3F0298-03, Summarizes Key Elements of App R Restart Issue Work Scope & Provides NRC W/Clarifications of Two SERs Which Resulted,In Part,From Mods Made During Present & Previous Outages

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Summarizes Key Elements of App R Restart Issue Work Scope & Provides NRC W/Clarifications of Two SERs Which Resulted,In Part,From Mods Made During Present & Previous Outages
ML20202F849
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 02/17/1998
From: Grazio R
FLORIDA POWER CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0298-03, 3F298-3, NUDOCS 9802200004
Download: ML20202F849 (29)
v • d • e


Text

_ . _ _ _ _ _ _ _ _ _ _ _ __ _ _ ______ ___ _____ ____ _ -_

Florida Power

==

February 17,1998 3F0298-03 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 0001 Sub)ct: 10 CFR 50, Appendix R Restart issue Closure and Clarifications of Two NRC Safety Evaluation Reports (SERs) for Sections Ill.G and III.L

References:

1. FPC to NRC letter, 3F0997-02, dated September 25, 1997 Response to NRC Request for Additional Information Related to Restart Matrix Item D ll, Appendix R (TAC No. M98943)
2. NRC to FPC letter, 3N0183 05, dated January 6,1983, Crystal River Unit 3, Safety Evaluation Report, Appendix R to 10 CFR 50
3. NRC to FPC letter, 3N098511, dated September 11, 1985, Safety Evaluation Report, Control Complex Dedicated Cooling System for Post.

Fire Alternate Shutdown Capability (Appendix R Section III.G.3),

Crystal River, Unit 3

4. FPC to NRC letter, 3F0997-06, dated September 18, 1997, Crystal River Unit 3 Resolution Plan for Nuclear Regulatory Commission Information Notice (IN) 92-18. " Potential for Loss of Remote Shutdown Capability During a Control Room Fire"
5. FPC to NRC letter, 3F1297-37, dated December 29,1997. " Reply to Notice of Violations, NRC Inspection Report No. 50-302/97-14. NRC to FPC letter, 3N1297-08, dated December 4,1997"

.M , s r* "#

Dear Sir:

In Reference 1 Florida Power Corporation (FPC), Crystal River Unit 3 (CR 3), provided the Nuclear Regulatory Commission (NRC) with a summary of Appendix R restart issue work scope descriptions. Cic,ure of these work tasks, including field work, testing, and procedural revisions, is now complete. The purpose of this letter is to summarize key elements of the Appendix R restart issue work scope and to provide the NRC with clarifications of two Safety Evaluation Reports (References 2 and 3) _which resulted, in part, from modifications made during this and previous outages. ,o u ,

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  • CrystalRiver. Floride 34425-6700 * (362)796 6480 e QO A IkMe Progress Corrnpany

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' U.S. Nuclear Regulatory Commission 3F0298 03 Page 2 The CR 3 Appendix R Fire Study has been reviewed and revalidated to assure compliance with Appendix R requirements. A major part of this review was conducted to assure compliance -

with 10 CFR 50 Appendix R, Sections Ill.G.3 and Ill.L for shutdown in the event of a fire in the Main Control Room (MCR), Cable Spreading Room (CSR), or the Control Complex IIVAC Equipment Room. This review identified the need to provide: additional electrical .

Isolation for the MCR and the Remote Shutdown Panel (RSP), reroute of cables out of the CSR, other electrical changes to selected safe shutdown equipment, and revisions to procedure AP 990, "Shutdo.vn From outside the Control Room.*'

The Appendix R Fire Study, AP 990, and OP 880, " Fire Service System," have been revised atxl approved under the 10 CFR 50,59 process, in Reference 4, FPC committed to a resolution program for Information Notice (IN) 92-18,

" Potential for Loss of Remote Shutdown Capability During a Control Room Fire," which included electrical modifications to numerous valves and a review of post fire shutdown flow paths to determine required procedural revisions. The valves originally plarmed for electrical modifications were provided in Attachment B of Reference 4. It is noted that during the final engineering and procedure reviews, five additional valves: MUV 24, MUV 26, MUV-567, MUV 53 and MUV 257 were modified, and three valves: MUV-40, MUV-41, and MUV-505 did not require modification. A revised / final table of the valves that were modified under the IN 92-18 resolution program is provided in Attachment A of this document.

Clarifications of Two NRC Safety Evaluation Reports (SERs)- Sections III.G.3 and Ill.L Restart related- Appendix R modifications and revisions to AP-990 and OP 880 meet the following criteria:

1. Compliance with applicable sections of Appendix R, Ill G and III.L,
2. _ Compliance with CR-3 License Condition 2.C.(9) "The licensee may make changes to the approved fire protection program. without prior approval of the Commission only if tho.e changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire "
3. Processed through the onsite 10 CFR 50.59 process with no Unreviewed Safety Questions (USQ),' and
4. Do not materially change or alter the conclusions in NRC Safety Evaluation Reports.

Some revisions to the CR-3 Appendix R compliance strategy were required in order to correct design errors, adequately modify AP 990 and OP-880, and implement the committed IN 92-18

' The NRC issued violation 50 302/97 14 13 regarding improper corrective action on determining the proper normal operating position of DHV 34 and DHV 35. FPC conducted a 10 CFR 50.59 review of these valves in the closed position and determined that this was a USQ. In Reference 5, FPC committed to submit a Licensing Amendment Request that will resolve this USQ. Additional information is provided in Attachment B under item G.11

R q

U U.S. Nuclear Regulatory Commission 3F0298-03 i Page 3 '

  • )

.l

-%lution program. Further, under ~ the CR 3 Thermo Lag Resolution Program, circuit -

reroutes, Mecatiss installations, and Re-analysis changes crediting offsite power in selected 111.G.1 and 111.G.2 fire areas'were also completed. These changes also resulted in changes to Appendix R compliance strateg!cs.-

Attachments B and C have been developed to show comments and clarifications resulting from changes in compliance strategies pertaining to the SERs it References 2 and 3. FPC requests the _NRC review these Attachments and either append the current SERs or issue revised SERs. -

This process will assure that the licensing documents of record for Appendix R Sections 111.0-anxi lil.L are current and accurately refket CR 3's compliance with tim ccde, if you have any questions regarding this letter, please contact Ms. Jherry Bernhoft, Manager, Nuclear Licensing at (352) 563-4566.

