ML20138P079
| ML20138P079 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/17/1985 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Berkow H Office of Nuclear Reactor Regulation |
| References | |
| P-85476, TAC-55287, NUDOCS 8512240285 | |
| Download: ML20138P079 (102) | |
Text
{{#Wiki_filter:q~ ? ?: j OPublic Service ~- 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 Company of Colorado December 17, 1985
, Fort St. Vrain Unit No. 1 P-85475 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
. Attn: Mr. H.N. Berkow, Project Director Standardization and Special Projects Directorate Docket No. 50-267
SUBJECT:
Request for Additional Information - Fort St. Vrain Building 10
REFERENCE:
NRC Letter Dated November 21, 1985, E.J. Butcher to 0.R. Lee (G-85469)
Dear Mr. Berkow:
The reference letter requested that PSC provide a legible copy of the safety etaluation, design criteria and design analysis for CN-1255, CN-1332, CN-1294, and CN-1391. Attachment 1 to this letter provides the requested information, and for eech CN, the latest revised copy is included. In addition, the reference letter requested information on any subsequent changes in battery design as indicated on CN-1391. The future changes referred to on Page B1 of CN-1391 are for a non-safety related battery. PSC reused the old 1C battery to provide uninterruptible power to the non-safety related data logging system. This in no way affects the safety related battery. If you have any questions on this information, please contact Mr. M. H. Holmes at (303) 480-6960. Very truly yours, g2240285051217 m
- p D. W. Warembourg, Manag/'er
- ADocK 05000267 PDR Nuclear Engineering Division DWW/ RAG:pa g9 Attachment i
'k cc: Mr. K. Heitner - 2 copies of letter and Attachment 1
g 6 i Attachment 1 to P-85476 CN-1255 CN-1332 CN-1294 CN-1391
- f. ISSUE ' STATES Pubiic w sr. VRAIN NUCLEAR GENE ITfp%ggo jg i A' ~ ' SerVICD PU;LIC SET.VICE COMPANY OF C OLORADO BY MgS89 COMPLETE PAGE /
,. CHANGE NOTICE REISSUE J/0#/AGCf[2f-97587 SYSTEM NO. U REISSUE # M SUBJECT OF 'RIGINAL CN _ Y$W SfA Y'*d $AfffA'ff / d A as e / A . S d 'A c d e r
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DESCRIPTION OF REISSUE hCMb4 emf U! b A Yf Yb 8 lM AfdsJa/At * , THIS CN REISSUE INVOLVES: YES NO 1 C X Change to the safety evaluation (Attach effective safety evaluation) 2 [ */ Change to the environmental evaluation (Attach effective environmental evaluation) 3 Z X Change in the safety related classification 4 Z X ' Change in design intent 5 Z Y Change in scope 6 Z V Change in installation / procurement responsibilities 7 Z X Change to the installation / procurement package 8 Z X Change to the design background information package 9 X Z Change to the document update package 10 Z ,R Change notice cancellation ATTACHMENTS: CHANGE NOTICE REISSUE COORDINATION SHEET: EFFECTIVE DATE 9,!9 8f SAFETY EVALUATION: OVERALL EFFECTIVE DATE 7.2.1 - B + ENVIRONMENTAL EVALUATION: OVERALL EFFECTIVE DATE T-/3-95 REVIEWS, APPROVALS: N INEERING 6 4 % A f i,ref' a 8[ tOATE PROD. A i$iG%AtyAt. 9 MI sOA fli
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- AND DATE M 6 4 0 SEP 2 4.19E YES Ng Ah*'M9 0 [Y Safety Significant
#* " 7 0 Unreviewed Safety Question O d[i2Unreviewed Environmental Ouestion y'yy..y nA Nr gy;n .
- REQUIRED FOR AFETY RELATED CHANGE NOTICES.
f as,2 Y 0 A ,,, FormiO 372 303381
e . PUBLIC SERVICE COMPANY OF COLORADO 6 FORT ST. VRAIN NUCLEAR GENERATING STATION hTCR / SCR / PC / TR NO IMI ( SAFETY EVALUATION PAGE 2
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TYPE: M OVERALL OCN SU8MITTAL OSETPOiNT CHANGE REPORT OTEST REQUEST OTEuPORARY CONFIGURATION REPORT OPROCEDURE CHANGE (FSAR) O OTHER CLASslFICATION: ARE THE SYSTEM (S) EQUIPMENT OR STRUCTURES INVOLVED, OR DOES THE ACTIVITY AFFECT: CLASSI E YES C NO ENGINEERED SAFEGUARD 2 YES O NO SAFE SHUTDOWN b YES C NO PLANT PROTECTIVE SYSTEM 3 YES O NO SAFETY RELATED E YES O NO SECURITY SYSTEM O vES 2 NO REMARKS 1 IS THE ACTIVITY IDENTIFIED IN THE F S A R OR TECH SPEC? A YES O NO LIST THE APPLICA8LE SECTIONS REVIEWED FSAR sections 8.2.3, 8.2.2.4, 8.3, 8.2.3.4, 8.4; Criterion 24 and 39; Tech Specs section SR 5.6.2 and LCO 4.6.1 2 DOES THE ACTIVITY REQUIRE THAT CHANGEISI BE MADE TO THE F S A R OR TECH SPEC? FA YES O NO LIST SECTIONS TO BE CHANGED AND THE CHANGES TO BE MADE. . M _ CHAD 9Ad_to_5b.oW_JLOW_.eQUlpfDeDt_Dumher for the battery. 'FSAR Ficure 8.2.14 must be 3 DETERMlNE WHETHER OR NOT THE ACTIVITY INVOLVED IS AN UNREVIEWED SAFETY OUESTION UTILIZING THE FOLLOWING GUIDELINES. (A) HAS THE PRO 8 ABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EOulPMENT lMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE F $ A R BEEN INCREASED? O YES 3 NO STATE BA$15: Replacement of existing batteries with batteries of _gretter_ capacity _can_not_ increase _the_probabOity_of_occ.urrer1Ce or the consecuences of any accident or malfunction evaluated in the_ESAR (8) HAS THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE F S A R BEEN CREATED? O YES b NO STATE BASIS: Since-the only change _to the DC_ system is replating existing batteries with new batteries of _ increased capacity. nn_ possibility of a new accident or malfunction has been createL (C) HAS THE MARGIN OF SAFETY, AS DEFINED IN THE BASl5 FOR ANY TECHNICAL SPECIFICATION OR IN THE F S A R 8EEN REDUCED? O YES QNO STATE BASIS. No maroin of safetv concernino _ battery capacity is defined.__Since this modification islin the conservative
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_firectinn increated hattary canacity chnuld infer greator nrntactinn fenm nypelnari
, _and thorpfnre less,_ chance _nr nc system failure.
DOES THE ACTIVITY APPEAR TO: INVOLVE AN UNREVIEWED SAFETY QUESTION O YES @ NO OE SAFETY SIGNIFICANT O YES @ NO gy . -> N"*/}= N
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PUBUC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION h OTH@ ( l ENVIRONMENTAL EVALUATION NO. PAGE 4
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som m pw CA TE GORY - - r r+ - e ' W * "+s a s ~w . -= + TYPE: % CN Overall 2 CN Submittal Other Are all moseurable nonradiological effects of this activity confined to the on-site areas previously disturbed during site preparanon, plant construction or previous plant operation? Qes Z No State baeie A'5 CN reA h A GXuTh% boihdo! lb /hU bhll0 NET in ~Rao hC. olnM't. ' u 1iLau. AfD stkln en &> AM iE 4t1fiAr. D(su ,_ a ya is the actrvity required to achieve compliance with Federal, State or local environmental regulations? 2 Yes MNo NOTE: If either answeris Yes, then sign and date form. If both answers are No, then complete remainder of this evaluation form. EVALUATIOl4 (Use addstional sheets drequired) - ' .- - . - . -
- 1. Is the activity identified in the final environmental statement (FES).
