2) to eliminate LCO 3.0.5, and (3) to put the conditions from LCO 3.0.5 into the LCO's for A.C.

Sources - Operating.

This is the version of the definition in the Vogtle TS.

This revision placed all the required Actions for inoperability of an A.C. source in the LCO for A.C.

Sources - Operating.

This also clarified the requirements for A.C. power in the Cold Shutdown and Refueling i

Modes.

A plant could meet its TS requirements for operation with reduced RCS inventory with one operable offsite power source and one operable diesel generator as long as power was supplied to both safety buses.

Both trains of RHR would be operable, j

receiving power from their respective safety electrical buses.

Based on this interpretation of the definition of OPERABILITY, j

the IIT concluded that Vogtle was in compliance with the TS requirements for operability of the RHR when the event occurred.

j This is the interpretation of the definition of OPERABILITY which is used by the Reactor Systems Branch (RSXB) and the Technical Specifications Branch (OTSB).

This is the interpretation that has been used by Vogtle since licensing.

The NRC has never told Vogtle that this interpretation is incorrect.

Many plants use this interpretation and have never been told by NRC that there is anything wrong with this interpretation.

Other plants have the original version of the definition of OPERABILITY, use the DPV interpretation, and have been told by NRC that the DPV interpretation is correct.

7

'A 9

AUG 161991

' In a memorandum from F. Rosa to E. Butcher dated February 2, 1984 (Enclosure 4), the Electrical Systems Branch (SELB) recommended that the current definition of OPERABILITY in the STS be revised to read "offsite and emergency electrical power."

SELB made this recommendation because the current definition does "not clearly t

convey that both offsite and onsite electric power are necessary i

for operability."

OTSB informed SELB that this interpretation is i

inconsistent with the position of OTSB and SRXB.

l Based on the above analysis, DOEA agrees with the IIT conclusion that Vogtle met its TS requirements for RHR operability at the time of the event.

However, there is substantial inconsistency in the TS requirements for diesel generator operability to support the RER in operation with reduced RCS inventory.

DOEA l

agrees that these requirements need to be clarified to achieve consistent application at all plants.

The guidance on operability which has been proposed for inclusion in Section 9900 of the Inspection Manual is a start.

The implementation of the i

new STS and the recommendations of the shutdown risk study will complete the clarification.

In OTSB's view, the preliminary l

information from the shutdown study strongly suggests that the electrical source configuration allowed by the STS in operation j

with reduced RCS inventory should be strengthened. Therefore, OTSB is eagerly awaiting the opportunity to implement the recommendations of the study when they are finalized.

har es

orsi, irect r Division of Operational Events Assessment Office of, Nuclear Reactor Regulation

Enclosures:

1.

Vogtle TS; def. of OPER LCO 3.8.1.1 LCO 3.8.1.2 LCO 3.9.8.2 2.

4/10/80 generic letter which includes original def. of OPER and 3.0.5 3.

6/11/80 generic letter 4.

2/2/89 memo Rosa to Butcher 5.

1/6/89 memo Rosa,to Virgilio 6.

4/10/83 memo Eisenhut to Norelius

?

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9 0

\\

i 4

l 1

l l

1 ENCLOSURE 1 O

e 9

i

t o

DEFINITIONS MEMBER (S) 0F THE PUBLIC

~

1.18 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupa-l tionally associated with the plant.

This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries, i

This, category does include persons who use portions of the site for recre-l ational, occupational, or other purposes not associated with the plant.

0FFSITE DOSE CALCULATION MANUAL j

1.19 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology l

l and parameters used in the calculation of offsite doses due to radioactive gasecus and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

OPERABLE - OPERABILITY 1.20 A system, subsystem, train, component or device shall be OPERABLE or l

have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its j

function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.21 An OPLRATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l

combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.22 PHYSICS TESTS shall be those tests performed to naasure the fundamental l

nuclear characteristics of the reactor core and related instrumentation:

(3) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l

1eakage) through a ronisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM i

1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l

j sampling, analyses, tests, and determinations to be made to ensure that proces-sing and packaging of solid radioactive wastes based on demonstrated processing of actual or sieulated wet solid wastes will be' accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State 1-4 Amendment No. 32 (Unit 1)

V0GTLE UNITS - 1 & 2 Amendment No. 12 (Unit 2)

Jul. 3 01990 u

a o

4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

Two physically independent circuits between the offsite transmission a.

network and the onsite Class 1E Distribution System, and b.

Two separate and independent diesel generators, each with:

1)

A day tank containing a minimum volume of 650 gallons of fuel (52% of instrument span) (LI-9018, LI-9019),

2)

A separate Fuel Storage System containing a minimum volume of 68,000 gallons of fuel (76% of instrument span) (LI-9024, 1

J LI-9025),and 3)

A separate fuel transfer pump, APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With one offsite circuit of the above required A.C. electrical power a.

sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

1 sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

If either diesel i

generator has not been successfully tested within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , demonstrate its OPERABILITY by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for each such diesel generator, separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the diesel generator is already operating.

Restore the offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With either diesel generator inoperable, demonstrate the OPERABILITY of the above required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within I hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

If the diesel generator became inoperable due to any cause other than preplanned preventive maintenance or testing, demonstrate

~

the OPERABILITY of the remaining OPERABLE diesel generator by perform-

)

ing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *#.

Restore the inoperable diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY.

The diesel shall not be rendered inoperable by activities performed to support testing pursuant to the ACTION Statement (e.g., an air roll).

YOGTLE, UNITS - 1 & 2 3/4 8-1

o O

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued)

With one offsite circuit and one diesel generator of the above required c.

A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. offsite source by performing Surveillance Re-quirement 4.8.1.1.1.a within I hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, and, if the diesel generator became inoperable due to any cause other than preplanned preventative maintenance or testing, demon-strate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within8gours*,unlesstheOPERABLEdieselgeneratorisalready operating.

Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore the other A.C. power source (offsite circuit or diesel generator) to OPERABLE status in accordance with the provisions of 3.8.1.1, ACTION Statement a or b, as appropriate, with the time requirement of that ACTIM Statement based on the time of initial loss of the remaining inoperable A.C. power source.

A successful test of diesel generator OPERABILITY per Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 performed under the ACTION Statement for an OPERABLE diesel generator or a restored to OPERABLE diesel generator satisfies the diesel generator test requirement of ACTION Statement a or b.

d.

With one diesel generator inoperable in addition to ACTION b. or c.

above, verify that:

1.

All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and When in MODE 1, 2, or 3, the steam-driven auxiliary feedw'ater 2.

pump is OPERABLE.

If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With two of the above required offsite A.C. circuits inoperable, demon-e.

strate the OPERABILITY of two diesel generators separately by perform-ing the requirgments of Specification 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANOBY within the next Following restoration of one offsite source, follow ACTION 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Statement a with the time requirement of that ACTION Statement based

This test is required to be completed regardless of when the inoperable EDG is restored to OPERABILITY.

The diesel shall not be rendered inoperable by activities performed to support testing pursuant to the ACTION Statement (e.g., an air roll).

vnnTlF llNTTS - 1 & 2 3/4 8-2

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) on the time of the initial loss of the remaining inoperable offsite a.c. circuit.

A successful test (5) ef diesel OPERABILITY per Surveil-lance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 performed under this ACTION Statement for the OPERABLE diesels satisfies the diesel generator test requirement for ACTION Statement a.

f.

With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing the require-ments of Specification 4.8.1.1.1.a. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SKCTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Following restoration of one diesel generator unit, follow ACTION Statement b with the time requirement of that ACTION Statement based on the time of initial loss of the remaining inoperable diesel generator.

A successful test of diesel OPERABILITY per Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 performed under this ACTION Statement for a restored to OPERABLE diesel satisfies the diesel generator test requirements of ACTION Statement b.

SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class 1E Distribution System shall be:

a.

Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, and indicated power availability.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a.

In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1)

Verifying the fuel level in the day tank (LI-9018, LI-9019),

2)

Verifying the fuel level in the fuel storage tank (LI-9024, LI-9025),

3)

Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, 4)

Verifying the diesel starts and that the generator voltage and frequency are 4160 + 170 -135 volts and 60 + 1.2 Hz within 11.4 seconds" after the start signal.

The dTesel generator shall be started for this test by using one of the following signals:

"All diesel generator starts for the purpose of surveillance testing as required by Specification 4.8.1.1.2 may be preceded by an engine prelube period as recommended by the manufacturer so that the mechanical stress and wear on the diesel engint. is minimized.

l V0GTLE UNITS - 1 & 2 3/4 8-3

s ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) a)

Manual, or b)

Simulated loss-of-of fsite power by itself, or i

c)

Simulated loss-of-offsite power in conjunction with an ESF Actuation test signal, or d)

An ESF Actuation test signal by itself.

5)

Verifying ths, generator is synchronized, loaded to an indicated 6800-7000 kW, and operates at this loaded condition for at least 60 minutes, and 6)

Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

7)

Verifying the pressure in at least one diesel generator airstart receiver (PI-9060, PI-9061, PI-9064, PI-9065) to be greater than or equal to 210 psig.

b.

At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from the day fuel tank; c.

At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks; d.

By sampling new fuel oil in accordance with ASTH-04057 prior to addition to storage tanks and:

1)

By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:

a)

An API Gravity of within 0.3 degrees at 60*F, or a specific gravity of within 0.0016 at 60/60'F, when compared to the

-supplier's certificate er an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or 4

equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees:

This band is meant as guidance to avoid routine overloading of the diesel generator.

Loads in excess of the band or momentary variations due to chang-ing bus 1 cads shall not invalidate the test.

All diesel generator starts for the purpose of surveillance testing as required by Specification 4.8.1.1.2 may be preceded by an engine prelube period as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.

i V0GTLE UNITS - 1 & 2 3/4 8-4

]

4 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) r i

b)

A kinematic viscosity at 40*C of greater than or equal to I

1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with supplier's certification; c)

A flash point equal to or greater than 125'F; and d)

A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82.

2)

By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with ASTM-0975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D2622-82.

At least once every 31 days by obtaining a sample of fuel oil in e.

accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-02276-78, Method A*

I f.

At least once per 92 days and from new fuel prior to addition to the.

l storage tank obtain a sample and verify that the neutralizatio number is less than 0.2 and the mercaptan content is less than 0.01%

i g.

At least once per 184 days by:

s 1)

Verifying the diesel starts

from ambient conditions and the generator voltage and frequency are 4160 + 170, -135 volts and 60 t 1.2 Hz within 11.4 seconds after the start signal. The diesel generator shall be started for this test by using one of the signals listed in Surveillance Requirement 4.8.1.1.2.a.4.

This test, if it is performed so it concides with the testing required by Surveillance Requirement 4.8.1.1.2.a.4, may also serve to concurrently meet those requirements as well.

