ML20236U836

Northeast Nuclear Energy Co Amend Application.* Util Files Copy of 980507 Response to NRC Staff Request for Addl Info on Amend Application.W/Certificate of Svc

ML20236U836
Person / Time
Site:	Millstone
Issue date:	07/29/1998
From:	Repka D NORTHEAST NUCLEAR ENERGY CO., WINSTON & STRAWN
To:	Atomic Safety and Licensing Board Panel
References
CON-#398-19372 LA, NUDOCS 9807310105
Download: ML20236U836 (59)

Text

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/S.37Z DOCKETED USHRC July 29,1998 i

W JUL 30 P3

40 UNITED STATES OF AMERICA f;f y0

i NUCLEARREGULATORYCOMMISSIO@FE} g ADJUDC/iKti. ETAFF BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

Northeast Nuclear Energy Company ) Docket No. 50-423-LA

)

(Millstone Nuclear Power Station, )

Unit No. 3) )

NORTHEAST NUCLEAR ENERGY COMPANY'S AMENDMENT APPLICATION

• In accordance with the Order of the Atomic Safety and Licensing Board of July 28, 1998, Northeast Nuclear Energy Company ("NNECO") hereby files a copy of the March 3,1998 license amendment application seeking to modify the licensing basis for the Recirculation Spray System regarding direct injection into the reactor coolant system following a design basis accident (Attachment A). In addition, for completeness, NNECO hereby files a copy ofits May 7,1998 l

1 9807310105 980729 PDR f  %

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response to an NRC Staff request for additional information on the " amendment application (Attachment B).

Respectfully submitted, k \

U ^

David A.Repka WINSTON & STRAWN 1400 L Street, N.W.

Washington, D.C. 20005-3502 (202)371-5726 ATTORNEYS FOR NORTHEAST NUCLEAR ENERGY COMPANY Dated in' Washington, D.C.

this 29th day of July,1998 l-I - - - - - - - - - - - - - o

W, i '

DOCKETED UNITED STATES OF AMERICA USNRC NUCLEAR REGULATORY COMMISSION l

L  % Jll. 30 P3 :40 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD L

OFRCE nr Srm ,m

^

RULENJW .a ,u -

i In the Matter of ) ADJUDICAliCM STAFF

)

Northeast Nuclear Energy Company ) Docket No. 50-423-LA 1

)

(Millstone Nuclear Power Station, )

Unit No. 3) )

CERTIFICATE OF SERVICE l

. I hereby cenify that copies of" NORTHEAST NUCLEAR ENERGY COMPANY'S AMENDMENT APPLICATION," in the above-captioned proceeding, have been served on the following by deposit in the United States mail, first class, this 29th day of July,1998.

Nancy Bunon, Esq. Thomas S. Moore 147 Cross Highway Chairman Redding Ridge, CT 06876 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Office of the Secretary Dr. Charles N. Kelber U.S. Nuclear Regulatory Commission Administrative Judge

.. Washington, DC 20555 Atomic Safety and Licensing Board Attn: Rulemaking.and Adjudications U.S. Nuclear Regulatory Commission i (original + two copies) Washington, DC 20555-0001 Adjudicatory File . Dr. Richard F. Cole l Atomic Safety and Licensing Board Panel Administrative Judge U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board  !

Washington, DC 20555 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 L

_ _ _ _ _ _ _ _ ~ _ _ _ . _ _ _ _ _ _ ._ . - _ _ _ _ . ___-__--_____ _ ______ _ ___ _ _ __ - _ _ _ _ -

l '.

i i  :

Office ofCommission Appellate Adjudication Richard G. Bachmann, Esq:.

U.S. Nuclear Regulatory Commission Office of the General Counsel Washington, DC 20555 U.S. Nuclear Regulatory Commission Washington, DC 20555 ud J Sk-David A. Repka \

Winston & Strawn Counsel for Northeast Nuclear Energy Company 1

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i I-ATTACHMENT A a

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Nuclear Energy mi. ne A,iear co-r simoon Nortlwane Nuclear Energy Genipany P.O. Ilos 128 Waterford. Cr 06385-0128 (860) 447 1791 Fax (860) 444-4277 I

& Northean Udh6ee System MAR - 31998 Docket No. 50-423 B17044 Re: 10CFR50.90 10CFR50.59(a)(2) l U.S. Nuclear Regulatory Commission Attention: Document Control Desk -

Washington, D.C. 20555 Millstone Nuclear Power Station, Unit No.3 Proposed License Amendment Request .

Recirculation Sorav System Direct Iniection Chance (PLAR 3-98-1)  !

Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby' proposes to amend Operating License NPF-49 by incorporating the attached proposed revision into Chapter 6 of the Millstone Unit No. 3 Final Safety Analysis Report (FSAR).  ;

The proposed revision to the Millstone Unit No. 3 licensing basis eliminates the  ;

requirement to have Recirculation Spray System (RSS) direct injection into the reactor I coolant system.

NNECO, in our February 16,1998 submittal,- provided the NRC with an integrated safety assessment for all the RSS changes which have been processed to date under ,

J 10CFR50.59. The information and conclusions in the February 16,1998 submittal and this letter provide the NRC with a description which provides the bases for reviewing

. the current operability of the RSS.

Description of Proposed Revision l NNECO, in a 198610CFR50.59 change (FSARCR 86-MP3-53), modified the system description in the Millstone Unit No. 3 FSAR to reflect the elimination of RSS direct injection into the reactor coolant system. A recent review of this 10CFR50.59 change

concluded that the 1986 evaluation incorrectly concluded that this did not involve an L unreviewed safety question. Despite this incorrect 10CFR50.59 conclusion, NNECO L has determined that with respect to this change, the RSS was operable and would have performed its intended function.

l I

083422 5 MV. I195

U.S. Nuclear Regulatory Commission l

B17044\Page 2 Accordingly, this submittal corrects the 1986 issue by providing a proposed revision to Chapter 6 of the Millstone Unit No. 3 FSAR for NRC review and approval. This l submittal specifically addresses the changes that occurred to the RSS in 1986 that was recently determined to involve a USQ.

NNECO has reviewed other 10CFR50.59 changes made to the RSS since 1986. Under the assumption that this proposed revision is incorporated into the FSAR, NNECO has found all other 10CFR50.59 conclusions to be correct.

Markuo of Proposed Revision A copy of the marked up 1986 FSAR pages are contained in Attachment 2. The l markup identifies only the changes that were contained in the 1986 FSAR revision i (FSARCR 86-MP3-53) that are addressed in this submittal and does not identify all the changes that were contained in 1986 FSAR revisions or all the revisions to these FSAR pages that have occurred since 1986.

Backaround. Safety Assessment. Significant Hazards Consideration and Environmental Considerations ,

i The Background, Safety Assessment, Significant Hazards Consideration and Environmental Considerations that support this proposed revision are contained in Attachments 3 and 4.

Plant Operations Review Committee and Nuclear Safety Assessment Board Review The Plant Operations Review dommittee and the Nuclear Safety Assessment Board have reviewed this proposed amendment request and concur with the determinations. -

~ ~

State Notification In accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment to ensure their awareness of this request.

Schedule Reauest for NRC Aporoval NNECO requests NRC review of this proposed revision by April 30,1998 and that the license amendment be effective upon issuance with implementation within sixty (60) days.

O.S. Nuclur R:;gulatory Commission B17044\Page 3 if the NRC Staff should have any other questions or comments regarding this submittal, i please contact Mr. D. Smith at (860) 437-5840.

Very truly yours, l NORTHEAST NUCLEAR ENERGY COMPANY N$

M. L. Bowling, Jr. s/

Millstone Unit No. 2 - Recovery Officer SWbrn, to and subscribed before me s > .

this'd day of h er h 1998

%e b Wotary t A Pubic bo. A a My Commission expires ho. 33.Gool ,

Attachments cc: H. J. Miller, Region I Administrator J. W. Andersen, NRC Project Manager, Millstone Unit No. 3 A. C. Cerne, Senior Resident inspector, Millstone Unit No. 3 W. D. Travers, Ph.D, Director, Special Projects Office Director . .