Z Yes ! No Or Supplementary Environmental Documents ISee Q-3)? Identify documents and document sections reviewed
- 2. Determine whether or not the activity involved is an unreviewed environmental question using the following guidelines.
(if the answer to any of the following questons is Yes, then this activity involves an unreviewed environmental question.) (Al Will this activity result in a significant increase in any adverse environmental impact previously evaluateo in the FES? Z Yes ! No State basis: , (B) Will this activity result in a significant change in the types, or a significant increase in the amounts of effluents. or a significant increase in the authorized power level? Z Yes Z No State basis-(C) Does this activity involve an environmental matter not previously reviewed and evaluated in the FES? Yes : No State basis: i l Does the activity indve an unreviewed environmental question? Z Yes : No t By I of1MA . 4[/NN Approved
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a. PUBLIC SERVICE COMPANY OF COLORADO CN /8 k FORT ST. VAAIN NUCLEAR GENERATING STATION BY CHECK LIST OF DESIGN VERIFICATION QUESTIONS FOR DESIGN REVIEW METHOD PAGE 8.3 Form sad.22.doeS YES g N/A O O -1. Were the inputs correctly selected and incorporated into design? TC O 2. Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent re-verifications when the deta led design activities are completed? IO O 3. Are the appropriate quality and quality assurance requirements specified? YO O 4. Are the applicable codes, standards and regulatory requirements including issue and addenda property identified and are their requirements for design met? IO O 5. Have applicable construction and operating experience been considered? IO O 6. Have the design interface requirements been satisfied? MO O 7. was an appropriate design method used? OO 8. is the output reasonable corvoared to inputs? IO O 9. Are the specified parts, equipment, and processes suitable for the required application? dO O 10. Are the specified materials compatible with each uther and the design environmental conditions to which the material will be exposed? O O 11. Have adequate maintenance features and requirements been specified? O O 12. Are accessibility and other design provisions adequate for performanca of needed maintenance and repair? I'O O 13. Has adequate accessibility been provided to perform the in service inspection expected to be required during the plant life? O O f 14. Has the design properly considered radiation exposure to the public and plant personnel? O O 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished? O O [ 16. Have adequate pre operational and subsequent periodic test requirements been appropriately specified? O O 17. Are adequate handling, storage, cleaning and shipping reqdrements specified?
.O C 18. Are adequate identification requirements specified?
O O '19. Are requirements for record preparation review, approval, retention, etc., adequately specified? NOTE: If the answer to any question is no, provide additional information and resolution below: RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED DURING THE DESIGN VERIFICATION PROCESS k d$ 1 &&hC/ % M fa. u a f M v/im 4"
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.' PUBLIC SERVICE COMPANY OF COLORADO CN / 39/4
, FORT ST. VRAIN NUCLEAR GENERATING STATION CHECK LIST OF DESIGN VERIFICATION QUESTIONS B61M
_FOR _ DESIGN REVIEW METHOD PAGEl$3.hl Form 344 22 4005 4 YES NO N/A IO O 1. Were the inputs correctly selected and incorporated into design? O O g 2. Are assumptions necessary to perform the design activity adequately described and reason necessary, are the assumptions identified for subsequent re-verifications when the detailed design act are completed? O O d 3. Are the appropriate quality and quality assurance requirements specified? O'O T 4. Are the applicable codes, standards and regulatory requ!.ements including issue and add identified and are their requirements for design met? - O O # 5. Have applicable construction and operating experience been considered? O O d 6. Have the design interface requirements been satisfied? O O T 7. was an appropriate design method used? O O d a. Is the output reasonable compared to inputs? O O T 9. Are the specified parts, equipment, and processes suitable for the required application? O O M 10. Are the specified materials compatible with each other and the design environmental conditions to w the material will be exposed? O O d 11. Have adequate maintenance features and requirements been specified? O O I 12. Are accessibility an' d other design provisions adequate for performance of needed mainten O O I 13. Has adequate accessibility been prvvided to perform ti e in-service inspection expected to be requ during the plant life?
; _O O d 14. Has the design properly considered radiation exposure to the public and plant personnel?
l I MO O 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verificatior. that design requirements have been satisfactorily accomplished? Q~ O I 16. Have adequate pre-operational and subsequent periodic test requirements been appropriately specif O O I 17. Are adequate handling, storage, cleaning and shipping requirements specified? 6O O 18. Are adequate identification requirements specified? FO O 19. Are requirements for record preparation review, app % val, retention, etc., adequately specified? t I NOTE: ' If the answer to any question is no, provide additional information and resolution below: RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED DURING THE DESIGN VERIFICATION PROCESS
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ISSUE STATUS
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..- FORT ST. VRAIN NUCLEAR GENERATING WMV'ED aesah'N/$$$ psy^
CHANGE NOTICE REISSUE PAGE / Form toi372 30 3Ji$1 i _ COMPLETE 0/O#/ ACCT.# ddnGar5__ IMIN((h SYSTEM NO. N REISSUE # -
. SUBJECT OF 0RlGINAL CN AdNW f//ffWS.587&MUf be$368ff4TS W/r# ffithA__
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/A~oM SeatJ/C C -
E THIS CN REISSUE INVOLVES: YES NO
-1 X Change to the safety evaluation (Attach effective safety evaluation) 2 Z X Change to the environments' wa'uation (Attach effective environmental evaluation) 3 ~~ V Change in the safety re%:ed class cation 4 Z g Change in design intent 5 Z k Change in scope 6 Z
- Change in insta36 tie r prewemcn responsibilities 7 ' _ R Change to the insta!!ation procu'ement package 8 .2 X Change to the design background information package 9 X 'l Cnange to the document upve t ackage 10 Z X Change notice cancellat.on ATTACHMENTS:
CH ANGE NOT.CE REISSUE COORDiN" Tic ' SSEET- - EFFECTIVE DATE /,!7.!W SAFETY EVALUATION: OVERAtL EFFECTIVE DATEj N ENVIRONMENTAL EVALUATION: OVERALL EFFECTIVE DATE 9/I8/88 i/ REVIEWS. APPROVALS: ENGINEERING _ _ . . _ . _ _ _ l!7!18 PROD.' . ._. .. _ /_ S ._ _, I
'PORC MTG. # AND DATE PORc 6 0 4 JAN.1_6t1985 YeS j y Safety S gnif cant ' . p g of - poc. t/pperf -_
l Unreviev ec Safety Quest.on Z 7 Unreviewed Environmenta: Quer :n QA - -- s,u . v .