Sincerely, yM N R.E. Grad Director, Nuclear Regulatory Affairs REG /jnb

- Attachments xc: - Regional Administrator,'. Region 11

' Senior Resident inspector:

NRR Project Manager i

+

Attachment A U.S. Nuclear Regulatory Commission Page1ofI 3F0298-03 Motor Oper _d Valves Modified to Protect Against Information Notice 92-18 Failures Normal Potential / Adverse Shutdown Valve Tag Valve Function Valve Fire Scenario IN-92-18 Condition Number Position Valve Failure Position DilP-I A Suction From BWST G osed Control Rm / Other Fire Areas Ogn and Incierable liot Shutdown DIIV-34 flot Shutdown DlIV-35 DIIP-1B Suction From BWST C osed Control Rm / Other Fire Areas Ord and intierable Open Other Fire Areas OgrJOosed Intier.ble flot Shutdown EFV-Il EFP-2 Isolation Valve EFP-1 Isolation Valve Open Control Rm / Other Fire Areas Open/Gosed hnierable flot Shutdown EFV-14 EFP-2 Isolation Valve Open Other Fire Areas Open/Gosed incierable llot Shutdown EFV-32 Open Control Rm / Other Fire Areas Open/Gosed hmierable flot Shutdown EFV-33 EFP-1 Isolation Valve Supply Block Valve to EFP-2 Open Other Fire Areas Gosed and Intigrable flot Shutdown MSV-55 MSV-56 Supply Block Valve to EFP-2 Open Other Fire Areas Gosed and incierable liot Shutdown IIPI CTRL VLV to RX Inlet Ixop A Closed Control Rm / Other Fire Areas Open/Cosed Imier ble Ilot Shutdown MUV-23

" VI V o t RX Inlet Loop A C osed Control Rm / Other Fire Areas Open/Cosed Inoperable Hot Shutdown MUV-24 11PI O osed Control Rm / Other Fire Areas Open/ Closed Inoperable flot Shutdown MUV-25 IIP 1 G k j LV to RX Inlet leop B .

IIPI CTRL VLV to RX Inlet Loop B G osed E Control Rm / Other Fire Areas Open/Gosed Inoperable Ilot Shutdown MUV-26 IIPI CTRL VLV to RX Loop A Open Control Rm / Other Fire Areas Gosed and Inoperable flot Shutdown MUV-27 Make Up Pump Recire. Iso. Open Control Rm / Other Fire Areas Gosed and Irmierable Ilot Shutdown MUV-53 Open Control Rm / Other Fire Areas Closed and Inoperable llot Shutdown MUV-58 High Pressure Suction from BWST BWST to Make Up Pumps Isolation Closed Other Fire Areas Closed and Incierable flot Shutdown MUV-73 Open Control Rrr. / Other Fire Areas Gosed and Inograble Ilot Shutdown MUV-257 Make Up Pump Recire. Iso.

Open Control Rm Fire Open and incierable flot Shutdown MUV-567 Letdown Cooler Outlet Main Iso O osed Control Rm / Other Fire Areas Gosed and Inty rable Cold Shutdown DlIV-4 DilR Outlet Isolation DIIV-5 LPI RB Isolation from DHP-I A G osed Control Rm / Other_ Fire Areas Gosed ami hnici.ble Cold Shutdown G osed Other Fire Areas Cosed and irggrable Cold Shutdown DlIV-6 LPI RB Isolation from DlIP-1B C osed Control P.m / Other Fire Areas OperJOosed Irn5(rabic -

Co!d Shutdown DlIV-39 DH Isolation to DHP-I A G osed Control Rm / Other Fire Areas Open/Gosed Inoperable Cold Shutdown DlIV-40 Dil Isolation to DHP-IB G osed Control Rm / Other Fire Areas Gosed am! Inoperable Cold Shutdown DifV-41 DHR Outlet RB Isolation Closed Control Rm / Other Fire Areas Open and Inoperable Cold Shutdown DlIV-42 DH Suction from RBS to DHP-I A G osed Other Fire Areas Open and Inoperable Cold Shutdown DIIV-43 Dil Suction from RB3 to DHP-1B DHV-110 Outle: of DHIIE-I A Open Control Rm / Other Fire Areas Gosed and Intierable Cold Shutdown Open Other Fire Areas Gosed and Inoperable Cold Shutdown DlIV-111 Outlet of DlIHE-1B

I I

  • U.S. Nuclear Regulatory Commission Attachment H 3F0298-03 Page 1 of 22 Comments and Clarifications on January 6,1983 SER COMMENTS AND CLARIFICATIONS NRC SER DATED JANUARY 6,1983 ON JANUARY 6,1983 SER

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CRYSTAL RIVER UNIT 3 SAFETY EVALUATION REPORT APPENDIX R TO 10 CFR 50 INTRODUCTION On February 19,1981, the fire protection rule for nuclear power plants,10 CFR 50.48 and >

Appendix R to 10 CFR 50 became effective.

The rule required all licensees of plants licensed prior to January 1,1979, to submit by March 19,1981: (1) plans and schedules for meeting the applicable requirements of Appendix R (2) a design description of any modification proposed to provide alternative safe shutdown capability pursuant to Section Ill.G.3 of Appendix R, and (3) exemption requests for which the tolling provisions of Section 50.48(c)(6) was to be invoked.

Section 111.0 of Appendix R, " Fire Protection of Safe Shutdown Capability" was retrofitted to all pre-1979 plants regardless of previous SER positions and resolutions.

By submittal dated October 29,1982, the licensee provided the description of the proposed modifications to the Crystal River plant to meet the requirements of Appendix R to 10 CFR 50, Section Ill.G. The proposed modification _will also resolve the open items _ _ ,

NOTE: Comments and clarifications are provided for SER statements that are glo_uh!c_unje x ;M

  • U.S. Nuclear Regulatory Commission Attachment B  ;

3F0298-03 Page 2 of 22 Comments and Clarifications on January 6,1983 SER 4

NRC SER DATED JANUARY 6,1983

. ANS ON JANUAR) 6,1983 SER concerning alternative shutdown from our previous (SER Page 2 of 13)

SER.

~ ~ "~~"'"~~~~~~~ ~ ~~~~~"~"

AdditionaIihiShia"tI$5~a'57c~iEiEcItIini$i~~

provided by submittals dated June 30.19R, Following issuance of the SER on January 6, <

pgqgnber 9.1982. and Destmist 17.1.98L 1983, FPC sent the following letters to the Our previous fire protection Safety Evaluation NRC providing comments and clarifications:

Report (SER) dated July 27,1979 concerning 1) FPC to NRC letter,3F028311, dated llranch Technical Position APCB 9.51 Fehmary 11,1983, " Appendix R to 10 CFR indicated that in certain plant areas redundant 50, Fire Protection" systems could be damaged by a single fire 2) FPC to NRC letter,3F058511, dated which would affect safe shutdown. The hiay 17,1985, " Safety Evaluation Report for licensee was requested to provide alternative Crystal River Unit 3 on Appendix R to 10 shutdown capability for areas which could not CFR 50 Section Ill.G." ,

be protected by fire barriers, fire detection 3) FPC to NRC letter,3F078512, dated July and fire suppression systems. 17,1985, " Alternative Shutdown Capability"

4) FPC to NRC letter,3F0185 01, dated +

Our evaluation of the licensee's submittals January 2,1985 " Control Complex follows: Dedicated Cooling System"

' ~ ~~~~ ~ ~ ~ ~ ~ ~ ~

if5fi551S USI!Ii FOR POST FIRE SAFE SilUTDOWN A. Systems Reauired for Safe Shutdown Safe shutdown is initiated from the control rocm by a manual scram of the control rods. Reagletro_elant invento_or and _ Ctl The Appendix R make up Dow path and gactivity conimLars_maintainss!_by_ene_of control philosophy for the post fire control of thf_1hIqcanaMUDJmsLpyrification pitmps the hiakeup pumps have changed, taking suction from the borated water storage tanks (BWST). Instead of one hiake up flow path as discussed in the October 29,1982 submittal, four injection paths are now available post fire via the injection valves htUV-23,24,25, or 26.