Bureau of Air Management

- Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127 i

)

l l

Docket No. 50-423  !

B17044 l-L l

?

l Attachment 1 Millstone Nuclear Power Station, Unit No. 3 ,

Regulatory Commitments j l March 1998 j.-

^

d.S. Nuclear Regulatory Commission j 817044%ttachment 1\Page 1 l

Attachment List of Regulatory Commitments The following table identifies those actions committed to by NNECO in this document.

Please notify the Manager - Regulatory Compliance at the Millstone Nuclear Power Station Unit No. 3. of.any questions regarding this document or any associated regulatory commitments.

Commitment Committed Date or Outage NONE N/A- . I i

. \

I i

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r MMPS-3 FSAR Tables 6.2-12, 6.2-13, and 6.2-14. The energy distribution in the l I NSSS is given in Table 6.2-15.

The containment experiences a subatmospheric peak pressure after the RWST has emptied and the quench spray terminates. The maximum subatmospheric peak pressure of -0.07 psig occurs after the PSDER with minimum ESF. The times of subatmospheric peak are given in Table 6.2-9, and the pressure transients are shown on Figure 6.2-9.

- This case is the DBA for maintaining subatmospheric pressure.

Lorc y The initial conditions used for the PSDER with minimum ESF are the same as those listed above for the 0.6 PSDER with minimum ESF. A chronology of events for the PSDER with minimum ESF is given in Table 6.2-16. The energy distribution in the containment is given in Table 6.2-17 at the same occurrences as listed above for the 0.6 PSDER. Hass and energy release data used for the PSDER with minimum ESF is given in Tables 6.2-18, 6.2-19, and 6.2-20. The energy distribution in the NSSS is given in Table 6.2-21. .

No other single failure besides the diesel generator failure is considered since all other single failures are bounded by the diesel generator failure.

The heat inputs from blowdown and spillage shown in the containment energy distribution tables (Tables 6.2-11 and 6.2-17) for the 0.6 PSDER and PSDER with minimum ESF are greater than the total .

energy output shown in the corresponding NSSS energy distributions

( tables (Tables 6.2-15 and 6.2-21). This is largely due to conservatively interpolating between the data points provided from the mass and energy release analysis. It is also due to a larger calculated safety injection flow from the refueling water storage tank than was assumed for the mass and energy release calculations.

As discussed in Section 6.2.1.3, the sensitivity of the actual calculated containment pressure transient to the transient assumed for the mass and energy release calculation was determined. Two .

bounding cases were evaluated, one assuming an 1,800 second containment depressurization time, and one assuming 3,600 seconds. -

The results, shown in Table 6.2-9, indicate that the cases are approximately the same in regards to both depressurization time and subatmospheric peak pressure. Since both cases show significant margin within the required depressurization time (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />), neither case is clearly limiting. However, an assumed 3,600 second depressurization time is slightly more limiting for containment a subatmospheric peak and, on that basis, is used for all containment j analyses.

b 2.1.1.3.7 Main Steam Pipe Break Results The pressure responses resulting from a main steam pipe break j accident are calculated with the containment analysis computer '

l program LOCTIC. The program is used to calculate the thermodynamic

! state of the containment due to the mass and energy addition to the containment atmosphere as described in Section 6.2.1.1.3.2. The Amendment 18 6.2-15 March 1986 l l

l

t Sn/ s e n. T k The DBA for maintaining subatmospheric peak pressure is examined for a modified flow path from the containment recirculation pumps, following switchover to the injection mode of operation. The modiftsd path l elleinstes the parallel flow path direct to the cold legs and returns that portion which is through the charging and safety injection pumps to the cold legs, The reanalysis of the subatmospheric peak pressure case considering the moJified flow path utilises a revised degraded containment recirculation pump curve. The. curve includes a reduction in the magnitude of the assumed pump degradation of the. design curve from to to 5 percent. The revised curve is shown in Figure 6.2-40a.

Results of the reanalysis of the subatmospheric peak pressure case show a maximum subatmospheric peak of -0.09 psig. The time of subatmosphecic peak is given in Table 6.2-9. The pressure and temperature transients are essentially the same for this case as in the previously discussed case which resulted in a -0.07 psig subatmospheric peak pressure.

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HMPS-3 FSAR l

Determination of Dryout Time During the blowdown following a steam line rupture, a point may be reached when all the initial fluid inventory of the affected steam generator, including that added from outside sources. will be j i

depleted. At that time, the blowdown rate out of the break will be limited by the rate at which water is added to the steam generator l from the auxiliary feedwater system. This point in the transient is I

! termed "dryout." The time of dryout can be determined as described in WCAP-8860.

Additional mass and energy flow from the affected steam generator to containment results from:

1. liquid flashing in the unisolated portion of the main feedwater pipe;

2. pumped main feedwater; and .

3. auxiliary feedwater flow before isolation.

l For DERs, the main feedwater flow to the affected steam generator is '

1 conservatively assumed to be at runout flow for the various power levels depicted in Table 6.2-59 from the time of the break until the main feedwater isolation valve (FWIV) receives a signal to close.

The unaffected steam generator FWIVs are then assumed to close

• l instantaneously while the affected steam generator FWIV is assumed to have a longer closing time.

For split breaks, a hydraulic analysis determined the main feedwater flow rate to the -affected steam generator as a function of steam l generator pressure and power level (see Table 6.2-59). The main steam line break mass and energy release analysis used the average feedwater flow rate based on the integrated herage steam generator .

pressure up to the time of feedwater isolation.

• All main feedwater is diverted to the affected steam generator for the duration of the valve closure time. During the FWIV closure" at ,

i. the affected steam generator, feedwater flow is conservatively l . .

increased to 4 times the value in Table 6.2-59 for partial DERs and l' full DERs only. For split breaks the feedwater flow .is not

• increased. 'No flow reduction is assumed to occur during the closure sequence. The main feedwater system flow at.various power levels, feedwater temperature, and FWIV closure times are listed in I

Table 6.2-59. The time of main feedwater isolation is dependent upon the time that the isolation signal setpoint is reached, which varies l

i with the break condition.

m A v imu m The Aauxiliary feedwater flow to the affected, steam generator is 42.4 lbm/sec for the partial and full DERs and 39 lbm/see for the split breaks. The flow is limited by passive flow control devices (cavitating venturis) installed in the line to each steam generator (Section 10.4.9).

i: Amendment 18 6.2-37 March 1986

_ - _ - - - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - . _________-__-___:__________________-______--________________-_______-____--_-_____ _ __A

(

MNPS-3 FSAR

6. Regulatory Guide 1.26 quality group standards. The systems are designed in accordance with ASME III, Class 2 and is designated Safety Class 2.

! 7. Regulatory Guide 1.29 for seismic classification. The systems are designed to Seismic Category 1

8. Regulatory' Guide 1.82 for the design of sumps for ECCS and

containment spray systems

9. The. containment atmosphere pressure is returned to subatmospheric within one hour after the DBA, assuming the worst single active failure occurs concurrently with the DBA l~ (see Section 6.2.1)., ,, ,

10. The ' systems are capable of operating in the post-accident

environment to' maintain a .subatmospheric conta'inment pressure for 30 days following the DBA.

11. The quench and containment recirculation spray headers are capable of delivering spray water to the containment atmosphere in sufficient quantity over a sufficient area of the containment and with an average droplet diameter to ensure adequate heat removal to accomplish design bases 1, 4, 9, and 10 above.

• 3.2 . The design of the containment recirculation system is g g,g sufficiently independent and redundant so that an active y, , , , , failure in the recirculation spray mode :: :: ::ti:: :r

- ; ::1 : frilur: in 2: ::1d 1:;; :::ir:21:.ti:  ::d: of the ECCS has no effect on its ability to cerform '

its engineered H*r 8K' safety function.

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• • 1. - 13. Instrumentation is provided to monitor the containment heat removal systems and system component performance under ~

L accident- conditions in accordance with Regulatory .

Guide 1.97.

14. Provisions are made to allow drainage of spray and emergency j core cooling water to the containment sump. The sources and l quantities _ of energy that must be removed from the ,

containment to meet the design bases are discussed in i Section 6.2.1.

