ML13246A327

Energy Northwest Documents Regarding an Engineering Change and 10 C.F.R. 50.59 Evaluation

ML13246A327
Person / Time
Site:	Columbia
Issue date:	04/10/2012
From:	Energy Northwest
To:	NRC/OIS/CSD/FPIB
References
FOIA/PA-2013-0317
Download: ML13246A327 (49)

Text

,all ENERGY EC Number

-- NERGYS DMS REFERENCE DOCUMENT nfa

-**NORTHWEST D

SE u~

Peopl,- Vision -Solutions

]INDEX SHEET aed cal 5.17.19 3

Primary Document Identification, D ocum entocu en N u mberS ee D

um n Type Sub-Type Document Number Number Revision Input References OutputReferences ADD DELETE Type Subtype Doc Number Sheet No ADD DELETE Type Subtype Doc Number Sheet No I

1T

]

PPI PPM OSP-ItHRRIST-Q702 2

PPI PPM OSP-RHR/IST-Q703 43 E.

PPI PPM oSPýRHRAST-Q704 7

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0 I,

[T _

[].r L3

]

Cl C3 l.'

14 1

]...

D

[

15

]

O 19

[]

0 U][

Note:

Note':

~W

~W m.-m 25405 R2

Ezngineezring C22ange EC :z~t DC2D~

CI1

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t J7V iI 2

I l l llll I l l I I l lll E N, E' "Y NORTHWEST

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br

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r A Iz:k+%

Ai L.

p.Lw i'

M4iles tone Date

?&"mPott namre 7 /

~-

~

~*7 M Raq By Syw t.i F'acSy DeOaarptziof Affwcted Ddcumets Z~ist Sub-Fac Lý~

• P Docu~nont Shoot

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n Cross Aefoez~cos Rof.

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Sub-Numbe r Number Descriýption

.'.4-

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f.r i:

Engineering Change Print Date:

04/10/2012 EC Number 0 000010922 000 "GY Faclit 02ENERG Type/Sub-type:

CMR NORTHWEST Page:

3 Topic:

PURPOSE/SCOPE Last Updated 5 y DGREGO From Panel TIME100 Last Updated Date:

04/09/2012

'Text Status

-_UNLOCKED The purpose of this CMR is to add the RHR.heat exchanger into the flow path for LPCI injection.

The analysis to determine the pressure reading on the test gauge currently assumes that all flow goes through the RHR heat exchanger bypass line. The system is always configured with the inlet and outlet valves of the heat exchanger open.

Therefore, the heat exchanger is part of the flow path.

c~

o.,.j

_K Q-

Topi.c Last Updated iBy Fircm Panel F:EiE'Ca Last Updated Date:

Text Status Talý,

vaýUe, Fol

-,ho;zld i:m 5511.

Zngr~ineering Change EC Number Facility Type/Sub-type:

0000010922 000 02 CMR Print Date:

04/10/2012 ENERGY NORTHWEST Page:.

_- A-Topic PREDICTED RESULTS Last Updated By DGREGO From Panel TIME100 Last Updated Date:

04/10/2012 Text Status UNLOCKED Crediting the RHR heat exchanger flow path in addition to the bypass decreases flow in. the bypass line from 100% down to 56.25%..

'Using a more conservative Cv value for.RHR-V-48A(B) combined with using the parallel flow path of the bypass and heat exchanger results in an increase in margin of 2.9 psi. It is prudent not to use ail of this marginh therefore

-80%,

or 2.25 psi will be incorporated into the procedures.

Therefore, the acceptance spreadsheets in OSP-RHR/IST-Q702 and.OSP-RHRiIST-Q703 will be reduced by 2.25 psi.

There are 6 outstanding Modified CMRs against this analysis.

This CMR modifies the results of CMR 10916, 3084 and 93-0588.

AR assignmnent 00261301 has been created to track incorporation of CMRs either into one CMR or into the base calculation..

ENERGY NORTHWEST PR Cti tr VERIFICATION CHECKLIST FOR CALCULATIONS AND CMRs Caltutation/CMVR 10922 was ver'fied ujsingo the fo~lowing melthocs:

0 Checklist saiow Revision 0 E]

Aternate Oa-culatom(ws)

Checklist Item Clear statement of pUrpose of analysis.............................................................

Methodoloj IS clearly Stated, sufficientfy detailed, and appropriate for the proposed application...........................................................................................

Is raw data (PDiS) being used as a calculatiorWOMR iitput?

0*Yes No.......................................

If yes, ensure the data was 9de0uayvaJlv ated (W1. DES-4-1 step 2.3).........

Does the analysiSicalculatirn:

Methodology (nmciuding criteria and assurr.tions) differ from that 0escrrib in t,,.e Pant or ISFSI FSAR or NRC Safety Evaluation Report, or are the results of tic ana~ysis/ca!cu!aton as described in the Plrent of ISFSI PSAR or NRC Safety Evaluation Report" M Y e s E N o..................................................

nvolve a chanqe to an SSC tnat adver-sey affects an FSAR descr-ttd Design F*nctior o-to an SSC or Cask design thal adverely affacts an ISFSI or Cask FSAR described Destgn Function?

nYou M)No.......................................

Invoive a cf ange to a procedure Ilat iaversely affects how FSAR described SSC destgn functions or how ISFSf or Cask FSAR described $SC design funcrions are perfor ed or controlled?

Z Y e s

' 1 N o...............................................................................................

If yes to any of the above questions, ensure that the requirements cf 10 CFR 50.59 ar*wdr. I0 CFR 72 48 have he.er processed In accordance wth S W P -L IC -0 2.............................................................................................

Does the analysisfceicula; ion result require revising artn* exxs{tirg output interfbce document as jdenitfied in DES 4-1, Attachment 7.3?

ID Yes

[I No............................................

f yes, ensu-e that the appropriate actions are taken to revise the outpu interface document per DES-4-1. Setion 3.1.8. (i.e., document change is initiated irn accordance with aoilolcable procedures).........................................

Does the Coulatiofl's Summary of Resaut (or CMR's Predicted Results) Mnclude a discussion of avaflabie margin4

[l Yes E No 0 Not Applicable.........

Loglcaý conr is ency of at4-n ysiS.............................................................................

Com;rnetefless of documentirng refererces.......................................................

rnom eeIr',ess 0{ irput Accuracy Wf.r~ut daa......................................

Czns.ltency of irp.t ata with approved cr:teria Corp!etemess ir. statirp assumpion.

IVa' dC y of assumpto s.......................................

Calcj ation sufficiently detaled................................................

A rihmnetical accuracy.......................

............. I.

Physical ýAnits soecited and correc¢ iy use..................................................

Reasonableness cf ojtouf conclusion..............................................................

SuperV'sor independency check (If acting as Verifer)....................................

Did not soe..7y analysis aporoach Did not n.re iCut specific arafysis o~t~cns Did n-ot es:abiisn analysis inputs.....................................

V jer Initials

/"

~

.~.,.

ENERGY NORTHWEST Page Conrd On Page VERIFICATION CHECKLIST FOR CALCULATIONS AND CMRs Verifier.'InItlals If a computer program was used:

Is the program appropriate for the proposed application?.............................

Haventhe program error notices been reviewed to determine if they pose any limitations for this application?.......

Is the program name, revision number, and date of run inscribed on the output?.............................

Is the program Identified on the Calculation Method Form?..............................

If so, is it listed in the Software Catalog or if the software Is-used for calculations completed by vendors Is It done under an approved (10 CFR Appendix B) software program ?............................

If a CMR or calculation revision was prepared against calculation E/I-02-90-01:

Verify that manual calculations agree With automatic-calculations on rows with modified input data Verify that changesin loading are reflected atithe upstream load centers.