During the Appendix R review of electrical isolation of the hiakeup system, it was detern*!ned that the makeup path through hiUV-31 and hiUV-27 was not properly isolated _ and/or separated to use as a dedicated NOTE: Comments and clarifications are provided for SER statements that are dpyklg,undgrlined.

. U.S. Nuclear Regulatory Commission Attachment 11 3F0298-03 Page 3 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983

' ^ ^

ON JANUARY 6,1983 SER injection path. This was reported to the NRC in LER 97 35 00. The four Makeup injection valves were selected as the dedicated l Appendix R Section Ill.L makeup paths. j 13ypass switches have been installed for these valves to assure their post fire operability following a fire in the Control Room, Cable Spreading Room, or Control Complex IIVAC room. Emergency lighting has been installed to provide post-fire operator access to the switches. In addition, MUV-23,24,25, and 26 have also been modified to assure their isolation from the above fire areas and climinate failures of the type identified in IN 9218, " Potential for Loss of Remote Shutdown Capability During a Control Room Fire".

These modifications meet the requirements of 10 CFR 50, Appendix R, Sections 111.0.3 and Ill.L. and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR 3 License Condition 2.C.(9), and do not materially change or alter the conclusions in NRC Safety Evaluation issued on January 6,1983.

AP 990, " Shutdown From Outside the Control Room", was revised to trip or verify tripped all three Makeup pumps. This aa;on is performed to minimize the possibility of pump damage resulting from potential spurious actuations in the suction and/or-discharge path. Following verification of suction and discharge valve alignment, the desirea Makeup pump is started from the appropriate 4 KV ES Switchgear.

Procedure AP-990 has been revised to implement this alternative control philosophy for aligning post-Grc makeup, verified on the simulator, Deld walked down to assure staffing requirements, and these changes have NOTE: Comments and clarifications are provided for SER statements that are dpublq,unded{ ped.

c

. U.S. Nuclear Regulatory Commission Attachment B 3F0295-03 Page 4 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983

^ '

, ON JANUAR) 6,1983 SER

) been processed through the onsite 10 CFR 50.59 process.

The revision to AP-990 for establishing and maintaining Makeup flow meets the requirements of 10 CFR 50, Appendix R,

, Sections 111.0.3 and Ill.L. and "would not 3 adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR-3 License Condition 2.C.(9), and does not materially change or alter the conclusions in NatC Safety Evaluation issued on January 6,1983.

Primary system pressure is maintained by the press _utiter_hesucts, the makeup pumps Clarification of the Pressurizer IIcaters and taking suction from the BWST ppmhitiq.d Spray is provided in Section Q below.

(SER Page 3 of 13) with leids now. athLths oressurizer Clarification of letdown flow isolation is gnraylqold shukinMLealyt provided in Section Q below.

For hot shutdown, decay heat removal is accomplished by the emergency feedwater pumps supplying water to the steam <

generators from the ERDdensate stetale Clarification of the dedicated alignment to the ljig Emergency Feedwater Tank is provided in gdQ below.

. - Iht.almasnhelicJ#manlyngre used to Clarification of the use and sequencing of the gniove healitenhs steam generators. Atmosr'Bric Dump Valves (ADVs) is For cold shutdown, decay heat removal is provided in Section Q below, accomplished by the decay heat removal system in conjunction with the decay heat closed cycle cooling system and the decay heat seawater cooling system.

Support for the above systems is provided by the reactor building cooling system, the nuclear service closed cycle cooling system, the nuclear service seawater

_, cooling, system, essential area llVAC NOTE: Comments and clarifications are provided for SER statements that are dpuMg_undgilingst.

. U.S. Nuclear Regulatory Commission Attachment B 3F0298 03 Page 5 of 22 .

Comments and Clarifications on January 6,1983 SER NRC SER DA1.ED JANUARY 6,1983 COMMENTS AND CLARIFICATIONS ON JANUARY 6,1983 SER systems and the 125 volt DC power system.

-. .-- .... .  : .~. . .. ........ . . . . . . . . . . . . . . . ..-

jg,thc.rvent of a loss of offsite nov'er. the Q For fire areas that comply with 10 CFR diesel cencrators will be utilized to nower 50, Appendix R, Sections 111 G.1 and Ill.G.2, ,

ths.A@Dutdown systems. and where the availability of offsite power has been demonstrated to be unaffected as a result of the fire, the use of offsite power may be  ;

credited for safe shutdown.

The commitment to assume a loss of offsite power was made in response to the Clarification of Generic letter 81-12. " Fire Protection Rule (45 FR 76602, November 19, 1980)", FPC letter 3F1082-32, dated October 29.1982 Attachment 3, Section 11,

" Assumptions" Number 4, FPC states " Loss of offsite power occurs simultaneously with, or subsequently to the fire". This was an over commitment, as 10 CFR 50, Appendix R, Sections Ill.G.1 and Ill.G.2 do not require this assumption, nor did the original ve d on of GL 81 12.

I I A key part of FPC's Thermo-Lag Resolution Program employs re analysis of the fire area safe shutdown analyses to teraove reliance on Thermo-Lag protected raceways. One of the re analysis strategies is to credit the availability of offsite power. FPC has held four meetings with the NRC wheNin our intent to take credit for the availability of offsite power was presented and discussed (February 28,1995, October 24,1995, May 20,1997, and June 11,1997). During these meetings the NRC Fire Protection Staff concurred that 10CFR50 Appendix R, Section Ill.G.1 and 111.G.2 compliance does not require the assumption of a loss of offsite power unless the fire or its effects can induce the loss of offsite power.

1n FPC,,t.o NRC letter 3F1295-05, dated NOTE: Comments and clarification: are provided for SER statements that are pppMp_undgjLnqd.

U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 6 of 22 Comments and Clarifications on January 6,1983 SER 4

NRC SER DATED JANUARY 6,1983 AND GAREAM.W ON J ANUARY 6,1983 SER December 21,1995 FPC submitted the CR 3 Thermo Lag Resolution Program. This submittal stated, under the re analysis section at item (b), that Thermo-Lag reduction would j be accomplished: "using offsite power sources for post fire safe shutdown." This analysis w6.:ompleted and implemented during the recent outage for Thermo Lag in fire areas where the availability of offsite power was demonstrated to be unaffected by the fire or the affects of the fire.

This revis( ' commitment meets the requirements of 10 CFR 50, Appendix R, Sections 111.0.1 and 111.0.2, and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR-3 License Condition 2.C.(9), and does not materially change or alter the conclusions in NRC Safety Evaluation issued on January 6,1983.

The shutdown system will be monitored and controlled from the control room or the dedigatnLshutdo.x1Lnod, local The panel is now referred to as the Remote control stations, switchgear, and motor Shutdown Panel, control centers.

(SER Page 4 of 13)

B. Areas Where Alternative Safe Shutdown is Proposed The licensee's proposed modification provides alternative shutdown capability for the control room, cable spreading room and control complex IIVAC equipment r00m.