15. The quench spray system is capable of. adding a boric acid / sodium hydroxide solution from the refueling water storage tank (RWST) and the chemical addition tank (CAT) to r- the containment sump for sump pH control between 7.0 and 7.5 while not exceeding a spray pH of 10.5. l 6.2.2.2 System Design The containme'nt heat removal systems consist of two parallel redundant quench spray subsystems feeding two parallel 360 degree ,

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l. 6.2-41 l . .

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HNPS-3 FSAR Py = Vapor pressure of sump liquid (saturation pressure at liquid temperature)

All parameters are expressed in feet of head. This expression can be simplified by making the conservative assumption that the vapor pressure of the pumped liquid is equal to the total containment pressure, as follows: ,

Available NPSH = Z - Hg ,

)

The following tabulation presents the determination of the minimum available NPSH following a DBA and a comparison with the required NPSH to demonstrate adequate margin. 0:th th; r;;ircui:ti:n :pr:y

d; n.d th

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ld ic; recircui ti r. ;;d n; =: luted. The parameter values used to evaluate theminimumavailabJeNPSHare taken at the -time the available NPSH is at a minimumf!!h+e- I is the -

time of initial pump start-up for the spray mode, 2nd th: tim: cf

vit:h=u fu th: 0000 r;;ircul: tie.; . de. In addition, water level is minimized by assuming minimum ESFr one quench spray pump in operation prior to initiation of the containment recirculation pumps.

Losses in the suction piping have been maximized by considering all contributors, including pipe bends and containment sump screen losses.

• The most limiting single active failure for evaluating the NPSH for ,

the containment recirculation pumps is the failure of one ESF train. . l This failure exhibits the highest calculated pump flow (system runout i flow) and correspondingly highest required NPSH. The required NPSH is selected from the pump manufacturer's test data. j Uncertainties, such as NPSH variation between similar pumps and testing inaccuracies, were considered but not included in the calculations due to the large margin between available and required NPSH.

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Recirculation C:ld ';g Spray Hode "::ircul:ti:n "d:

Elevation head (ft) (Z) 23.4 -99 6-Pipe losses (f t) (H ) 6.7 -t-e-Available NPSH (ft) (Z-H ) 16.7 21.7 Pump flow (gpm) 3,880 4,200 Required NPSH (ft) 7.0 0.0 Margin 9.7 12.7 The results shown in the above tabulation are not sensitive to break location or initial containment conditions. The assumption that the vapor pressure of the liquid in the sump is as great as the containment total pressure eliminates any effects on the calculation l of the availability of ECCS spillage temperature, initial containment

! pressure and temperature, RWST temperature and service water I temperature.

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IINPS-3 FSAR I

l The' safety injection pumps deliver water to the RCS from the RWST i

during the injection phase and from the containment sump via the ,

I containment recirculation pumps during the recirculation phase. Each  !

high head safety injection pump is driven directly by an induction

! motor. The pump _ lubricating -oil coolers are cooled by the safety injection pump seal cooling subsystem (Section 9.2.2.5).

A minimum flow bypass .line is provided on each pump discharge to recirculate flow to the RWST in the event that the pumps are started {

with the normal flow paths blocked. This line also permits pump )

testing during normal plant operation. Two parallel valves in series l l

l with a third. downstream of a common header, are provided in this line. These valves are manually closed from the control room as part of the ECCS realignment from the injection to the recirculation mode.

A pump performance curve is shown on Figure 6.3-5. --

Containment Recirculation Pumps .  ;

The containment recirculation pumps (Section 6.2.2) are provided for containment- structure depressurization and later during the 3 recirculation mode for core heat removal. The pumps will provide 4ew-  !

5::t pr:::;r; safety injection dic;;tly  ;.. via the charging and' safety injection pumps during recirculation. -

6.3.2.2.4 Containment Recirculation Coolers ,

The containment recirculation coolers (Section 6.2.2) are shell and tube type heat exchangers serving to cool recirculated water flowing through the shell side from the containment recirculation pumps.

Service water acts as the cooling medium flowing through the tube side of the cooler.

• 6.3.2.2.5 Valves The design parameters for all ECCS valves are consistent with the .

design parameters of their respective systems 'as described in

• Table 6.3-1. Relief valve design parameters are listed *in

• Table 6.3-2.

'The IEEE 323 Environmental Qualification Program for all ECCS valves will be completed prior to fuel load.

The design features used to ' minimize valve leakage include l

1. Where possible, use of packless valves

2. Other valves which are normally open, except check valves and those which perform a control function, with backseats to limit stem leakage

3. Normally closed globe valves installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (potentially radioactive) water

4. Enclosed relief valves with a closed bonnet Amendment 4 6.3-7 September 1983 I-l L

MNPS-3 FSAR .

i detect and isolate such-leaks in the emergency core cooling flow path within 30 minutes.

Potential Boron Precipitation Boron precipitation in the . reactor vessel can be prevented by a back-flush of cooling water through the. core to reduce boil-off and resulting concentration of boric acid in the water remaining in the reactor vessel.

T%so

-Th::: flow paths are available for. hot leg recirculation of sump l water. Each safety injection pump can discharge to two hot legs with suction taken from the containment recirculation pump discharge. -En- -

ffftic;, th; :::ir:21:ti:: 7 ;: ::: fi :i ;;: th :r-h

:r. ::: lin: - ' inj::: =t:: th : ;;h %: h:t 1d:. '" :" * - '

Loss of one pump or one flow path will not prevent hot- leg recirculation, since two redundant flow paths are available for use.

Safety Grade Cold Shutdown Function During a safety grade cold shutdown the ECCS is relied upon to provide one of the two redundant flow paths for boration and makeup.

The~ ECCS high . head injection header provides this function. The ,

redundant flow path is the nonnal charging header which is part of the chessical and volume control system. Two independent subsystems each consisting of a charging pump and the associated valves and l piping are provided and are powered by redundant emergency buses in a i manner. that ensures that at least one subsystem is always operable.

A solenoid valve provided in_ each subsystem and located in the chemical and volume control system ensures that the remote throttling capability necessary for RCS inventory control and . shutdown is available. Provisions are also included in the ECCS design to ensure j that the accumulators can be either isolated or vented so that RCS '

depressurization can be accomplished. Details of the cold shutdown -

design are discussed in Section 5.4.7. , ,

. 6.3.2.6 Protection Provisions The provisions taken to protect the system from damage that might result from dynamic effects associated with postulated rupture 'of piping, are discussed in Section 3.6. The provisions.taken to l protect the system from missiles are discussed in Section 3.5. The l provisions to protect the system from seismic' damage are discussed in Sections 3.7, 3.9, and = 3.

• 0. Thermal stresses on the RCS are discussed in Section 5.2 6.3.2.7 Provisions for Performance Testing Test lines are provided for performance testing of the ECCS system as ,

well as individual components. These test lines and instrumentation  !

are shown on Figures 6.2-37 and 6.3-2. All pumps have miniflow lines l for use in testing operability. Additional information on testing j can be found in Section 6.3.4.2.

6.3-15 l

l.

MNPS-3 FSAR conservatively estimated to be 2 minutes. Assuming the higher outflow for 2 minutes, and then a constant outflow of 7.710 gpm for 8 minutes, approximately 06.800 gallons of water would be used during the assumed 10 minute switchover time.

Initistion of switchover is conservatively assumed to occur at the low-low level setpoint with allowance for negative instrument error (tank elevation = 23 feet 5 inches). This level has approximately 479,000 gallons of water remaining within the RWST. The amount of water remaining within the RWST at the completion of switchover would then be approximately 402,000 gallons (tank elevation = 19 feet 8 inches), assuming both residual heat removal pumps automatically stop, and approximately 392,000 gallons (tank elevation = 19 feet 2 inches), assuming one pump fails to stop. Adequate margin in available.NPSH for therECCS pumps- exists at this. lower RWST level (Refer to Section 6.3.2.2.3).