"All engine starts for the purpose of surveillance testing as required by l

Specification 4.8.1.1.2 may be preceded by an engine prelube period as recommended by the manufacturer to minimize mechanical stress on the diesel l

engine.

j

Mercaptan content shall not be required to be verified within specification for new fuel prior to its addition, for up to 15,000 gallons of fuel added to the tank, if the last tank sample had a mercaptan content of less than All subsequent new fuel addition will require mercaptan content veri-0.007%.

fication prior to its addition until the tank contents are verified to be less i

than 0.007%.

1 e

V0GTLE UNITS - 1 & 2 3/4 8-5

^

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L i a

a ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) j l

2)

Verifying the generator is synchronized, loaded to an indicated value of 6100 - 7000 kW*** in less than or equal to 60 seconds, and operates with a load of 6800-7000 kW*** for at least 60 minutes.

This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.5, may also i

serve to concurrently meet those requirements as well.

h.

At least once per 18 months,** during shutdown, by:

1)

Subjecting the diesel to an inspection in accordance with proce-dures prepared in conjunction with its manufacturers' recommenda-tions for this class of standby service; 2)

Verifying the diesel generator capability to reject a load of greater than or equal to 671 kW (motor-driven auxiliary feedwater pump) while maintaining voltage at 4160 + 240, -410 volts and speed of less than 484 rps (less than nominal speed plus 75%

of the difference between nominal speed and the Overspeed Trip Setpoint); and recovering voltage to within 4160 + 170, -410 volts within 3 seconds.

3)

Verifying ?.he diesel generator capability to reject a load of 7000 kW without tripping.

The generator voltage shall not j

exceed 5000 volts during and following the load rejection; 4)

Simulating a loss-of-offsite power by itself, and:

a)

Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b)

Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected j

loads within 11.5 seconds," energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads.

After energization, the steady-state voltage and frequency of the emergency busses i

shall be maintained at 4160 +170, -410 volts and.60 + 1.2 Hz j

during this test.

5)

Verifying that on an ESF Actuation test signal, without loss-of-offsite power, the diesel generator starts

on the auto-start signal and operates on standby for greater than or equal to 5 minutes.

The generator voltage and frequency shall be 4160

+170, -135 volts and 60 t 1.2 Hz within 11.4 seconds after the 1

All engine starts for the purpose of surveillance testing as required by Specification 4.8.1.1.2 may be preceded by an engine prelube period as recom-mended by the manufacturer to minimize mechanical stress and wear on the diesel engine.

- For any start of a diesel, the diesel must be operated with a load in accor-dance with the manufacturer's recommendations.

- - This band is meant as guidance to avoid routine overloading of the engine.

Loads in excess of this band or momentary variations due to changing bus loads shall not invalidate this test.

V0GTLE UNITS - 1 & 2 3/4 8-6 f

4 SURVEILLANCE REQUIREMENTS auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test; 6)

Simulating a loss-of-offsite power in conjunction with an ESF Actuation test signal, and:

a)

Verifying deenergization of the emergency busses and load shedding from the emergency busses; b)

Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 11.5 seconds," energizes the auto-connected emergency (accident) loads through the load sequencer and l

operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +170,

-410 volts and 60 + 1.2 Hz during this test; and c)

Verifying that all automatic diesel generator trips, except engine overspeed, low lube oil pressure, high jacket water l

temperatures ### and generator differential, are automatically ll bypassed upon loss of voltage on the emergency bus concurrent 1

with a Safety Injection Actuation signal.

7)

Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded to an indicated 7600 to 7700 kW,"* and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to an indicated 6800-7000 kW.**

The generator voltage and frequency shall be 4160 + 170, - 135 volts and 60 1 1.2 Hz within 11.4 seconds after the start signal; the steady-state generator voltage and frequency s and6021.2Hzduringthistest.gallbe 4160 + 170, -410 volts Within 5 mimates after com-plating this 24-hour test, perform Specification 4.8.1.1.2h.6)b);,,

8)

Verifying that the auto-connected loads to each diesel generator do not exceed the continuous rating of 7000 kW; 9)

Verifying the diesel generator's capability to:

"All engines starts for the purpose of surveillance testing as required by Specification 4.8.1.1.2 may be preceded by an engine. prelube period as

}

recommended by the manufacturer to minimize mechanical stress and wear on j

the diesel engine.

- This band is meant as guidance to avoid routine overloading of the engine.

Loads in excess of this band or momentary variations due to changing bus loads shall not invalidate the test.

Failure to maintain voltage and frequency requirements due to grid disturbances does not render a 24-hour test as a failure.

1. If Specification 4.8.1.1.2h.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.

Instead, the diesel generator may be operated at the load required by Surveillance Requirement 4.8.1.1.2.a5 kW for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temperature has stabilized.

ffThe high jacket water temperature trip may be bypassed.

l'

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I

4 A

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) a)

Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simu-lated restoration of offsite power, b)

Transfer its loads to the offsite power source, and c)

Be restored to its standby status.

i

10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;

11) Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the installed cross-connection lines;

12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within + 10% of its

~

design interval; 1.

At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 440 rps in less than or equal to 11.4 seconds; and j.

At least once per 10 years by:

1)

Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite i

solution, or equivalent, and 2)

Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection N0 of the ASME J

Code at a test pressure equal to 110% of the system design pressure.

4.8.1.1.3 Reports - All diesel generator failures, valid or nonvalid, shall be reported to the Commission in a Special Report pursuant to Specification 6.8.2 within 30 days.

Reports of diesel generator failures shall include the informa-tion recommended in Regulatory Position C.3.b of Regulatory Guide 1.108 Revi-

sion 1. August 1977.

If the number of failures.in'the last 100 valid tests on a per nuclear unit basis is greater than or' equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108. Revision 1. August 1977.

i V0GTLE UNITS - 1 & 2 3/4 8-8

a J

I TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE Number of Failures Number of Failures in in Last 100 Valid Last 20 Valid Tests

Tests

Test Frequency 51 14 Once per 31 days 1 2**

15 Once per 7 days

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Criteria for determining number of failures and number of valid tests shall j

be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.

For the purposes of determining the required test frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed, provided that the overhaul, including appro-priate post-maintenance operation and testing, is specifically approved by the manufacturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutiv'e tests in a single series.

Ten of these tests shall be in accordance with the routine Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 and four tests in accordance with the 184-day testing requirement of Surveillance Requirement 4.8.1.1.2.f.

If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to transvalue the failure count to zero requires NRC approval.

- The associated test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one.

V0GTLE UNITS - 1 & 2 3/4 8-9

J 4

ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUT 00WN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

One circuit between the offsite transmission network and the Onsite a.

Class 1E Distribution System, and b.

One diesel generator with:

1)

A day tank containing a minimum volume of 650 gallons (52% of instrument span) (LI-9018, LI-9019) of fuel, 2)

A fuel storage system containing a minimum volume of 68,000 gallons of fuel (76% of instrument span) (LI-9024, LI-9025), and 3)

A fuel transfer pump.

APPLICABILITY:

MODES 5 and 6.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive re-activity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and provide relief capability for the Reactor Coolant System in accordance with Specification 3.4.9.3.

In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERA 8LE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated

)

OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5), and 4.8.1.1.3.

4 VDGTLE UNITS - 1 & 2 3/4 8-10

e i

J REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION Two independent residual heat removal (RHR) trains shall be OPERABLE, 3.9.8.2 and at least one RHR train shall be in operation.*

APPLICABILITY:

MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet.

ACTION-With less than the required RHR trains OPERABLE, immediately initiate a.

corrective action to return the required RHR trains to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible, i

With no RHR train in operation, suspend all operations involving a b.

reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR train to operation.

Close all containment penetrations providing i

direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS At least one RHR train shall be verified in operation and circulating 4.9.8.2 reactor coolant at a flow rate (FIC-0618A, FIC-0619A) of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

" Prior to initial criticality, the RHR train may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 2-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.

3/4 9-9 V0GTLE UNITS - 1 & 2

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ENCLOSURE 2 t

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Umtso stats NUCLE AR REGULATORY COMMIS$10N

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ALL POWER REACTOR LICENSEES Gentlemen:

It has recently come to our attention that there may be some misunderstanding regarding the use of the tem OPIMBLE as it ap criterion for safety systems in power reactors. plies to the single failure The purpose of this letter is to clarify the mening of this term and to request licensees to take specific actions to assure that it is appropriately applied at their facilities.

II Infomation Notice No, 79 35. ' Control of Maintenance and Essential Equipmen also contained infomattin on this subject.

The NRC's Standard Technical Specifications (STS) were formulated to preserve the single failure criterica for systems that are relied upon in the safety analysis report.

By and large the single failure criterion is preserved by 9

specifying Limiting conditions for Operation (LCOs) that require all redundarit components of safety related systems to be OPIMILE.

When the required redundancy is not maintained, either due to eculpment failure or maintenance cutage. tetion is required within a specifisc time to change the operating modr tri de plant to place,it in a ssfe condition.,The specified time to take actio.. vsually called the equipment out of. service time is a temporary relaxation of the single failure criterion, which, consistent with overall

system reliability considerations.'provides a limited time to fix equipment er.ciherwise make it OPIMILE.

If equipment can be returned to CPIMBLE status within the specified time, plant shutdown is not required.

LCOs are specified for each safety related system in the plant and with few exceptions, the ACTION ststements address single outages of com,ponents, trains

'cr subsystems. Ecr any particular system the LC0 does not address multiple of any support systems a such as, por does,it address the effects of outages outages of redundant components electrical power or cooling water. that are reited upon to maintain the OPIMl!LITY of the earticular system.

This is because of the large number of combinations of these types of outages that are possible.

Instead, the STS employ general specifications and an explicit definition of the term CPERA8LE to encompass all such cases. These provisions.

have been formulated to assure that no set of equipment outages would be allowed to persist that would result in the facility being in an untrotected condition..These specifications are contained in the enclosed Model Technical

$pecificatiini. 'lllustrative examples of how these specifications apply are containediin the associated lases.

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-2 April 10,1980 Because of the importance of assuring safety system availability,*the staff has concludtd that all facility technical specifications should contain these requi egtr, and that appropriate procedures should be implemented to assure that theaegessary records, such as plant logs or similar documents, are reviewed.

to detemine compliance with these specifications (1) prom tly upon discovering a component from service. ystem to be inoperable, and (2) a component, train, or subs prier to removing Therefore, we request that you (1) submit proposed changes to your technical Model TechnictI within 30 days, that incorporate the requirements of the enclosed specifications Specifications, and (2) implement the above described procedures to assure compliance with your proposed changes within 30 days thereaf ter.

With regard to technical specification changes, we recognize that the terminology used in the enclosed Model Technical Specifications may not directly apply to plants without STS, Y,herefore the OPEMTIONA!. MODE or CONDITION definitions are also included in the enclosure.

If you do not have STS you should redify the teminology to make it consistent with your particular facility technical specifications.

l If you have any questions, please contact us.

Sincerely.

