ML20080J060

Application for Amend to License DPR-53,reflecting Mod to Hydrogen Purge Sys Outlet Isolation Valves MOV-6900 & MOV- 6901,consisting of Addition of Safety Injection Actuation Signal to Valve operators.Marked-up Tech Specs Encl

ML20080J060
Person / Time
Site:	Calvert Cliffs
Issue date:	09/20/1983
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	John Miller Office of Nuclear Reactor Regulation
References
NUDOCS 8309260154
Download: ML20080J060 (9)

Text

BALTI M ORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL, JR.

vict PRESIDENT SUPPLY September 20,1983 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: Mr. 3. R. Miller, C aef Operating Reacto, s Branch #3 Division of Licens: ag

SUBJECT:

Calvert Clifis Nuclear Power Plant Unit No.1, Docket No. 50-317 Request for Amendment Gentlemen:

The Baltimore Gas and Electric Company hereby requests an Amendment to its Operating License No. DPR-53 for Calvert Cliffs Unit No.1, to reflect a planned i

modification to the hydrogen purge system outlet isolation valves MOV-6900 and MOV-6901. This modification consists of the addition of a Safety Injection Actuation Signal (SIAS) to the valve operators.

i Pursuant to the procedural requirements of 10 CFR 50 paragraphs 50.59, 50.91 and 50.92, this letter forwards our notification of the proposed license amendment and provides a determination with regard to significant hazards considerations. In addition, with regard to operation of the modified portion of the system in MODES 1 through 4, a preliminary determination on significant hazards is submitted.

l More detailed information based on the safety analyses being performed by our architechtural engineer will be forwarded in a supplement to this request at a later date, l

and will allow a final determination of significant hazards for operation in MODES 1 l through 4.

Our present intention is to perform the physical modifications during the upcoming Unit I refueling outage beginning October 1,1983. We plan to modify the valves and perform functional testing that will permit opening and shutting the valves while in MODES 5 and

6. The valves will be locked closed and administratively maintained in the closed position prior to entry into MODE 4, following completion of the outage. This condition will be maintained by proposed Technical Specification 3/4.6.1.8 during MODES 1 through 4 until receipt of the approved Technical Specifications.

8309260154 830920 PDR ADOCK 05000317 P PDR (,C,{

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r Mr. 3. R. Miller September 20,1983 Page 2 PROPOSED CHANGES (BG&E FCR 83-104)

1. Remove old page 3/4 6-23 and replace with marked-up page 3/4 6-23 incorporating the changes, which consist of the addition of "SIAS" under the column entitled

" isolation channels", a change in the name of the function to " containment vent isolation" and the insertion of " < 20" for isolation time for both MOV-6900 and MOV-6901.

2. Add a new Technical Specification 3/4.9.14 consisting of a limiting condition for operation, surveillance requirements, and Technical Specification bases as shown on the attached proposed Technical Specification pages. This additional Technical Specification assures that the containment vent isolation valves are maintained shut during core alterations or movement of irradiated fuel within the containment.

3. Add new Technical Specification 3/4.6.1.8 as shown on the attached proposed Technical Specification page. This Technical Specification assures that the isolation valves will be maintained shut in MODES 1-4 until receipt of the final NRC Safety Evaluation regarding operational use of the containment vent system.

DISCUSSION AND DETERMINATION OF SIGNIFICANT HAZARDS During normal operation the containment must occasionally be vented to maintain containment internal pressure below the Technical Specification upper limit of 1.8 psig.

Presently, and in the past, this has been performed by venting the containment atmosphere through the containment normal sump isolation valves which exhaust to the Emergency Core Cooling System (ECCS) pump room sump, and subsequently to the environment via the ECCS pump room ventilation system exhaust filter train. While this method of venting the containment has been acceptable in the past and is carefully controlled and monitored, an operationally easier and more useful method is possible through the use of a portion of the existing Hydrogen Purge system. The portion of this system to be used as a vent path consists of a four inch diameter hydrogen purge outlet line, which terminates in the penetration room in the main plant vent piping upstream of the penetration room exhaust filter train.

The function of Post Loss of Coolant Accident (LOCA) hydrogen control is more than adequately performed by the installed, redundant, safety grade hydrogen recombiners.