Verify that other calculation cells have been checked as deemed appropriate by the verifier using manual calculations to assure accuracy has been m aintained.................................................................................................

Discussion of Results/Conclusions are: still valid after calculation update. (i.e.,

bus & transformer rating not exceeded, etc.)............................

If the calculation involves an instrument, verify that the calculated values are within the instrum ent display range....................

Other elements considered:

Verifier Initials If separate Verifiers were used for validating these functions or a portion of these functions, each sign and initial below.

Based on the foregoing, the Calculation/CMR is adequate for the purpose intended.

Verifier Signature(s)/Date Verifier Initials 25280 RIO Page 2 of 2

Pieparod By, D%,i Gregory 'iN"4; CN4R 1P~J2~

Path 1: Branch line around the RHR A" heat exchanger

.Note: To be consistent vwith the ba~e calKuIaiorT, W`;Ct MUod'.:fx 'A ot oý trie deterni rion tlhaLt "A" and q,-' were qimilnr, this CM1'R rymr-cc wt byp~s loop A,evie~w &vj corndurted to confir-

-ha~t the byp~s-c 100P dferi cnc~s vver ri P~tt; i; tii 18 1 'ýt-rt at star1 o# rec, end at -.,-.d of teeŽ.

Cal~cu ta~n 5 :71 ists the ;HR pressure n the heat evcŽ-riger as 260 ~v L2.wfnon5..1 7.eC lists Utle krR vs~ur- ',! the hea~t asc 7rnje'~542' s

t'rr*( +~ 1466 (psi' Notp: Th-: rtnu",!; nt the a-lysi% err j insensith/ce to Tre pressure r

• -~

GE Sec 2L-

'(1~,14 "Frt!;s Daii"~ fo;r modQ A-1 Pis ir S,ýCUh' d pipe I8"RHW 1)--

p*ýisi

ý)avilnu te

ý.;ac fl~ rate.

HByvs,. gymA V 1=

DisI 4

r Close~ly rnat~cniP FSAFR stait~nem t oF Lrypaj.G 101,, -,hra-.jgh izxfht

~

y X1 6"I

-7>

(

,~'>

j p

~o1c M~4~'\\

i

-k(,-

m

.-rjr1 fliZirm resistancus, from. C-VI ISO 2,15-00,32S4 Eiriz'Ch tee - VraLx from CtM1F 93K)5&

ii --

Prepared By; Dwayne Gregory-t>--

A)e )z.

Verified By: John Feilman

(/t i'Il Piping resistances from CVI ISO 215-00,3037 Pipe := (]Sft + 2.5in) - (2ft + 3in) = 12.958 ft Pipe:= Pipe-+ [(I 5ft + 7in) - (2ft + 3in) - (2ft + 3in)] =24.042 ft Pipc := Pipe + I( 13f + 9in) - (2ft + 3in) - (2ft + 3in)- 50inj = 29.125 ft Pipe := Pipe + [(lOft) - (2ft + 3in) - (2ft + 3in)] = 34.625 ft Pipe := Pipe + [(lOft + 8in) - (2ft + 3in) - (I 1.25in)] = 42.104 ft Pipe;= Pipe +

3ft + 2 + L,

-0 (11.25in

=44.401 ft CMR 10922 4 lhow.90 := 411.

.3-=

= 0.797 14.93Cc Kelbow.45 := ft. 18

'1

= 0,133 D18 use 22.37 ft from 93-0588 use 14.93 ft from 93-0588 RFIR-V-48A CV:= 3511 Used 3.511 to match 93-0588., Vendor confirmed that the valves were retested and do not have the Cv found on the valve drawing.

p Bypass 2.psi = 1.389 psi Avalve' 2

lb CY ft3 Piping resistances from CVI ISO 215-00,3229 Kec.h := Ktc.b.h + 0.72 = 1.63 from 93-0588 tee branch to run K = 0.72 Determination of dynamic pressure loss in this flow path:

'66PBypuss:= '6ýpvalve +

~~Byass = ~Pa~ve+

1 ee.b + Kelbow.9O + Kelbow.4s.v

= 2,031 psi x

prepated 13y: Dwayne Ggegofy w'ý6 Voreod BY: Jo6rn FaUmen Path 2 - Through Heat Exchanger Pining reit,ýe

-om CVI.'SG 215-00,3284 CWMR 109h22 KCjr

= 1r~fx 0,054A Use L for-rur, toe frorn, 93-0i;,88 pjpU~;'- (L.Sft '- 41m, - GMf + 31mn

= 21IYsl L.7fl Pi jn res~stan~.es f rom CVI IS0 213 5-001,30 7 riptu:

Piro *t

• ,1M

)

ý2r

' 1=

8 A3.1 ft Jlptr.:= PiN~ +

+t -11r) - 1 ft 3111; - (2ft + Jilt, -- 3

ý; 5.5 fI pipe :' Pipe + T1im 1 1" + 2 ii 2n- ýin) - cfl 3iw'= =4Xi.5isfn Pirw!= fPpc -t

-t-[, 6tip) - ( Ift.-f )DI ln -

I.25j ~

=52'$f ptpc~p:= 60

.5

=

j~,~0.707

+

i

=1 i),721

'e,>,'

f!

l14'

Prepared By: Dwayne Gregory Verified By: John Fellman (T

m V//

/,I --

CMR 10922 Kexandr:=2.6.sin(i) ( -l

)

2.291 x. 10 18 inch End of piping resistances from CVI ISO 215-00,3077 (7450 - Bypass) 2 APhx := 8.3psi.

-- 1.589psi TEMA sheet states 8.3.psi, 7450 gpm 7e4502

• Piping resistances from CVI ISO 215-00,3228 Pipe2 0 := Pipe20 + [(3ft) - (2O + 6in)) = I ft 20 inch Pipc2 0 := Pipe. 0 + I (30) - (2ft + 6in)] = 1. ft For a 20 inch, 90 degree Long Radius Elbow L = LF + LL Reference 7, page 8-1 where L

• LF LL

= Equivalent Length

= Equivalent Length of Pipe Bend, measured along the axis

= Equivalent Length of Pipe, KLL D LL~ =f KL

= local resistance coefficient d

= Pipe inside diameter f

= Moody diagram friction factor DB

diameter of pipe bend along axis Reference 1 - center to face value "A" Rb:= 30.in d20:= 19in Schedule 30 RHR(.1)-2-2 ASME 11-2 d20 in f20 := 0.012 inch crane 2,7, Rb LF:

4

Prepared Ely Dwayne GreZ tr CMF# 10922 A IFoi-90 cýegrec elbows, per Raleerene 7, p~

8 2 E mo ~i s For sted1 pipe, Reference 4 0'!

F Refenrete 7, Paqe 89-2 dK 1 i K1

! 1 1 20 22"11 11,ilbeW l+

I-L 2 5. 979 EquwvqVent leycqtr of 20 ricrim Elbow tu~r M

ft

~

to 8 inch

-ir tiujC t

+

it-2nJ=P.91 Nate: reduccri t; U intch length P

-w. Pitve

!'(Wr

+- 6in) - (21 3in) - f,20 + 3io) - (Mot) -63.19~3 t

11ifie ~

1 ie-I

+ý i -

(2 (1 3in -

k ~2rv Am 8!,)!