M d(L 1 DarrellG.E$s.enhut. Acting Director eDivision of cperating Reactors e

Office of Nuclear Reactor Regulation Inclosure:

Model Technical Specifiestions b

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MODEL TECHNICAL SPECIF! CAT 10'Q PRES $UR17ED WATER REACTORS 1.0 des 1N1 Tits 5 OPEUBLE-TPIR. ABILITY 1.6 A system, subrystem, train, component or device shall be OPIMBLE or have OPEMBILITY *.an it is capable of perfoming its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instru-mentation, controls, nomal and ener ey electrical power sources, cooling or seal water lubrication or other aux ary equipment that are required for the system, subsystem, train, component or device to perfom its function (s) are also capabit of perfoming their related support function (s).

3/4 t1MITIN* CORD 1710N5TOROPEUTION(GENERAL) i 3/4.0 APPLIC ABf LITY_

LIMITIN3 CONDITION FOR OPEUTION 3.0.3 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of circumstances in excess of those i

l addressed in the specification, the unit shall be placed in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in at least HOT SHUTDOW within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and in at least COLD SHLf7DOW within the followias 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months unless corrective measures are completed that pemit operation under the permissible ACTION statements for the specified time interval as measured from initial discovery or until the reactor is placed in a MODE in which the specification is not applicable. Exceptions to these i

requirements shall be stated in the individual specifications, i

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3.0.5 When a system. subsystem, train, component er device is datermined to be inoperable soley because its emergency power source is inoperable, er solely because its normal power source is inoperable it may be considered OPIMBLE

]

for the pu ese of satisfying the requirements of its applicable Limiting i

Condition or Operation. provided:

1 1 its corres nding normal er amergen power source is OPIRABLIl and al' fitsredunantsystem(s), subsystem

),

train (s), component (s) and dev (s) are OPERABLE. er linewise satisf the requirements of this specification. Unless both conditions (1) and ( ) are satisfied, the unit shall be placed in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in at least HOT 5HVfDOW within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least COLD 5H'JT00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This specification is not applicable in MODES

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5 or 6.

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3/a.0 a#PLICABILITE

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Bast 5 This specification delineates the ACTION to be taken for circumstances 3.0.3 not directly provided for in the ACTION statements and whose occurrence would violate For example. Specification 3.5.1 requires each the intent of the specification.

Recctor Coolant System acceulator to be CPIRABLE and provides explicit ACTION re:utremnts if one accrulater is inoperable. Under the tems of Specification 3.0.3. if were than one accumulator is inoperable, the unit is r As a further example. Specification 3.6.2.1 requires two Contairrient Sprt,y Systems to be OPERABLE and provides explicit ACTION requirem 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

spray system is inoperable:the required Centaiment Spray Systems are inoperab

j in at least HOT STANDBY within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , in at least HDT SWTDOW within the It is following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least COLD SWTDOW in the next 30 h by promptly initiating and carrying out the appropriate ACTION statement.

3.0.5 This specification delineates what additional cenditions must be satisfied.to pemit operation to continue, consistant with the ACTION statements It for power sources, when a normal or emergency power source is not OPIRABLE.

specifically prohibits operation when one division is inoperable because its nonna1 or mergency power source is inoperable and a system, subsyste. train.

. component or device in another division is (noperable for,another reason.

The provitions of this specification pemit the ACTION statements associated with components, or devices to be consistent indivic'ual systems, subsystems, trains,iated electrical power source.

It allows with the ACTION statements of the assoc operation to be governed by the time Ifmits of the ACTION statement associated with the Limiting Condition for Operation for the normal or em ndividual ACTION statements for each system subsystem, train, component or device thH is detemined to be inoperable solely because of the inoperability of its acrsal of emergency power soutet.

For example. Specification 3.8.1.1 tequires in part that two eme generators be CPIRABLE.

If the definition of time when one emertency diesel generator is not OPERABLE subsystems, trains, components and devices supplied b ACTION statements for each of the applicable Limiting Conditions for Opera} ion.

However, the pvevisions of Speelfication 3.0.5 pemit the f

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eeergency diesel generator instead, provided the other specifie rust be OPIMBLf., and all redundant syster.s. Subsyste.s satisfied.

d have at least one morral er one emergency power source OPEPABLE)gn function anIf they are not s capable of perftming their desi in accordance with this specification.

As a further example. Spec-ification 3.8.1.1 requires in part that two physically 5

independent circuits between the offsite transmission network and the onsiteT Class It distribution system be OPEPABLE.

hour out.cf. service time when both nquired offsite circuit's j

3.0.5, all systems, subsystems, trains. components and devices supplied by the J

both of the offsite circuits, would also be 4

inoperable nomal power sources, invoking the applicable ACTION statements for inoperable. This would dictateHowever, the provisions of Specification 3.0.5 each of the applicable LCOs.

pemit the time limits for conticued operation to be consistent with the ACTION statenent for the inoperable nemal power sources instead, provided the o j

specified conditions are sstisfied.

division the emergency power source must be OPEMBLE (as aust be the components j

supplied by the avrgency power source) and all redundant systems, subsystems.

trains cer ponents and devices in the other division mus In other words, both energency and have an emergency power source OPEPABLE).

poetr sources must be OPEMBLE and all redundant syststs, subsystems, trains If these components and devices in both divisions must also be OPER specification.

In MODES 5 or 6 Specification 3.0.5 is not applicable, and thus the individual ACTION statevnts for each applicable Limiting Condition for Operation in these

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MDES must be adhered to.

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DEFINITION OF WESTIN3H00$E PW T'

OPERATIONAL E DES I

REACT!YITY 5 RATED AVERAGE COOLANT MODE CONDITION, Keff THEPp.AL POWER

TEMPE RATURE l

1 99

> 5%

1 3500F 0

1.

POWIR OPIRATIDH 1

99

< 55 1 350er 0

2.

STARTUP 3.

ET STAEBY

< 0.99 0

1 3500F

4. ET SmW

< 0.99 0

350eF > Tavg

> 2000F 5.

COLD 3HtTTDOW

< 0.99 0

1 200er 1

95 -

0 1 140er 0

6.

RIFUILIM**

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Exclucing cecay heat.

"Reacter vessel head unbolted or removed and fuel in the vessel.

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ENCLOSURE 3 6

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UNif ED STATES NUCLEAR REGULATORY COMMISSION

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June 11. 1980 TO ALL OPERATING PRESSURIZE 0 WATER REACTORS (PWR'S)

Gentlemen:

This letter transmits the request that you amend the Technical Specifications (TSs) for your facility with respect to reactor decay heat removal capability.

The basis for our request is founded in a number of events that have occurred i

at operating PWR facilities where decay heat removal capability has been seriously degraded due to inadequate administrative controls utilized when the plants were in shutdown modes of operation. One of these events occurred at the Davis-Besse. Unit No.1 plant on April 19. 1980 which was described in IE Information Notice 80-20 dated May 8.1980.

In IE Bulletin 80-12 dated May 9,1980. you were requested to innediately implement administrative con-trols which would ensure that proper means are available to provide redundant methods of decay heat removal. While the function of the bulletin was to i

effect immeolate action with regard to this problem we consider it necessary that an amendment of your Itcense be made to provide for permanent 1cng term assurance that redundancy in decay beat removal capability will be maintained.

I You are requested to propose TS changes for your facility that provide for redundancy in decay heat removal capability for your plant (s) in all modes of operation. To assist you in preparing your Amittal, we have enclosed a copy of Model TSs which would provide an acce xable resolution of our Your proposal should use the enclosure as a guide and should concern.

include an appropriate Safety Analysis as a basis.

It is requested that you submit your proposed TSs with the basis within If you have any questions about this 120 days of receipt of this letter.

matter, please contact your Project Manager.

Sincerely, DarrellG.E{se*nh rector Division of Licensing r

Enclosure:

Model TSs concerning Decay Heat Removal Capability

a

'O' TO ALL OPERATING FRESSURIZED WATER REACTORS (PWR'S)

Gentleren:

This letter transmits the request that you amend the Technical Spectfications (T5s) for your fact 11ty with respect to reactor decay heat removal capability.

The basis for our request is founded in a number of events that have occurred at operating PWR facilities where decay heat removal capability has been seriously degraded due to inadequate administrative controls utilfred when the plants wre in shutdown modes of operation. One of these events occurred at the Davis-Besse. Unit No.1 plant on April 19, 1980, which was described in IE Infomation Notice 80-20 dated May 8.1980. In IE Bullett 80-12 dated May 9,1980, you were requested to imediately implement administrative con-trols which sculd ensure that proper means are available to provide redundant methods of decay heat removal. While the function of the bulletin ws to effect irunedtata action with regard to this problem, we consider it necessary that an asundnant of your 1(conse be made to provide for pemanent long term assurance that redundancy in decay heat renoval capability will be maintained.

You are requested to propose 75 changes for your facility that provide for fn all modes redundancy in decay heat removal capability for your Mant(s)have enclosed of operation. To assist you in preparing your submi'3.41, we a copy of Model T5s which would provide an acceptabt. resot4tfon of our Your proposal should use the enclosure as a guide and should concem.

include an appropriate Safety Analysis as a basis.

It is requested that you substit your proposed T3s with the basis within 120 days of receipt of this letter.

If you nave any questions about this matter, please contact your Project Manager.

Sincerely, Darrell G. Eisenhut Director Division of Licensing

Enclosure:

Model T$s concoming Decay Heat Removal capability 600n80 $7#I

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NUCLEAR REGULATORY COMMISSION I

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%, r a.....f June 11, 1980 MEMCRANDUM FOR: Thomas M. Novak, Assistant Director for Operating Reactors Division of License, NRR Gus C. Lainas. Assistant Director for Safety Assessment Division of Licensing. NRR FRCM:

Darrell G. Eisenhut, Director Division of Licensing. NRR

SUBJECT:

GENERIC LETTER CONCERNING DECAY HEAT REMOVAL CAPABILITY Attached is a generic letter to all operating PWR's which requests licensees to amend the Technical Specifications (TS) for their facilities concerning decay heat removal capability. Also attached are model T5s for each of the three PWR vendor types of plants. The letter, with the appropriata version of the model T5s shos:1d be sent to Itcensees by each Operating ceactor Branch within the next week.

The attimated total nanpower expenditure for review of submitted TSs is 0.1 The lead engineer assigned manyear per reactor site or about 5.0 manysars.He will initiate TACS for all is Daniel Garner. (Room 334, ext. 27435).

facilities and will forward sheets to the Project Managers for completion.

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a Otuiston of icensing, NRR Attachments: As stated

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o Distribution:

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tral file DGarr.er OR R!ngram NRC PDR RReid NRR Rdg.

MDORANDLH FOR: Thomas M. Novak Assistant Director for Operating Reactors Division of Licensing NRR Gus C. Lainas Assistant Dimeter for Dib NNng, NRR FROM:

Darrell G. Eisenhut, Director Division of Licensing, NRA

SUBJECT:

GDWtIC LETTER CONCERNING DECAY HEAT RDCYAL CAPAGILITY j

Attached is a generic letter to all operating PWR's wt.fch requests licensees to arund the Technical Specifications (TS) for their facilities concerning decay heat raraovel cawb111ty. Also attached are sedel T5s for each of the three PWR vendor types cf plants. The letter, with the appropriate version of the rodel TS: should Lt sent to licensees by each Operating Reactor Branch within the next week.