C. Remainine Plant Areas By letter dated October 29,1982, the licensee stated that the Crystal River Unit

3 plant would be modified to comply with

_,,,, Section 11LQ2 of Appendix R except_as Other non-Alternative Shutdown areas are in NOTE: Comments and clarifications are provided for SER statements that are dquhle_ugdqtynes.

l

. U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 7 of 22 i Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983

' W R ANS -

ON JANUARY 6,1983 SER noted above. Plant modifications are compliance with ill.G.2. '

needed to comply with Section Ill.G.2.

The proposed plant modifications include: ,

(1) the addition of three-hour barriers to separate redundant equipment,(2) relocation of cabling, (3) locking out power to valves to prevent fire induced spurious operation arn! (4) providing one.

hour rated barriers enclosing cable trays and conduit.

D. Alternative Safe Shutdown System The alternative shutdown capability will consist of a dedicated shutdown panel, redundant dedicated shutdown auxiliary equipment cabinets, redundant dedicated shutdown relay cabinets, local control stations, and motor control centers.

~ ~

~~1[x_designMiie~aIEative shpN$$n M Es~ stated iriiiIe~55E~a255I6~CFR 50, gashility provides the Appendix R Sections Ill.G.3 and Ill.L.

cornpliance is achieved by providing one train (SER Page 5 of 13) of systems needed for safe shutdown. The CR-3 Remote Shutdown Panel and procedures ggahuity of ooeratine at least ons_tra!ILef meet this requirement primarily with Train B lhe systems needed for safe shygg} equipment. The Remote Shutdown Panel listed in Section A above. In aMilMILibs contains additional Train A capability and desien provides for redunkngyjor most process monitoring which has been provided systems used for safe sigios The for operator flexibility and exceeds the dedicated shutdown panel (DSP) will be requirements of lil.G.3 and III.L. Not all the located in a fire area designated as the redundant capability and control provided by _

dedicated shutdown room. One set of the panel, for functions over and above the cabinets, a dedicated shutdown auxiliary minimum required, have been protected for equipment cabinet and a relay cabinet, is fire scenarios, llence. the CR-3 design meets located in the switchgear room "A" and the requirements of Appendix R lli.G.3 and the other set of cabinets is located in Ill.L with one train of shutdown equipment, switchgear room "B" The design of the The original October 29,1982 submittal and dedicated shutdown panel provides the SER issued by the NRC on January 6, electrical isolation from the control room 1983 require clarification that the CR-3 design by transfer switches located in the meets the Appendix R 111.G.3 and Ill L dedicated shutdown auxiliary equipment requirements with one train of safe shutdown cabinets, equipment. This same SER clarification was

._ _ . _ . _ . . _ _...._ . submitted to the NRC in FPC letter 3F0785_

NOTE: Comments and clarifications are provided for SER statements that are siquhlg.underlinSA.

l U.S. Nuclear Regulatory Commission e Attachment H 3F0298-03 Page 8 of 22

, Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983 A A A IO.NS ON JANUARY 6,1983 SER 12 dated July 17,1985. No subsequent correspondence was received from the NRC i on this SER clari0 cation, hence it is again

, provided for NRC clari0 cation. The CR 3 design is in compliance with 10 CFR 50, Appendix R. Section 111.0.3 and Ill.),

requirements with one train of safe shutdown equipment, and does not materially change or t alter the conclusions in NRC Safety 4 Evaluation issued on January 6,1983.

