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Forwards Response to 840210 Request for Addl Info on 820907 Proposed License Amend to Change Tech Specs Re Containment Integrity & Closure of MSIVs
	ML20083P708
Person / Time
Site:	Brunswick
Issue date:	04/13/1984
From:	Cutter A; CAROLINA POWER & LIGHT CO.
To:	Vassallo D; Office of Nuclear Reactor Regulation
References
	NLS-84-169, NUDOCS 8404200128
	Download: ML20083P708 (13)
	v • d • e

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o CD&L Carolina Power & L:ght Company SERIAL: NLS-84-169 APR 131984 Director of Nuclear Reactor Regulation Attention: Mr. D. B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION MISCELLANEOUS TECHNICAL SPECIFICATION REVISIONS '

Dear Mr. Vassallo:

In your letter dated February 10, 1984, Carolina Power & Light Company (CP&L) was requested to provided additional information concerning our request for license amendment dated September 7,1982. Attached are CP&L's responses to the questions raised by your Staff.

Should you have any further questions, please contact a member of our Licensing Staff.

Youra.very t ly, -

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A. B. Cutter - Vice resident Nuclear Engineering &' Licensing MAT /ccc (9839 MAT)

Attachments cc: Mr. D. O. Myers (NRC-BSEP)

Mr. J. P. O'Reilly (NRC-RII)

Mr. M. Grotenhuis (NRC)

DR DO O __

411 Fayetteville Street

P. O. Box 1551 e Raleigh, N. C. 27602 L

ATTACIDfENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION MISCELLANEOUS TECHNICAL SPECIFICATION REVISIONS (9839 MAT /ccc)

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Enclosure 1 4

P.EOUEST FOR ADDITIONAL INFORMATION CAROLINA POWER & LIGHT C0"PANY BRUNSWICK UNITS 1 AND 2 nocKET f 05. Sn-3?5 ann 50-3?4 Re: CP&L P.eouest for License Amendment Dated Sectemoer 7,1022; Attacncents 7 and 8

'0 The oroposed changes to the technical soecifications given in attachment 7 deal with containment inteority and specifically with the relaxation of surveillance requirements for verifvino the closure of ecuipment hatches and for verifying the closure 'of valves and flanges in high radiation areas.

7.1 You have recuested the deletion of the recuirement for verifying that all eouipment hatches are closed and sealed at least once per 31 days.

State the. objective of this change. Demonstrate the means by which verification of sealing and closure will be verified and state the

frequency of the verification.

. 7.2 You have proposed that certain. valves and blind flances in high radi-i ation areas be verified to be closed during each COLD SHUTDnWN greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months (but not more often than once per 92 days). The current technical specifications recuire such verification at least once oer 31 days. Please state the objective of this change and give the safety considerations that provide a basis for relaxing this recuirement.

He nave previously considered this relaxation but it has not been granted.

i Snow that tnere is a particular need for this change at the Brunswick facilities.

1.0 Ir. at acnmen- 2, the orcoosed chance to the technical soecifications ceals with tne recuirecents for closure of the main steam isolation valves. The current soecifications reouire the coerability of inter-

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lecks that cause closure of these' valves upon occurrence of a low condenser vacuum when reactor steam pressure is greater than 500 osig.

4 The modified soecification would recuire closure of.the isolation valves I

' uenn occurrence of a low condenser vacuum at any reactor pressure but would permit bypassing this requirement when all turbine stoo valves ,

are closed. Bypassing the reouirement for closure o# the main steam e '-

isolation valves could allow the inadvertent pressurization of- the 5

turbine condenser via the ' turbine bypass valves and 'the consequential release of radioactive materials to the turbine building. The safety censiderations of this inadvertent nressurization have not been ad-dressed. Furthermore the ourpose of this change and its consecuences are not clear.

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8.1 State the objective of this change.

8.2 State the intended purpose of this bypass and discuss the procedures that would be established to control its use.

8.3 Consider the inadvertent use of this bypass including its consequences and the means for avoiding inadvertent bypassing.