Prepared By: Dwayne Gregory *> "tile *ti-Verified By: John Feilman

~

~

'11~

CMR 10922 Piping resistances from CVI ISO 215-00,3229 Pipe:= Pipe + [(4ft + 4.5in) - (13.5in)] = 73.109 ft f, 6.0Oft Ktcc.r.:= Kice~r + ft. 1 W 6=06f =,.108 D18 End of piping run - now merged back with bypass line K18 := Ktee.r + Ktec.b + Kuxpander + KrtScducer + Kv,3A + Kv47A + Kelbow.90 + Kelbow.45 2.563 (7450 - Bypass),gpm 4.475 ft 7=

4.475-(7450 - Bypass).gpm ft K2 0 :=KeTo200

-3,688-K2 0 := Klbow2o.q9o

0. 197 AP:= APhx+ p, K.18 + rA. Ppe8 18+

D18 )J 2 P,{K 2 0 +

Pipý2 0 V20 2 f2 o0 - o

= 2.03 psi APBypass = 2.031 psi These two values need to match

Pteparea5y. Ow~ayne Gregry.ýq,;

GMFI VY-4fý2 verm1ted Bly: Jotvl Fetlmann Path 3:, Branch, line around the, PHR h~eat, exchanger, no heat exchanger flow

Tire is zill is ;roi-h start at stam r~ ftc~, end at end of t 'e.

Ulckj

.lr /.1 4j lll;is Ltt RHR mrfs-,tur in !%e he,.!*, exd-Tarior z% Z&CD ric,

<i.AI zuýat-nn

/.40 lists the~ PHR pressutre ir, rric hea* ev-\\han',,jS 2-:4 fý6),14,961L P;;3 Note: Tre res-utt of the an~!i ar il"2 ~

~

insensit ve to U1t. Lrpnssre t.m I

t

'Sch~duJl std pipe 213"RHR(11)-2-1

!terato beiow to daermine the bypass flow. Iteration based on bolh patli havring the same f ow rate.

flypas:

7410 Full fkwv% throuqh bypas.~

-P stl if I

g o "

I I ! 9 '* o.'ý: 2ý'

T ~en from Cra-'e

Prepared By: Dwayne Gregory F_

• G /_CMR 10922 Verified By: John Fellman Wf*--LffJi--L4 Piping resistances from CVI ISO 215-00,3284 Kt.:0.90 Branch tee - value from CMR 93-0588 Piping resistances from CVI ISO 215,00,3037 Pipe := (15ft + 2-5in) - (2ft + 31n) - 12.958 ft Pipe := Pipe + [(15ft + 7in) - (2ft +3in) - (2ft + 3in)] = 24.042.ft Pipe := Pipe + [(I 3ft + 9in) -. (2ft + 3in) - (2ft + 3in).- 50inj = 29.125 ft Pipe := Pipe + [(loft) - (2ft + 3in) -- (2ft + 3in)] =U4.625 ft Pipe := Pipe +.[( loft + 8in) - (2f' + 3in) - (I 1.25in)] = 42.104 ft Pipe:

Pi

[e+

3ft + (2+

I in -(I 1.25in) 44,401 ft Pipe:

Pipe+ L ~

16j n]

4, 22.37f1 Kelbow.90:=

• t.18.-

= 0.797 use 22.37 ft from 93-0588 Dig

= " 14.93f'I KcIbow.45:= tl.--

8

= 0.133 use 14.93 ft from 93-0588 RHR-V-48A Cv:= 3511 changed from 4280 to 3511 to match calc p

(Bypass 2 PPvae:

F... spsi

=,4.39 psi 6

2,' 4 -

\\lb fc3 Piping resistances from CVI ISO 215-00,3229 Ktee.b :=

t ee.b + 0.72 = 1,63 from 93-0588 tee branch to run K 0.72 Determination of pressure loss in this flow path:

APBypass2 :=APvalve + P"(ft.i18

+ Ktee-b + Kelb90 + Klbw4 6A 9psi D18 5)

APBypass2 -

y 4.4 psi Extra margin from crediting path through heat exchanger

Prepared By. Owqy'e G pgr CIVIR DQZ22 This CMR ir'~corporaters a coflervatiSm of the CV finr RHIR-V.-48A/B. Tnr-wv,.

daw~ing 5tates a Cv of 428CJ irrstead of the 3511 usied nr CMIZ 93-05,.48.

,ven'dor bazs confirm, Od that the Cv w-a ratested after the v~ive drawimr, wVZj!

issued aid lower Cv values were fot..rd. Therefore, th& vendor cr~duede~

that the Cv' is likely 1511 I~s. their 193I1 lette rsti*Lu Instead cf the~ 4*280 s-tlown on their draw'rig. UýIirg the Cv value of 3511 aCoos IAS2 psi to tht

-,rr: 4 4 ;X~,%

-1,6Ps

?

- CM 46, Pvro ý, 2 - -41, Bypass Fi-aI extra mrgn~rr avoilabke from zreditirnq the pai&, *,riruuih hF.-i

= P SUVEILANCETEST DISCHARGE PRESSURE CORRECTION TABLES RHR-P-2A 1 RHR-P-28 RHR-P.2C Required Required Required Fkow, gpm I*,e V

.9 DlsharVge Discharge.

ii Pressure.;, i.

Prossurs', Rsi 7490.: j 1265 126 124A5 7491 12650

!26"59 124.39 7492 126:5 5 653

_.24.34 7493 612549 4124,28 7494 126:4

,12-4.23

' 7495

-124.17 7496 12.3 126,26

l24*1 7497 125.27 1219 124.06 7498 128.1

.120:13 124.01.*

7499 1.t 126.06

.123:95 7500 125.10 1

123.90 7501 T54 125.93 0123.84 7502 125-99 2

123,78 7503 1.25:93 "125.79 123.73

.704-12588 125:73 123.57

.7505 125;66 23,6 7506 125.70

12-5.59 0123.56 7507 125.71 125 12350 7508

'125,656.

125;46 12345

.7509 125.39 123.39 7510 1

125:32 123.34 7511 125:48 125.25 123.28 7512 1

125.19 2 3.22 7513

1 71.2"

-123.17

-7514

.:125.31 125,05 123.11 "1

• 7515 2124.98 123.06 7516 125.20 124:91

.123.00

.7517 25:4 2485

,,2294.

7518

-:15r0 124;78 122;'89 7519 12543 12471.3 7520

'__24

' 122;77 7521 2

124,57 122.72 7522 124.:86 24,51.

12256 7523 i1244 1 22461

.7524 4.37 "

12 7525 14 122249 7526 I

124.23 i 22.44 7527

' 1 58 124.15 122:38 7528 t

2 t22:32 7529 12447 124.03 122.27

.7530 241 122.21 7631 124.35 123*8 122.15 M53 1

123.82 "122.10 7533 "12424 123;75 2_',04 7534 124,18 123.68 121:98 7535 1"24.3 3

121:92 7536 1

123.54 12487 7537 7 123.47 121181 7538 12*9 1

121.75 7539 12333 121.70 7540 123.88 123.27 121.Z4 7541 1217 23.20 121M5 7542 12.7 125.'13" 115 7543 123.68 123;06 121:,47 7544 123152

'122:99 121:41 7545 123.56

12.

21.35 754 123.51 122:85 121.30 7547 1212278 121.

7548 12339 122.7121.1

RHo PUMP SURVEILLANCE TST tISCHARGE PRESSURE CORRECTION TA LES

Flvw, 1 sc Dicr 57r~

ihaa PM~rvjru 1.

oa{

pos~rms psi Fi-essw',. cs.

12.64121 12 7550 i 23. 2.

F1 6

l2f, 67 751 123 22 122.5K 121,D'1 715 2&1IC.

12242 Y,20 56.

75j123.11 122.36 2,8 7-71123.05 12.2 1.2c.84 7555 122:9z.

") 21 2.. 7.1

745p, 122.94

'22. 14 120,7?

65"?

122.M.

22.