We, therefore, intend to delete reference to the hydrogen purge system in the Updated Final Safety Analysis Report (FSAR), and install a blind flange on the air replenishment

( portion of the system (penetration 48b). The new function of the H2Purge outlet as a

containment vent system will be addressed in the next revision to the FSAR.

l Presently, the hydrogen purge outlet containment isolation valves (MOV-6900 and i MOV-6901) are maintained closed in accordance with Technical Specifications to prevent unnecessary exposure of the public if an accident resulting in airborne radioactivity release were to occtm. b order to allow venting of the containment through these valves during MODES 1 through 4, yet still provide the necessary protection to the public in the event of an accident, the physical modification will provide for closure of these valves on a Safety injection Actuation Signal (SIAS).

k

Mr. 3. R. Miller September 20,1983 Page 3 Normally, during power operations, these valves will remain shut and only be opened for brief periods of time to allow venting. Initiation of SlAS from either of two diverse parameters (containment pressure high or pressurizer pressure low) would close the isolation valves within 20 seconds, ensuring that the site boundary doses are maintained within a small fraction of the limits of 10 CFR 100.

The proposed Technical Specification changes would then, update Table 3.6-1 to reflect the addition of an SIAS closure signal to the hydrogen purge outlet isolation valves (to be renamed containment vent isolation valves) and provide limitations to prevent their use during MODES 1-4, or during core alterations or movement of irradiated fuel within the -

containment. These restrictions on operation of the containment vent valves preclude any need for installation of a Containment Radiation Signal (CRS) closure feature or for any safety analysis regarding postulated accidents during core alterations or fuel handling operations. To meet licensing requirements the above restrictions also ensure the isolation valves will be maintained shut in MODES 1-4, until receipt of the final NRC Safety Evaluation regarding operational use of the Containment Vent System.

Based on current review of the presently analyzed accidents in the Updated Final Safety Analysis Report, we find this proposed change does not involve a significant increase in the probability or consequences of these accidents, nor does it create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change does not involve a significant reduction in the margin of safety as defined in the Technical Specification bases. With additional restrictions imposed by proposed Technical Specification 3/4.6.1.8, the proposed change falls under the examples of amendments not likely to involve a Significant Hazards Consideration, as listed in Federal Register Volume 48, No. 67 dated Wednesday, April 6,1983. We conclude that the proposed change does not involve a Significant Hazards Consideration.

A preliminary evaluation based on the above review of Chapter 14 of the updated Final Safety Analysis Report has yielded a determination that operation of the Containment Vent System in MODES 1-4 does not involve a Significant Hazards Consideration. This evaluation will be supplemented following our review of the safety analysis currently performed by our Architectural Engineer. We will forward the results of our review in a final determination which supplements our submittal.

SAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Safety and Off-Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public. The forthcoming supplemental detailed safety analysis will be submitted following additional safety committee reviews.

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( 1 1

l Mr. J. R. Miller September 20,1983 Page 4 FEE DETERMINATION We request that you consider this change request as a supplement to the Unit I re-load application dated August 22,1983.

We have determined, pursuant to 10 CFR Part 170, paragraph 170.22, that this Amendment request consists of a Class III amendment for Calvert Cliffs Unit No.1.

Accordingly, we are including Baltimore Gas and Electric Company Check No. B297515 in the amount of $4,000.00 to cover the fee for this request.

Very truly yours, x w AEL/ JET /sjb 3

STATE OF MARYLAND :

TO WIT:

CITY OF BALTIMORE :

Arthur E. Lundvall, Jr., being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge,information, and belief; and that he was authorized to provide the response on behalf of said Corporation.

WITNESS my Hand and Notarial Seal:

Y/[e Notary n// bPublic? a dw' My Commission Expires ht)/[</f[b if f '

cc: J. A. Biddison, Jr., Esquire G. F. Trowbridge, Esquire D. H. Jaff e, NRC R. E. Architzel, NRC R. E. Corcoran, Chief (DHMH)

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ALTIMORE GAS AND ELECTRIC COMPANY .e.:o coy 93645-01

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300 , ,,o m.e'er ,a r.mcr ast m:E" . 6ce cits' 0:ssdusi - ~ nor Avount 3U9082 09 12 83 $4,000.00 C1 css 3 Technical Specification Change Unit 1 Table 3.6-1 adds SIAS signal to hydrogen purge valves' i

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09 13 83 TOTALS >', $4,000.00 DETACH THIS PORTION BEFORE DEPOSiflNG CHECK

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$20 Es: BALTIMORE GAS AND ELECTRIC COMPANY BALTIMORE, MARYLAND 21203

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. PAY EXACTLY **4.000.00** $4.000.00** B297515.