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, 'h. : A i .D Cm A*.GE NOTICES t
t . PUBLIC SERVICE COMPANY OF COLORADO
- FORT ST. VRAIN NUCLEAR GENERATING STATION CN / TCR / SCR / PC / TR
-g " NO 1255 B ai8 SAFETY EVALUATION PAGE 2 I TYPE: 3 CN OVERALL OCN SUBMITTAL CSETPOINT CHANGE REPORT OTEST REQUEST OTEMPORARY CONFIGURATION REPORT CPROCEDURE CHANGE (FSAR) OOTHER CLASSlFICATION: ARE THE SYSTEM (S) EQUIPMENT OR STRUCTURES INVOLVED. OR DOES THE ACTIVITY AFFECT: CLASS 1 3 YES C NO ENGINEERED SAFEGUARD b YES C NO SAFE SHUTDOWN 3 YES O NO PLANT PROTECTIVE SYSTEM O YES O NO SAFETY RELATED 3 YES C NO SECURITY SYSTEM C YES h NO REMARKS 1 15 THE ACTIVITY IDENTIFiE') IN THE F S A R OR TECH SPEC? d YES [. N O LIST THE APPLICABLE SECTIONS REVIEWED. F$AD Cactinne 7_A_3, 8_2_2_3, 8_2_3_4, 8_2,5_2, _ 8.2.2.4. 8.2.3.5: Tech Snec_ tro d_62 FRAR Rectinn R 9 1.~ Finnroe
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8.2-28, and R_2-?Q_ {Ref_ aro Enr Rovisinn ?) 2 DOES THE ACTIVITY REQUIRE THAT CHANGE (S) BE MADE TO THE F S A R OR TECH SPEC? @ YES O NO LIST SECTIONS TO BE CHANGED AND THE CHANGES TO BE MADE. See attached 3 DETERMINE WHETHER OR NOT THE ACTIVITY INVOI.VED IS AN UNRE5 tEWED SAFETY OUESTION UTILIZING THE FOLLOWING GUIDELINES. (A) HAS THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE F S A R BEEN INCREASED 7 C YES U NO STATE BASIS:CN-1255 installs an independent instrument powar trans-former to p_toyide_back-up_ power _for_each inverter / static _ transfer switch (1A and 18) whict in turn Drovides opwer to non-interruptible_ buses _la_and_18. Power is transferred from the oreferred source to the_ alternate _ source _via_ static _tra_nsfer function replacing manual transfer switches in the previous configuration. The_elec_trical power source for the new Instrument Power Transformers 1A-1 and 1B- Lis_ Essential 480LSwgr_ Bus _ Land _ Bus 3 respectively. The old electrical configuration allowed backup power (see attached) (B) HAS THE POS$1BILITY OF AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EVALUATED PREVIOUSLY IN THE F S A R BEEN CREATED? C YES @ NO STATE BASIS: The system functions as before to orovide oower_.to_vitalequipment_but._wittL_ increased reliability as described in 3(Ah therefore. the possibility of a new_ accident _or malfunction hai_not been createc (C) HAS THE MARGIN OF SAFETY, AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION OR IN THE F S A R BEEN REDUCED 7 0 YES T NO STATE BASIS: Increasina the reliabilitv and flexibility of the non-interruptible. power supply _sys. tem __can__o_nly increase _the system's margin of safety described _in100 4.6. DOES THE ACTIVITY APPEAR TO: INVOLVE AN UNREVIEWED SAFETY OUESTION O YES @ NO BE SAFETY SIGNIFICANT 3 YES U NO Ab B7AAA
- APPROVED b
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- R E QUI R E D NLY FOR CHANGE NOTICE FORM ( A) 372 02 3382 w
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~Aage 2 et Question 2:
FSAR Sections' 7.4.3 8.2.2.3, 8.2.3.4, 8.2.5.2, 8.2.2.4; Figures 8.2-26, 27,. 28, and 29 should reflect the -new equipment and configuration. (Revision 2 has incorporated these changes.) In-the basis for -Tech Spec LCO 4.6, the statement " Backup electric power for the non-interruptible a-c instrument loads -is provided by bus ties from Instrument Bus No. 3 to Instrument Buses . I and 2 which - are normally fed . by the two - Instrument Power Inverters" must be revised. .. O fO Y 9-e
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. 3.- A.. '(Continued. . . ) .
for~ Non-interruptible buses 1A and-1B to be provided from the.same source. The inverters:(1A and 18).are also being-upgraded-from 14.5 KVA to 25 KVA which will prevent system overload and allow . sufficient margin for future modifications. Power to Interruptible Bus'3 will also be provided through a static transfer switch utiliz-
. ing instrument power transformers.1A and 18.:
The static transfer switch provides automatic th'rowover capability-1 between either of two essential sources. The addition of two distribution panels. Instrument Power Buses IA-1 and 18-1 supplied
- from Inverter / Static Transfer Switches 1A-and IB, respectively, pro-
- vides-for added flexibility in future design modifications.
Each item changed within this CN. adds to the . reliability and flext-bility of- a vital. electrical' power system within FSV. The conse-quences.of loss of power. analyzed in FSAR Section 8.2.2.3 have been
-decreased by the added instrument power system reliability.
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> FORT ST. VRAIN NUCLEAR GENERATING STATION h 0THER ME ENVIRONMENTAL EVALUATION No.
PAGE f. PoRM 372 364310 T(PE: O CN Submittal Other [ CN Overall
'Are all measurable nonradiological effects of this actnnty confined to the on-site areas previously disturbed during site proporsuon, plant construction or previous plant operation? (Yes O No State basis M d Mdt N- M "' EA @ M M N iAS AW N f
- remove unna att aus . w - +. ok. T-4) 2 all dau_nA nrwiks6, sMfui6el i
a .s;4s cums. U ' ' I r is the activity required to achieve compliance with Federal, State or local emntonmental regulations?, C Yes g No NOTE: If eitheransweris Yes, then sign and date form. If both answers are No, then complete remainder of this evaluation form. , AtUA110C(Use additional sheets if required) ~ TT ' ~ ~ ~ { 1. Is the activity identified in the final emnronmental statement (FES) C Yes C No Or Supplementary Environmental Documents (See Q-3)? Identify documents and document secuons reviewed
?
A l
' Determine whether or not the activity involved is an unreviewed environmental question using the fo!!owing guidelines. l
'2.
l (if the answer to any of the following questions is Yes, then this activity involves an unreviewed environmental I l question.) IAJ Wal mis activity rest.t in a significant increase in ans 3cverse enviro.. mets .mpact s.,rev:cus.y naiudtec in the l l FES? C Yes : No State basis: i i I i (B) Will this activity result in a significant change in the types, or a significant increase in the amounts of effluents. or a ';
' significant increase in the authorized power level? C Yes Z No State basis:
I i (C) Does this activity involve an environmental matter not previously reviewed and evaluated in the FES? 2 Yes C No State basis: I Does the activity involve an unreviewed environmental question? . Z Yes Z No Byi ) Ot iM-- N/ N
' ' "Jata-Approved
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- REQUIRED ONLY FOR CHANGE NOTICE (Q-3
PUBLIC SERVICE COMPANY OF COLORADO FCAT ST VRAIN NUCLEAR GENERATING STATION CM /2656
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- 4*Sf INSTRUCTIONS Add New Pages Substitute Page(s) / for Page(s) 84.2& V-724:.
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FORM 344 22 4082 1h FORT ST. VRAIN CHANGE NOTICE NO. /(([d l (ENG 2)
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. FORT ST.-VRAIN NUCLEAR GENERATING STATION' Mechanical O NUCLEAR PROJECT DEPARTMENT Structural O
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'PUSLIC SERVICE'C::MPANY CF CCLCRACO Electrical T FORT ST. VRAIN NUCLEAR GENERATING STATICN Mechanical G NUCLEAR PROJECT CEPARTMENT . Structural G Safety Raiated g .
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SAFETY RELATED DESIGN ANALYSIS SHEET 1 PAGE 8-2 itorm 344 22-4083
% ELECTRICAL C MECHANICAL C STRUCTURAL
- f. SEISMIC ANALYSIS
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- 11. STRESS ANALYSIS N
lit. PIPING / HANGER ANALYSIS Nfb s IV. HYDRAULfC. PNEUMATIC ANALYSIS I t E N G 21
CN /fSS BY /4 R/cidV PAGE s-J. y SAFETY RELATED DESIGN ANALYSIS (continued) Sheet 2 of '5
-V. Thermal Analysis:
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' VI. Nuclear Analysis:-
AJ/A. l VII. Fire Protection Analysis: FA446 Af/DG //YJ74ufP I
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4 SAFETY RELATED. DESIGN ANALYSIS (continued) Sheet 3-of 3-
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-X. Accessibil'ity for In-Service Inspection, Maintenance and Repair:
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' FOR DESIGN REVIEW ETH00 X NO.N/A'
[(;) L (. )l 1. -'Were the inputs correctly selected and incorporated into design? -
~( d (')?(:) 5. _ Are assumptions necessary to perform the design activity
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adequately; described and reasonable?~ Where necessary .are the assumptions identified for subsequent re-verifications when the detailed design activities are completed? (d-(?)l() 3.- ^re the appropriate quality and quality assurance requirements specified?
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-including issue and addenda properly fied ident$ulatory requirement and are their requirements for design met?
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'([(,)(-)- 7. Was an appropriate design method used?
- ( ('-).() '8._:Is the output reasonable compared to inputs?