~ ~

~~~555.cslinIf'the didicatedl55En y I5urirlg the rIcent $((Er~di~x~ E~~~~~~~~

oanel provides electrical isolation from the compliance review, it was determined that a gnatrol room by trarpfer switches located clarification should be made as described in the dedjettedltmtdown auxiliary below.

eauloment cahlatis Thus. a fire in either the dedicated shutdown room or thg A Gre in the Remote Shutdown Panel room, control room will not result in loss of CC-108102, would result in one train of control of the systems needed_for saf3 shutdown equipment being available in the shutdown at the other locatlog A fire in control room, either of the switchgear rooms would result in loss of only one train of systems The CR 3 design is in compliance with 10 needed for safe shutdown. CFR .10, Appendix R. Section 111.0.1 and 111.0.2 requirements for all areas not using EVALUATION alternative shutdown under 111.0.3 and Ill.L.

A. Performance Goals and this clarification does not change or alter For post fire shutdown, the performance the conclusions in NRC Safety Evaluation goals of the alternative safe shutdown issued on January 6,1983.

capability will be met using (SER Page 6 of 13) the systems listed in Section A above.

Reactivity control will initially be provided by a manual scram of the control rods from the control room. Continued shutdown reactivity control is provided by the makeup and purincation pumps taking suction from the borated water storage tank (BWST).

' Two makeynrymps_ ate used for safg Q These statements are accurate, but only shutdown and are started at the __,,,_

one Makeup Pump is required.j'he following_

' NOTE: Comments and clarifications are provided for SER statements that are dqubje_urldSt[ined.

i l

. U.S. Nuclear Regulatory Commission Attachment B I 3F0298-03 Page 9 of 22 l Comments and Clarifications on January 6,1983 SER i

. t NRC SER DATED JANUARY 6,1983

^" ^ ^

ON JANUAR) 6,1983 SER  ;

switchnear. An optional _ third makeup additicnal clarification is also provided.

oumn is also available for thutdown.

FPC recently agreed to modify the CR 3 design to accommodate r. new type of valve failure that was identified in Information  ;

4 Notice 92-18. FPC submitted this commitment to the NRC in letter 3F0997 06, on September 18,1997. The NRC responded back in letter 3N1097 28, dated October 20, 1997 approving the CR 3 design approach to IN 9218. ,

r The statement in the SER regarding the Makeup Pumps remalm, accurate, however because of this new valve failure mode, if suction valves should fall, then two pumps

. would be available when operating in the Decay lleat Removal mode. Only one Makeup Pump is required for safe shutdown.

Makeup valves required to support safe shutdown are operable from the Remote Shutdown Panel or can be operated manually in the field. This is in compliance with Appendix R requirements.

The original design covered under our January 6,~ 1983 NRC SER accommodated fire induced spurious operation of safe shutdown valves. If a valve went open or closed the operator could operate the vala cither from the Remote Shutdown Panel u manually in the field. The new failure mode requires consideration for the valve spurious action bypassing the torque and/or limit switch and being damaged to an extent that the valve could not be repositioned remotely or manually, Numerous valves were rewired at the MCCs to remedy this failure rnode. Some of the valves were not required to be modified as other shutdown flow paths and methods were available which meet Appendix R requirements. Procedure AP-990 " Shutdown

.. _. ._._ From O. utside the Control Room" has been NOTE: Comments and clarifications are provided for SER statements that are doyMuwn,deflingd.

. U.S. Nuclear Regulatory Commission Attachment B 3FC298 03 Page 10 of 22 Comments and Clarifications on January 6,1983 SER OWms M GAREAM.W NHC SER DATED J ANUARY 6.1983 ON JANUARY 6,1983 SER revised to accommodate this new failure mode, verified on tne simulator, field walked down to assure staffing requirements, and these changes have been processed through fac onsite 10 CFR 50.59 process.

This new failure has been accommodated in the Appendit R design and the revision to AP 990, and are in compliance with 10 CFR 50 Appendix R Section 111.0.3 and Ill.L requirements, and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR 3 License Condition 2.C.(9), and do not materially change or alter the conclusions in NRC Safety Evaluation issued on January 6, 1983.

Ild3E[sliir~thc.1U3keyp_and The required Post l ire c$ilii$l5$i"Eakeup ""'

ggdfisat!ntuntemXt contielltd_Alths and Purification system valves are isolated MSL and on the Remote Shutdown Panel except for MUV-18 which is used to isolate the seal injection line, in the event that this valve fails, manual operator action is called upon in AP 990 to isolate seal injection manually.

Emergency Lighting is available for these actions. See also Section Q above for discussion of Makeup Flow Paths.

"~~ ~ ~ ' ~ ~ " ~ ~ ~ ~ ~ ~ ~ " ~ ~ ~

Reaci$i~c's$lant inientor[and priIiE"r5~~~

system pressure are maintained utilizing the makeup pumps and the lejdown_ lint fu6 This statement requires clarification as it The valvtslD_lbe,1cl(lown line_ ate implies that letdown is used in llot Standby gentrolled_nLibe_IlSE and during the cooldown to Cold Shutdown.

Letdown is isolated post fire from the remote shutdown panel. A new downstream letdown isolation valve has been installed, MUV-567, and it has been provided with IN 9218 protection to assure that a damaging spurious failure can not occur.

Procedure AP-990 " Shutdown From Outside the Control Room" has been rewritten to NOTE: Comments and clarifications are provided for SER statemen'.s that are dgukle_un_dqtlined.

._ _____,s_.____.____-__..-_.__ _ _ _ _ _

i

. U.S. Nuclear Regulatory Commission Attachment 11 3F0298-03 Page 11 of 22 Comments and Clarifications on January 6,1983 SER I

NRC SER DATED JANUARY 6,1983

' ^ ^

ON JANUAR) 6,1983 SER accommodate isolation of letdown, verified on i the simulator, field walked down to assure .

staffing requirements, and these changes have been processed through the onsite 10 CFR 50.59 process.

FPC SER clarification letter 3F0585 II, dated

May 17,1985 also stated that letdown is
isolated following a Hrc causing evacuation of the Main Control Room. Since no subsequent ,

correspondence was received from the NRC on this SER clarification, it is again provided.

I Isolation of letdown has been accommodated i in the Appendix R design and procedures, this CR 3 design feature is in compliance whh 10 CFR 50, Appendix R Section Ill.G.3 and Ill.L requirements, and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR 3 License Condition 2.C.(9), and does not materially change or alter the conclusions in NRC Safety Evaluation issued on January 6, 1983.

4 Additionally, the option of usinn a croun Qd These SER statements require of pressgjzer heaters witislLarc_f.00JI0lled clarification as they imply that pressurizer alth0_DSEl53vailable_ to the operator. heaters will be used in llot Standby and Some_ repairs may be necessary to provide during the cooldown to Cold Shutdown, and control of the pressurirer hIa!sIF2 For cold that pressurizer spray will be used following shutdown, DIcnurktr spray tht9Maldhe entry onto the Decay lleat Removal system.

dcs.ay_hcaLIctn0XalJWDHLCMitgetion to the spray gl.ycmenttbeJnade available Pressurizer heaters may be damaged in the with sonic.Icopirs. to provide additional fire and therefore are not credited to be pigssure control, available, if the pressurizer heaters are available, the operator can use them at his discretion, liowever, no credit is taken for the pressurizer heaters in Procedure AP-990

" Shutdown From Outside the Control Room".

This procedure has been verified on the simulator, field walked down to assure stafHng requirements, and this change has been processed _ through the onsite 10 CFR,,,

NOTE: Comments and clarifications are provided for SER statements that are do_uhjpyndsfrjjngd.

o U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 12 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983 0.WEmON JANUAR)AW KARIEAUOM 6,1983 SER 50.59 process.

An SER clarification was also provided to the

NRC in FPC letter 3F058511, dated May 17,  :

1985 stating that post shutdown credit is not ,

assumed for the pressurizer heaters. No

subsequent correspondence was received from the NRC, hence this clarification is again provided.

Pressurizer spray is not required for post fire safe shutdown. If pressurizer spray is available while on the Decay lleat Removal system, the operator can use it at his

, discretion.

This spray now path may be damaged in the fire and is not credited as available. No credit is taken for the pressurizer spray in Procedure AP 990 " Shutdown From Outside the Control Room". This procedure has been verified on the simulator, field walked down to assure staffing requirements, and this change has been processed through the onsite 10 CFR 50.59 process.