8.4 Please provide clear diagrams of all isolation logic to be bypassed and show how reactor operation would be affected.

4

4 RESPONSE TO ITEM 7.1

. The requirement for verifying that all equipment hatches are closed and sealed at least once per 31 days is maintained by Technical Specification (TS) 4.6.1.1.a. which deals with Primary Containment Integrity surveillance requirements. CP&L believes that the statement "all penetrations" includes the equipment hatches. Therefore, reiteration of this requirement is not necessary.

RESPONSE TO ITEM 7.2 i The proposed change, allowing certain valves and blind flanges in high radiation areas to be verified closed during each. COLD SHUTDOWN of greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , was made with ALARA considerations in mind. Requiring this

surveillance to be performed once every 31 days unnecessarily exposes personnel to high radiation fields. Administrative controls are already in place which ensure that proper alignment is maintained. Valves and flanges in i four high radiation areas are affected by the proposed change. These areas '

. include: 1) the TIP Room; 2) the RWCU Penetration Room; 3) the MSIV Pit; and i 4) the Drywell Head Area.

i f Shield plugs prevent access to the MSIV Pit and the Drywell Head Area.

Procedures currently exist which require valve and flange. position verification prior to installation of the shield plugs for the Drywell Head Area. Administrative controls are in place which require the shield plugs for the MSIV Pit to be in place when the plant is in OPERATIONAL CONDITIONS 1, 2, i and 3. Valve and flange positions are verified during system lineup. . Access is limited by the shield plugs, thereby ensuring correct valve alignments.

Therefore, redundant verification of valve alignments every 31 days is not necessary.

The TIP Room contains the test connections between the MSIVs, the feedwater check valves, and the steamline drain valves all of which are locked in the closed position. Administrative controls ensure that the TIP. room remains locked. In addition, the valve position is verified prior to startup. These measures provide adequate assurance that the valves are positioned correctly yielding the 31 day surveillance requirement unnecessary.

Administrative controls ensure that the RWCU. Penetration Room remains locked at all times with a work permit containing a detailed job description required for entry. Valves requiring surveillance in this area are the LLRT test connections, the 0-Ring test connections,'and body drain valves. . Alignment of these valves is verified during each system lineup. These measures, taken together, ensure correct valve alignment, .thus the 31 days surveillance i

requirement is not necessary.

RESPONSE TO ITEM 8.1 The objective of this - change is to provide the plant with greater flexibility

.in allowing dif ferent plant conditions during ' reactor heatup. The change

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simply removes an administrative limit on maximum reactor pressure-allowed

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( during heatup without availability of main condenser vacuum.provided the l turbine.stop valves.are closed. Etis limit is not required for the.BWR/4' product line (see GE' SIL No.7107. dated October 31, 1974 - Attachment 2).-

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, I RESPONSE TO ITEM 8.2 The purpose of the vacuum bypass switch is to allow reactor heatup prior to obtaining condenser vacuum. Operating the bypass switch will override the low vacuum MSIV (Group 1) isolation signal only when the turbine stop valves are closed and the reactor mode switch is not in the RUN position. In all other conditions, operating the bypass switch will have no effect. The pressure limit c1 operating the bypass switch has been demonstrated as not required.

Once this limit has been removed from the TS and operating manual, no additional procedures will need to be established to control its use because existing control logic will prevent inadvertent use when the condenser is

! vulnerable to accidental pressurization.

RESPONSE TO ITEM 8.3 Inadvertent use of the low vacuum bypass switches is prevented by interlocking with Reactor Protection System indications of plant mode and turbine stop ,

valve positions. Accidental pressurization of the condenser via the turbine bypass cannot occur because the turbine control logic will lock the bypass valves shut on low condenser vacuum signal sensed and actuated at the same setpoint as the TS instrument B21-PT-N056 (Condenser Vacuum - Low Pressure

'- Transmitter). Operator action or decision does not enter into the condenser protection scheme. The operator only decides whether to pressurize beyond the MSIVs to the turbine stop valves. In either situation the condenser is 4

protected from accidental pressurization and the resulting release of radioactive materials by automatic action of the reactor and turbine control functions. The main steam lines from the MSlVs to the turbine stop valves are protected from overpressurization by the safety relief valves. Relief valve settings are maintained below the design pressure rating of these steam lines.