The estimated total manpower expendituresfor nyiew of suberitted T5s is 0.1 canyear per reactor site or about 5.0 manyears. The lead engineer assigned is Daniel Garner, floca 334, ext. 2744 He will initiate TACS for all facilities and will forwaN sheets to the Project Managers for cospletion.

Orf rical signed W Darrell G. Eisenhut Director Division of Licensing, NRA Attactments: As stated Ch

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STARTUP aND PCWER CPEE.AT*CN t **!T!'tG C**;01T1CN ~CR CPESATICN in each loop 5cth reactor c:al. ant loops and both reac ce coolant pu::pt 3.4.1.1 shall be in operation.

l APot1CABILITY: "COES 1 ano 2'.

ACTICH:

ump not in operatien STARTw"P and PCWER

'n'ith one reactor coolan:

OP!UTION r.ay be initiated and :nay proceed prDvided THE7."At PCW is restricted to less than ( ):

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the se: points for the following trips have been reduced to the values specified in Specification 2.2.1 for operation with three resc er c:elant p.eps operating:

i (Nuclear Overtower). Nuclear Over:ower based on RCS fl 1.

% clear overpower based on p=; menit:rs).

2.

3.

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I pav!!LLANCIRIM:.!w!N*S Tne above required reactor coolant loops shall be verified to be in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.1 Oceration and circulating reactor coolant at least once per The Reee :r ?-::ective Instr *;mnta. ipa channels specified in the a de verifiec to have had their trip setpoints changed 4.4.1.2 f rete:or ACTICN statement acove snal)to the vaives 5:ecified in $cecif coolant pu :s c;erating either:

=;: ::::bination if.

a hears af ter swit:r.ing = a diffe en Vi

'n a.

tr e s w * :n i s ra f.e w hil e :: e r a

ir.g. c r' P-i:r to react:r :. isicality if the swit:n is Fade wnile shutdcen.

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'see 5pec.a l.es: Izeep;i:n 2.10.4 3rd-5~5

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R!aC*02. C00'.1NT SY$*EM 3/a.a WOT $7A?C5Y 702 C7 Etat *0N LIM! TING CON 0*.i1CM The reactor c:olant loops listed below shall be CPERABLE:

3.4.1.2 t.

React:r Ceolant Loop (A) and at least one associated react:r 1.

c:olant pump.

Reactor C:olant Loop (3) and at leastspne associated reactor 2.

coolant pump.

itas: one of the above Resctor Coolant L::;s shall be in b.

A:operation'.

(

..CCE 3 A78LitA31L171:

ACTIC4:

loops OPERABLE restore

a. With less than the above required reactor coolan:the requ CT EMUTDCWN I

ni:hin the nex 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i With no reactor coolant locp in operation, suseerd b.

and i==ediately ini:ia:e action to rpturn the required System

elant 1 coo to cceration.

$U 'li.LisCE RE}"* JE*ENTS l te

ans :Se above required reactor coolant put:s if net in ope-ttie k

alignments to te OPERA 2LE once per 7 days by verifying c:rrect brea er A:

4. 4.1. 2.1 determine:

cwer availacility.

i

eas: ene ecoling loop shall be verified to be in operation and indicate:

4.4.1.2.2 A:

reactor ::cian; at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 ided (1) 90 rea:-:r colar.; ;.::s e.ay be ce-ener;i:ed for vp to i h:ur provv:uld cause teer.e-1:vre is taintained a

'A.:

e sti:-s tre stati *.ed ina i

system :: :r. c:rcer.:rt:i:n. and '2) ::re ev:le:

t*/~ :el:w sa urn: ton tar:ert:are.

i

'tss:

BEN *h h

3/:.: tr:: :7 00ct:NT SY! !M SHUT 30uN 204 OP ES AT

O*:

(:u!T*NG CON 0!T*0N wo of *.he coolant loops listed :elow shall be OPERA!LE:

a.

A: 1ees:

3. 2. '.'3 (A) and its associa:ec steam gene s se Reacter Coolant Lee:

loss: one asscciated resc::r coolan: ;ums.

1 anc a:

Reacter Coolant Loop (S) and its associated steam generator least one associated reac:cr coolant pumD, 2.

anc 4:

temeval L:op (A).'

3.

Decay Hea:

4 Oecay Heat timeval Loop (3).'

At least one of the above coolant Icers shall be in opert:ic3.

b.

aPOLICABIL'TY: MODES a and 5 ACT!C){:

Vi:n less inan the above equired coolant locos CPEP.ABLI. immediately initiate corrective action to return the reevired coolant icc:s to a.

OPERABLE sta:us as socn as possfale; te in COLD SHUTDOWN witnin i

20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

1cco in operation sus;end til cetrati ns invclving a reduc-'en in boren :encentration of the teactor C:o W'.th me :colan b.

'~ iolint 1ccc to oceration.

c SURVEf t'.aNCI S.!OU11E*ENTS The recuired cecay heat removal 1:co(s) sna11 be catermined OPERA 5 2.4.1.3.1 ser 50etification 4.0.5.

The recuired rtac cr coolant pumo(s), if not in coeration, shall be !

breaker alignmen:s 4.4.1.3.2

ste-mined to be OPERABLE.once per 7 days by veridying :erree:

l and indi:ated ;ower availability.

The recutred s: cam generat:r(s) shall be determinec :PERABLI by ver

/

)').

4.4.1.3.3 sec:ndary s :e level :o se gres:er than or e: val is (

verified to be in coera*iCn and Cir*ulatin9 A.S.I 3.A At leas

ene c:olant loco shall be a: lets: once ser 12 hcurs.

[

eac:or ::cian:or emergercy : ewer seur:e may :e 'ne:e acie :n MCOI 5.

!re no rr.a s f-r

- A*.1 *escier c oiant :um:s and tecay heat removal pum:s -ay te de-emergi:ed cermitted the: wculd causa r.*'o*ien to i n:ur previces 'i) no coeratiens at:sys:em coran concentration, and (2) core : :

cf sne *eacice c: clan:

tem:erature is main:2ined a: least iC*F below sa:uration tem:erature.

! &* *- ST S I

I

! :'. P. : '. -:!)AT :NS LO*' '* ?io L EVEL L:F1 IN- ::::0!? t *N FOR CPE:.ATICN i

3.i.E.2 7.o ince:endent ;HR Iceps shall be OPERABLE.'

MOC[ 6 when the water level above the top of !..e tr*aciated AP:8. : C'!! L ! 7Y :

fuel assem.clies seated within the reactor pressare vessel is less than 23 feet.

l j

r-,

m..i.

'4ith less than the recuired OHR ICCos 07!RASLE, in'eciately initiate f

a.

errective action to return the required iceps to OpgAAILE status as seen as pessible.

he previsiens of Specification 3.0.2 are not applicable.

b.

e...:.; ;... : y... e r.... 3 :er.N7 5

~he retuirec DHR iceps snal) be deter.ir.ed CP!;A3LE :er 5:ecif' cation 2.7.5.2 A.3.5.

e :

.. :r + t ;2.:y ::.e-s:.r:e 17 :e ' ::t-2 :t e '..

i i

,3/1.4

! CT*2 CCCL N7 SYSTEM

!ASES 3/a.a.1 CCOL NT LOCp5 AND CCCLANT CIRC'JLAT!CN

erate ith both reacter coolant loops in The plant is ces'gned: :

during all nort.41 ope-2:i:ns

e stion, and maintain Cn3R above (1.32/1.30)Wi: 5 one reactor cociant pumo net in ope and an:ici;ated ;ransie'nts.Icop, inERMAL POWER is restricted by the Nuclear Over i

en SCS Flew and AII AL POWER !MBALANCE and the Nuclear Over;cwer Based en S.m u

one

nit:rs trip, ensuring tha t the NBR will be maintained above (1.32/1.20) at the maximum possible THER"AL POWER fcr the nu.mber of

?

whic*tver is more restrictive.

In MCDE 3. a single etact:r c:clant Ices rovices sufficient heat remeval ct:ati's'*y for et?Cvirg decay heat Acwever, single failure cCnsidert:f cas re: sire that two 'ce:s be CPERABLE.

In.50 DES a ane 5, a single reactor coolant 1ces or CHR Icep pr sufficien: hes:

Psice-a: ices recuire that a: less: two iceps te C*ERABLE. Thus, i' the reic :e :: alan: ic :s are not CPERASLE, this see: 1ficati:n requires two OkR 1:c:s to te OPERA 5LI.

The 03e'Itien Cf One ".eaC ce C clant Pump or Cne CHF,;mo previcts ide:

Vate s:-4:idiesti:n and preduc * : adual -et::iv ty f

sw :: e-ssre mixing, ;*twen:

t

Parges curing teren c:r.cen:rsti:n recuctions in ;$e. eac er : olant Sjstam.

~'s -et:t vity c-ar;e -ete ess:c'atec with bcr:n -tfuc-i:n will, there': t, te e': Sin ;r.e ca:ati'.. y of :cerater rec:gnition anc ::ntrol.

e t

)

I*..

IT3 l

* ' * " ~ ~. - l

- - - - : J '

L

a!FUit:NG *2!:::1CNS.

j EA5!5 DECAY FI AT R! OVAL AND C00thd7 C tC'.JL AT*0't I

3/a.9.3 sufficient cooling es;acity is available to re. :ve decay i d during :se IITUEUNG water in the reactor ;ressure vessel below 140'F as ree.u.ecircula: ion is -aintained t

CCE. and (Il sufficient c:clar.:c:re to. mini.1:e the effect f a bcr:n dilution in

oren strati-5 ficati:n.

The -tcuiretent to have t'.<o DMR Icces OPEVELE when there is les feet of wa:er accve :te c:re ensures tha:Ic o will act result in a cerroitte 1 W1:n

arge heat

Me rese::e sessei nead removec and 23 feet of water abcve :te c:re, a~hus, in sink is a.ai'.a:1e for cere ceciing.Te is ;r: viced to initiate etergency ;r:cetures

t atir.g ::4 ic:p, ace:,, ate :

c:cl :ne c:re.

t 9

.....g a &* *e i

~---__..,,g;,y; w

O OR COOLANT SYSTEM

@D 4%

6 tT g

COOLANT LOOPS AND COOLANT CIRCULATION T[

'[e /

ANO CWER OPEPATION

~

3 CONDITION FOR OPERATION Soth reactor coolant loops and both reactor coolant per.ps in each loop e in operation.

!!LITY: MODES I and 2*.

With ene reactor coolant pump not in operation, STARTUP and POWER OPER* TION may be initiated and may proceed provided THERS.AL POWER l

is residcted to less than ( )t of RATED THERFAL POWER and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the setpoints for the following trips have been reduced to i

the values specified in Specification 2.2.1 for operation with three r

reactor coolant pu.ps operating:

1 (Nuclear Overpewer).

WN l

2.

(Nuclear Overpewer based on RCS flow and AX!AL Pt%T.R IPSALANCE).

3.

('Nelear Overpewer based on pump monitors).