The changes in AP 990 are in compliance witn 10 CFR 50, Appendix R, Section Ill.G.3 and Ill.L requirements, and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR 3 License Condition 2.C.(9), and do not materially change or alter the conclusions in NRC Safety Evaluation issued an January 6.1983.

(SER Page 7 of 13)

~

Decay heat removal will inigally_Lw Q Clarification of the availability and the provided by the atmospheric dump valves sequence of operation of the .\DVs is and the turbine-driven emergency provided below.

feedwater pump.

Decay heat will initially be removed by the

. a_ _ . -

. Main Steam Safety Valv_es and the turbine _,,,,,,,

b'OTE: Comments and clarifications are provided for SER statements that are pquhlunderl[ngd.

, - ,n- -

_ ..e., .-- - , . . - - - - - ,.n.-- .. - -n,

, U.S. Nuclear Regulatory Commission Attachment B i 3F0298-03 Page 13 of 22 Comments and Clari0 cations on January 6,1983 SER NRC SER DATED JANUARY 6,1983 OMNTS MD pMMCANS ON JANUAR) 6,1983 SER driven emergency feedwater pump. The ADVs are controllable from the Remote Shutdown Panel. FPC's 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooldown calculation provides for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> hold at hot standby conditions with steam relief via tbc Main Steam Safety Valves. AP-990,

" Shutdown From Outside the Control Room",

provides the necessary steps fer re-establishing instrument air and operation of thc ADVs for cooldown to Cold Shut, wn Conditions. This procedure has beer erified on the simulator Held walked down to assure staffing requirements, and this change has been processed through the onsite 10 CFR 50.59 process.

The previous version of AP 990 dispatched an operator to the ADVs locally to operate the valves manually until instrument air was restored, and then the ADVs would be operated from the Remote Shutdown Panel.

It was discovered that the same fire could cause a loss of IIVAC to the ADV areas and that the environmental conditions would not allow for safe operator access to the ADVs.

This condition was reported under LER 97-010-00, dated May 20,1997. Crediting Ge Main Steam Safety Valves for maintaining Hot Standby conditions until air is restored to the ADVs (up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) as revised in AP-990 is in compliance with 10 CFR 50, Appendix R, Section 111.G.3 and Ill.L requirements, and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR 3 License Condition 2.C.(9), and does not materially change or alter the conclusions in Nhc Safety Evaluation issued on January 6, 1983.

~

Lontrol e of steam to_ the entemeng1 dThe final design and installation of the feedwater nump and_the__ atmospheric Emergency Feedwater Integrated Control pumn valves is provided at the DSP. (EFIC)ystem was completed after the NOTE: Comments and clarifications are provided for SER statements that are dguAlgyndqdlippd.

, U.S. Nuclear Regulatory Commission Attachment 11 3F0298-03 Page 14 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983 " ^

ON JANUAR) 6,1983 SER Additionally, the motor driven emergency January 6,1983 NRC Safety Evaluation.

feedwater pump can be started at the Following r. fire that requires evacuation of

- switchgear. the control room, the EFIC system will automatically control Emergency Feedwater flow for the turbine driveu emergency feedwater pump. The EFIC system design was installed to Appendix R criteria to provide this automatic emergency feedwater control feature following post fire evacuation of the control room. At the operator's discretion, Emergency Feedwater can be manually controlled from the Remote Shutdown Panel and in the field. This feature of the CR-3 design is in compliance with 10 CFR 50, Appendix R, Section Ill.G.3 and 111.1. requirements, and does not change or

, alter the conclusions in NRC Safety l Evaluation issued on January 6,1983.

See Section Q above for discussion of ADVs Flow control for the_smengnerfcgdwater See Section Q above for discussion of ELsitinjsR0YWallLth1JhSL For cold Emergency Feedwater Control shutdown, the decay heat removal system, the decay heat closed cycle cooling system and the decay heat seawater cooling system are utilized for decay heat removal.

The pumps for thest.31stnnspah See Section G12 below for discussion of slauN3Lthe switchgeatvillidng_rtraJts repairs.

lo_the_c.pntrol nownt_citquit. Flow control for the decay heat removal system is provided at the DSP. The cooling water ,

systems will be controlled locally.

Processing monitoring for safe shutdown will be provided by the instrumentation at i ti c dedicated shutdown panel.

. .The following variables are monitored at  !

the DSP: pressurizer level, rressurizer -

temperature, reactor _ coolant pressure, NOTE: Comments and clarifications are provided for SER statements that are dgyhje underlined.

i

,- U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 15 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983

^ ^

ANUAR 6 FR reactor coolant hot leg temperature,

'(SER Page 8 of 13) reactor coolant cold leg temperature, steam generator pressure and steam generator level. The DSP also includes monitoring of emergency feedwater Dow, makeup Dow, condensate storace tant L19 The dedicated Emergency Feedwater jggj and BWST level. Tank (EFT) was installed after the January 6, 1983 SER was issued. The Emergency Feedwater system is aligned to the EFT with <

the valves normally locked open. The Condensate Storage Tank and Ilot Well are also available manually via post-fire procedure AP-990, " Shutdown From Outside the Control Room", as backup sources.

Using the EFT as the dedicated feedwater source per procedure AP-990 is in compliance with 10 CFR 50, Appendix R, Section III.G.3 and III.L requirements. and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR-3 License Condition 2.C.(9), and does not materially change or alter the conclusions in NRC Safety Evaluation issued on January 6,1983.

))owever. the DSP does net include a As committed in FPC letter, 3F0283-11, source range neutron Hux monitor. We dated February 11,1983, a source range reauire that the licensee provide a source neutron flux monitor was installed and is rance neutron Hux monitor electrically available on the Remote Shutdown Panel, independent of the control roorJ,_and cable spreadine room._ (It should be noted that this instrumentation does not have to be safety grade, but only meet the requirements of Section III.L6 of Appendix R).

Support systems required for safe shutdti vn include the nuclear service closed cycle cooling system (NSCCC), the NOTE: Comments and clarifications are provided for SER statements that are slogble_ underlined.

, U.S. Nuclear Regulatory Commission At'achment B 3F0298 Pt.ge 16 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983 ^

ON JANUARY 6,1983 SER nuclear service seawater cooling system (NSSC), the reactor building cooling system, essential area llVAC systems, the 125 volt DC power system and the diesel generators.

The NSCCC pumps and valves can be controlled at the DSP. The NSCCC and This should be NSSC.

the reactor building cooling system will be operated from the switchgear and motor control centers.

The diesel cenerator wilLbg provided with .Cl1 This is accurate and remains a CR-3 the capability to automatically initiate on Appendix R design feature. Clarification of

}oss of offsite power. and two potential spurious actuations and operator actions is provided below.

(SER Page 9 of 13)

The Appendix R design for isolating the to be electrically independent of the EDGs from a control room, or cable g_ontrol room and calle spreading room._ spreading roo.n fire assures EDG independence by isolating the circuits with switches in the A and B ES 4 KV Switchgear Rooms. Between the time that a control room evacuation is declared and the switches are used, a spurious action is assumed to occur which could cause either an EDG breaker trip or an exciter trip. The EDG lock-out resets for the breakers are located in their respective ES 4 KV Switchgear Rooms and the Exciter resets are located in the Diesel Generator rooms. Emergency Lighting is available to allow operators to respond to these potential spurious actuations. The corrective actions for these potential spurious actuations are included in AP-990. This procedure has been verified on the simulator, field walked down to asure staffing requirements, and this change has been processed through the onsite 10 CFR 50.59 process.

Accommodating these potential spurious actuations per procedure AP-990 is in compliance with 10 CFR 50, Appendix R.

NOTE: Comments and clarifications are provided for SER statements that are doubjpynderlined.

. U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 17 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983 COMMENTS AND CLARIFICATIONS ON JANUARY 6,1983 SER Section III.G.3 and III.L requirements, and "would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire" under CR-3 License Condition 2.C.(9), and does not materially change or alter the conclusions in NRC Safety Evaluation issued on January 6,1983.