RESPONSE TO ITEM 8.4 See Attachment 3.

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- SIL No. 107 October 31, 1974 .. ,

Category 3 File Tab "A" INCREASING FLEXIBILITY OF REACTOR STARTUPS Most current BWR technical specifications require a reactor scram if vessel pressure exceeds 600 psi.with the reactor mode switch in startup and the Main Steam Isolation Valves (MSIVs) closed. This current scram logic is the result of experience gained during the startup of an earlier GE/BWR in 1966 when, at that time, operators found it was difficult to control reactor power above approximately 600 psi without pressure control.

DISCUSSION '

There has recently been considerable interest shown in the capability to heat up a reactor in a " bottled-up" condition to rated pressure. The advantages of this capability are:

1. To be able to heat the reactor vessel and internals without running the feed pumps.

2. To be able to heat the reactor vessel in parallel with maintenance work being performed on the main turbine and condenser. ,

Item I should be of particular inter.est to those GE/BWRs with turbine-driven feed pumps, and Item 2 provides a higher degrt.e of flexibility in operation and maintenance. -- -

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,s In response to this interest, a special test was written and performed as part of the startup program at a current GE/BWR 4 design. The purpose of this test was to detect any marginal stability parameters which may exist in a idore current modal BWR at rat'ed pressure with the MSIV closed. .

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The documented test results indicate that sequential heating of the reactor vessel, main steam lines, and main turbine are within the capability of the '

plant tested. That is, the test results indica. e that the condition experi-t enced at the older GE/BWR was not encountered at the startup of the current design plant. .

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" Bottled-up" operation is a mode that can provide additional flexibility.

This mode of operation should be considered as an addition to (but not as a l

replacement for) the more conventional pressure aHd steamflow heat-up with l

the MSIVs open as well as closed. MSIV closure scram logic is therefore only

i necessary when the reactor is in the RUN mode.

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NUCLE AR ENERGY DIVislONS

BWR SERVICES e SAN JOSE. LIFO'R,NIA 95125 No WARRANTY OR REPRESENTATION EXPRESSED on IMPLIED 13 MADE WITH .

RESPECT To THE ACCUllACY. COMPLETENESS oR USEFULNEsS of THIS INFoR.

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SIL No. 107 &

October 31, 1974 , ,

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REC 0$ MENDED ACTION Due to specific differences between various design generations (BWR/1, BWR/2, BWR/3 and BWR/4), we are listing two distinct categories of action.

For BWR/1, 2 and_3_'lants . .,

Any utility with a plant in this group who wishes co further' explore the possibility of " bottled-up" ~startups should contact its normal General Electric service representative for a quotation concerning ,

engineering analysis and testirg./ ,

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For BWR/4 Plants t

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. We recommend for any utilityiwitn a, plant in this group.the remo'.21 of that* portion of the reactor protection system logic which presently requires a sc am 1.f the reactor pressure exceeds a set point (commonly 600 psig', when the reactor mode switch is in "startup" and the MSIVs are closed. L, test performed at the current design plant (a BWR/4) has proven that tr.is logic is no longer required on this product Mne.

- It should be noted that if this recommendation is, imp'lemented, 3 tech spec change _f,E required. . .

We would request that if this change 'is incorporated, General, Electric be notified so that the . design specs and transient data sheets can be updated.

GE will be pleased to pr. ovide engineering assistance fo'r implementing this change and modifying plant tech specs. Your normal _GE service r.epresentative -

should be contacted fo:' a quotation. .

7 Prepared by: L. F. Karan s ',

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Text

Dear Mr. Vassallo:

Reference:

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