I e

l 11 be i

1 Nt! REOUIREMENTS

,ing Th above required reactor coolant loops shall be verified to be in 1

and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The Reactor Protective Instrumentalisa channels specified in the applicable stement above sh4H be verified to have had their trip setpoints changed

's specified in Specification 2.2.1 for the applicable number of reactor i

' cs operating either:

dithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after switching to a different pump coc61 nation if the switch is made while operating, or Srior to reactor criticality if the switch is made while shutdown.

ia Test Exception 3.10.4 T

8007180 k

W t

Y hYT REFUELING OPERA 1!ONS

$$dA 3/4.9.8 DECAY HEAT REMOVAL AND COOLANT RECIRCULATION M hlMbI U

)D ALL WATER LEVELS LIMITING CO?c! TION FOR OPERATION 3.9.8.1 At least one decay heat removal (DHR) loop shall be in coeratien.

APPLIC AB11.!TY : MODE 6.

ACTION:

With less than one DHR loop in operation, except as provided in b below, susoend a.

all operations involving an increase in the reactor decay heat lead or a recuction Close all containment pene-in boren concentration of the Reactor Coolant System.

trations providing direct access fmm the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The DHR loop may be re.eved from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months period i

b.

during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure,

vessel (hot) legs.

The provisions of Specification 3.0.3 are not applicable.

l c.

SURVE1Lt. ANCE REOUIREMENTS t

oolant at a flow rate of greater than or equal to (2800) gpm 9

BW STS

RI UELING CPERATIONS EASES DECAY HEAT RENOVAL AND COOLANT CIRCULAT10*i J/A.9.8 The requirement that at least one DHR lo p be in operation ensures that (1) sufficient ecoling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUEL!tiG l'CDE, and (2) sufficient coolant circulation Is maintained through the reactor c:re to minimize the effect of a boron dilution incident and ;revent boren strati-fication.

The requirement to have two DHR locos OPE *>SLE when

sere is less than 23 of the epersting CHR feet of water above the core ensures that a single failur :

With lo:p will not result in a complete loss of decay heat re" sval capability.

tre reactor vessel head removed :-d 23 feet of water abo e the core, a lar;e heat sir.k is a<ailable for core cooling. Thus, in the event of a failure of the c:erating CHR loop, adequate time is provided to initiste e-ergency procedures to cool the core.

i

! 1 i

l t

h 3 D.*-f * $

1

\\

OEFINITION OF BARC0CK I'WILC01 PWR OPERATIONAL M00E5

' REACT!YITY 5 0F RATED AYERAGE COOLANT CPE:ATIONAL MODE CONDIT!cN, K,ff THERM.AL poker

TEMPERATURE

> 55

>(305)CF 1.

POWER OPERATION

> 0.99 4 55

> (305)CF 2.

STARTUP 3,0.99 3.

Foi STAN05Y

0.99 0

> (305)CF HOT SMUTOCWN

< 0.99 0

(305)cF>Tayg>200eF 5.

CCLD IM'JT30WN

< 0.99 0

< 2000F 6.

REFUELING

< 0.95 0

,1400F

Lacluc1ng cecay heat.

React:. vessel head unbolted or removed and fuel in the vessel.

s

e..

-- c,c

- =.

O.~

3/a.4 R! ACTOR COOLANT SYSTEM 3/a.4.1 COOLANT LOOPS AND COOLANT CIRCULATION _

STARTUP AND POWER OPERATION LIMIT!NG CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant ;v ps in each loop e

shall be in operation.

AF8L1CABILITY:

1 and 2.*

ACTION:

'a'ith less than the above required reactor coolant pumps in :peration, be in at least HOT STA!!DBY within I hour.

SURVE!LLANCE RE0UIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'See Special Test Exception 3.10.3.

L007180(,C!/

CE-STS

,lp

' ~ ~

- = -

RE!.CTOR C00tAf4T SYSTEM HOT STA!C3Y i

L M1 TINE, CONDIT!ON FCR OPEDATION The reactor coolant loops listed below shall be OPE 2.A!LE:

3.4.1.2 a.

1.

Reactor Coolant Loop (A) and at least one associated 1

reactor coolant pump.

2.

Reactor Coolant Loop (S) and at least one assoc'ated reactor coolant pump.

At least ore of the above teactor Cec 1snt L:cos saall te in b.

operat en'.

' F. l C ; 31 t l T Y : M00! 3

CTION:

With less than the ateve required reactor coolant locps c;erible, a.

restore the required 1 cops to OPERABLE status witMn 72 5:urs or be in NOT SHUTCOWN within the next 12 heur:;.

b.

With no reactor coolant loop in operation, suspend all c e ati:ns involving a reduc. tion in bcron concentration of the Reacter Coolant System and 1. mediately initiate corrective action to return the required 1 cop to operati,on.

Stayt;ttA9CE D!cuittytyTS 4.4.1.2.1 At least the above required reactor coolant pumps, if -ot in c eration, shall be determined to be OPERABLE once per 7 days ty.erify-ing correct breaker alignerts and indicated power aveilability.

4,4.1.2.2 At least one cooling 1 cop shall be verif'ed to be in c;eratien and circulating reactor coolant at least ence per 12 heves.

l

'A11 reactor c:clant p ros.ay be de.e er;i:<: fte,o to 1 ' car ;r:v'de: (1)no operations are permitted that. wid cause 'i?Lti*n of the eeJct;r c:c'!"*

l system d'$ren concertrat!cn. 39d (2) e. ore at14t te re sture 's t'**.s'*e'.

It lesst *G'F telcw saturst'on te ;4rt*,re.

CE.STS mi j

I-K! ACTOR COOLANT SYSTEM SHUTDC'4N LIMiiiNG COND! TION FOR Op!D.aTION At least two of the coolant iceps listed belcw shall be CPER !LE:

3.4.1.3 a.

1.

Rea: tor Coolant Loop (A) and its associated steam genera;:r and at least one associated reactor ecclant pump, 2.

Reactor Coolant Loop (B) and its associated steam ge.e ator and at least one asse: dated reactor coclant pump.

3.

Shutd:wn Cooling Loop (A)$

4 Shutdown Cooling '. cop (B)#

5 s".all te in c;erati:n*.

b.

At least one of the ateve c:c' ant 1:o:

AP:L!CtBILITY:

MODE 5 4 " and $**

at:1T::

With less than the above required coolant loops CPERABLE, a.

immediately initiate corrective action to return the required coolant loops 't6 0PERABCE'st'a'tu's' is soon as possible; be in COLD SHUTCOWN yi, thin 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

With no coolant loop in operation, suspend all operations invc1ving b.

a reduction in boren concentration of the Reactor Coolant Sy' stem i

and immediately initiate corrective action to return the required

~ ~ ~ '

toolant loof t6 6peration.

SUU.'E!LLaNCE REOUIREMENTS The re:g' ed shutdown cooling locp(s) shall be determined CPEE*3LE 4.4.1.3.1 per 5;ecification 4.0.5.

The re u' red reactor coolant ;s.p(s), if not in c:eratien, shall be J.4.1.3.2 1 breaker ali;t e-ts deterr.ined to be GFIP.ABLE cnce per 7 days by verifying corre:

and ir.dicated p.cwer availability.

l p;-ps.4/ te de-ereet':ed fcr j

3 Ail reacter coolant pumps and decay beat remova Jis:ilst' n up to I howr ;rovided (1) no cpeartions are ;ermitted t6at 'aculd C:

I of tr.e reactor ceolant system teren ca.ncentrat'co, and (2) ctre out'et tet;erature is sintained at least 10 F helcw satjration it.cerat.re.

0

. p shall *.ot be started.ith ere er mere of t'e R;5 :e'd

A reactor ct:':*:

2.-i:e aster leg ic ;erats es ! ass than or et.a1 to (275)?F uctess 1) the ;re:(90

.e-2

.t. e cf r

h vet; e is less e :C 5 s t e am j e ".s :* a '.. P i s l e s s th s ". (

6 '4 II lic.e ec:h cf 1"e SCS :0'd le7 tee;eratures.

The *

r I " r e".er~ee c/ ;taer s c ;"*.e s / be ' *t:t' at

e '. ' *: E 5.

CI-5*S

I

?

e REACTCR COOLANT SYSTEM j

.RIACTOR COOL ANT LOOPS AND C00L ANT C!RCutaTIOM S'J8vE!LLANCE DE0 Vite *ENTS (Continued)

The required steam generator (s) shall be determined CPERASLE by 4.4.1.3.2 verifying the secondary side water level to be !

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

At least one coolant loop shall be verified to be in operation 4.A.1.3.4 and circut,ating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I i

)

E-!!S

l

\\

RE*'Ji'.ING 08!RAf f 0NS 3/4.9.8 SHUTDOWN COOL f NG AND COOL ANT CIRCUL ATION.

j ALL WATER LEVELS LIF!T.NG CCNDITION FOR OPERATION l

3.9.E1 At le;st one shutdown cooling loep shall be in ope ation.

A77 tic'3*LITY: MODE 6 ACT!D'l:

a.

With less than one shutdc n cooling 1:op in :pe-ation, except as providad in b. below, suspend all operations involvi9g an increase in the reactor decay heat load or a redsction in bcron concentration of the Reactor Coolant System.

Close all contairment penetrations providing direct access from the centainment atmosphere to the outside atmosphere within a howes.

b.

The shutdown cooling loop may be remcved from operation for uo to I heur per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel het legs.

c.

The provisions of Specif t:ation 3.0.3 are not applicable.

S*JR'.'!!LL *NCE Ri?U"REMENTS 4.9.t.1 At least one shutdown cooling 1 cop s5all be verified to be in c;eratien and circulating reactor coolant at a flow rate of greater than or equal to (3000) gpm at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

CE-5~5

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

[

p;rgitts3 optq:;;oss LOW WATER LEVEL L***T*NG CONDITION FOR 08E:.ATION 3.9.8.2 Two independent shutdown cooling loc;s shall be OPE 8ASLE.*

A P'_ItaSIL ITY :

MODE 6 when the wate-level ab ve the top of the irradiated fuel assr-blies seated within the reactor presssre vessel is less than 23 feet.

A C T I O*. :

With less than the required shutdown cooling loeps OPERABLE.

a.

immediately initiate. corrective action to return toeps to CFERABLE status as soon as possible.

b.

The pe:visiens of Speci'4 cation 3.0.3 are not a;;11:able.

4

$UP'.'E*LL*NCE SESVit!*EN'S 2.g.e.2 The ee: ires shutdown cooling loe;s shall de geter-ine: 0;[t:5,g s

er 5;e: ificatfor. 4.0.5.

!be 9;r al cr e"erienc/ power sosrce may te (*c;e Able # r each s*.st'e., c:elfr.g 1:ep, i

cE-::s

3/4.4 EEACTOR COOLANT SYSTEM EASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULAT!CN The plant is designed to operate with both reactor coolant loops and associated reactor ecolant pumps in operation, and maintain ONER above 1.30 during all normal operations and anticipated transients.

In M00E 3 a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations reRuire that two loops be OPERASLE.