~ ~ ~~~ ~ ~

A dedicated IIVAC svstem will t e The SER for this installation was issued in provided for ventilation and cooline of the letter NRC to FPC,3N0985-11, dated dedic;ited shutdown panel nower sunnly September 11,1985, " Safety Evaluation and switchcear. The licensee aas not Report, Control Complex Dedicated Cooling pompleted the desien of the dedicated System". A clarification / revision of this SER llVAC system: and thus. the IIVAC is provided in Attachment C to this document, system review is not included in this TW%

B. 22-Ilour Requirement The alternative shutdown systems have the capability of achieving cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The alternative shutdown systems can accomplish cold shutdown using only onsite power sources.

C. Renairs The licensee intends to utilize repairs to C.12 Cooldown to Cold Shutdown conditions restore control oflome optional couipment (s 200 'F) using AP-990, " Shutdown From for usace durine hot shutdown such as the Outside the Control Room", does not require pressurizer heaters. Additionally, the equipment repairs. The option of conducting licensee will utilize repairs to restore repairs is available in the procedure but not contro11the decay heat removal system _ required. AP-990 has been verified on the and other components used for cold simulator, field walked down to assure shutdown. The renairs cenerally include staffing requirements, and has been processed wire removal. installation of temporary through the onsite 10 CFR 50.59 process.

circuits or wirine at the motor control o center (SER Page 10 of 13) pf switchgear. in order to p ovide local -

gontrol of equipment. The licensee has developed procedures for these repairs and

_._ a!LRLaterial needed for the repairs will be NOTE: Comments and clarifications are provided for SER statements that are double underlined.

o U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 18 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983

' ^ ^

ON JANUAR) 6,1983 SER stored onsite..

D. Associated Circuits The licensee provided the results of their associated circuits review for the control room and the cable spreading room. The results identified the associated circuits of concern in these areas and the proposed methods for protecting the safe shutdown capability from tire-induced failures of these circuits. The proposed methods for protecting the safe shutdown capability are consistent with the guidelines provided by us.

1. Power Source Case - The licensee's analysis concluded that power circuits which share a common power bus with the power circuits of the dedicated shutdown panel are provided with coordinated fuses and breakers or are isolated from the power source via transfer switches.

(SER Page 11 of 13)

2. Spurious Sinnal Case - The licensee's analysis identified a number of circuits whose fire-induced failures may adversely affect the safe shutdown capability. The licensee has proposed methods for protecting the safe shutdown capability. Over 80 valves will be electrically isolated from the Cd3 In FPC to NRC letter 3F1082-32, dated control room and will be controllable October 29,1982, Table 1 on Page 45, lists from the DSP. equipment to be isolated from the Main Control Room (MCR) and be controllable from the Remote Shutdown Panel (RSP). As provided in this table, valves DHV-42 and DHV-43, from the Reactor Building Sump, were to be isolated from the MCR and controllable from the RSP.

I NOTE: Comments and clarifications are provided for SER statements that are clouble underlined.

. U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 Page 19 of 22-Comments and Clarifications on January 6,1983 SER COMMENTS AND CLARIFICATIONS NRC SER DATED JANUARY 6,1983 ON JANUARY 6,1983 SER This commitment has been classed as INACTIVE by FPC based on the following:

Valves DHV-42 and DHV-43 are redundant to and in series with valves DHV-34 and DHV-35, which are controllable from the RSP. Post-fire spurious actuation of these -

valves could result in draining the Borated Water Storage Tank (BWST) to the Reactor Building Sump if remedial operator action is not taken. All four * .ves are maintained in the closed position owing normal operation.

Valves DHV-34 and DHV-35 are controllable from the RSP and can be isolated from the MCR and Cable Spreading Room in the event of a fire, Since control is provided for valves DHV-34 and DHV-35 at the RSP, control of valves DHV-42 and DHV-43 from the RSP is not required to meet Appendix R requirements.

Control of valves DHV-34 and DHV-35, in conjunction with the BWST level indication prosided at the RSP, assures that post-fire isolation capability is available in the event of a spurious action on any of theses four valves.

It is noted that in FPC to NRC letter 3F0176-03, dated January 13, 1976, FPC committed to have DHV-34 and DHV-35 in the normally open position to preclude the possibility of a water hammer caused by injection of water into a poten*ially dry piping system.

Following a 1985 review of potential Appendix R spurious actuations, procedure OP-404, " Decay Heat Removal System", was revised and DHV-34 and DHV-35 were placed in the normally closed position.

On September 26,1996 Problem Report PR964401 was issued which identified the licensing / design basis issue regarding the need to analyze and determine the proper normal NOTE: Comments and clarifications are provided for SER statements that are double _undedned.

4

.- U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 -

Page 20 of 22 Comments and Clarifications on January 6,1983 SER NRC SER DATED JANUARY 6,1983 COMMENTS AND CLARIFICATIONS ON JANUARY 6,1983 SER operating position for DlIV-34 and DHV-35.

In NRC to FPC letter 3N1297-08, dated December 4,1997, the NRC issued violation 50-302/97-14-13 citing improper corrective action of this PR9M401 issue. FPC conducted a 10 CFR 50.59 review of these valves in the closed position and determined that this was a USQ. In FPC to NRC letter, 3F1297-37, dated December 29,1997, FPC committed to submit a Licensing Amendment Request that will resolve this USQ.

Valves for cold shuldmvD_ systems may See Section Cd2 above for discussion of utilire goair to provide local control repairs.

as stated above Additionally, a number of circuits will be rerouted independent of the control room and cable spreading room. For instrumentation circuits, voltage to current converters will be installed.

The alternative shutdown capability relies on automatic start of the diesel generator on loss of offsite power.

~ ~

Thus. for each diesel cenir~ator. the This is accurate and remains a CR-3 Appendix control room controls for the R design feature. Clarification of two covernor one of the redundant stan potential spurious actuations and operator gircuits. and stop circuits will be actions is provided in Section C.11 above, isolated from the control room and '

cable spreadinn room. Further, for prevention of a possible fire-induced LOCA, the power for one of the redundant electrically controlled valves at the high/ low pressure interface of the reactor (SER Page 12 of 13) coolant loop and decay heat drop line, will be locked closed by locking out the breaker at the Motor Control Center.

NOTE: Comments and clarifications are provided for SER statements that are double underlined.

. - U.S. Nuclear Regulatory Commission Attachment B 3F0298-03 - Page 21 of 22 Comments and Clarifications on knuary 6,= 1983 SER NRC SER DATED JANUARY 6,1983 A D RAREAMNS ON JANUARY 6,1983 SER -

3. , Common Enclosure Case - The

- licensce's analysis identified the

- associated circuits which share a common enclosure with the alternative shutdown circuits. These circuits are isolated by transfer switches at either -

the dedicated shutdown auxiliary -

panels or at the dedicated shutdown panel.

E. Safe Shutdown Procedures and Manpower The licensee will revise existing --

procedures EM-101, " Fire Protection Plan" and EP-113. " Plant Shutdown from Outside Control Center" prior to operation of the dedicated shutdown panel. The manpower necessary for safe shutdown using the dedicated shutdown panel will be available. No fire brigade members are included in the shutdown manpower requirements.

-The licensee 5ii submit Technical ppendix R related Technical Specifications -

EEEdiaMig for the dedicated shutdown - were removed and placed in the Fire panel prior to operation of the panel. Protection Plan in License Amendment No.

147 issued on January 22,1993. It is noted, however, that the Appendix R Remote -

Shutdown Panel equipment is to be updated in the Fire Protection Plan 90 Days after Restart

__,,,_,, _ from the current Outage. _

(SER Page 13 of 13)