~

'Following the completion of the switchover sequence, two of the four containment recirculation pumps would take suction from the .

containment sump and deliver borated water d*.ec d y :.. G. O c 4 0 -

le;r_ i pertien ef t'e r"r':ulstier pr- di :hr;: f1:u :: 1d i:

Tof88 cr:d i: pr: id suction 8the two charging pumps and the two safety injection pumps, which-weeM-elee deliver directly to the RCS cold legs. 'As part of the r,witchover procedures, the suctions of the charging and safety injection pumps are cross-connected in the event

• of failure of either recirculation pump.

Section 7.5 lists the process information available in the control

, room to assist the operator in performing the svitchover actions.

s I Amendment 15 6.3-16b September 1985 l

\

MNPS-3 FSAR TABLE 6.3-6 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS

• FLOW PATH L ; !!;;d Indication of Loss T. :ircul:tien of Flow Path Alternate Flow' Path Tre... : nt ir..cr.t  ;. : ..ul:tien ef water "ie th indcp;nd:nt, e- - * - -

-et:utture :" p -in th: engineered -ler herd fler path' * * - * ~

t: 1: '::d inj::--  ::feti f::tur:: utilizing th :th r ti .. ;. . der vie- -huilding v;17; pit  ::nt:innent ::- -

the tre centri w  :::: -circul:td:n ;" 7 :nd

nt s ;ir;ulatien--

:1 r

=;.

:.nd :::1 r:

High Head Recirculation From containment Accumulation of water Via alternate high -

structure sump to in the engineered head flow utilizing the hign head safety features the other con-

. injection header building valve pit tainment recirculation via two contain- area or auxiliary pump and cooler and ment recirculation building sump alternate safety pumps and coolers injection pump and the safety injection pumps

. NOTE: -

~

• Long term phase ~

t I

t

(

1 1

1 1 of 1

HNPS-3 FSAR TABLE 6.3-7 SWITCHOVER PROCEDURE

• I A. From Injection to Cold Leg Recirculation L The 'following manual operator actions are required to terminate the :

1. injection mode and establish the recirculation mode. It should be noted that the RHR~ pumps have been stopped automatically on receipt of a RWST low-low level signal Step.1 Valve realignment prior to containment recirculation pump switchover from spray to cold leg recirculation.

1; a..' Co.os a RHR ro,e' caso ues su resrow vns.uns (fros e/s) 6 p. Close RHR pump suction valves from the "RWST (1-8812 A/B).

c . )(. Close tne RHR pump discharge cross-connecting values (1-

~

'8716 A/B).

d f. Close containment recirculation pumps A and B discharge valves to. containment recirculation spray headers (H0V 20 '

A/B ) . .

ei [. Open containment recirculation pumps A and B discharge

! valves to the charging / safety injection system (1-8837 .

A/B).

Step 2 Close the-safety injection pump miniflow valves 1-8813 (1-8814

! and 1-8920).

Step 3 Align safety injection and charging pumps to the containment j recirculation pump discharge.

i

a. Open contclnment recirculation pump A to the charging pumps suction header valves (1-8804 A). -

b. .open containment recirculation pump B discharge to safety

• l injection pumps section header (1-8804 B). l 1

c. Open charging - safety injection suction cross connection j valves (1-8807 A/B). j Step 4 . Isolate the refueling water storage tank.

I a.- Close charging pump suction RWST valves (1-LCV-112 D/E).

b. - Close safety injection pump suction RWST valve (1-8806).  !

l

[

k l

l 1 of 2 '

1 HNPS-3 FSAR - ]

TABLE 6.3-7 (Cont) {

l B. From Cold Leg Recirculation to Hot Leg Recirculation Ot;p i *lign ;;ntairc;nt r::irculati:n p' ;

• t: d: liv:r dir :tl'i t:-

ihu RC3 Lhivoyh the het 1;; inj;;ti;n h::d:r :nd t: centinu: te d: liv:r t; th: ;ucti:n :f th: ::ntrifu; 1 ch:r;in; p" r: and th Jo L 1 s.3 ec tie.. pu...p; -sie t!.; ; c.....; n Cher;ing/0 pu;p-

ti:n :re:::ver line.

c. 01;;; th; ;;1d 1;; i;;1;ti;n v;l;; (1 0000 .'.) .

S. 0;:r cr:::ti: i: 1:tien v:1v: (1-0715 '.).

c. Op;n het 1;; i;;10ti;n v;17; (1 00'.0).

Step 2 Align centeir.;;nt recirculation pump 0 te deliver te th 203 vi; th; safety injection pumps / To DEwusa re r** 8CI 8.vArss og c.usex ccesso

a. CloseAthe cold leg isolation valves (1-8809 44, A/S) ,

b. Stop safety injection pump 1. -

c. Close safety injection pump crosstie isolation valve for .

separation of safety injection pumps (1-8821 A).

d. Open hot leg isolation valve (1-8802 A).

e. Start safety injection pump 1.

i

f. Stop scfety injection pump 2. )

g. Close cold leg isolation valve and crosstie isolation j valve (1-8835/8821 B). - l l

h. Open hot leg isolation valve (1-8802 B). *

. I

1. Start safety injection pump 2.  ;

l l

- l t i l

NOTE:

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n m o1 S t rooSvur Rci eg roSof no roooo csmoiB 1 P a . e' rso o

N M

6 E

c uroC miC /r a C 2l ue lurM e l p c R pi u R Ii tR f mnq Scn .trl ov pH s s n v c pi s l uroChdroeoa pc iil rt t pR C19 i

' g ol eeuc i nm s

. L B E f af f f ui e f H e of o oh a iaff enianuae mtg iaffhhl eeuCouoi FoooSMroLrCoNhtv Fool otvipll F o o t tF R b gR C pt v A -

T e

s a

h P

n C - C C nR nR nR io o o o t if if if a to to to r a a a e

p lug s l s l s ug ug O ce rl s ce rl s ce rl t S p i p p C

C icto eoo cto eoo i

cto E Rhl Rhl eoo

• Rbl e

d o

M o _ on o e tn to trg r od d en u s~ n sen svid l I na l sa liki i

a ape

'l em l om lt ru al e aeol F Fod Fcd F!:wf d d e e n t t op) a a re ms re ) ituu t

n ev . s pl ev aFo 8u pl l(g e oaA2o oa . u o n

v20g v5 cal o r 08o r 3 r a p oe88l oe8 ipn m

o oa tt8- a tt8 cma 1 n oa- eu C Mg1(a M g1 RPB.

0 1 2 3 3 3 m

r

MNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont)

SI pumps which deliver to the reactor through their cold leg connections and to both of the CC pumps which deliver to the reactor through their cold leg connections. Th: On pump :1: delic:r: f1;u directly ts th: :::ter thr ugh tu: ::1d leg cine: the dir:52rge

r::: ::nn::t 1:1::: ::: clered 'her m: kin; the tr:r:fer fr r injectier te retirculetie-Mode C - Hot Leg Recirculation This mode presents the process conditions for the case of hot leg recirculation, assuming containment recirculation pump A operating, centrifugal charging (CC) pumps A and B operating, and safety injection (SI) pumps A and B operating.

In this mode, the safeguards pumps again operate in series with only the CR pump taking suction from the containment sump. The recirculated coolant is then delivered by the CR pump to both of the CC pumps which continue to deliver to the reactor through their cold leg connections and to both of the SI pumps which deliver to the ,

reactor through their hot leg connections. ,

l 1

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Amendment 2 2 of 18 April 1983

6 4 l

HNPS-3 FSAR  !

l l

NOTES TO FIGURE 6,3-1 (Cont) f VALVE ALIGNMENT-CHART OPERATIONAL MODES Valve No. _A _B .C.

1 0 C C l 2 0 C C' 3 O C C 4 0 C C 5 0 C C 6 0 0 C 7 0 0 C ~

8 C C 0 9 C C 0 10 C C C 11 C C C 14 C C C 15 C C C 16 C C C 17 C C C .

18 0 0 0 19 0 0 0 20 C 0 0  !

21 C 0 0 I 22 0 C C 23 0 C 4-C 24- 0 +C C 25- C C 4- C 26 0 +C C 27 C C C 28 0 C C 29 C 0 0 . -

30 C C C l . 31 C C C 32 0 0 0 0 = open C = Closed 3 of 18

NNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont)

MODE B - COLD LEG RECIRCULATION (PUMP NO,J[ OPERATING).