( ( ) (2). 9. Are the specified parts, equipment, and processes suitable for
.the required' application?
h ( ( -) ( )- 10. _Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed?-
-+ (d()'(111.-Haveadequatemaintenancefeaturesandrequirementsbeen specified?
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- 12. : Are accessibility and other design provisions adequate for performance of needed. maintenance and repair?
( ( )-~( )' 13. ': Has adequate accessibility been provided to rform the in-service inspection expected to be required d ring the plant life? (:)L( ) .( 14. Has the desi n properly considered radiation exposure to the public and p ant personnel?- ( ()-() 15. . Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have beten satisfactorily accomplished?
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.(-) ( ) ( 16. Have adequate pre-operational and subsequent periodic test requirements been appropriately specified?
( ) (-) 17. Are adequate-handling, storage, cleaning and shipping require-ments specified?
-(/[( ) ( )~ 18. - Are adequate identification requirements specified?
(y) ( ) (_) 19. ' Are requirements for record preparation review, approval, retentien. etc. , adequately specified? Note: If the answer to any question is no, provide adcitional information and resolutten below RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED CURING THE DESIGN VERIFICATION PROCESS L, w ~kg,.sk.-,w.s2,.
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PU Bl.lc S E RVIC E COMPANY OF COLORADO CN If66 3,' FORT ST. VRAIN NUCt. EAR GENERATING STATION Bf N//fI,/ianft, V P. N M ! B ~ n CHECK 1.lST OF DESIGN VERIFICATION QUESTIONS ""'*****"**** FOR DESIGN REVIEW MET 1100 Perm 344 22 40e8 YES .NO N/AMW ' IO O hihe inputs correctiv selected and incorporated into desion? O O d I.' ' Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are completeo? OO 3. ' Are the appropriate quality and quality assurance requirements specified? IO O 4. Are the applicable codes, standards and regulatory requirements including issue and addenda property identified and are their requirements for design met? IO O 5. Have applicable construction and operating experience been considered? IO O 6. Have the design interface requirements been satisfied?
- IO O 7. was an appropriate design method used?
O O s. Is the output reasonable compared to inputs? OOd 9. Are the specified parts, equipment, and processes suitable for the required application? O O T 10. Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed? O O I11. Have adequate maintenance features and requirements been specified? O O I 12. Are accessibility and other design provisions adequate for performance of needed maintenance and repair? O O # 13. Has adequate accessibility been provided to perform the in service inspection expected to be required during the plant life? O O d 14. - Has the design properly considered radiation exposure to the public and plant personnel? Are the acceptance criteria incorporated in the design documents sufficient to allow verification that
- O O . 15.
design requirements have been satisfactorily accomplished? l . O O [ 16. Have adecuate pre-operational and subsequent periodic test requirements been appropriately specified? O O f 17. Are adequate handling, storage, cleaning and shipping requirements specified?
.O O # 18. Are adequate identification recuirements specified?
O O 19 Are requirements for record preparation review, approval, retention, etc., adequately specified? NOTE: If the answer to any question is no, provide additional information and resolution below: RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED DURING THE DESIGN VERIFICATION PROCESS citM///9 eM D- &4 rf L L/* s.&, n,f
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. CHECK LIST OF DESIGN VERIFICATION QUESTIONS PAGE M.fC -l FOR DESIGN REVIEW METHOD Form 344 22 4o95 YES NO N/A
.W O O 1. Were the inputs correctly selected and incorporated into design?
O O [ 2. Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are completed? O O # 3. Are the appropriate quality and quality assurance requirements specified? OO 4. Are the applicable codes, standards and regulatory requirements including issue and addenda properly identified and are their requirements for design met? O O M 5. Have applicable construction and operating experience been considered? IO O 6. Have the design interface requirements been satisfied? YO O 7. Was an appropriate design method used? 6O O 8. Is the output reasonable compared to inputs? O O I 9. Are the specified parts, equipment, and processes suitable for the required application? O 10 6 10. Are the specified materials compatible with each other and the design snvironmental conditions to which the material will be exposed? O O I 11. Have adequate maintenance features and requirements been specified? O O d 12. Are accessibility and other design provisions adequate for performance of needed maintenance and repair? O O I 13. Has adequate accessibility been provided to perform the in service inspection expected to be required during the plant life? O O I 14. Has the design properly considered radiation exposure to the public and plant personnel? O O d 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished?
-O O [ 16. Have adequate pre-operational and subsequent periodic test requirements been appropriately specified?
O O [ 17. Are adequate handling, storage, cleaning and shipping requirements specified? O O 18. . Are adequate identification requirements specified? [O O 19. Are requirements for record preparation review, approval, retention, etc., adequately specified? NOTE: If the answer to any question is no, provide additional information and resolution below: RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED DURING THE DESIGN VERIFICATION PROCESS 6/fC/N . pg = shk
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-, m ISSUE STATUS ,- !- l , 's PR FORT ST. VRAIN NUCLEAR Gt:NenAuNUSTAF Urv ggg OR**
SerViCJ ru:Lic senvice c:MPANY OPQp 7.Go d , B Y_ ILEMbhLL l PAGE I ._,l CHANGE NOTICE REISSU E COMPLETE @ ACCT.# dFrtAtt 4 _ SYSTEM NO. E REISSUE # EI l SUBJECT OF ORIGINAL CN 1EEIPLLE - ESSEk'il&L SulHCEEM. s E1.E"10C6L CDMutCtton W 1 MLDLEs SMS'IDA i 41LsD - 440 40L1 TD.MLSFWNiiRJ. 4Rb CulttEMt uf%t t Irm. I 11EBC.TDQ.5. '- l DESCRIPTION OF REISSUE Orbalii. ' bot.VNSG S $NtTIEh filB R YGfh405 isSU 25 ; i i THIS CN RElSSUE INVOLVES. YES NO 1- Z 3 Change to the safety evaluation (Attach effective safety evaluation) 2 Z- E Change to the environmental evaluation (Attach effective environmental evaluation) 3 Z E Change in the safety related classification 4 Z B Change in design intent 5 Z E Charge in scope 6 Z E Change in installation procurement responsibilities , 7 Z-E Change to int instalation / procurement package 8 E Change to tne ces:ga background information package 9 5 : Change to tne de:ument upaate package 10 ~ E Change notte cancellation ,
. JVb ~ ATTACHMENTS:
CHANGE NOTICE REISSUG COORDINATION SHEET: EFFECTIVE DATE 2 M AFETY EVALUATION: OsE _ EFFECTIVE DATE 11 M ..
/ ENV;i C'.YENTAL EVALUr : OVERALL EFFECTIVE DATE 174 b _ _ . . .