CONCLUSION Based on our review, we conclude that the perfonnance goals for accomplishing safe shutdown in the event of a fire, i.e., reactivity control, inventory control, decay heat removal, pressure control, process monitoring and support function are met by the proposed alternative shutdown capability with the As committed in FPC letter, 3F0283-11, exception of the capability to monitor neutron dated February 11,1983, a source range flux. We reauire that the licensee provide a neutron flux monitor was installed and is NOTE: Comments and clarifications are provided for SER statements that are dpgle_u,njejiAed.

I

-e U.S. Nuclear Regulatory Commission Attachment B-

= 3F0298 _

Page 22 of 22 '

. Comments and Clarifications on January 6,1983 SER -

NRC SER DATED JANUARY 6,1983

^ ^

ON JANUARY 6,1983 SER g3nrce ranne neutron flux monitor electrically available on the Remote Shutdown Panel, indcoendent of the control room and cable spreading room This instrumentation does not have to be safety-nrade. but only meet the-reuuirements of Section III.L.6 of Anoendix-R. The iustification for our reauliement is orovided below.

Additionally. the licensee has committed to - The SER for this installation was issued in orovide a dedicated HVAC system that meets letter NRC to FPC, 3N0985-11, dated.-

the reauirements of Section Ill.G of Anoendix September 11 1985, " Safety Evaluation R. However. the licensee has not comoleted Report, Control Complex Dedicated Cooling the desian of the HVAC system. We will System". A clari0 cation / revision of this SER provide a separate evaluation for the HVAC is provided in Attachment C to this document.

sy.stgg Therefore, we conclude that the licensee's alternative shutdown capability for the control room, the cable spreading room and the control complex HVAC equipment room complies with the requirements of -

Section III.G.3 and III.L of Appendix R oending the licensee's commitment to orovide As committed in FPC letter, 3F0283-11, a source ranne neutron flux monitor, dated February 11,1983, a source range -

neutron flux monitor was installed and is -

available on the Remote Shutdown Panel.

-j 1

NOTE: Comments and clarifications are provided for SER statements that are do_ugynderlined.

e s'

U. S. Nuclear Regulatory Commission Attachment C 3F0298-03 Page 1 of 3 Comments and Clarifications on September 11,1985 SER NRC SER DATED SEPTEMBER 11,1985 COMMENTS AND CLARIFICATIONS ON SEITEMBER 11,1985 SER SAFETY EVALUATION REPORT CONTROL COMPLEX DEDICATED COOLING SYSTIN FOR POST-FIRE ALTERNATE SIIUTDOWN CAPABILITY (APPENDIX R, SECTION III.G.3)

CRYSTAL RIVER, UNIT 3 In our safety evaluation report input dated In FPC to NRC letter 3F0185-01, dated December 28,1982 for the post-fire alternate January 2,1985, FPC submitted the design shutdown capability (Appendix R, Section for the Control Complex Dedicated Cooling Ill.G.3) for Crystal River, Unit 3, we noted System, commonly called the Appendix R that the licensee had committed to provide a chiller. On September 11,1985, the NRC dedicated ventilation and cooling system for issued an SER approving the Appendix R control complex equipment as an alternate to chiller under Section III.G.3 and III.L of the normal control complex IIVAC system. Appendix R. The basic design, now paths, The control complex at Crystal River 3 is a and system operation are discussed in the FPC building which houses the control room, cable submittal and are summarized in the NRC spreading room and various electrical SER.

equipment rooms. This modification was required because a majority of the normal In 1990, the Turbine Building (TB) control complex IIVAC system components Switchgear (SWGR) Room Cooling was (e.g., fans, dampers, chilled water cooling added to the Appendix R chiller system. The units, etc.) are located in one fire area Appendix R chiller is now normally in (control complex, elevation 164'-0"). Further, operation supplying cooling for the TB SWGR the existing air handling system consists of a room. In the event of a fire requiring the single duct feeding scveral rooms in series Appendix R Chiller to perform its dedicated containing redundant safe shutdown electrical function, the TB SWGR room is isolated and equipment and therefore, closure of a single the Control Complex Dedicated Cooling fire damper in response to a fire or single function is placed in service using approved failure would result in the loss of cooling air post-fire procedure AP-990, " Shutdown From Dow to these rooms. Thus, in order to ensure Outside the Control Room

that areas within the control complex containing required safe shutdown equipment Originally the TB SWGR room was cooled by can be adequately cooled in the event of a the Control Complex Cooling system.

postulated fire, the licensee by letter dated However, it was discovered that this could January 2,1985 provided the conceptual compromise the Control Complex Cooling design of a separate dedicated cooling system portion of the system in the event of an in order to comply with the criteria of earthquake, since the TB SWGR rooms are Appendix R. The following control complex not seismically designed. MAR 89-03-06-01 areas will be served by the dedicated cooling was implemented to utilize the Appendix R l

l

  • U. S. Nuclear Regulatory Commission Attachment C 3F0298-03 Page 2 of 3 Comments and Clarifications on September 11,1985 SER-NRC SER DATED SEPTEMBER 11, 1985 COMMENTS AND CLARIFICATIONS ON SEITEMBER 11,1985 SER system: Chiller for providing a chilled water supply to the TB SWGR room during normal operation,
a. The new remote shutdown room at el.

108'-0'; Appendix R Section Ill.G.3 and Ill.L does

b. Divisions A and B 4160V switchgear not preclude the use of new/ additional rooms at el.108'-0"; equipment, added for Appendix R
c. Divisions A and B inverter rooms at el. compliance, from having a normal .

108'-0; operating / secondary purpose. As long as the

d. Divisions A and B battery charger rooms Appendix R chiller is dedicated post-fire to at el,108'- 0"; perform its III.G.3 and Ill.L functions,
c. Divisions A and B 480V switchgear rooms compliance with Appendix R is achieved.

at el.1M'-0; and AP-990 isolates the TB SWGR rooms and

f. Divisions A, B. C, D emergency places the Appendix R chiller into service to feedwater initiation and control (EFIC) room perform its dedicated post-fire function. AP-at cl.124'-0". 990 has been field walked down to assure the reasonableness and timing of these Appendix Each of the above control complex areas R chiller manual operating requirements. The required for post fire safe shutdown CR-3 Appendix R Chiller design remains in is cooled by an individual fan coil cooling unit compliance with 10 CFR 50, Appendix R, located in the respective areas. An air cooled Section 111,G.3 and III.L requirements, and chilled water unit including piping, circulating the normal use of the Appendix R Chiller, and pump and expansion tank located outside the "would not adversely affect the ability to control complex on the roof of the clean achieve and maintain safe shutdown in the machine shop provides chilled water to each event of a fire" under CR-3 License Condition fan coil unit. The chilled water piping is 2.C.(9), and this change does not materially seismically supported within the control change or alter the conclusions in NRC's complex to ensure that damage will not occur Safety Evaluation Issued on September 11, to safety-related equipment in a seismic event. 1985.

The system cooling capacity is designed to handle the total heat load of the required shutdown equipment.

Electrical power to the chilled water unit.

circulating water pump and local fan coil units will be provided from the essential onsite power sources. Manual control switches and status indication for the chilled water unit circulating water pump and local fan coil units will be provided at a local control panel in the turbine building at el.145' independent of areas where a fire may cause loss of the control complex normal IIVAC system. In the

o g.-

U. S. Nucle:r Regulatory Commission Attachment C 3F0298-03 _

Page 3 of 3.

Comments and Clarifications on September 11,1985 SER ,

NRC SER DATED SEl"fEh1BER 11,1985 COhlh1ENTS AND CLARIFICATIONS ON SEFFEh1BER 11,1985 SER event a fire disables the control complex normal IIVAC system, the dedicated IIVAC system will be started from its local control panel. Manual initiation of the dedicated IIVAC will occur prior to exceeding temperatures necessary to assure proper function of the safe shutdown equipment.

Emergency lighting will be provided at this local control panel and along the access and egress routes to its location.

Based on the above, we conclude that the design of the control complex dedicated cooling system meets the requirements of Section III.G.3 and III,L of Appendix R and is, therefore, acceptable. Further, we reconfirm our previous conclusion in the January 6,1983 safety evaluation report input regarding the acceptability of the alternate shutdown capability at Crystal River, Unit 3.

. _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _

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