I Pressure Temperature, Flow Volume Location Fluid _

(psig{ (*F) (Gpm)(*8 (lb/sec) (gal) 1 Refueling ATM Tank 50 - -

<5,000 Water 2 Refueling -

50 0 0 -

Water 3 Refueling -

,50 0 0 -

Water -

4 Refueling -

50 0 0 -

Water 5 Refueling -

50 0 0 -

Water 6 Recirc' -

182 0 0 -

• Coolant 8 Refueling -

50 0 0 -

Water

- O o 9 Recire 95- 182 t-296- t99- -

Coolant l 10 Recirc N 182 N ,- .

Coolant i

. 7b7 7tl 95 11 Recirc 1,105 182 499 $$- -

Coolant 12 Refueling -

50 0 0 -

Water da ran -6 -

ro o o 13 rgg

-M- 449- 496 50- -

gen niw -

5'o o o 14 rggg 1,10; +ee- 4 39- -Se- -

l. At reference conditions 212*F and 0 psig

• Minimum allowable volume at normal operating conditions .

J i

9 of 18 4

J

HNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont)

MODE B - COLD LEG RECIRCULATION (PUMP NO. g OPERATING) 1 Pressure Temperature, Flow Volume Location Fluid (psig) (*F) (Gpm)(18 (lb/sec) (gal) 15 Refueling -

50 0 0 -

Water 16 Refueling -

50 0 0 -

Water 7 tl 9f 17 Recirc 1,000 182 -6%- 149- -

Coolant 17F 24 18 Recire 182 -Et9- -

Coolant O 678 3 */

19 Recirc X 182 -279tE- -968- -

Coolant 0 178 24

• 20 Recirc N 182 -B19- & -

Coolant 21 Nitrogen 0 Ambient 0 0 -

22 Nitrogen 0 Ambient 0' O -

23 Nitrogen 0 Ambient 0 0 -

24 Recisc -

212 0 0 -

Coolant g a rerw=c, 50 '

25 -

-462- 0 0 -

. R. .$.f!4

.c . .. .

Rarp rum (O 26 necu; -

-199- 0 0 -

Ce.l i.;

Ar rec.ws To 27 gg w.......

462- 0 0- -

Reva w ~s -

50 0 0 28 49- 462- h566- M -

"::ir ait. .

i

_w f w.. .e..

j V

Rs.rus uec CO 29 ":ggrg - 462- 0 0 -

%ww.....

1. At reference conditions 212*F and 0 psig j

• Minimum allowable volume at normal operating conditions '

10 of 18 l

l t

1

]

I

. 4 l

MNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont)

MODE B - COLD LEG RECIRCULATION (PUMP NO Jf OPERATING) i Pressure Temperature, Flow Volume Location Fluid (psig) (*F) (Gpm)(3) (lb/sec) (gal)

Athenas - Co o o 30 P g j4 40- -tee- ',714 tie- -

Rakemns So 31 Pg . p, ,

-tee- 0 0 -

Rrence.ws -

(

0 0 -

[idHU<

33 - - -

50 0 0 -

.. M.4..'r4. .

34 Reactor -

212 0 0 -

Coolant .

35 Refueling -

50 0 0 - -

Water 36 Refueling -

50 0 0 -

Water  ;

37 Refueling -

50 0 0 -

Water l

38 Refueling -

50 0 0 -

Water - '

i 39 Refueling -

50 0 0

, , Water Ac t, 182. gleo 17s" j 40 Refueling / -e- -

Water 41 Refueling -

50 0 0 -

Water 42 Refueling -

50 0 0 -

Water

1. At reference conditions 212*F and 0 psig

• Minimum allowable volume at normal operating conditions 11 of 18

MNPS-3 FSAR -

NOTES TO FIGURE 6.3-1 (Cont)

MODE B - COLD LEG RECIRCULATION (PUMPNO./ OPERATING) i Pressure Temperature, Flow volume Location Fluid (psig) (*F) (Gpm)(13 (1b/sec) Qay As

~

SO O O 44 n c%$Lws

r- Cn Pu p -tet- 'r;600- 405- -

k. ..N. .!. .( .n.-.a..,..

Arc,i Iil ggao e 7(

45 ngf;.g<c,r.; 206 -G- - -

- W r 46 Refueling -bew- ,50 0 0 ,

Water -Prc: ur:

Refueling 47 tow 50 0 0 -

Water P. c a r.. ;

48 Refueling h w- 50 . 0 0 -

Water Fre: ure 49 Refueling bew- 50 0 0 -

Water Pr;;;ur:

50 Refueling h 50 0 0 -

Water Fre: cur 51 Refueling Lew 50 0 0 -

Water Pre: ure i

l 52 Recirc - 182 0 0 -

, l Coolant ,

l l 53 Recirc -

182 0 0 -

l Coolant i .to G. Ilse 17f 54 Recirc / 182 4-- 4- -

Coolant 539 S~11 to 55 Recirc +r H 9- 182 -444- 46- -

Coolant .

106 'lll tf l 56 Recirc -+99- 182 -896- -itt- - l Coolant l Rob S*11 to  !

57 Recire 182 -+t9- -

Coolant j

1. At reference conditions 212'F and 0 psig

• Minimum allowable volume at normal operating conditions .

I 12 of 18

I NNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont) l MODE B - COLD LEG RECIRCULATION (PUMPNO.[8 OPERATING) l Pressure Temperature, flow Volume Location Fluid (psig) (*F) (Gpm)(13 (lb/sec) (gal)

Ran_puus E:: fo o 0 l 58 r- -142- +t9 -

1 M. . ..* F.i.t. .

Rtfatwos -

50 o o 59 R::ir: + dis- +99 4t9- -

n. .W.M. .k. . .

Rssws 6'39 fo 74 to  ;

60 1,010 462- -if4- M -

,N E - 4..'r?.!. '

61 Recirc 0 182 -

Coolant C3 't S*A S 7o 62 Recirc 1,000 182 9t4- 99- -

Coolant S*14 C.13 70 64 Recirc 1,200 182 4t+ W -

Coolant (31 838 tt 65 Recirc W 182 17G.; tt -

l Coolant i

o 838 87 66 Recire N 182 4-76-9 24 -

j Coolant * '

l i

l

1. At reference conditions 212*F and 0 psig

• Minimum allowable volume at normal operating conditions 13 of 18 -

I I

l

HNPS-3 FSAR .

NOTES TO FIGURE G.3-l'(Cont)

H0DE C - HOT LEG RECIRCULATION (PUNP NO. 1 OPERATING)

Pressure Temperature, Flow Volume Location Fluid (psig) (*F) (Gpm)81* (lb/sec) (gal) 1 Re fueling ATM Tank 50 - - <5,000 Water 2 R_efueling -

50 0 0 -

Water 3 Refueling -

50 0 0 -

+

Water 4 Refueling -

50 . 0 0 -

Water 5 Refueling -

50 0 0 -

Water 6 Recirc -

182 0 0 -

Coolant 8 Refueling -

50 0 0 -

Water 160 691 13 9 Recirc -i!4- 182 6 40- -

Coolant -

l- -

140 611 15 l

10 Recirc 182 460- $9- . -

Coolant 7 43 611 13 11 Recire -fts- 182 -660- 90- -

Coolant 12 Refueling -

50 0 0 -

Water i 14e 499 T3 j 13. Recirc 46- 182 -66& -

Coolant '

7TJ 61f 93 14 Recire -Vit 182 464 - {

Coolant

.15 Refueling -

50 0 0 -

Water

1. At reference conditions 212*F and 0 psig 14 of 18

MNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont)

MODE C - HOT LEG RECIRCULATION

{

(PUMP NO. 1 OPERATING) '

Pressure Temperature, Flow Volume Location Fluid (psig) ('F) (Gpm)i18 (lb/sec) (gal) 16 Refueling -

50 0 0 -

1 Water l

l 17 Recirc -

182 0 0 -

Coolant 18 Recire -

,L82 0 0 -

Coolant .