REbE.s ADCROVALS: t C .', E E :NG_
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.I_ Unreviewed Safety Question
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- REQu' RED FOR SAFETY RELATED CHANGE NOTICES. l
7 , FORT ST. VRAIN NUCLEAR GENERATING STATION T S R/FC/TR na PUBLIC CERVICE C2MPANY CF COLCRADO PAGE 'l SAFETY EVALUATION ' ' C A TEGORY TYPts C CNSuBMe,TAL C SETPOINTCHANGE REPORT C TESTREQUEST E CNOVERALL C PROCEDURECHANGE IFSARI C OTHER C TEMnORARY CONFIGURATION REPORT et aan'81 CATION: ARE THE SYSTEM (S) EQUIPMENT OR STRUCTURES INVOLVED, OR DOES THE ACTIVITY AFFICT: CLASSI 3 YES C NO ENGINEERED SAFEGUARD C YES 3 NO SAFE SHUTOOWN 3 YES C NO PLANT PROTECTIVE SYSTEM C YES 3 NO SAFETY RELATED 3 YES C NO SECURITY SYSTEM C YES 3 NO REMARKS Ungrading 4160/480-Volt Distribution System EVALUATION Use Addstsonal Sheets If R . red l
- 1. DOES THIS ACTIVITY AFFECT STRUCTURES. SYSTEMS. COMPONENTS. EQUIPMENT. TESTS. EXPERIMENTS OR PROCEoVRES I DESCRIBED IN THE FSAR OR TECH SPECS) 3 YES C NO UST THE APPUCA8LE SECTIONS REVIEWED: All sections of the FSAR and Technical SDecificationS were reviewed for electric cower recuirements associated with the 480-volt essential buses. FSAR Sections 1.2.2.9. 8.2.2. 8.2.5. Ficures 8.2-17. 8.2-18. 8.2-22. and i
8.2.25 and Technical Soecification LCO 4.6.1 are acolicable. DOES THE ACTivlTY REQUIRE THAT CHANGEtSi BE M ADE TO THE FSAR OR TECH SPEC? C fES 3 NO - 2. UST SECTIONS TO BE CHANGED AND THE CHANGES TO BE MADE: FSAR Section 8.2.2 was chanced in undated FSAR Revision 2 to reflect the installation of the new 4160/480 y transformers and the new 480 V AC essential switchaear. N; further revisions to the FSAR are necessarv. No changes to the Technical SDacifications are recuired as a result of this activity. 1 DETERMINS WHETHER OR NOT THE ACTIVITY INVOLVED IS AN UNREVIEWED SAFETY CUESTION UTIUZING THE FOLLO I Al NAS THE PROBABluTY CF OCCURRENCE OR THE CONSECUENCES OF AN ACC' DENT OR MALFUNCTION OF EQUIPMEN iMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE FS AR BEEN INCREASED 7 ! C YES E NO STATE BASIS.The uagrading of the 4160/480 Volt distribution svttom with larger canacity transformers and associated 480' V AC essential switchcear will not effect the coerability of the electric system so as to increase the Droba-bility of occurrence or the consecuences of an accident or malfunction of eou1Dment imoortant to safety. The relocation of the transformers (from indoors to Outdoors) and reelacement of ones of larcer cacacity orovide: the electric system (4160/480-volt) (Continued On attached page)
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181 HAS THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN ANY EV ALUATED PAEVI IN THE FSAR SEEN CREATED? C YES 3 NO STATE BASIS: The canability of the electric evetam tn meet nlant enuinment recuirements has not been changed by the ungrading of l the alan /AAA vnit transfnrmart and attnciated switchgene. The reduction in the heat Inad in the AAn valt twitehnear rene due to relocation of 4160/480 volt transformers will result in a better environment for the essential electrical equipment (Continued. ..) IC) H AS THE MARGIN OF SAFETY. AS DEFINEDIN THE B ASIS FOR ANY TECHNICAL SPECIFICATION OR IN TH STATE BASIS: No margin of safety is defined in the Technical l l C YES E NO ! 9nacificatient or FSAR concernina the 4160/480-volt transformers and associated twitehnmar. The overall niant margin of safety will be increased due to the uagrading i of the tvetam at a retult of better voltace reculation and ecuiement Drotection l r nnerti na'ti nn , DOES THE TlVITY APPEAR 70-. NVOLVE AN UNREVIEWED SAFETY CUESTION A YES 1 NO l 8 ANT C YES b NO /
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.with reduced. voltage drops (better regulation) to.the safety-related
. equipment and the capacity to meet bus-tie load demands as well as possible future needs.
3-(b)' housed in this room. The relocation and change from an indoor, dry
-type, self-cooled transformer to an outdoor, oil-filled, forced air--
-cooled transformer equipped with a fire suppression deluge system has
- not created the potential for a transformer fire affecting plant safety-related equipment. The transformers are located approximately feet from the nearest. facility (Building 10) housing' safety-related
- equipment. Each transformer bay is segregated by a concrete fire wall and surrounding dike with each bay designed to contain the approximately
- 750 gallons of transformer oil plus 20 minutes of 100 gpm of deluge.
Building 10 is constructed of concrete with each area housing safety-related equipment surrounded by a 3-hour fire-rated wall. This prevents
.any potential for a fire originating from these transformers affecting any safety-related equipment.
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B FUBLIC SERVICE COMPANY OF COLORADO e FORT ST. VRAIN NUC1. EAR GENERATING STATICN CNI M
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ENVIRONMENTAL EVALUATION NO. tA % PAGE 1 Y oom m-tsaa a Issae:[eioIiDaW M
- M a wf/>4k h %'4 W hWI W M ?5td3sA*iCM-ME Mt-M M,_I TYPE: 5. CN Cverall C CN Submittal Other Are all measurable nonradiological effects of this actvity confined to the on site areas previously disturted dur:r.g site preparation, plant construction or previous plant operation? 3.Yes C No State basis
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is the actvity required to achieve compliance with Federal, State or !ccat environmental regulations? Z Yes No NOTE: If either answeris Yes. then sign and date form. If both answers are No, then complete remainder of this evaluation form. 4 avm U M e O N . - 4ti 7-E r.rG w & %=-1 ;rtis t . .t. n-Mn.. . M + d f? 2 &AMk W % % % GE M *]
- 1. Is the activity identified in the final environmental statement (FES) Yes C No Or Supplementary Environmental Occuments (See Q 3)? .
Identify documents and document sections reviewed
- 2. Determine whether or not the activity involved is an unreviewed environ -ental cuestion using the following guidelines.
(if the answer to any of the fe!!owing questions is Yes, then tt:s acovity involves an unreviewed environmental Question.) (A) Will this activity result in a significant increase in any adverse env:ronenental. impact previous:y evaluated in the FES7 Yes Z No State basis: (S) Will this actvity result in a significant enange in tne types, or a significant increase in tne amounts of effluents, er a significant increase in the authorized power levep Yes : No State basis: (C) Does this activity involve an environmental matter not previous:y reviewed ano evaluated in tne FES7 l Yes C No State basis: l Oces the activ"y involve an ur evie- d environmental cuestion? Yes M No M 4*" 8 deproved * /2- / bU By
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COORDINATION SHEET FOR CHANGE NOTICE DESIGN OUTPUT PACKAGES
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Date '7/1/da"~
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Internal Design Interfece Reviews - Engeg. Engrg. Not Engrg. Not Complete Complete Required By By 7. A LJ+w
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Date 7 S-b C E Elec. Mech. MO O O O E By # Ma <-+ * -. Date 7-NC Struc.- 0 0 E By RN Date 7 ~2 2.-8 Licen. f f .. Approved [-$ c Effective '7[2 '1 9[ - PAGE SCHEDULE 36 -* 13 ' 6 (, 6 823A ' AN9 + Atts di79A 'AlisA 4' 3p.IA 6G.8 -9 C 6.5 ^#84 ' 6/f7 A As fe d 6157 -+ 4.1W V7 23 67-*&13 6/// dt#7
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DESIGN 1 l BACKGROUND I 1 INFORMATION l e . FORM 244 22-4052 t
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FORT ST. VRAIN CHANGE NOTICE NO. Ibk ! (ENG 2)
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PUBLIC SERVICE COMPANY OF COLORADO CN lECt 4 FORT ST. VOAIN NUCLEAR GENERATlWG STATICES BY bMM DESIGN INPUT REQUIREMENTS PAGE b l poem 344 22 4000 ' 5- ELECTRICAL
~
O MECHANICAL C STRUCTURAL E 3AFETY RELATED C NON. SAFETY RELATED The existing 4160-480 Volt,1000/1333 KVA transformers are indoor dry
- type located within the essential switchgear. The high loss, high operating temperature characteristics of these units have caused numerous temperature related problems for other equipment located in this room. The transformers themselves have failed four times in the past due to high loading and high temperature. The new transformers will be installed on CN-1605C; all electrical connections will- be done by this CN. I The 480 Volt essential switchgear is presently equipped with a 1600 Ampere bus and main circuit breakers. The 1600 Ampere circuit breakers have a continuous rating of 1333 KVA, an exact match for the transformers.
The switchgear cannot be upgraded due to the bus size and cubicles which will. not accept higher-interrupting capacity circuit breakers. The normal' load on this system is 1000-1300 KVA, which is at the upper limit of both the transformers and switchgear. The existing arrangement of three buses and two tie breakers was originally designed to provide backup to any but in the event of problems or maintenance
, on one of the 4160 Volt feeders.- This is not possible with the loads we now have.