2C8 7 5 5~

12zsz 143' J210 1.559

~

22, 72' 1.Us 1 201,55 756C 2.P12,5.t3

• 16 122.65 1','

120.43 7552 122.60 2l'

__1____Ow 1 4122,2 12 1. S'-

C,__

756___

t2 2 -' 1)

,12 6

120 C

F 1.1 7,1

.756-9

~

2. 23

'22r 2 2.0.9 121.03 11-385 7 C7;:

3 7 2 1 0.

11T70 12' 1.£ -17

" 9,1

.73 574_____

111 8 b 96 Y21.90I 11206 19.56 1z, 761-I ;->F7 i zj-q

?

e 77 121.62 120 5-112,18 758 12.1:L7 7204-J 7-S,92 141.,45

!.202P 1 91 7162 lzj;4t 120.21

-1 5.

584_____

121:241 12C. 14 ur -

75$51212.21 120S-V'

£21.22 120.00 (18 "

7587

.1 131 VS 9.9 1891 75841 121. 11 M65 11.85

/58 1i 15

,18 75SO7

0. 98 1__IL__7a 75001 1,

4 19,32_

7IW~~6

~

'8 9

____1_

120.

1M4 76014 ii,. tit L

I (c3 25_

_____6

'7 7__

'-:2 L

7

`1PUMP SURVEILLANCE TEST DISCHARGE PRESSURE CORRECTION TABLES RHR-P-2A RHR-P-2B RHR-P,2C

• ReqUired Requh:ed Requ red Flow, gpm Discharge 0ischarge Discharge P70ssu9 * -,pL Pressur'ei Pressure', psi 7609 11!9.90 18,0" 117.61

.76i0 T1T§84 f1*8.23 11.5*

7611 119.79 11816 117A9 7612 119.73 118.08 1

.7613 1

1 1

r"7814

" 1!*1 179 117.31 7615 '

1!9.

4'.:

117.25 7518.

• 1W.50'.

17.19 S

119:44

.17.71 117.13 761a 1*3*.

T76 71,*,

7619

'119.32 i

117.56 11,7;01 7620 119.27 1-7.48 t16"95 762.t1192

.117;r'4.1 116.89 7,62 1.19.15

.117;3 11_______

7623 119.09 117.26 116.76

.7624 t119,03 1.17.15 116.70 7625 118:98 117.11 j

116A34 7628 118.92 117.03 116.58 7627 118.88 116.95 118.52 7*62 118.o 116.88 118.48 7629 118.74 1

116.40 7830

~1,169 116,73 116.34 7631 1T16 6

116128 7832 1'16.57

.1,

.116.22 7633 1161 1

TT6.50 116,1 7634 1T!4 114510 75

'118.'40, 11.35F t, 1604 76368 118627.

11-.g5.

7837 1

115s92 7638 118.21 8

5 7839 1.18.6

'116h0*!

115.79

7840,

1 t8

11 i15j9

.11573 7641 118.05 118.89 1.15.67 7642 117"29 1

115.61 7643 1

5,116,:

7644 '

1787

-115.6 115.49 7645 117,81 115.58 11b.43 74 117.76 115:50 116.38

7647 1

1 115.30 7848 '

177;6

!* 1.

1 i.24

.7849 11758.

11.287 5

7650 11752 115.19 11T'2 Notes:

1. The test acceptance values listed are to be-compared to the TDAS reading. These acceptance values take into account that there are flow losses between the pump discharge and the location of the TDAS instrument and as a result the values listed are less than actual pump discharge pressures. NOTE: The acceptance values con'responding to the Tech. Spec. requirements of 7450 gpm

@ 26 psld wetwell to RPV before the correction is factored in are:

RHR-P-2A - 131.65 psi, RHR-P-2B - 131.65 psi. RHRP-2C - 128;8 psi.

DIC 413.3A ENERGY NORTHWEST LICENSING DOCUMENT-CHANGE NOTICE FORM NUMBER LDCN - FSAR 025 REV. 0 AR 230095230095BRIEF DESCRIPTION OF ACTIVITY: (3 Changing the FPC temperature limit for abnormal operation from 155F to 175F and removing a statement from 9.1.3 that is inconsistent with Table 9.1-6.

REVIEWS AND APPROVALS: Signature indicates authorization to process the subject change into an amendment.

@) POCReview Req'd 0 Yes.

No G) POC Meeting No.

CNSRB Review Req'd before submittal C] Yes.

No CNSRB Meeting No.

Plant General Manager Approval 0 Yes El No Ext # 4707 Originator Dwayne Gregory Mail Drop PE24 230095-01 Date:

AR Asslgnmnem # or Signature Originator Supervisor/Manager Jim Brower 230095-09 Date:

AR A.algnmoul for Signature Regulatory ProgramsManagement 230095-10 Date:

AR Assignment #

Sg n

.tre 0D POC Recommendation for PGM Approval Date:

• Signature Plant General Manager Approval Date:

Signaure (2) Activity Initiating the Change

@) Forms Attached -. Check all that aoolv

-Initiative to support an 18 day outage.

0D AD 10-1432 0 POC Review Criteria 0 Operability Impact Determination Screening/Worksheet Related or Impacted LBDs LDCN-FSAR-05-005 C] 50.59 Screen _]

7248'Screen 0 50.59 Evaluation 5059-11-003 C] 72.48 Evaluation 7248-

[

Licensing Basis Impact Evaluation LE-

© Hold LDCN Imolementation For Explain:

License Amendment

] Yes 0 No Plant Implementation C] Yes 0 No Other Change Management E] Yes 0

No (D Change management actions (include procedure revisions)

Action AR Assignment Resp Org/Person Revise ABN-FPC-LOSS to use 175F instead of 155F.

Revise SOP-FPC-START to use 175F instead of 155F.

Revise SOP-FPC-DEMIN to use 175F instead of 155F.

Revise SOP-FUELPOOL-DRAIN to use 175F instead of 155F.

Revise DRD 327 to note that the allowable temperature is 175F, not 155F.

(Tab to insert new row) 24341 R20 Page 1 of 3

NUMBER LDCN - FSAR 025 REV. 0 I

(Tab to insert new row)

In your review, please ensure:

The reviews required by section 3.1.4 of SWP-LIC-03 have been performed.

The package was reviewed by appropriate subject matter experts, as necessary, to ensure a thorough review.

Comments are attached to the LDCN package or provided electronically via AR assignment.

NOTE: Also include in your review, any forms listed in Section 9 of the LDCN.

(Tab to insert new row)

Date comments provided Section #

Reviewer AR assignment # or Signature (May leave blank If tracked by AR) 9.1.3 Steve Kartchner 230095-02 9.1.3 Tom Morales

.230095-03 9.1.3 Tony Huiatt 230095-04 Table 9.1-6 Steve Kartchner 230095-02 Table 9.1-6 Tom Morales 230095'03 Table 9.1-6 Tony HWiatt 230095-04 I

.1-

+

+

I

-~

+

+

+

.1-

.1

~1~ +

+

+

• 1-

+ (3)

Date resolution of comments Comment Resolution provided by AR assignment # or Signature leave ided Dwayne (May l blank if tracked by AR)

Dwayne Gregory if230095-18 i

Acceptance of Comment Resolution Date resolution of comments and LDCN approval accepted Reviewer/Approver Name AR assignment # or Signature (May leave blank if tracked by AR)

Steve Kartchner 230095-15 Tom Morales 230095-16 Tony Huiatt 230095-17 24341 R20 Page 2 of 3

NUMBER LDCN - FSAR 025 REV. 0 AR 230095230095CHANGES Licensing Change Document Brief Description of Each Change Justification for Each Change Change Source Document of Each No.

Section No.