In M00E5 4 and 5. a single reactor coolant loop or shuttewn cooling loop

rovices sufficient heat removal capability for removing decay heat; but single failure consideratiens require that at least two loops be CFE31!LE.

Thus, if the reactor coolant loops are not OPERABLE. this specificaticn requires two shutdown cooling loops to be OPERASLE.

The operation of one Reactor Coolant Pemp or one shutdown cociing ; ump provides adequate flow to ensure mining, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the teactor Coolant System.

The reactivity change rate associated with boren redsctions will, therefore, be within the capability of operator recognition and contecl.

The restrictions on starting a Reactor Coolant Pump during MOOCS 4 and 5 with one or more RCS cold legs less than or equal to (275):F are pecvided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR rart 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendia G by either (1) restricting the water volume in the pressuricer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water tet;erature of each steam ger.erator is less than (46)0F above each of the RCS cold leg tem;eratures.

CI-STS F '

RE UILING 0 !:ATIONS BASES 3/4.9.8 COOL ANT CIRC'JL ATION the requirement that at least one shutdcwn cooling loop be in operation ensures that (1) sufficient cooling capacity is available to recove decay heat C

and raintain the water in the reactor pressure vessel below 140 F as re;uired during the REFUELING MODE, and (2) sufficient coolant circulation is maintair.ed thrcsgh the reactor core to minimize the ef fects of a boron dilution incident and ;revent boren stratification.

The retuire ent to have two shutdcwn cooling loops CFEEAELE. hen there is less than 23 feet of water above the core, ensures that a singte failure of the e,;erating shutdcwn cooling loop will not result in a cor.plete loss of decay With the reactor vessel head rescved and 23 feet of heat removal capability,

=ater above the cere, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdcwn cooling loop, ade;uate time is provided to initiate emergency procedures to cool the core, C -5~5

D.

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 P.E CTOR COOLANT LOOPS AND COOLANT CIRCULAT:CN STARTUP AND PCWER OPERATION l

LIMITING CC..'DITION FOR OPERATION 3.4.1.1 All reactor coolant loops shall be in operatien.

APPLICA3!LITY:

MODES 1 and 2.*

ACTICN:

With less than the abeve required reactor coolant loops in cperation, be in at least HOT STANC5Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

1 SURVEILLANCE REQUIREMENT The above required reactor coolant loops shall be verified to be in 4.4.1.1 c;eration and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

-;.a 5;ec3al Tes: E<:eptien 3.10.4 f

8007180507

'hSI5 t

J

RE*CTOR COOLANT SYSTEM HOT STANDBY _

LIM 171i:G CONDITION FOR OPERATION At least two of the reactor coolant loeps listed below shall be 3.4.1.2 a.

OPERABLE:

Reactor Ceolant Lo p (A) and its associated steam generat:r 1.

and reactor coolant pump, Reactor Coolant Loop (B) and its associated staam generat:r 2.

and reactor coolant pump, Reactor Coolant Loop (C) and its, associated steam generster 3.

and reactor coolant pump, Reactor Coolant Loop (D) and its assogiated steam generat:r

~

4.

and reactor coolant pump.

At least one of the above ecolant loops shall be in operation.'

b.

AFPLICA3fLITY: MOCE 3 ACTION:

With less than the above required reactor ecolant loops OPERABLE, restore the required loops to OPERASLE status within 72 h urs or be a.

in HOT SHU7DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , With no reactor coolant loop in operation, suspend all cperations involving a reduction in boren concentration of the Reactor Coolant b.

". System and immediately initiate corrective action to ret 0rn tha l

required coolant 1 cop to operation.

SURVEILLANCE CEOUIREMENTS At leas.t the above required reactor coolant ;umes, if not in c;eration, shall be determined to be OPERASLE once per 7 days by verifying 3.4.1.2.1 I

c:rrect breaker alignments and indicated power availability.

At least cre c: cling 1cep shall be verified to be in c;trati:n a.a.2.2.2 and circulating reactor coolant at least once ;er 12 hears.

-All reacter c:clant ;.mps may be da-energized ~for up to 15:ar pr,vited (1) no 19 j

c;aratict.s 2re pt-mit:2d that s::14 cause dilutien of the 144st 10;F telcw saturaticn te.perature.

W-STS c

REACTCR COOLANT SYSTEM SHUTOC'.iN LIMITING CONDITION FOR OPERATION At least two of the coolant loeps listed below shall be CPERA3LE:

3.4.1.3 a.

1.

Reactor Coolant Loop (A) and its associated steam generator and reactor coolant pump,*

2.

F.eactor Coolant Loop (B) and its associated steam generator and reactor coolant pump,*

3.

Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump,*

4.

Reactor Coolant Loop (D) and its associated steam generator and reactor coolant pump,*

i 5.

Residual Heat Removal Loop (A),"

6.

Residual Heat Removal Loop (B)."

b.

At least one of the above coolant loops shall be in operation."*

APPLICABILITY: MCDES 4 and 5 ACTION:

With less than the above required loops OPERASLE, immediately a.

initiate corrective action to return the required loops to OPERABLE status as soon as pcssible; be in COLD SiiUTDCWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.

With no coolant loop in operation, suspend all cperations involving a reduction in boron cencentration of the Reactor Coolant System and immediately initiate corrective acticn to return the required coolant loop to operation.

!I

'A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equhl to (275)0F unless 1) the pressuri:er water

,i cubic feet or 2) the secondary water temperature of volume is less than each steam generator is less than 0F above each of the RCS cold leg temperatures.

"The normal or emergency power scurce may te ineperable in MC;E 5.

'"All reacter coclsnt pumps and decay heat re. oval pum;s may be d: reagi:sd for up to 1. cur provided 1) no operatiens are pem.itted that :.c. d cause S

dilu f en of ee reactor coolant system 5:r:n cer.:entration and :) ;:re Outlet tam;erature is maintained at least 10 F teicw saturation ti.;iriture.

0

1-575 i

e

T RE*CTOR COOLANT SYSTEM _

SURVEILLANCE REQUIDEMENTS The required residual heat removal loop (s) shall be determined CFERABLE 4.4.1.3.1 per Specification 4.0.5.

The required reactor coolant pump (s), if not in c;eratien, shall be determined to be OPERABLE once per 7 days by verifying ccrrect breaker sli;n.en 4.4.1.3.2 and indicated pewer availability.

The required steam generator (s) shall be determined CFEEASLE by

): at least verifying secondary side level to be greater than or equal to (

4.4'.1.3.3 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

~

At least ene ecolant Icep shall be verified to be in c;eratien and 4.4.1.3.4 circulating reacter coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i e

o S

1 1

i I

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REFUELIN3 OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOY AL AND COOL ANT CIRCUL A710N ALL WATER LEVELS LIMITI.*G CCNDITICN FCR CPEP.ATION

' O.&l At least ene residual heat remcval (RHR) Icep shall be in cperatien.

AP P L i c e o ! '. ,-

MODE 6 ACTION:

With less than ene is

..1 heat removal loop in operation.

a.

except as provided in b te: w. euspend all cperatiens involving an increase in the reactor decay r,s.r.*-ad 'or a reduction in beren cencentration of the Reacter Ccolant Q, '.

C1cse all centainc.ent penetrations providing direct access frem ;"' rantainment atmosphere to the cu: side atr. sphere within 4, hours, b.

The residual heat removal loop may be removed frem operat;:a

e up to I hour per 8 hcur period during the performance of CCRE ALTERAi.'..:

in the vicinity of the reactor pressure vessel (het) legs.

The provisions of Specification 3.0.3 are not applicable.

c.

SURVEILLANCE REOUIREMENTS a.9.&1 At least ene residual heat renoval 1 cop shall be verified to te in cperation and circulating reacter ceclant at a flew rate of greater tSan er equal to (23C0) gpm at least ence per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

e E-575

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.E UE'.:!:3 G7E:ATI CNS_

2/1.9.3 REi-;UAL wEAT t! OVAL AND CCCL !iT CIRC'JL aTICr{

L L ~':TER LE'IEL S

:v !T:':3 C::;DITICN T R CPEF.AT:CN

3. 7. E.1 1: icas: ene residual heat rem: val (RNR) loep snail te 'n ::erati:n.

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L

CA37tiTY

MCCE 6 CTICN:

With

ess than one residual 5 eat removal lecp in ::erati:n, a.

as ;revided in b below, suspend all c:eratiers inv:1ving an exce::

increase in the reactor decay heat load 'er a reduction 4 9 ter:n Cicse til ::m:afe. men:

c:ncen:ratien of the Reactor Coolant System.

penetratiens providing direct access frem the c:ntai-ment a:m:s;te-e

the Out:1de attes;here within 4. hours.

r The residual heat removal icep,may be removet fr:m ::erett:n f:r to to b.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months rer 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of C RE ALTER'TIONS in the vicinity of the react:r ;ressure vessel (het) legs.

The provisi:ns of 5;ecification 3.0.3 are net a:oit:atie.

c.

/R'!!!L*.J..'CI *ECUitt:975 i

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J. 9. 2.1 at leas

ne residual heat re. oval loop shall be veriff ed to be in

ertti:n and cir:ulati g reac:cr c:olant.at a fic < -ate of grea:er *an Or et.a* to (23CO) ;;m at itsst :nce per a 5:urs.

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e REFL'EL1!:S OPERATIONS _

LCW WATER LEVEL LIMITING CONDITION FOR OFERAT!0N.

Two independent Residual Heat Removal (RHR) loops shall be 0FERA3tE.*

3.9.8.2 MODE 6 when the water level above the top of the irradiated APPLICABILITY:

fuel assemblies seated within the reactor pressure vessel is less than 23 feet.

ACTICN:

With less than the required RHR lo:ps OPERABLE, i ediately initiate corrective action to return the required RHR iceps to CFERABLE a.

status as soon as possible.

The provisions of Specification 3.0.3 are not a;plicable.

.i b.

I i

9 SURVE!LL'NCE RECUIREMENTS I

I The required Residual 1: eat Renoval loops shall be determined CFEFABLE l

4.9.; -

per Spec.'. tion 4.0.5.

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ir.e normal or emeriency ;ov.tr scurce may be ir.:;eratie f:r es:h R..

no.

E-STS

)

l 3/4.4 REACTOR COOLANT SYSTEM 1

BASES REACTOR COOLANT LOOPS AND CCOLANT CIRCULATION 3/4.4.1 The plant is designed to operate with all reactor c:clant loeps in operat and malatain DN3R abcve 1.30 during all normal operations and anticipated In MCOE', I and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANC5Y within transients.

I hour.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability fcr removir.g decay heat; hewever, single failure consideratiens

~

require that two 1ceps be OPERABLE.