Replacement of the transfortres and essential switchgear will allow the 480 Volt system to operate as originally designed, i.e., operate two buses from one transformer. It will also give added capacity for future growth. The new switchgear is the same height and length as the existing; however,
-it is 10" deeper. This will cause no problems since there is adequate room between the barrier walls.
The additional weight varies from 2100 to 2900 pounds; however, since the switchgear covers a larger floor area the increase will not overstress the 480 Volt room floor slab. L. The switchgear has been seismically qualified by 'the manufacturer; the I test reports will be placed in the Seismic Books. PasPAnnoavMda.es sieATuns MI M> & N enoVED SY -
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- SIGNATURE DATE (ENG 2)
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-PUBLIC SERVICE COMPANY OF COLORADO CN V2.Cp.k
- : FORT ST. VRAIN D'UCLEAR GENERATING STATIOM BY bMDk'L i DESIGN INPUT REQUIREMENTS PAGE S-2
.f Form 344 22 4060
& ELECTRICAL s O MECHANICAL O STRUCTURAL T SAFETY RELATED 0 NON SAFETY RELATED The bus duct supports have been calculated to be acceptable for both vertical and horizontal loading.
Due to the additional transformer capacity, the current transformers in the 4160 Volt feeder cubicles must be replaced and the ameters rescaled. The fire protection deluge system for the transformer area installed as part of CN-1605C will-be wired as part of this CN. New cable from the 4160 Volt load centers to the new transformers will be installed in a duct bank that was installed on CN-1605A. This cable has been selected to allow for future growth of the system. Current limiting reactors installed by CN-1629 will be connected on this CN. These will be used to limit the fault current between the essential load centers and the essential motor control centers. Spray shields on existing switchgear will be rebuilt for use on the new switchgear. These-shields serve to protect the switchgear from the M [Eed $ system. L
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, PUBLIC SERVICE COMPANY OF COLORADO CN 17.
FORT ST. VRAIN NUCLEAR GENERATING STATION
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DESIGN ANALYSIS PAGE __ S -3 Form 344 22 4062
& ELECTRICAL {
s O MECHANICAL 8 O STRUCTURAL ' 82-SAFETY RELATED ' O NON-SAFETY RELATED I e; The transformers which were installed on CN-1605C will be connected on this CN. .The transformers are equipped with winding temperature re'ay, l ' pressure relief device, rapid rise relay, liquid temperature and liquid level. relays. The rapid rise relay will 1) trip and block close the associated main circuit breaker and 2) initiate operation of the deluge system. , i i The existing 200:5 current transformers in the 4160 Volt cubicles that I feed the 480 Volt load center transformers will be replaced with 500:5 i current transformers and, the associated ammeters rescaled accordingly; ; The existing 480 Volt essential load center will be replaced with Brown-Boveri ', (formerly ITE) K-line switch gear. It is equipped with a'4000 Ampere bus rated at 85 Ka Rms. ' The feeder circuit breakers are fused 600 Ampere, 200 Ka Rms. The tie '
-breakers are fused 3000 Ampere, 200 Ka Rms and the diesel generator circuit j breakers are unfused 3000 Ampere, 85 Ka Rms. The main circuit breakers- -
are 4000 Ampere, 85-Ka Rms. , The load centers have 13 spare circuit breakers, both M.0. and E.0., ' and 15 future compartments; this is a total .for all three buses. ' New. relaying will be installed on the EHC pumps, three relays per pump. Two of the relays will alarm at 150% full load current while the third will trip the circuit breaker at 200% full load current. Current transformers and ammeters will be-installed on each motor. control center feeder breaker. 'They will be on one phase only and used for local indication at the switchgear. The 480 Volt bus duct is non-segregated phase metal enclosed ITE type , XL-U, 4000 Ampere, 250 Ka Rms minimum. Installation of the outdoor portion i will be done on CN-1605C, the indoor portion on this CN. ! t
' Bus tie cable ~will be upgraded from 3-1250 MCM cables per phase to 4-1000 l
MCM cables per phase. This will raise the bus tie capacity from approximately ' 1400 Amperes to 1800 Amperes. l t The fire protection system which was installed on CN-1605C will be wired ! on this CN. ' 6e e-r e reucsue of kl PREPARED B yyy y ATU'R E DATE f (ENG 2)
~~-
PUBLIC SERVICE C O M P A N'Y ' O F COLORADO CN \2,%k FORT ST. VRAIN NUCLEAR GENERATING STATION BY Om tl
.., - DESIGN INPUT REQUIREMENTS PAGE3 L lk Form 344-22 4060 '
% ELECTRICAL C MECHANICAL C STRUCTURAL 9, SAFETY RELATED C NON SAFETY RELATED 12.ERV EML 4 Ce%ES - (%n, %13 3 M'l 4 ,%15)
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.d. FORT ST. VRAIN NUCLEAR GENERATING STATION
, ,C'T PubllC CN 129d
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- E PUBUC SERVICE COMPANY OF COLORADO VY*V SGIVIC9= BY VfN" CALL DESIGN INPUT REQUIREMENTS DAGE ~B 32A 3 ELECTRICAL G MECHANICAL 0 STRUCTURAL 3 SAFETY RELATED C NON SAFETY RELATED -
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Form (A> 344-22 4000 (ENG-2) ' I
r-PUBLIC SERVICE COMPANY OF COLORADO CN I2.#Ik-FORT ST. VRAIN NUCLEAR GENERATING STATION gy{ SAFETY RELATED DESIGN ANALYSIS SHEET 1 PAGE ~8-4 Porm 344 22 4083
- 8. ELECTRICAL
- O MECHANICAL
~
O STRUCTURAL f I
- i. SElSMIC ANALYSIS The switchgear has been seismically qualified; the test report will be placed in the Seismic Book. The bus du::t supports in the 480 Volt room have been desi duct. &c ed for dead weight
- 10. load and seismic loading of the bus
.ges 38 l-4r. -t i t. STRESS ANALYSIS f Stresses for the additional loads created by the switchgear are within acceptable stress levels by engineering judgment and by field inspection of existing equipment.
Per telecon with Sargent & Lundy on 12/12/83, the floor in the 480 Volt Room was designed for 350#/ft2 . Max. estimated floor load due to new switchgear is - 250#/fta, therefore no stress problems are anticipated.
' lit. P!PfNG/ HANGER ANALYSIS This analysis does not apply to this CN since no piping systems are involved.
I V.- HYDRAULIC / PNEUMATIC ANALYSIS Does not apply to this CN. (ENG 2) i _ _ . . . I
PUBLIC ~ SERVICE COMPANY OF COLORADO . CN L 2. FORT ST. VRA1M NUCLEAR GENERATING STATIOM SAFETY RELATED DESIGN ANALYSIS- SHEET 2 BY h & &tt ! i [ PAGE_ S-5 Porm 344 22 4064 V. THERMAL ANALYSIS l' Removing the transformers from the 480 Volt room will remove a considerable source of heat, adding to the reliability of all equipment in the ro,om and reducing the load :n the HVAC system. M'mWA
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. VI. NUCLEAR ANALYS:S Does not apply.to this CN.