Type Change 9.1.3 & Table Changed the temperature limit from 155F back to This change corrects an error that 5059 RS-001 9.1-6 the original FSAR value of 175F.

changed. the capability value to a limit and EVAL EVAL 5059-11-003 eliminated the real limit. Using the design temperature.as the limit is acceptable per NRC documents; 2

9.1.3 Removed "normal" from "normal refueling" and The'l 75F limit for a single failure is FA N/A removed, "i.e."

applicable for normal operation and all refueling activities with the exception of a full core offload. Full core offload is excluded since full core off load does not include single failure. This is a change required to provide consistency between different sections of the FSAR.

3 9.1.3 Removed the paragraph discussing that the results LDCN-FSAR-05-005 removed the results FA LDCN-FSAR-05-005, pages 13 of bounding analysis are included in FSAR Table of bounding analysis from Table 9.1-6.

through 15 (changes 25 9.1-6, as the bounding analysis are no longer Removing this paragraph provides through 42) discussed in Table 9.1-6.

consistency between different sections of the FSAR. The FSAR table Still lists limits, FSAR 9.1.3.3, page 91-31, 2 nd but there is no need to change this paragraph paragraph from discussing results to discussing limits since this discussion already exists in the FSAR.

(Tab to insert new row) 24341 R20 Page 3 of 3

LICENSING DOCUMENT CHANGE NOTICE FORM INSTRUCTIONS FOR COMPLETION OF FORM NO. 24341 NOTE:

Refer to SWP-LIC-03 for document specific additional instructions.

ENTRY INFORMATION 1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

Enter the LDCN.and revision number obtained from Regulatory Programs. Listed below are Licensing Basis Documents and abbreviations to use in LDCN number:

TS

= Technical Specification 72212

= ISFSI 10 CFR 72.212 Evaluation TSB

= Technical Specification Bases OQAPD

= Operational Quality Assurance Program OL

= Operating License Description (50.54a)

LCS

= Licensee Control Specifications PSP

= Physical Security Plan (50.54p)

FSAR

= Final Safety Analysis Report EP Emergency Plan (50.50q)

ODCM

= Offsite Dose Calculation Manual COLR

= Core Operating Limits Report IFSAR

= ISFSI FSAR Add AR number as provided by Regulatory Programs.

Provide a brief description of activity.

Determine whether POC,.CNSRB and/or PGM review/approval is required. Complete AR assignment or sign as originator.and route.'for Supervisor's/Manager's concurrence for further processing. Supervisor/Manager approves after ensurino aoorooriate technical reviews and resolution or comments have been completed.

Document POC/CNSRB review (meeting number) if applicable (by POC/CNSRB Coordinator).

Regulatory Programs Management ensures all comments are-resolved and approves.

The POC Chairman will complete an AR assignment or provide signature recommending PGM approval if required.

Identify the activity that prompted initiation of this LDCN by driving document number. Identify any related or impacted LBDs byLDCN number or specific section of impacted.LBD.

Identify and attach all applicable documents/Forms generated tosupport the LDCN.

Answer YES or NO and explain the reason the LDCN should be placed on hold. Examples include: The change is being prepared in suooort of a proposed or as a result of a License Amendment or Technical Specification change; another LDCN;.awaiting change management actions such as procedure changes; awaiting plant implementation of a design change; or, setpoint changes, surveillance requirements or operability requirements.

List any change management actions associated with this change, the AR.# and assignment # created to track the action and the responsible organization/person responsible for performing the action. Notify procedure group lead(s) or sponsor(s) of required procedure changes.

Appropriate reviewers are assigned by originator and confirmed by originator supervision.

NOTE: Technical Reviewers and additional reviewers may be added, or alternative reviewers specified, at management's discretion.

Document resolution of reviewer comments. Approvals of comment resolution via AR assignment completion or signatures on the LDCN form indicate "approval" of the LDCN package.

This field (block 13) may be used to document multiple reviews and/or approval reviewer comments and resolutions.

Justification for dissenting comments provided during the LDCN review and approval process (i.e., voiding/revising, etc.) can also be provided in this section.

Enter the change number for each change.

Enter the Licensing Document Section/Subsection affected by the change.

Enter a succinct description of each change in the space provided. This will facilitate the review by permitting the review to focus on the proposed change.

Provide an adequate justification for each change such that a reviewer can understand and evaluate the rationale, for the change.

Provide the change type as determined on the AD form. If a 50.59 Screen is performed for this item, put SCREEN in this block. If a 50.59 Evaluation is performed, put EVAL in this block.

Identify specific source documents for existing, new or revised information.

The LDCN review package must contain at least the following:

1)

LDCN Form 24341,

2)

A copy of any required documents (or links to required files), as noted in Block 9.

3)

A copy of marked up LBD pages intended for review.

14.

15.

16.

17.

18.

19.

DEFINITION:

24341 R20

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 details). The fuel pool is maintained at or below 125°F during normal plant operations. The fuel pool temperature may rise above this value during refueling activities or during an anticipated operational transient of the loss of one train of the FPC system. The RHR system can also be manually aligned into several configurations to provide supplemental cooling of the fuel pool. One of these RHR configurations is the RHR/FPC assist mode of RHR (see Section 9.1.3.2.2 for additional details).

The maximum heat load is present in the spent fuel pool during refueling activities when recently irradiated fuel bundles are discharged from the reactor core,to the fuel pool. The magnitude of this heat load is contingent upon the cycle-specific refueling activities, i.e., the number of the bundles discharged, the burnup of the discharged bundles, and -the decay time of each bundle when it is placed in the fuel pool. The heat load associated with a planned discharge can be calculated with the ORIGEN-ARP computer code assuming a 2% thermal power uncertainty or other acceptable means to account for code bias and uncertainties.

During refueling activities, the fuel pool temperature is managed by controlling the number and schedule of fuel assemblies discharged, controlling the number of heat removal systems in service, and controlling the temperatures of the systems (RCC or SW) used to remove the heat from the FPC heat exchangers.

The fuel pool cooling system was originally designed to maintain the pool at a temperature of less than or equal to 125°F during refueling activities with both trains of fuel pool cooling in operation and RCC cooling water at 95 *F. The decay heat load assumed for the original design basis (normal offload) was based on the original licensed power of 3323 MW-thermal, a one-year fuel cycle, and a quarter core offload with a 20-day decay period.

Since the original design, the licensed power was increased to 3486 MW-thermal and the operating cycle was revised from a one-year cycle to a two-year cycle. These changes resulted in an increased heat load in the fuel pool, particularly during refueling outages. For the current design (3486 MW-thermal and a two-year cycle), a 150'F fuel pool temperature limit (see Table 9.1-6) applies to normal refuelpi tivities for the scenario of both trains of fuel pool cooling in operation.

O s

c a

apo snd TheFPC system is also designed to paovned*afdcent cooling for an anrvcapated operaeional transient of the loss of one train of thr system. For adde tois the fhe maximum bulk fuel atvteanter temperature is limited t e

1 F. This limit applies to both normal operation and "oemt~l~(e fueling activitiest(ýxclu ing ia full core offload).

Thsyt ems biit t stifytese temperature limits could be challenged based on outage-specific plans or activities. Outage-specific calculations are performed, as needed, to ensure acceptance criteria limits are maintained and adequate decay heat removal capability exists.

Management of the rate and magnitude of the heat load added to the fuel pool during refueling activities and the temperature of the credited heat sink (i.e., RCC or SW) are considered.

l 9.1-26

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 TTo permit tlhis use of the RHR

system, normally locked closed valves interconnecting the RHRT

[system to the FPC system are opened after verification of the installation of the flanged spool lpieces. A full core offload is.a non-routine evolution. A single failure is not.postulated to.

occur during this evolution and credit for RHR/FPC assist is allowed. During this evolution, sufficient cooling should be provided to maintain the fuel pool temperature at less than 145 OF.

The licensing basis limit is 175 0F.