In MODES 4 and 5, a single reacter coolant loop or RHR loop provides suf ficient heat removal capability for removing decay heat; but single failure Thus, if the considerations require that at least two loops be OPERABLE.

reacter :colant 1 cops are not OPERABLE, this specification reRufres two RHR loops to be OPERABLE.

flow to ens.sre mixing, prevent stratification and prcduc changes during boren concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boren reduction will, theref:re, be l

within 3he capability of operator re:cgnition and control.

with one or The restrictions on starting a Reactor Coolant Pump more RCS cold legs less than or equal to (275)0F are provided to prevent RC5 pressure transients, caused by energy additions from the secondary system, i

The RC5 will which could exceed the limits of Appendix G to 10 CFR Part 50.

be pr:tected against overpressure transients and will not exceed the limits of A;pendix G by either (1) restricting the water volume in the pressurizer and 1

restricting starting of the RCPs to when the saccr.dary each steam generator is less than (

)oF above each of the RCS cold leg tec.p era tures.

'f-575

n o

FIFUILING OPERATIONS E'5ES 3/A.9.8 ESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement t' hat at least one residual heat removel (RHR) loop be in operation ensures that (1) sufficient ecoling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel belcw 140 f as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boren dilution incident and prevent boren stratification.

The -equirement to have two RHR f oeps OPERAELE when there is less than 23 feet of water above the core ensures that a single failure of the cperating AMR Icon will not result in a complete loss of residual heat remeval capability.

With the ruactor vessel head remaved and 23 feet of water above the core, a large heat sink is available for core cooling.

Thus, in the event of a f ailure of the operating RHR loo ~p, adequate time is provided to initiate emergency procadores to cool the core.

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f-STS j

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ENCLOSURE 4

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-[

gg 2 1983 1

MEMORANDUM FOR:

Edward J. Butcher, Chief Technical Specification Branch Division of Operational Events Assessment i

FROM:

Faust Rosa, Chief Electrical Systems Branch Division of Engineering & Systems Technology

SUBJECT:

REVIEW 0FNEWSTANDARDTECHNICALSPECIFICATIONS(STS)

References:

1.

Memorandum from C. E. Rossi to L. Shao (and others),

this subject, dated November 23, 1968.

l 2.

Memorandum from F. Rosa to E. Butcher, this subject, dated December 27, 1988.

3.

Menorandum from F. Rosa to M. Virgilio concerning Review of Fermi 2 Unit 2 Technical Specification dated January 6,1989.

/

i As a result of our recent acceptance of the industry owners group proposed new Standard Technical Specification (STS) definition for Operable-Operability (Ref. 2) and our subsequent recommendation that a different definition be considered in the development of STS (Ref. 3), Millara Wohl of your staff requested clarifica-tion as to our position on the definition that should be included in the new STS.

l The definition for Operable-Operability as stated in the current STS for GE, CE, and B&W is as follows:

r OPERABLE - OPERABILITY i

A system, subsystem, train, component or device shall be OPERABLE-or have OPERABILITY when it is capable of perC. ming its specified l

function (s),andwhenallnecessaryattendantinstrumentation.

controls, normal and emergency electric power sources, cooling or seal water, lubricetion or other auxiliary equipment that are i

required for the system, subsystem, train, component, or device to perform its function (s) are also capable of perfoming their related supportfunction(s).

Contact:

J. Knor. SELB/ DEST X23285 t

i r

E. Butcher ThedefinitionstatedinthecurrentSTSforEisasfollows:

i OPERABLE. OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the s stem, subsystem, train, component, or device to perform its function (s are also capable of performing their related support function ( ).

In regard to the terminology "...all necessary attendant...norval and emergency electric power sources..." and "all attendant... electric power," underlined in the above definitions, the industry owners group (Ref.1) proposed that the termi.

nology be changed to "...all necessary attendant... electrical power sources...".

In our review of Fermi Technical Specification (Ref. 3), we concluded that the terminology did not clearly convey that both offsite and onsite electric power are necessary for operability. We thus reconnended that the terminology be changed to "...all necessary attendant...offsite and emergency electrical power...".

Therefore, we recommend that this terminology be included in the new STS.

ee.!ri-;1313:220Y' 7:u. : : : '

Faust Rosa, Chief Electrical Systems Branch Division of Engineering & Systems Technology cc:

C. E. Rossi Distribution:

L. Shao Central Files A. Thadani SELB Rdg M. Wohl J. Knox (PF)(2)

J. E. Knight F. Rosa IE PREVIOUS CONCURRENCE 25ELB/ DEST * :5C/5ELB/ DEST *:BC/5EM T:

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t E. Butcher,

The definition stated in he current STS for W is as fo ows:

OPERABLE.

ERABILITY A system, subsystem..tra n, component or devic shall be OPERABLE or have OPERABILITY when t is capable of per rming its specified function (s), and when all ecessary attendan instrumentation, controls, electrical power, cooling or seal ater, lubrication or l

cther auxiliary equipment th t are require for the s stem, subsystem, train, component, or device t perform it function (s are also capable of perfoming their re sted supp tfunction().

In regard to the terminology "...all ecess y attendant... normal and emergency electric power sources..." and "all a and t... electric power," underlined ',i the above definitions, the industry ow rs group (Ref. 1)' proposed that the termi.

nology be changed to "...all necessary t endant... electrical power sources...".

In our review of Fermi Technical Specif1 tion (Ref. 3), we concluded that the teminology did not clearly convey that

'th offsite and onsite electric power are necessary for operability. We thu re ommended that the terminology be changed to "...all necessary attendan...o site and emergency electrical power...".

Based on the above considerations, is the LB position that the definition proposed by the industry owners gr for Oper le.0perability in the new STS is not clear as to what is mean b pecessary e etric power sources. The steaning of the term "necessary"(p eils further el rification.We reconnend that the industry owners groups be re ;e Lted to provi this clarification, j

k Faust Rosa, Chief Electrical Systems ranch Division of Enginee ing & Systems Technology i

cc:

C. E. Rossi Distribution:

L. Shao Central Files SELBRdg(PF)(2)

A. Thadani J. Knox M. Wohl J. E. Knight F. Rosa ILB/ DEST

50/5ELB/ DEST :BC/5ELB/ DEST:

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5 ENCLOSURE 5

V r G > #1 35p5 ggy5 # #

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~g UNITED STATES

.P NUCLEAR REGULATORY COMMISSION o

3-

.j msmworou, o. c. nosss

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o 0..-g Docket No.:

50-341 JM MEMORANDUM FOR:

Martin J. Virgilio, Project Director Project Directorate 111-1

. Division of Reactor Projects III/IV/Y & Special Projects FROM:

Faust Rosa, Chief Electrical Systems Branch Division of Engineering & Systems Technology

SUBJECT:

FERMI UNIT 2 - TIA REVIEW OF TECHNICAL SPECIFICATIONS WHICH ALLOWED ELECTRICAL CONFIGURATION SUPPLYING POWER TO THE OPERABLE ECCS PUMPS IN COLD SHUTDOWN Plant Name:

Fermi Unit 2 Utility:

Detroit Edison Company Docket No.:

50-341 TAC No.:

71089 TIA Resp. Directorate: PD III-1/DRP Project Manager:

John Stang Review Branch:

SELB/ DEST Review Status:

Complete By Memo from W. Rogers, RI/RIII to R. Cooper, SC/RIII dated October 19, 1988, Region III (Under TIA) requested NRR to review Fermi Unit 2 Technical Speci-fication (T/S) 3.8.1.2 to determine ether it should be further clarified to indicate that the operable (not on ma ntenance) ECCS subsystem "be powered by the operable onsite A.C. electrical ~p wer source" in Modes 4 and 5.

(We inter-pret the quoted portion to mean that the operable onsite AC source should be available to the operable ECCS subsystems, not actually powering the systems.)

The Electrical Systems Branch (SELB) reviewed the referenced T/S provision for the operability requirement of A.C. electrical power with respect to the ECCS subsystems and concurs that the operable ECCS subsystems should have available /

an operable onsite A.C. electrical. power source (as well as an offsite source) 1 in all modes. We also agree that the T/S are not clear in this regard unless the Operable-Operability definition (T/S 1.25) is interpreted accordingly, as it should be. Therefore, we recomend that this definition be revised as follows.V (revisionisunderlined).

OPERABLE - OPERASILITY J'

.p

. 'W,<,

> * 'Q p.,r.M 1.25 A system, subsystem, train, component or device shall be

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OPERABLE or have OPERABILITY when it is capable of ier-f

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forming its specified function (s) and when all necessary attendant instrumentation, controls, offsite and emergency. p < g. g-

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Contact:

g. - to D. Tondi, SC/SELB/ DEST

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electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perfom its func-tion (s) are also capable of performing their related support function (s).

No other clarifications to the T/S are necessary.

In the Fermi 2 design, both 4.16Kv safety buses of both Division I and Division 11 can be connected to either of the two offsite power circuits. Therefore, T/S 3.5.2 and T/S 3.8.1.2 can be satisfied by any two of the four subsystems (2 Core Spray and 2 Low Pressure Coolant Injection) that are powered from either division, provided both safety buses of that division are connected to an operable offsite power circuit and their dedicated diesel generators are operable and available in event of loss of the offsite circuit. Also, T/S 3.5.2 does not preclude operation of either or both buses from their dedicated diesel generators provided an operable off-site circuit is available in event of loss of the diesel.

The revised definition (T/S 1.25) also applies for meeting the requirements of s

T/S 3.4.9.2.

That is, the shutdown cooling mode loops of the Residual Heat Removal System that are required to be operable must meet the same offsite/onsite power requirements cited in the above paragraph, regardless of whether credit is taken for an alternate decay heat removal method.

In sumary, we recomend revision of the definition of Operable-Operability as indicated above.

It is noted that this revision is already incorporated in some Technical Specifications, e.g., LaSalle and North Anna, 1

By copy of this memorandum this recommendation is also transmitted to the Technical Specifications Branch (E. Butcher) for their consideration in development of Standard Technical Specifications.

Faust Rosa, Chief Electrical Systems Branch Division of Engineering & Systems Technology cc:

L. Shao C. E. Rossi A. Thadant E. Butcher W.Rodgers(RIII)

J. Stang R. Cooper (R111)

T. Quay

2

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NUCLEAR REGULATORY COMMissl0N

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DCT 3 7 ggg 7#M 7/069 MEMORANDUM FOR:

M. J. Virgilio Acting Assistant Director for Regions III and V. NRR FROM:

E. G. Greenman, Director, Division of Reactor Projects Region III

SUBJECT:

REQUEST FOR TECHNICAL ASSISTANCE - TECHNICAL SPECIFICATION INTERPRETATIONS - (AIT #0384)

I During inspection activities at the Fermi 2 site, the resident inspectors have encountered several issues associated with Technical Specifications interpretations for which we are requesting NRR input. To a limited extent, these issues have been discussed previously between the resident staff and NRR staff members. The 4 tenet are fully diteutted in the attachmente t_n thft semorandum and include nuestions en the emerability of RHR/LPCI numes. lack of use and procedural 12ation of backup manual scram breakers by the licensee, and a deficiency in the Technical Specifications that does not require operable ECCS pumps to be powered by the operable A.C. electrical power source while in cold shutdown.

Your efforts and clarification will be appreciated. If there are any

~

2, at (313)please contact Mr. W. G. Rogers, Senior Resident Inspe questions, 586-2798.