19 Recirc -

182 0 0 -

l Coolant 20 Recire -

182 0 0 -

Coolant 21 Nitrogen -

Ambient 0 0 - -

22 Nitrogen 0 Ambient 0 0 -

23 Nitrogen 0 Ambient 0 0 -

24 Recirc -

212 0 0 -

Coolant 4Aaws fo 25 -

6 0 0 N.4 Anmsw . r0 g3ggg, e .

26 -

o o. .

Rtk6uns fc r

27 ggggg, ' -

e o o ,

I Assas.no re l 28  :'; -i- c -

9+0E- 0 0 -

m..k..at1" i

l Renes < So l 29 pggj 6 0 0 -

Ashtoems (o 30 -

-4+BB- 0 0 -

-Reel C$1*[ +e$_"

tann.as to 31 c ;i c:

.c443 O o .

waren

1. At reference conditions 212*F and 0 psig 15 of 18

HNPS-3 FSAR -  !

l

)

NOTES TO FIGURE 6.3-1 (Cont)

MODE * - HOT LEG RECIRCULATION (PUMP NO. 1 OPERATING) l Pressure Temperature, Flow Volume Location Fluid _

(psig) (*F) (Gpm)(18 (lb/sec) (gal) i COOla.it j Rtcu tss~t C )

32 --

@o 0 0 -

2 - -i r ?.4

,,N.

4. t.

33 Rermsw-s n::ir-n 442-o 3,102

'o

-404- -

a

., w m . .i.t.. . .h . .

34 Recirc -

212 0 0 -

Coolant R a re tsen e [0 35 n -

tee- 0 0 -

M.... ' A. . . .

Rerun w t to 36 -

,, _ ' r

-tee- 0 0 -

n M_ .a r#.

Assua6ianc fo 37 'rc- -

-lea- 0 0 -

n;R.aest www . . . .

Reresses To 38 0 'r: -

-4483- 0 0 -

sovM.a ... r tt.

Ramesi,s 5"O 39 -

-tes- 0 0 -

n; M. .hE. .M.

l'18 '2, 3 'i l 182 40 Recirc 182 2,130 -99+- - -

• Coolant Rumas oe - $~0 41 n ;ic- -tet 0 0 -

c_h_ _M_ E. 5 Aerenwe -

5"o 42 Retir- 69 -tee- 0 0 -

WA.r. A. ,4 R efunws fo 44 n--ir; 4446- 0 0 -

e_N_ _ _f _TN_ - .

175' 2 3 'f I 3 # 2, 45 Recirc Cn "c; 182 0,000 999 -

Coolant di :htrge-l 1. At reference conditions 212*F and 0 psig 16 of 18

HNPS-3 FSAR NOTES TO FIGURE 6.3-1 (Cont)

MODE C - HOT LEG RECIRCULATION (PUMP NO. 1 OPERATING)

Pressure Temperature, Flow Volume Location Fluid (psig) (*F) (Gpm)(1) (lb/sec) (gal)

Rathee4 ' --

- 5'o o o 46 7- -140- 0,;;; -+ts -

n M..- 6. .t.M.. .

grows To o o 47 S::b- -tea- +;M4- -209 -

, mar 30

-ww--www 743 a a- . Gif TJ 48 Recirc -645-- 182 -666- 60- -

Coolant .

o 3 ro 47 49 Recire f 182 -tT90t 959 -

Coolant 743 LTT TJ 50 Recirc -645- 182 -666- -Ge -

Coolant ,

0 In V7 51 Recire / 182 -330 -

Coolant 52 Recirc -

182 0 0 -

Coolant 53 Recire -

182 0 0 -

Coolant i 214l 170 SIL 54 Recirc / 182 -2,1 0 -966 -

Coolant ,

1330 471 G2 55 Recire 1,010 182 -+t9- - -

.' Coolant

= .

870 8400 st7 56 Recirc -+36 182 -t--See- -t Mr -

Coolant 170 4 71 G3 57 Recirc 182 -+t9-  % -

Coolant 170 4 71 63

58 Recirc -SS- 182 -+t9-  % -

i Coolant i 1130 4 71 63 l 59 Recirc 1,510 182 -+t9- 4fr -

Coolant 133e ss?

60 Recire +r5t6- 182 +f+ 16 -

1. At reference conditions 212*F and 0 psig 17 of 18 .

l i

l HNPS-3 FSAR .

NOTES '!3 FIGURE 6.3-1 (Cont)

MODE C - HOT LEG RECIRCULATION (PUMP NO. 1 OPERATING)

Pressure Temperature, Flow Volume Location Fluid _

(psig) ('F) [Gpm)(18 (lb/sec) (gal)

Coolant lit l El Recirc 0 182 -te+- 16 -

Coolant III8 f28 lei 62 Recirc. 1,455 182 -944- -95 -

s j

1 Coolant ,, 1

- 1 1330 #At set 64 Recirc 1,0^0 182 -929 M -

Coolant 1310 ter n7 65 Recirc 4 ROOS. 182 4Hh+ -E+ -

Coolant 0 285~ 17 66 Recirc -See 182 -tHhfr -24 - -

Coolant  : .

l I.

I l

1 i

1. At reference conditions 212'F and 0 psig 18 of 18 l

j

Docket No. 50-423 B17044 l

w ',fars *- ew r Attachment 3 Millstone Nuclear Power Station, Unit No. 3 Background and Safety Assessment l

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U.S. Nuclar R:guictory Commission B17044\ Attachment 3\Page 1 Millstone Nuclear Power Station, Unit No. 3 Background and Safety Assessment Backaround Northeast Nuclear Energy Company (NNECO),. in a 1986 10CFR50.59 change (FSARCR 86-MP3-53), modified the system description in the Millstone Unit No. 3 Final Safety Analysis Report (FSAR) to reflect the elimination of Recirculation Spray System (RSS) direct injection into the reactor coolant system. A recent review of this 10CFR50.59 change concluded that the 1986 evaluation incorrectly concluded that this did .not involve an unreviewed safety question (USQ). Despite this - incorrect 10CFR50.59 conclusion,' NNECO has determined that with respect to this change, the

, RSS was operable and would have performed its intended function.

Accordingly, this submittal corrects the 1986 issue by providing a proposed revision to Chapter 6 of the Millstone Unit No. 3 FSAR for NRC review and approval. This submittal specifically addresses the change that occurred to the RSS in 1986 that was recently determined to involve a USQ.

NNECO has reviewed other 10CFR50.59 changes made to the RSS since 1986. Under the assumption that this proposed revision is incorporated into the FSAR, NNECO has

• found all other 10CFR50.59 conclusions to be correct.

Recirculation Sorav System Oriainal Desian j The RSS, together with the Quench Spray System (QSS), is designed to provide long-term cooling of the containment and the core after the design basis accident. The RSS i goes into operation approximately 11 minutes after a Containment Depressurization Signal is actuated. All four RSS pumps take suction on the containment sump and I

- deliver to the two RSS spray headers. This phase of operation is referred to as the ,

injection phase and is meant to assist the QSS in the depressurization of the

containment after a Loss of Coolant Accident (LOCA).

The cold-leg recirculation phase starts approximately 33 minutes (with both trains of Emergency Core Cooling System (ECCS) running) after the Containment j Depressurization Actuation (CDA) signal, when the low-low level in the refueling water storage tank is reached. In this alignment two RSS pumps supply:

2 charging (CHS) pumps to 4 cold legs, )

2 Intermediate Head Safety injection (SlH) pumps to 4 cold legs, )

e direct injection to 4 cold legs via Motor Operated Valves (MOVs) 8809A and B.

4 l

iJ.S. Nuclear Regulatory Commission B17044\ Attachment 3\Page 2 .

Several hours into the LOCA, the hot-leg recirculation is initiated, to prevent boron precipitation in the core. The hot leg recirculation function was provided by 2 RSS

- pumps supplying:

{

l 2 CHS pumps to 4 cold legs, '

. 2 SlH pumps to 4 hot legs,-

e isolation of cold leg direct injection, e direct injection to 2 hot legs via MOV 8840.