The fuel pool heat exchangers are normally cooled by the RCC system to contain released Steradioactivity Sse mana in the *event th ulof a polfuel pool waeheat tmeareexchanger elw25tube failure. F DurigOFwenrmoig normalth operations,noia l Iheat load 'from the fue 1oo.[

Durning normal refueling.outage conditions,.the system is evaluated to ensure it has e capability to maintain the fuel pool water temperature below 150'F. This limit applies to core offloads up to, but not including a full core offload, and is ase the of 2 FPC trains in service.. The system maintains fuel pool water 45o the event that only one pump and one heat exchanger are I 7"Favailable. Dependgin*g* Dfthe heat load in the pool during refueling activities, RCC or SW temperature is controlled to ensure the FPC system can perform its design functions within the acceptance criteria shown in Table 9.1-6.

Following a seismic event or major plant disturbance the SW system is available to cool the fuel pool (by means of FPC or RHR-B heat exchangers) to preclude boiling of the fuel pool water. The SW pressure is higher than the fuel pool pressure; thus, any leakage will be into the fuel pool system. In addition, radiation monitors of the SW return line detect any gross failure in the heat exchangers.

The fuel pool design precludes any condition which could allow the fuel pool to be drained below the pool gate between the reactor well and the fuel pool. Two diffusers are placed in both the reactor well and the fuel pool to distribute cooled return water efficiently. Diffusers are placed to minimize stratification of either temperature or contamination. Valving is provided to prevent water from being siphoned out of the pool. [ All1 piping connected to the fuel pool and reactor well, except for drains and liner drains, are Seismic Category 1, including any normally closed manual or normally open automatic valves that provide isolation from the Seismic Category II portion of the system. Drain and liner drain piping connected to the fuel pool, reactor well, and dryer separator pool are Seismic Category 11 supported to Seismic Category I requirement~s. s~ifnc~e -te fuel pool system is at low temperature and pressure (moderate energy system) postulated breaks in the Seismic Category I portion are limited to cracks.

Fuel pool cooling can be established and monitored from the control room following a design basis LOCA. I ne of the two FPC trains is adequate to prevent fuel pool boiling by a large margin. I However, during normal plant operation, one or both trains operate to maintain 125°F pool water. Should one of the tp4b unavajL1ethe second train operates to maintain pool water temperature below4

__* -55 alue is applicable for an

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 anticipated operational transient involving the loss of one FPC loop. If the fuel pool heat load is such that the -resulting temperature transient exceeds 125 IF, procedural guidance :is in place to restore the fuel. pool temperature to < 125TF. Due to the large thermal capacity of the fuel pool sufficient -operator time is available for the operator to take necessary corrective action to supplement cooling.

The results of bo g fuel pool c anlses are ded in Table! 9

.These A makeup water valve controlled by skimmer surge tank level switches supplies water from the condensate transfer system to the fuel pool*.to replace losses. I The backup source of makeup"I

• at'--*

s fo---m h:

eimcatg--"-"-'

I ysfe-class-' 3 SW syst'e'-'m. This source.supplies makeup" for. long-term pool water losses.

Each. filter demineralizer is capable of continuous operation at a normal fuel pool water flow rate of 575. gpm or a maximum fuel pool water flow rate of 1000 gpm and will maintain water conditions as specified in Section 9.1.3.2.

A radiological evaluation of the cleanup system is presented in Chapter 12.

From the foregoing analysis, it is concluded that the FPC system meets its design basis and satisfies -the requirements of Regulatory Guide 1.13, Revision 1.

9.1.3.4 Testing and Inspection Requirements Special testing is not required for the system except as noted below because, when fuel is stored in the pool, at least one pump, heat exchanger, and filter demineralizer are routinely in operation.

Routine, periodic visual inspection of system components, instrumentation, and alarms are adequate to verify system operability. Likewise, the interconnecting valves between the FPC system and the RCC, SW, and RHR systems are periodically inspected to verify their operability.

9.1.4 FUEL HANDLING SYSTEM 9.1.4.1 Design Bases The fuel handling system is designed to provide a safe and effective means for transporting and handling fuel from the time it reaches the plant until it leaves the plant after postirradiation cooling. Safe handling of fuel includes design considerations for maintaining occupational radiation exposures as low as reasonably achievable during transportation and handling.

9.1-32

Table 9.1-6 Bounding Fuel Pool Cooling Events RHR Loops in FPC FPC Loops Assist Mode Fuel Cycle Acceptance Criteriaa Length (OF)

Fuel Pool Heat Load/Scenario (Months)

Available Credited Available Credited Normal refueling (non-transient condition)

Normal refuelingc (anticipated operational transient condition) 24 24 2

1 2

1 1

1 0

150 07~/ 4 JL I n W-Full core offload refuelingd 24 2

0.

1 1

145/175 Normal Operationse (non-24 2

2 0

0 125 transient/non-accident condition)

Normal Operationsf (anticipated operational transient condition) 24 1

1 0

0

-i4 - I IS T,F I-Normal Operationsg (design 24 2

1 0

0 125 (pre-accident) basis LOCA condition) 175 (ýpost-accident)

C) 0, 0

0

Table 9.1-6 Bounding Fuel Pool Cooling Events (Continued)

Performance Data Table Notes:

a Outage-specific/Cycle-specific calculations are performed, as needed, to ensure acceptance criteria limits are maintained.

b A normal refueling does not involve a full core offload. The normal refueling temperature limits go into effect upon placing the "Ti,.

plant in cold shutdown and remain in effect until plant start-up. The 150°F acceptance criterion is applicable for the case of both FPC loops available.

,m-c A normal refueling does not involve a full core offload.

4,Pf ceptance criterion is applicable for an anticipated operational transient of the loss of one FPC loop.

d A full core offload is a non-routine evolution. Although no single failure is postulated to occur during this evolution, only credit for RHR/FPC assist is taken. During this evolution, sufficient cooling is provided to maintain the fuel pool temperature at less 0

than 145°F. The licensing basis limit is 175°F.

W e Normal Operations is defined as any plant condition other than a refueling outage. -The normal operation temperature limits V 0 become applicable upon plant start-up following a refueling outage. The 125°F value supports the initial (i.e., preýaccident) fuel pool temperature assumption in the design basis LOCA analysis.

f Normal Operations is defined as any plant condition other than a refueling outage. Th 155Fe ue is applicable for an anticipated operational transient involving the loss of one FPC loop. If the fuel pool heat load is such that the resulting temperature transient exceeds 125 0F, procedural guidance is in place to restore the fuel pool temperature to < 125 0F.

g Normal Operations is defined as any plant condition other than a refueling outage. The 125°F and 175 0F values support the

°°D CS initial fuel pool temperature assumption and the calculated peak fuel pool temperatures, respectively, in the design basis LOCA analysis. For this event, RCC is assumed to be lost and SW is manually aligned to the FPC system heat exchangers. An equilibrium temperature of 90'F is assumed for the SW.

a%

Implementing Activity:

Refer to SWP-IRP-01 ENERGY NORTHWEST LDCN-FSAR-10-025 POC REVIEW CRITERIA NOT*: The following questions help determine If the proposed activity is required to be approved by POC. prior to performing the activity. A YES answer to any of the following questions requires POC reView/approval of the proposed activity.

C3 Yes

[

No Could the proposed activity adversely affect reactivity? (Reactivity is affected by such items as RCS temperature, Main Steam flow, Main Steam :pressure, RRC Pump speed, 'Nuclear Instrumentation, Rod Control System, Fuel, and Fuel Components.)

-l Yes

[

No Could the proposed activity result in a significant Plant transient?

Q Yes No Could the proposed activity feedback into control circuits important to stable Plant operation (e.g.,

Feedwater Control, Control Rods)?

C3 Yes 0

No Could the proposed activity result in a malfunction which could feedback into a system :relied upon for.

stable Plant operations?