Edward G. Greenman, Director i

Division of Reactor Projects i

Attachments.

1.

Memorande, W. Rogers to R. Cooper of 10/19/88, Operability of RHR/LPCI Pumps in Cold shutdown 1

(w/ attachments 1-10) 2.

Memorandum, W. Rogers to R. Cooper of 10/19/88 Use of the Backup Manual cramBreaker(w/ attachment 1)

Memorandum, W. Rogers to R. Cooper of 10/19/88, Technical Specification Allowed Electrical Configuration supplying Power to the Operable ECCS Pumps in Cold Shutdown See Attached Distribution

L.

a a

i Distribution 2

ggy g 9 ggg cc w/ attachments:

i G. Holahan, NRR T. Quay, NRR F J. Stang. NRR H. Miller, RIII W. Rogers, SRI Fermi 9

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~T9t C. *l - ; l C 69 UnitTEo STATES q

.m ase NUCLcAR REGULATORY COMMIS$10N nacione m 3

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sos = sLLyn, stumois sesat OCT 191909 MEMORANDUM FOR: Rfchard Cooper, Section Chief FROM:

Walt Rogers, Senior Resident Inspector

$UBJECT:

TECHNICAL SPECIFICATION ALLOWED L1ECTRICAL CONFIGURATION SUPPLYING POWER TO THE OPERABLE Etis PUMP 5 IN COLD SHUT 00WN

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, e y ?. 4.s-r.s The Technica157cification for Fermi 2 requires certein equipment while in cold shutdown. Technical Specification 3.5.2 requires two low pressure ECCS sykjysteir(.to be operable in Modes 4 and 5.

This allows 2 core spray subsyste :s, 2 LPCI subsystems or 1 core spray /1 LPCI subsystem to be operable and comply with Technical Specification 3.5.2.

Technical specification 3.8.1.2 requires one division of onsite A.C. electrical power operable in Modes 4 and 5.

However, the Technical Specifications do not require that at least one ECCS subsystem be powered by the operable onsite A.C. electrical power source.

In Inspection Report 87031 I identified that the licensee did in fact have this configuration. The situation was brought to license senior management attention, who indicated that they would try to minimize such a configuration.

I pursued this matter with NRR (Marty Virgillo) who indicated that this was an oversight in the Technical Specifications.

I consider the onsite A.C. oower sourc6 Technical Specification deftetent for Modes 4 and 5 if it does not provide power to at least ans suhavstem ef I recommend that the NRC anoressivolv nursue a Technten1 taartfientian ECCs.

change to reevire such a tie between the ECCS and the antite A r electeiral newer source.

Finally. T think the same consideration should be niven RHR shutdown cooline q

licensee is esercisina the action statement'allew;d ur.dar when the Technical Boecification 3.4.9.2 to have alternate decaw heat r;;.sval methods.

Specifical y. when an alternate decay heat removal method is taken credit for the other normal RHR shutdown cooling system should be powered from an operable onsite A.C. power source.

Sincerely,

((4

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Walt Rogers

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4 ENCLOSURE 6 9

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3/4.0(9)

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A. 5 i0 120 MEMORANDUM FOR: Charles E. Norelius, Director Division of Project and Resident Programs FROM:

Darrell G. Eisenhut, Director Division of Licensing

SUBJECT:

WRR POSITION ON COMPONENT OPERABILITY WHEN A DIESEL GENERATOR IS INOPERABLE

REFERENCE:

Memo from C. E. Morelius to D. G. Eisenhut, dated February 16, 1983;

Subject:

Request for Technical Assistance - Technical Specification Interpretation".

4 Your memorandum to me dated February'16,1983 (see reference) requested an interpretation by NRR on the subject of operability. The requested inter-pretation was whether the loss of emergency power to a system would render that system inoperable for the purpose of satisfying another system LCO.

Your memorandum included a specific example dealing with the core spray system and the high pressure coolant injection systes at the Duane Arnold f acility.

It is our position that, in general, a systes may be considered operable for the purpose of satisfying its own LCO and that of another system if only its emergency power supply is inoperable. This position assumes that all the provisiens of Technical Specification 3.0.5 in Enclosure 1 of my April 10, 1980 letter to All. Power Reactor Licensees are also satisfied, i.e., a system may be considered operable for the purpose cf satisfying its applicable LCO when its emergency power source is inoperable provided the systas's corresponding nomal power source is operable, and its redundant train is also operable. These provisions have been incorporated into the Duane Arnold Technical Specifications as a clarification to the definitiori of Limitin!1 Conditions for Operation.

We realize that this position may result in a plant not being capable of fully satisfying the single failure criterion while operating in the degraded mode. However, we consider such operation to be acceptable since it would be of limited duration and the pronability of an accident occurring with a concurrent f ailure of the remaining operable l

system is remote.

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Contact:

D. Brinkr.an, x24707 i

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ith the Cort Spray Syster

r. ycur memoranout. you specifically astec:

cegraceo by loss cf its emergency oower source, is the Core Spray Syster to oe consicereo coeraole tc meet tne Hign Fressure Coolant Injection System LCO?" Duane Arneio Tecnnical Specification 3.5.0.2 is applicaole to tnis example; it permits reactor operation to continue for up to seven days providing that during such seven days all active components of the ADS subsystem, the RCIC system, and LPCI subsystem and both core spray subsystems are operable.

In accordance with our position, both core spray subsystans would be considered operable.

It should be noted, however, that our position is not intended to supersede the provisions of any technical specification which specifically requires the operability of diesel generators. For example, Duane Arnold Technical Specification 3.5.A.2 permits reactor operation to continue for up to seven days with one core spray subsystes inoperable provided the other core spray subsystem, the active components of the LPCI subsystem and the diesel generators are operable. Therefore, if one core spray subsystem and one diesel generator were inoperable, our position would not be applicable and continued operation would not be acceptable since Technical Specification 3.5.A.2 specifically requires the diesel generators to be op erable.-

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. Eisenhut, t'irector

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Division of Licensing P

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UNITED STATss h wn J'

NUCLEAR REGULATORY COMMIS$10N o

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e, sLqw ELLym. Luseoas ents?

FEB 161983 i

D. G. Eisenhut Director, Division of Licensing, NRR MEMORANDUM FOR:

C. E. Norelius, Director Division of Project and FROM:

Resident Programs TECHNICAL REQUEST FOR TECHNICAL ASSISTANCE

SUBJECT:

SPECIFICATION INTERPRETATION (AITS F03008283) of our Senior Resident inspectors requesting Attached is a memorandum fr.cm or.:

a Technical Specification interpretation by NRR regarding the subject of.

The purpose of ry memorandum is to request that interpretation.

operability.

In your letter dated April 10, 1980, to "All Power Reactors", all licensees were requested to submit Technical Specification changes to change the definition of operable to read:

"A system, subsystem, train, con @onent'or device shall be OPERA 8LE or have OPERABILITY when it is capable of perforuing Implicit in this definition shall be the its specified function (s).

assumption that all necessary attendant instrumentation, controls, dem and emercenev electrical power sources, cooling or seal water, lubrication or other auxiliary equipment snat are required for the system, subsystem, train, component or device to perform its function (s) are also capable of

\\

performing their related support function (s)".

"When a The definition of operable was further clarified as follows:

system, subsystem, train, component or device is determined to be inocerable solely because its emeroency oower source is inoperable, or solely because i

its norinal power source is inoperacle, it may be considered OPERABLE for

)

the puroese of satisfyino the recuirements of its applicable Limitinc (1) itt corresoondino normal or Concition for Operation, orovided:

all of its redundant systqMs),

and (2 emergency (oower sourcer is OPER/iBLE: subsystem s), train (sj, comp itkewise

)

satisfy the requirements of this specificatio '". (esphasis added)

It is very clear from the above that system, subsystem, train, component or device is not inoperable for the purpose of satisfying the requirements of its LCO if the' system, subsystem, etc., has merely lost its emergency power However, it is not clear to us whether the loss of emercenev cower source.

system. <nwevetem. ete_.,,

to a system, subsystem, etc., would rencer that LCO.

inoperable for tne purpose or satisfying another system, subsystem, etc.,

9

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i.

D. G. Eisenhut With the Core The attached memorandum addresses this specific example:

Spray Svstem deeraded bv loss of its emereeney power source, is the core _

Scray system to te consicerec operacie to meet tne nign rressure Coolant _

injection 5vstem LCOi We would appreciate a review of this issue by your staff and a response by arH 1 15.1983. Please contact Roger Walker of Iqy staff on FTS 384-2565 tf you nave any questions regarding this matter.

4.4.YYA C. E. Norelius. Director Division of Project and Resident Programs

Attachment:

As stated cc:

N. J. Chrissotimos, SRI Quad Cities Station 6

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March 22,1990 R P Mcdonald t ee, i.

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ELV-01469 0312 Docket No.

50-424 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C.

20555 Gentlemen:

V0GTLE ELECTRIC GENERATING PLANT WAJVER OF COMPLI ANCE 1his letter is a follow-up written request for a waiver of compliance for Georgia Power Company's Vogtle Electric Generating Plant Unit 1.

A one-time Technical Specification waiver of complianca was requested to make Technical Specification 3.0.4 not applicable to Techn.el Specification 3.8.1.2.

This waiver was required to allow entry into Mode 5 with the operability of Olesel Generator IA and its associated load sequencer unverified.

The Plant Review Board has reviewed and approved this waiver.

This waiver was necessary because recent failures of the Unit 1. Train A Diesel Generator and its associated load sequencer rendered their operability questionable. Even though an extensive investigation is being conducted, the specific cause of the failure of this equipment has not been identified.

The current Reactor Coolant System water level is at mid-loop.

This waiver allows tensioning of the Reactor Pressure Vessel head which also allows filling and venting of the Reactor Coolant System.

Filling and venting the RCS will result in an increase in RCS water inventory and make the steam generators available for heat removal should they be required. As additional compensatory action, on-shift operators have been briefed on the condittoil of the equipment and how to respond by manually initiating a Diesel Generator emergency start should it be required.

This is a one-time request due to present plant conditions and equipment status.

Since Technical Specification 3.8.1.2 action requirements are the same for Modes 5 and 6 the probability of occurrence and consequences of an accident are not The increased by this request and no significant safety hazards are involved.

additional inventory and addition of the steam generators as an available heat sink tmproves the margin of safety.

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licolp.i l'ot, t1 a.

U. S. Nuclear Regulatory Commission ELV-01469 Pace Two i

Since no change in plant design occurs as a result of this waiver of compliance and since the plant is being placed in an improved condition with respect to water inventory and available heat sinks, no adverse environmental effects are involved.

Sincerely, h

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R. P. Mcdonald RPH/HWM/gm xt: Georata Power Comoany Mr. C. K. McCoy Mr. G. Bockhold, Jr.

Mr. R. M. Odom Mr. P. D. Rushton NORMS U. S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. T. A. Reed, Licensing Project Manager, NRR Mr. R. F. Aiello Senior Resident Inspector, Vogtle l

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_ _ _ _,

ML20071M155

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