Vll. FIRE PROTECTION ANALYSIS , This analysis will not change since the new switchgear is the same manufacturer as the existing, the only difference being fuses on the circuit breakers. The bus duct is not combustible and will contribute no more as an ignition source than the present under floor cable. All new cables will be in conduit or flamastic coated. The removal of the transformers will also remove much of the internal heat in the switchgear. Vill. ENVIRONMENTAL ANALYSIS Since the switchgear is a replacement item no different conditions will exist. The new switchgear will have a spray shield installed similar to the existing shield. IX. , COMPATIBILITY OF MATERIALS, EQUIPMENT, AND PROCESSES Installation of new switchgear requires the use of current limiting reactors on each essential motor control center. All equipment is compatible and being used for its designed use. L - (ENG-2) e
a PUBLIC SERVICE C O M P A'N Y OF' COLORADO CN 12 M FORT ST VRAIN NI)CISAR GENERATING STATION SY (SN1ttt.\ SAFETY RELATED DESIGN ANALYSIS SHEET 3 PAGE T3-fo Form 344 22-4069 i X. ACCESSIBILITY FOR IN SERVICE INSPECTION. MAINTENANCE, AND REPAIR Accessibility for maintenance, inspection and repair will be no less
~
than with the existing switchgear. However, a slight improvement may be noted due to the increased depth of the compartments. I i XI. SEPARATION AND SEGREGATION ANALYSIS This analysis has not been altered with the installation of the new switchgear since it is a replacement item and the barrier walls have not been modified. Xil. ELECTRICAL ANALYSIS The new switchgear will give better circuit protection due to solid state trip heads and fused breakers allowing for better coordination of f rcuit protection. Connection of the current limiting reactors will give an approxima'te 1". voltage drop at the moto) control center. The current
-limiting reactors will have a box attached to the top for use as a conduit entrance. For cable routing purposes this will be considered as part of the curren$. limiting reactor.
Xill. ACCIDENT ANALYSIS Oces'not apply to this CN, however additional protective relaying for this switchgear will be installed by CN-1622. PREPAREO BY OATE (ENG 2) t ,-
PUBLIC SERVICE COMPANY OF. COLORADO CN /2 FORT ST. VRAIN. NUCLEAR GENERATING STATION
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DESIGN ANALYSIS PAGE b [a,IA Form 344 22-4062 6 ELECTRICAL O MECHANICAL O STRUCTURAL
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, ) SerVICD PUBLIC SERVICE COMPANY OF COLORADO BY d~AIDA Ll DESIGN ANALYSIS PAGE Eb b3 E ELECTRICAL 0 MECHANICAL O STRUCTURAL E SAFETY RELATED C NON-SAFETY RELATED ,
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CHECK LIST OF DESIGN VERIFICATION QUESTIONS PAGE M FOR DESIGN REVIEW METHOD Forrn 344 22 4095 YES NO N/A 1, were the inputs correctly selected and incorporated into design? E O O 8 O O 2. Are assumptions necessary to perform the design activity adequately described and reasonable? Where, necessary, are the assumptions identified for subsequent re verifications when the detailed design activities are completed?
'S O O 3. Are the appropriate quality and quality assurance requirements specified?
E OO 4. Are the applicable codes, standards and regulatory requirements including issue and addenda property identified and are their requirements for design met? E O O 5. Have applicable construction and operating experience been considered? 3 O O 6. Have the design interface requirements been satisfied? S O O 7. was an appropriate design method used? C5 O O 8. Is the output reasonable compared to inputs?
'E O O 9. Are the specified parts, equipment, and processes suitable for the required application?
E O O 10. Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed? E O O .11. Have adequate maintenance features and requirements been specified? E O O 12. Are accessibility and other design provisions adequate for performance of needed maintenance and repair? E O O 13. Has adequate accessibility been provided to perform the in-service inspection expected to be required during the plant life? O O E 14. Has the design properly considered radiation exposure to the public and plant personnel? E O O 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished? E O O 16. Have adequate pre-operational and subsequent periodic test requirements been appropriately specified? RO L] 17. Are adequate handling, storage, cleaning and shipping requirements specified? E O O 18. Are adequate identification requirements specified? E O O 19. Are requirements for record preparation review, approval, retention, etc., adequately specified? NOTE: If the answer to any question is no, provide additional information and resolution below: r RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED DURING THE DESIGN VERIFICATION PROCESS
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20.$&~s DESCRIPTION OF REISSUE h0Cf./Ald"D1" UA$AM AA.)W- , ALLt*7A//df/0L) - l*4ddr Ab. 3ddlkt W4f JA/C/4,IAllV LNlrA $(/1~ /b Ar Aftf76D /n 4*Al./33Z6, 7#ff CV b".Zff06 (f) CHA1]Cf5 " D &~l & ~ 7 / 8 1 ) ** rss -rws.s e'A A_' d- n "fAAA/A.JtC-dv/7~ " di 7Mr t*A As r ns .MP4 ?%Wr eAdtd~- IdVLl' fMMY M M. f./ff:0 s/ z*d- /'2 72 A !. w a:a w e a s- .P e r Qs.) creu'e: nw pdwg 7We strA u~ n s9M. t. THIS CN REISSUE INVOLVES: YES NO 1 Z X Change to the safety evaluation (Attach effective safety evaluation) 2 Z E Change to the environmental evaluation (Attach effective environmental evaluaton) 3 Z E Change in the safety related classification 4 Z X Change in design intent 5 Z E Change in scope 6 Z X Change in installation / procurement responsibilities 7 Z E Change to the installation i procurement package 8 Z X Change to the design background information package 9 % Z Change to the document update package 8 10 Z 3 Change notice cancellation (/ ATTACHMENTS: CHANGE NOTICE REISSUE COORDINATION SHEET: EFFECTIVE DATE 8 ' 7 - P.5~ SAFETY EVALUATION: OVERALL EFFECTIVE DATE h'- b ~~S M ENVIRONMENTAL EVALUATION: OVERALL EFFECTIVE DATE ~~ 30 -d 3 VIEWS, APPROVALS: UCLEAR
. ENGINEERING
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2 PS- . .r e. PROD. i b h l0, TAM M i y 3.a r. . . o ra,
'PORC MTG # AND DATE PORC 6 0 7 FER A .1Q95 a veS Np fetySignificant I Z _. reviewed Safety Question l Z Z Unreviewed Environmental Question aA ERt X -
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- REQUIRED FOR SAFETY RELATED CHANGE NOTICES.
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. PUBLIC SERVICE COMPANY OF COLORADO f
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** N O. 1117 R SAFETY EVALUATION PAGE 7 TYPE: -
UCN OVERALL CCN SUBMITTAL CSETPOINT CHANGE REPORT CTEST REQUEST CTEMPORARY CONFIGURATION REPORT CPROCEDURE CHANGE IFSARI 20THER CLASSIFICATION: ARE THE SYSTEMIS) EQUIPMENT OR STRUCTURES INVOLVED. OR OOES THE ACTIVITY AFFECT. CLA$s 1 3 YES 2 NO ENGINEERED SAFEGUAAD 2 VES 1 NO ' SAFE SHUTOOWN 3vES 2 NO PLANT PACTECTivE SYSTEM 2 vES 1 NC SAFETY RELATED a vES 2 NO SECURITY SYSTEM 2 VES I NC l REMARKS I 15 THE ACTIVITY IDENTISIED IN THE F S A R OR TECH SPEC? a.a Y E S 2 NO LIST THE APPLICA8LE SECTIONS AEviEWEO. FSAR Sections 7.4.3. 8.2.2.3. 8.2.3.4. 8.2.5.2. 8.2.2.4. 8.2.3.5. 8.2.3. Figures 8.2-26. 27. 28. 29. Tech Soec LCO 4.6 (Ret'ers to Revision 2) .._ _ _ 2 OCES THE ACTIVITY REQUIRE TH %T CHANGELS) BE MACE TO THE F S A R OR TECH SPEC? ] YES , NO LIST SECTIONS TO BE CHANGEO AND THE CHANGES TO BE MACE. FSAR Sections 7.4.3. 8.2.2.3. R.? 1.4, R 7 3.7 8.2.2.4, Elgures_8.2-26. 27 J 8. 29. & Tech Saec. LCO 4.6 thnuld bg_ caviled _tQ reflect this modification. (Revision 2 has incornorated thace changes )__ _ _. _ _ _ _ _ _ _ _ _ _ _ Rafaranca .to_the PPS battery. and_iti_. battery charger /inve.cter should be dalatad fen = tc0_4JL ta_ incorporata_the_ af. facts of titis_modif.ica. tion.. J OETERMINE WHETHER CA NOT T*E ACTIVITY INVOLVED IS AN UNAEV EWED SAFETY OUESTION UTILizlNG THE FOLLOWING GulOELINES.