17 Yes

[

No Could the proposed activity substantially reduce the ability of Operators to control or monitor the Plant?

ED Yes Z No Does the proposed activity involve the addition or deletion of any temporary loads on a Class 1 E electrical distribution system?

E] Yes Z No Is the proposed activity a high risk evolution which could affect nuclear safety?

[

Yes

[] No Does the proposed activity require a 10 CFR 50.59 or 10 CFR 72.48 Evaluation?

ED Yes Z No Does the proposed activity change the. Plant or ISFSI Technical Specifications or Operating License?

17 Yes

[

No Does the proposed activity change the Licensee Controlled Specifications (LCS)?

El Yes 19.:No Does the proposed activity change the Offsite Dose Calculation Manual (ODCM)?

M Yes

[

No Does the proposed activity change the Process Control Program? (SWP-RMP-01 & SWP-RMP-02)

EJ Yes

-:No-Does the -proposed activity change the Security Plan?

Ml Yes.

No Does. the proposed.activity change the EmergencyPlan?

El Yes Z No Does the proposed activity change the Core Operating Limits Report: (COLR)?

EYes

[No Does the proposed activity affect the Refueling Outage Shutdown Safety Plan?

El Yes

.Z No Could the proposed activity result in a Planned Special Exposure? (GEN-RPP-08)

[3 Yes Z

No Could the proposed activity result in a configuration specific Incremental Core Damage Frequency

_ DF) GT 1 E-31yr, or ORAM-Sentinel Plant Risk Level RED?

E" Yes Z

No Does the proposed activity affect post maintenance operability testing for safety-related components?

(Revisions to established surveillance procedures do not generally require POC review unless they meet one of the other review criteria.)

Ml Yes

[

No Could the proposed activity cause an uncontrolled draining of the RPV cavity/vessel?

El Yes Z

No Does the proposed activity involve a new procedure that defines a regulatory-based program?

El Yes

[

No Does the proposed activity involve a procedure change that adds, deletes, or otherwise alters POC review/approval requirements?

El Yes Z

No Does the proposed activity involve a procedure change that includes a substantial change to fundamental station practices (i.e., Work Management Process, Reactivity Management, PER Process, etc.) or substantial revision to a regulatory-based program?

E] Yes

[

No Is the proposed activity a new (Revision 0) Engineering Test procedure revision (Volume 8)?

NOTE: Cancelations/deactivations do not require POC review/approval, Ml Yes

[

No Is the proposed activity an Infrequently Performed Test procedure revision (Volume 18)?

NOTE.: Cancelationsldeactivations do not require POC review/approval..

Prepared By:

Dwayne Gregory 2/22/2011 Print Name

• gnat Date 26232 R6

ENERGY OPERABILITY IMPACT DETERMINATION SCREENING NORTHWEST NOTE: For individuals implementing the OlD process for the first time, or if you are an Infrequent performer and People. Vision. Solutions ie-ed a refresher, a CBT training module on the OlD process is available on the ENNET at Computer Based Training.

Activity Description (e.g., PPM, TMR#): LDCN-FSAR-10-025 Driving Document: 230095 Performed By: Dwayne Gregory ritae4 Date: 12/2/2010 Print Name SigP

,ur ACTIVITY SCREEN QUESTIONS IRESPONtt ACTION REQUIRED Procedure Revision/

Change 1

Does the change to this procedure direct or allow a change in the physical configuration of a Safety-Related SSC or any SSC that is applicable to a Plant or ISFSI Technical Specification (including SSCs that support Tech Spec SSC operability)?

L-I Yes Proceed to question la.

El No Proceed to question 2.

la.

Does the change to this procedure direct or allow this change during an

[] Yes Proceed to question lb.

operational condition (mode) in which the SSC is required to be Operable?

El No Proceed to question 2.

lb.

Does the change to this procedure direct entry into the applicable Tech Spec E] Yes Proceed to question 2.

LCO Condition and Action?

-I No Complete Operability Impact Determination Worksheet (Form No. 26294) and attach completed worksheet to procedure revision/change.

2.

Does the change to this procedure direct a Safety-Related component I] Yes Proceed to question 2a.

configuration change from that set forth in the applicable System Operating I] No Proceed to question 3.

Procedures (normal or abnormal) or specified in a design requirements document?

2a.

Does the change to this procedure direct or allow this change during an I] Yes Proceed to question 2b.

operational condition (mode) in which the SSC is required to be Operable?

I] No Proceed to question 3.

2b.

Does the change to this procedure direct entry into the applicable Tech Spec E] Yes Proceed to question 3.

LCO Condition and Action?

[I No Complete Operability Impact Determination Worksheet and attach completed worksheet to procedure revision.

3.

Does the change to this procedure allow administrative control or manual

[] Yes Proceed to question 3a.

actions to replace automatic functions performed by a Safety-Related SSC?

[] No No further Operability Impact Determination actions necessary. Proceed with process change per SWP-PROR02.

3a.

Does the change to this procedure direct or allow this change during an

[] Yes Proceed-to question 3b.

operational condition (mode) in which the SSC is required to be Operable? "

] No No further Operability Impact Determination actions necessary. Proceed with process change per SWP-PRO-02.

3b.

Does the change to this procedure direct entry into the applicable Tech Spec LCO Condition and Action?

0l Yes No further Operability Impact Determination actions necessary. Proceed with process change per SWP-PRO-02.

El No Complete Operability Impact Determination Worksheet and attach completed worksheet to procedure revision.

26290 R1 Page 1 of 3

ENERGY OPERABILITY IMPACT DETERMINATION SCREENING NORTHWEST NOTE: For individuals implementing the OlD process for the first time, or if you are an infrequent performer and People-Vision-SOluions.

need a refresher, a CBT training module on the OlD process is available on the ENNET at Computer Based Training.

Activity Description (e~g., PPM, TMR#): LDCN-FSAR-10-025.

Drivini Document: 230095 ACTIVITY SCREEN QUESTIONS RESPONSE.

ACTION REQUIRED Temporary

4.

Does the proposed TMR change the physical configuration of a El Yes Complete Operability Impact Determination Worksheet Modification Safety-Related SSC or any SSC for which there is a Plant or ISFSI Technical and attach completed worksheet to the Temporary Specification?

Modification Record.

El No No further Operability Impact Determination actions necessary. Proceed with TMR implementation per PPM 1.3.9.

Trouble-

5.

Does the proposed troubleshooting plan direct changing the physical El Yes Complete Operability Impact Determination Worksheet Shooting configuration of a Safety-Related SSC or any SSC that a Plant or ISFSI and attach completed worksheet to the troubleshooting Plan Technical Specification is applicable to?

plan.

El No No further Operability Impact Determination actions (Check "No" if applicable SSCs will be declared inoperable during conduct of necessary. Proceed with troubleshooting plan activity.)

implementation per PPM 1.3.42.

Temporary

6.

Does the proposed Temporary Alteration or Comp Measure change the E] Yes Complete Operability Impact Determination Worksheet Alterations in physical configuration of a Safety-Related SSC or any SSC that a Plant or and attach completed worksheet to the applicable support of ISFSI Technical Specification is applicable to?

implementing activity document IF the applicable Tech maintenance Spec LCO Condition and Action is not directed to be or Comp entered.

Actions to El No Proceed to question 7.

support

7.

Does the proposed Temporary Alteration or Comp Measure allow El Yes Complete Operability Impact Determination Worksheet Operability administrative control or manual actions to replace automatic actions?

and attach completed worksheet to the applicable Assessments implementing activity document.

EI No No further Operability Impact Determination actions necessary. Proceed With the proposed activity in accordance with goveming process procedures.

Implementing

8.