During the transfer to hot leg recirculation, the operators were instructed to isolate cold leg direct injection by closing MOVs 8809A and B, and establish hot leg direct injection 4

'by opening MOV 8840 .

.* 'r 198610CFR50.59 Chance for Elimination of Direct Iniection -

During pre-operational testing of the RSS in 1985, excessive tube vibration in the RSS heat exchangers was observed during certain modes of operation. The specific test which resulted in excessive flow included operation of one RSS pump feeding the two charging (CHS) pumps, the two SlH pumps and the hot leg direct injection path, it was determined that excessive tube vibration could occur for heat exchanger flows in , i excess of 4600 gpm. The corrective action was to eliminate RSS direct injection to

~

. reduce RSS heat exchanger flow. System analysis demonstrated that the flow provided by the two SlH pumps and two CHS pumps in the recirculation phase was sufficient to assure the minimum flow required for core cooling. - Therefore, the direct injection into -

the cold (and hot legs) was not required. i Emergency Operating Procedures (EOP) 35 ES-1.3, " Transfer to Cold Leg L I Recirculation" and EOP 35 ES-1.4, " Transfer to Hot Leg Recirculation," were revised to terminate flow from the RSS pumps directly to the RCS immediately after transfer to -l cold leg recirculation. As a result, the RSS pump supplied flow only to the, suction of . l the SIH pumps and the CHS pumps in a piggyback mode of operation. During hot leg recirculation,' the SlH pumps were aligned to inject into the hot legs. However, provisions in the EOP's were retained to open the valves for direct cold-leg injection as a contingency action, should one be required.

The 1986 cold leg recirculation alignment is summarized below:

. 2 CHS pumps to 4 cold legs, e 2 SlH pumps to 4 cold legs, e isolation of cold leg injection valves MOV 8809A and B.

Similarly, the hot leg recirculation alignment was changed to:

. 2 CHS pumps to 4 cold legs, l 4

U.S. Nuciscr Ragulatory Commission B17044\ Attachment 3\Page 3 .

• 2 SlH pumps to 4 hot legs, e isolation of cold leg direct injection, e no direct hot leg injection via MOV 8840 i

in 1986,' these changes were evaluated in accordance with 10CFR50.59. The evaluation concluded that no USQ was created. However, based on a recent review of this 10CFR50.59 change it has been determined that the evaluation of this change should have concluded that this was a USQ.

1998 Safety Assessment The 1986 changes to the EOPs and the Millstone Unit 3 design basis to eliminate the direct injection function provided by the RSS was a significant modification of system operation. The modification changed the RSS direct core cooling safety function from a redundant safety function to a contingent safety function in the event of failures of the primary injection flow paths. The closure of the direct injection valves was in response to the system pre-operational testing which identified excessive tube vibration in the RSS heat exchangers when they were subject to flow exceeding 4600 gpm. i Nonetheless, the modification did not eliminate direct cold-leg injection from the EOP procedures. It was recognized that cold leg direct injection would be needed in case )

both the SlH and the CHS pumps failed during long-term recovery actions after a )

LOCA, in a beyond-the-design -basis situation. The valves remained intact '

operationally, but their function was changed to the isolation of the direct injection flow paths. Subsequent evaluation of this modification concluded that the EOP guidance ~ ~

would have supported the core cooling function for the mitigation of a limited passive

' failure. The safety evaluations, written in 1986 to support this change, did not specifically address the limited passive failure scenario. .

The original design of the RSS system included direct injection during the recirculation

.- phase, without operator action. The design should have allowed full flow through the  !

RSS heat exchangers without excessive tube vibration. The 1986 change used l operator action to isolate the direct injection flow paths to avoid the high vibration l problem. After 1986, direct injection was available, but required operator action.

t Because of the additional operator actions, elimination of RSS direct injection became i- an unreviewed safety question.

The evaluation has determined that the change was safe. This was because:

. The results of the design basis analyses were verified to be acceptable without the direct injection in that the modified alignment delivered sufficient flow to meet the long term cooling requirements after a LOCA, and the results of the containment l

9

U.S. Nucl: r Regulatory Commission B17044\ Attachment 3\Page 4 analysis show that the design basis of maintaining subatmospheric containment pressure was unchanged.

j The increase in probability of malfunction of equipment due to the use of operator actions is acceptable. The EOP's provide clear guidance on the isolation of the

' direct injection, as well as, the re-establishment of cold leg direct injection if required. The operators have been fully trained on these procedures. Therefore, the likelihood of failure is low.

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Attachment 4 .

Millstone Nuclear Power Station, Unit No. 3 1 Significant Hazards Consideration and Environmental Considerations 4

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March 1998 i.

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4 6 U.S. Nuclear Regulatory Commission B17044\ Attachment 4\Page 1 Millstone Nuclear Power Station, Unit No.3 Significant Hazards Consideration and Environmental Considerations

. Significant Hazards Consideration Northeast Nuclear Energy Company (NNECO) has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve an

'l SHC because the revision would not: l w , .. , .

1. Involve a significant increase in the probability or consequence of an accident previously evaluated. .

The change to the Emergency Operating Procedures (EOP) to eliminate the use j of Recirculation Spray System (RSS) direct injection during cold and hot leg recirculation does not effect the probability of any accident. The elimination of the requirement to have RSS directly injection into the reactor coolant system did not increase the consequences of the previously evaluated accidents These -

consequences were evaluated based on very conservative assumptions-concerning the containment pressure after the design basis Loss of Coolant Accident (LOCA), containment integrated leakage rates, and the fraction of the  !

sprayed volume. None of these assumptions were affected by the elimination of l the direct cold-leg injection, j Therefore, the proposed revision does not involve a significant increase in the probability or consequeqe of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident -

previously evaluated. . -

The modification to the RSS did not create the possibility of a new or different accident from those previously analyzed. The change involved elimination of the  !

direct injection flow path from the design basis of the system but did not involve  !

physical modifications to the system itself. The operability of the affected valves within the direct injection alignments remained unchanged and these paths were still available to the operators for contingencies beyond the design basis. The

- EOPs provided clear and explicit guidance to that effect.

l Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated. l

U.S. Nuctsar Regulatory Commission B17044\ Attachment 4\Page 2

3. Involve a significant reduction in a margin of safety.

1 In considering the impact on the margin of safety as defined in the bases of the Technical Specifications, the impact of the change on the design basis analysis of the fission product barriers must be evaluated.

The minimum Emergency Core Cooling System flow requirement for long-term core cooling is that the modified alignment deliver sufficient flow to satisfy the inventory lost to the boil off in the vessel due to the decay heat and the extended boiling from hot metal in the downcomer agd the lower plenum. The analysis determined that these requirements were being met.

The elimination of the direct injection resulted in a flow reduction through the RSS heat exchanger, from approximately 4000 gpm to 1200 gpm, thus reducing the rate of the heat transfer from the containment to the service water system.

The design basis of the containment heat removal systems (circa 1986) is that the containment pres?ure will decreaseio subatmospheric within one hour after the Design Basis Acciaent to compensate for the reduction in heat removal from the containment, a smaller allowable RSS pump degradation was assumed in -

the revised containment analysis. The original RSS pump performance curve was based on a 10 percent reduction in developed head from the design curve.

For the modification, a 5 percent reduction was used. The results of the analysis show that with these changes the design basis of maintaining subatmospheric containment pressure was met.

Based on the above, elimination of the direct injection did not reduce the margin of safety because there was no violation of the acceptance limits and no ~

weakening of the protective boundaries.

, Therefore, the proposed revision does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

Environmental Considerations NNECO- has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations'. The proposed revision does not involve an SHC, does not significantly increase the type and amounts of effluents that may be  ;

released offsite, nor significantly increase individual or cumulative occupational l l radiation exposures. Base:' on the foregoing, NNECO concludes that the proposed i l

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U.S. Nucisar R:gul: tory Commission 817044%ttachment 4\Page 3--

revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion

- from the requirements for environmental review.