! Al M AS THE P A08 A81LITY 05 OCCURRENCE CA THE CONSEQUENCES OF AN ACCIDENT OR V ALFu'.C?iC'.
OF ECulP'.'ENT IVpO A TANT TO SAFETY FAEviOUSLY EV ALU ATED IN TME F S A A BEEN INCAE SE0' 2 vES T NO sT4TE a4 sis __Ihis change _ increases _the _capacit.y_ and rel_iability .o-the _ Don.-interruptible bus 1C system by installing a backup power transformer and new battery cbarger, and the 125 V_D C. system by upgrading battery charger.10. An addition; DC source _has_ been made available to the Battery / Charger / Inverter _1C._ configuration. By increasing __the capacity and reliabil.ity. and . adding a dedicated backup _ supply from 480V EssentiaLBus #1.via Inst. PwrmTransfonner IC-1 and an inverter / static transfer switch _ the overall integrity of the (see attached sheet...) 81 M AS TmE POSSIBILITY OF AN ACCIOE'. 08 'lAL=UNCTICN OF A OlF =E:EN T Ty pE Tw a% a*. E.L. _*i; aAEviouSLv iN TwE F s 2 A sEE . C= E are-' =S 7c s :Ts s: Sis The discussion abov in item 3(A) concerning increased reliability and capacity also applies Tiere. Since the system reliaEiility and capacity has been enhanced by this modificatien, the possibility a new adicFent or malfunction of equipment as a result of a non-interruptible power
, supply failure has not been created.
.C -as 7-E ' asiN ,= :_=ET> _ s := = N E: iN -E s: Sis 3 :a AN 5 :e.. :A L sa E::=.:_ - -
s2= s EE . = s ; _ ;E-- ~ =; I NO STATE s25:s Increasing the reliability and
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! capacity of the non-interruptible power supply system increases it,s margin of safety as defined in the basis for LCO 4.6. 00ES T-E ACTO.7* Ap p E a 8 TO t'.% C_VE AN LN A EVIEWED SAF ETY Ot ESTION [ YES .3 N C 35 SA*ETs SIGNisiCANT .X,VES __ N O 8v 2 MMA *APPAOVED p .... . m . . i / / c 4,e f V se -pas 'f ~ dio"ei.8
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. non-interruptible power supply system is improved. The probability of occurrence or the. consequences of an accident or malfunction of. equipment important to safety previously evaluated in the FSAR as a result of a non-interruptible power supply failure has been definitely
- reduced by this nodification.
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r . TYPE: C CN Submittal Cmer hCN Overes Are all measurabie nonrediologscal effects of this acavrty c nfired to de en site areas previously disturned during sta properation, pient conommcoon or pronous plant operanoni se ! No N AAAdikt.a.% cm W a Ect f3 e.a.wheed' 01 alA u meud,
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- 1. Is the acenty idenafled in the finsi emnronmental statement i 8E51 Yes ! No Or Suosiememory Envwonmental Occuments t See Q 3J Idemify documems and document secnons reviewed l i
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PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION CW . /88[/* BY[//f8f/ ~ v . COORDINATION SHEET FOR
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internal Design Interface Reviews Engrg. Engrg. Not Engrg. Not Complete Comolete Required By By - 7 '7h Date I!?- if Elec. X U C Date /-2437- Mech. O O E By - /f rygaer Date j-JP M Struc. O O 3 By M % YZvs. Date 2. ~S Licen. Appr ed Effective Date 2 !7!8.7 PAGE SCHEDULE A-/A 6-is 4.zfe c.sg 4.gu y; 4- A / 6-2 44*8 q.6p.ig y. t. I A 23 C.z./ 4 4o.1g 4.ya Y. s. tn
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Sheet 2 of _3 V. . Thermal Analysis:
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CHECK LIST 0F DESIGN VERIFICATION QUES' IONS - Cent'd. 1 YES NO,N/A
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j - t l- (/j()() 17 Are adequate handling, storage, cleaning and shipping require-i- ments specified?
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(/)()() 18. Are* adequate identification requirements specified? j '(/)()(-)' 19. Are requiranants for record preparation review, approval, retention, etc., adequately specified? g:- If the answer to any question is no, provide additional information and resolution below: RE30tuTION'0F DESIGN DEFTCIENCICS UNCOVERED DURING TNE DESIGN VERIFICATION PROCESS ll- O c . ( hI
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.O O 2. Are assumptions necessary to perform the design activity adequately described and reasonable? Where
!I necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are completed? l Or O 3. Are the appropriate quality and quality assurance requirements specified?
OO 4. Are the applicable codes standards and regulatory requirements including issue and addenda property identified and are their requirements for design met? IO O s. Have applicable construction and operating experience been considered? 6O O G. Have the dasign interface requirements been satisfied? IC O 7. Was an acerooriate cesign et*cd useo? 2' C O 8. Is tne output reasonaole comparea to inputs? 6'O O 9. Are the specified parts, equipment, and processes suitable for the required application? Z C C 10. Ace :n a spd.f.cc mataris.5 :.ro.ticie nitn each other and tne design environmental conditions to which the material will be exposed? IC C 11. Have adequate maintenance features and requirements been specified? [C C ' O. Ars 3 : ::ib::: y a-d other ::::; . Orr. :: ns adequate :r performance 8 of needed maintenance and repair? O C 13. Has acacuate accessibility been provided to perform the in service inspection expected to be required curing the plant life 7 O C I 14 Has the design properly considered radiation exposure to the public and plant personnel? YC C 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that casign requirements have teen satisfactorily accomplished? IC C 1c. Have adecuata pre ocerational ano subsecuent periodic test requicements been appropriately specified? IC C 17. Are adecuate hanol;ng, st: rage, ::eaning and shipping requirements specified? IO C 18. Are adequate identification requiraments specified? u ] [ 19. Are requirements for record creparation review, approval, retention, etc., adequately specified? j NOTE: If the answer to any question is no. provide additional information and resolution below: RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED OURING THE DESIGN VERIFICATION PROCESS tG f~ws ras 'A s b f.tsy,, A w wsos /~ A D ra,n % ~3r E ]<7MM A'en o s*f94^! XMt /6 Alnat fhrfacsfy 4Jes da~sss 9 sis't A ".%S*C. Es*$*wrt** : Cwds'M 'A c. Anom yss.,t Cao
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OO y Y - soie the inputs correctiv seiected and incorporated into design?
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-O O 62. Ate assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are completed?
O O d ' 3. Are the appropriate quality and quality assurance requirements specified? O O d 4. Are the applicable codes, standards and regulatory requirements including issue and addenda properly identified and are their requirements for design met? O O a' 5. Have applicable construction and operating experience been considered? O O 5 6. Have the design interface requirements been satisfied? - O O 3 ?. Was an appropriate design method used? O O Z 8. Is the output reasonable compared to inputs?
-O O 6 9. Are the specified parts, equipment, and processes suitable for the required application?
O O 3 10. Are the specified materials compatible with each other and the dec.ign environmental conditions to which the mcterial will be exposed?
-O'O O' 11, Have adequate maintenance features and requirements been specified?
O O 3 12. Are accessibility and other design provisions adequate for performance of needed maintenanos and repair? O O [ 6 13. Has adequate accessibility been provided to perform the in-service inspection expected to te : equi:od during the plant life? O .O 3 14. Has the design properly considered radiation exposure to the public and plant personnel? E O - 0 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished? O O d 16. Have adequate pre-operational and subsequent periodic test requirements been appropriately specified? O O M 17. Are adequate handling, storage, cleaning and shipping requirements specified? 5O O 18. . Are adequate identification requirements specified? 5O O 19. Are requirements for record preparation review, approval, retention, etc., adequately specified? NOTE: If the answer to any question is no, provide additional information and resolution below: RESOLUTION OF DESIGN DEFICIENCIES UNCOVERED DURING THE- DESIGN VERIFICATION PROCESS
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- design requirements have been satisfactorily accomplished?
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