US NUCLEAR REGULATORY **

TO THE COMMISSION **

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CONTAINilENT ISOLATION VALVES .

ISOLMION VALVE.  ! W ATION PCHEIRAil0N ISOLATION IDENTIFICATION NO. FUltCTION_ TiHF fSEC0 fibs)  %

7 NO. CliANNEL _

/

3

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230-1 Fire Protection NA

. 44 NA flA NA 238-1 ,

y HA /

r. NA H0V-6200 *

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I Ilydrogen Sample Outlet fiA ,-

ttA SV-6540A 47A

SV-6507A NA /

NA '

/

Ilydrogen Sample Outlet HA 4 470 NA SV-6540E / 9 NA O HA SV-6507E s g a . a

m. 4 SV-6540F Hydrogen Sample Outlet tiA (I ~

U 47C NA HA '.

tiA SV-6507f '

liydrogen Sample Return fiA 470 flA SV-6540G NA SV-6507G HA /

/

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kradefed heI witAin the c oNlasvi,1 est.

ItC TION : With one or inste e.w/aasnrest vent .isclofica val.es open, 41ned6kly suspend s// dpetaMas s !vokayi cwi

/1LTERAT/ oms ee snovem ea t of irradiiz1cd be/ vidah /Ae.

cow briartext.

snMEllLANCf AfaulM/1ENTS

4. 9, /Y 7he cashinctes t vent isolo Ma veNes sAall be de kenixed fa be closed us1hok '72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.s p nbr h Me duer coa E chref of sad a.t er le<rf cacee$per m,, gest 7 days hrsodtl ko!l u e l jw$ n'a 4lt e.

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r REPJELING OPERATIONS I 3AsEs 3/4.9.10 and 3/4.9.11 '.'ATER LEVEL-REACTOR VESSEL AND SPENT PJEL 2 COL 2ATER LE'/EL The restrictions en minimum water level ensure that sufficient water depth is available to rencve 99". of the assumed 100 icdine sac activity released frca the rupture of an irradiated fuct assembly.

The minimum water depth is consistent with the assumptions of the accident analysis.

3/a.9.12 SPENT FUEL PCOL VENTILATION SYSTEM The limitations on the spent fuel pool ventilation system ensure /

thatbeall will radioactive filtered inrougn the material released HEPA filters frcm an and charcoal irradiated adsorber i

cr ce ~ fuel asse The OPERASILITY of this system and tne discharge to the atmcsphere. c' resulting iodine removal capacity are consistent with the assumpticr.:

the accident analyses.

3/4.9.13 SFEMT FUEL CASK HANDLING CRANE The restriction on movement of the spent fuel shipping cask within one cask length of any fuel assembly ensures that in the event this icac is dropped (1) the stored spent fuel assccblics will not be damaged ,

and (2) any possible distortion of fuel in the storage racks will c.ot result is a critical array.

3/4.9.iy exrmMmpr vtar 15<)terisa vmves "The OPEM6Ill7Y and closare res4ra %s og lhe cea ta wmca c vcat isolaNon values are sdfs'cient t.o res%'ct radbac Hue. imt A. fuel element raphsrc based upaa Me. lack o f csw er an.u-a -

release fron th e. R E f u Et.iN G tt o b t. ,

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< f >* &Wf $4/f f f Cffd ef d $ P1/1/ Yd/Y/ I head .cui/cAe r /ecked is /de c/0 sed porHrba.

I APM/cn8/a rf : rionts I,.2,3, up y ACTits : ui /A one sr beIh esaistSwen t vea t irelt hi a t . i efea , clore the. spes vs/se(s) iss?Aik sse Acue :.,dc io ai /ess/ //s7 staxoBY ur%k the sex + 6 Aou<s saj w C0L.O SHurpoWAl ao%'a 4he follomky 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

/ttau1/itrt CRT5 Su R tiEIL LAN e C - ___.

con teknest vent i.colalion volves sh1l:

'l. 6 . I. T 7*Ae he delern&cd closed a. + les s + once. per 3o ds y s ,

%<. no+or s,wesfors by ve<ifyag na+ pauser +v is

? renoved and Me vs/ve s ta& ale sAu t ,

t CPLuf t1 T* CetffS - um 7 I Jh b -4 e 71mndwn <;

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