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MAY - 7 1998 I Qocket No. 50-423 817212 Re

10CFR50.90 10CFR50.59 (a)(2)

U.S. Nuclear Regulatory Commission l Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 3 Proposed License Arnendment Request Recirculation Spray System Direct injection Change (PLAR 3-98-1)

Resoonse to Reauest for Additional Information Northeast Nuclear Energy Company (NNECO), in a letter dated March 3,1998, f'

proposed an amendment to Chapter 6 of the Millstone Unit No. 3 Final Safety Analysis Report. The NRC in r. letter dated May 7,1998, requested additional information to support their review of s submittal. Attachment 2 contains NNECO's responses to the NRC questions.

Attachment i identifies that no commitments are contained within this letter. If the NRC Staff should have any questions or comments regarding this submittal, please contact Mr. D. Smith at (860) 437-5840.

Very truly yours, l NORTHEAST NU EAR ENERGY COMPANY N N M. H. Brothers Vice President - Operations cc- H. J. Miller, Region i Administrator W. D. Travers, Ph.D., Director, Special Projects Office J. W. Andersen, NRC Project Manager, Millstone Unit No. 3 A. C. Come, Senior Resident inspector, Millstone Unit No. 3 Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127 fEHJ24 NFV 12M I

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Docket No. 50- 423 f B17212 )

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l Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Proposed Ucense Amendment Request Recirculation Spray System Direct injection Change (PLAR 3 98-1)

Response to Request for Additional Information NNECO's Commitments -

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i May 1998 l

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U.S. Nucircr R:gul: tory Commission 817212Att: chm:nt 1\Pige 1 Enclosure Ust of Regulatory Commitments The following table identifies actions committed to by NNECO in this document. Please i notify the Manager - Regulatory Compliance at the Millstone Nuclear Power Station Unit No. 3 of any questions regarding this document or any associated regulatory commitments.

Commitment Committed Date or Outage

NONE

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l i Q2cket No. 50-423 l B17212

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l Attachment 2 r

. Millstone Nuclear Power Station, Unit No. 3 Proposed Ucense Amendment Request Recirculation Spray System Direct injection Change (PLAR 3-98-1)

Response to Request for Additional Information i

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I May 1998 l

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.f$Y. 8.1998 '332SPM MILLSTerE BUDGETS E6044ce464 NO.654 P.8 U.S. Nuclear Regulitory Commission <

B17212 Attachment 2\Page 1  ;

l Question 1 -

In your. reanalysis, the assumed pump degradation was reduced frora 10 percent to l 5 percent. Discuss the procedures you have in place for the monitoring of pump  ;

degradation and how these procedures support this change. l NNECO's Resoonse to Question 1 The revised degraded pump curves (flow versus pressure differential) were utilized

[ during the performance of Surveillance Procedure EN - 31121. This surveillance l

! procedure was utilized to verify that the pump did not degrade beyond the flow i i

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assumed in the accident analysis.

p Question 2 in your submittal, the proposed change will still require direct cold leg injection in case of a long-term passive failure. In the 1986 Final Safety Analysis Report change, you

! completely deleted references to the Recirculation Spray System (RSS) direct injection path. Please include a discussion of this configuration for completeness or justify its exclusion.

I NNECO's Response to Question 2 L

NNECO indicated in the February 16,1998 submittal, that the 1986 modification did not  ;

eliminate direct cold leg injection from the Emergency Operating Procedures (EOP). It was recognized at that time, that cold leg direct injection would be needed in case both i Safety injection (SlH) and the Charging (CHS) pumps failed during lo1g-term recovery I actions after a Loss of Coolant Accident (LOCA), in the beyond design basis situation.  !

Review of the 1986 Final Safety Analysis Report (FSAR) change made to support the -!

modification indicates that it was thought that sufficient redundancy existed in the j Emergency Core Cooling System (ECCS) components and associated flow paths to accommodate a limiting long term passive failure without reliance on the direct injection flow path. The safety evaluations, written to support the 1986 change, did not

' specifically address the limited passive failure scenario.

A subsequent 1998 evaluation of the 1986 modification concluded that the EOP guidance would have supported the core cooling function for the mitigation of a limited passive failure.

FSAR Section 6.3.2.5, ECCS System Reliability, addresses limited passive failure in

1some detail. The structural failure of a static component that limits the component's effectiveness in carrying out its long term design function is considered a passive failure. Examples include cracks in pipes, sprung flanges, valve pack ng leaks or pump I

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U.S. Nucirrr Rrgulltory Commission I B17212 Attachment 2\Paga 2 ]

seal failures. A single failure analysis is presented in Table 6.3-6. FSAR Table 6.3 6 has been enhanced by a 1997 FSAR change (MP3-97-0569) to addresses the Limiting Passive Failure and the use of the direct injection flow path. Accordingly, use of the direct injection flow path for a Limited Passive Failure is included in the current FSAR.

Question 3 On page 6.2-50a, you deleted the discussion on evaluation of Not Positive Suction Head (NPSH) for the recirculation mode of the RSS. Confirm that the recirculation spray mode is the limiting case for required NPSH.

NNECO's Response to Question 3 l

i r An evaluation of the RSS pump NPSHe was performed for the RSS spray, cold leg recirculation and hot leg recirculation modes. It was confirmed that the limiting case f was the spray mode. l it should be noted that dur'ng the preparation of the response to this question, review cf PLAR 3-98-001 (i.e. the existing 1986 FSAR change request) determined that the NPSH data provided was inaccurate. In the 1986 time frame, additional analyses were completed to evaluate the impact of LOCA induced containment sump screen blockage on RSS pump NPSH. However, the results of these analyses were not j

incorporated into the FSAR prior to the elimination of the direct injection mode of ECCS cold leg recirculation. The correct values for the Spray Phase of operation are documented in a February 26,1986 calculation revision and are provided below; i

ist Stage Imp El(ft) -47.33 Floor Water Level (ft) -23.1 i Z (ft), El HeMI 24.23 L H (ft), Pipe Loss 10.96  !

. Z-H (ft), NPSHavail 13.27 Q (gPm) 4475 NPSHrea 11 Margin (ft) 2.27 Question 4 '

in your submittal you stated that the results of the design basis analyses were verified to be acceptable without the direct injection in that the modified alignment delivered sufficient flow to meet the long-term cooling requirement after loss of Coolant Accident, and the results of the containment analysis show that the design basis of maintaining a

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B17212Attrchment 2\Paga 3 y ,

substmospheric containment pressure was unchanged. Discuss in more detail the

evaluations / analyses performed to support these conclusions.

NNECO's Response to Question 4 l

Lono Term Coolino Requirements

. The minimum ECCS flow requirement for long-term core cooling following a LOCA is

. the alignment in the rec ircu lation mo de without direct injection on minimum Engineered Safety Features (ESF). Minimum ESF is defined as one charging pump and one ECCS pump. Enough ECCS flow must be provided to exceed RCS boil-off in accordance with 10CFR50 Appendix K requirements and with Westinghouse internal criteria. To consider whether these requirements I criteria were met, three basic cases were considered:

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1) cold leg recirculation mode following a cold leg break, L 2) cold leg recirculation mode following a hot leg break, and

3) hot leg recirculation mode following a hot leg break .

The most limiting break was determined to be the cold leg break white the plant is in the cold leg recirculation mode. For this case, at the time when recirculation would be initiated, Reactor Coolant System (RCS) boil-off was calculated to be 68.9 lbm / sec while total recirculation flow was 148.8 lbm I sec. Of this total recirculation flow,111.6 lbm / sec is calculated to enter the core while 37.2 lbm / see was calculated to be lost out of the break. Since the total recirculation flow to the core is grea;er than the RCS boil-off with margin, the proposed ECCS alignment is acceptable.

The RCS boil-off is calculated based on ANS decay heat in accordance with 10CFR 50 Appendix K at 20 minutes. The calculation of bollaff is conservative since the l

' minimum time to recirculation is greater than 30 minutes and the boil-off requirements

1 are reduced as time elapses.

i The long term cooling analysis assumes no direct mixing of the recirculation flow with the core until after the switchover time to hot leg recirculation. This assumption allows l j the maximum amount of boron to build-up in the core. De proposed alignment has no effect on the time when hot leg switchover should occur and has no effect on the amount of boron calculated to build-up in the core.

Based on the. above analysis work, all requirements for long term cooling were determined to have been met.

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