During implementation of the proposed modification, will any Safety-Related El Yes Complete Operability Impact Determination Worksheet Plant Design SSC or any supporting SSC that a Plant or ISFSI Technical Specification is and attach completed worksheet to the applicable Change (PDC),

applicable to, be altered while it is being credited as being Tech Spec implementing activity document.

Minor PDC, Operable?

El No No further Operability impact Determination actions Equivalent Change or necessary. Proceed with implementing the proposed Minor Alteration activity in accordance With governing process procedures.

LBD Change

9. Does the proposed LBD change introduce a new criteria for assessing SSC El Yes Complete Operability Impact Determination Worksheet Operability, for a Safety-Related SSC or any SSC that a Plant or ISFSI and attach completed worksheet to the LBD change.

Technical Specification is applicable to? (N/A if activity is fully scoped under a ED No No further Operability Impact Determination actions Tech Spec Amendment Request.)

or necessary. Proceed with LBD change per El N/A SWP-LIC-03.

2690R Pg 2.of-26290 R1 Page 2 of 3

O ENERGY OPERABILITY IMPACT DETERMINATION SCREENING NORTHWEST NOTE: For individuals implementing the OlD process for the first time, or if you are an infrequent performer and People. Vision-Solutions need a refresher, a tCBT training module on.the O!D)process is available on the ENNET at Computer'Based Training.

Activity Description (e.g., PPM, TMR#): LDCN-FSAR-10-025 Driving Document: 230095 Justification for not requiring completion of the Operability Impact Determination Worksheet (provide brief explanation as to why the applicable screening questions were checked "No"):

The LBD change does not introduce a new criteria. The LBD increases the allowable temperature for the fuel pool during a loss of one division of cooling from 155F to the design temperature of 175F, consistent with the guidance provided in RS-01.

This change does not impact the criteria for assessing Operability, which is the capability to preclude fuel pool boiling.

26290 R1 Page 3 of 3

Page:lot

~

e,.

~Control No.

Page: I of 2 ENERGY and Implementing Document Number(s) I Rev.

SNORTHWEST Ao-10-1432 Rev: 0 People Vision

• Solutions AR 230095230095APPLICABILITY DETERMINATION FOR LDCN-FSAR-10-025 LICENSE,BASIS CHANGES CAUTION:

Qualification (LDAA or LDAB) is required to:Use This Form Prepared By (Print)

Prepared By (Signatlre)

Date GREGORY, DE 7:>

03109111 15:39 Reviewed By :(Print)

Reviewed Bt (Signaturer De MORALES, TP a

03110111 08:39 I. Brief Description Of Activity:

(What Is being changed and why)

What:

1. The spent fuel pool temperature limit in FSAR Section 9.1.3,and FSAR Table 9.1-6 Is being revised from 155F to 175F.

2. The statement that the results of bounding fuel pool cooling analyses are provided in Table 9.1-6 is being removed from section 9.1.3.3 of the FSAR.

Why:

1. Columbia is currently limited to approximately 30 day or-longer outages because of fuel pool cooling limitations.

The increase in allowable temperature from 155F to 175F will permit shorter outages to be performed.

2. LDCN-FSAR-05-005 removed the results of bounding analyses from Table 9.1-6; therefore, this statement is no longer valid.

Regulatory Applicability Determination: See the 10 CFR 50.59 Resource Manual (RM), Section 4.2 and/or NEI.96-07, Appendix B, Section B4.1 for additional guidance to determine if 10 CFR 50.59 and/or 10CFR 72.48 applies tothe activity. See the reference procedure for guidance to determine if the activityoaffects the program plan, procedure or manual identified below.

Address eachaspect of the activity. If the answer is YES to any portion ofthe activity, apply the identified process(es) to that portion ofthe activity.

Note that it is not unusual to have more than one process apply to a given activity.

II. Regulatory Requirements and Controls (RR):

Section 4.2.1 Of the RM Does the proposed activity Inpact the:

Section 4.2.1__ fthe

1. Technical Specifications, Operating License or Cask CoC (10 CFR 50.90)

[Z]No []EYes if Yes, process per SWP-LIC-03

2. Operational Quality Assurance Program Description (10 CFR 50.54(a))

_'ZNo EYes If Yes, process per SWP-LIC-03

3. Physical Security Plan or Security Training and Qualification Plan (10 CFR.50.54(p))

• (]No EYes If Yes, process per SWP-LIC403

4. Emergency Plan (10 CFR 50.54(q))

'ZNo [JYes If Yes, process perSWP-LIC-03

5. IST ProgramPlan ( 10 CFR50.55a(f))

.[]No E]Yes If Yes, process.per SWP-IST-01

6. ISIProgram Plan (10.CFR 50.55a(g))

[D No E]*Yes If Yes, process per swP-ISI-01

7. PCLRT Program Plan (Tech. Spec55.12)

R'No 0-Yes If Yes, process per PPM 1.5.5

8. Other Programs f] No E]Yes if Yes, process per SWP-LIC-03 Examples are: Fire Protection Program, Environmental Plan, ODCM, COLR or other applicable controlling procedure Ill. Maintenance Activities (MA) (Does Not Apply to the ISFSI)

Section 4.2.2 of the RM Does the proposed activity involve:

1. Maintenance which restores SSCs to their original or new approved design condition?

Z]No [EYes if Yes, process per SWP-MAI-01 or other applicable controlling procedure

2. A temporary alteration supporting maintenance during an outage or that will be in effect ZNo E]Yes if Yes, process per controlling during at-power operations for 90 days or less?

procedure: e.g, PPM 1.3.9, PPM 10.2.53, GEN-RPP-14 IV. FSAR Modifications (FA)

Section 4.2.3 of the RM Does the proposed activity Involve a change to the:

1. FSAR (including documents incorporated by reference ) that is excluded from requirements 1E No I0jYes if Yes, process per SWP-LIC-03 to perform a 50.59 or 72.48 review as identified in section 4.2.3 of the Resource Manual.

V. Procedure Governing the Conduct of Operations (AC)

Section 4.2.4 of the RM Does the proposed activity Involves:

S 4

f

1. Managerial or administrative procedures or process governing the conduct of facility ZNo []Yes if Yes, process per SWP-PRO-02 operations subject to the control of 10 CFR 50, Appendix B.

or other applicable controlling i..

procedure

2. Regulatory commitment not covered by another regulation based change process.

r] No E]Yes if Yes, process per SWP-LIC-01 VI. Environmental Impact Evaluation (EE)

Does the proposed activity:

1. Affect the environment, or alter non-radiological plant effluents or rated power level?

w]No EYes If Yes, process a License Basis (See SWP-LIC-02)

Impact Evaation in accordance VII. Administrative Activities (AA)

Section 4.2.6 of the RM Does the proposed activity involve a:

1. A change that is an administrative activity subject to the controls of 10 CFR 50, Appendix B.

_]No E]Yes If Yes, process per SWP-PRO-02 (See SWP-LIC-03, Attachment 7.8) or other applicable controlling I__

Iprocedure

Page: 2of 2 EControl No.

ENERGY andImplementing Document Number(s) I Rev.

O NORTHWEST AD-10-1432 Rev: 0 People. Vision -Solutions AR 230095230095APPLICABILITY DETERMINATION FOR LDCN-FSAR-10-025 LICENSE BASIS CHANGES

] All aspects of the activity, are controlled by one or more of the processes above, therefore, neither a 50.59 or 72;48 review is required. (NO)

(*] Aspects of the activity are not controlled by any Of the processes above and are associated withthe plant. Therefore a 50.59review is required and should be initiated by completing the 50.59 Screen. (5059)

-]Aspects of the activity are not controlled by any of the processes above and are associated with the ISFSI. Therefore a 72.48 review is requiredand should be initiated by completing the 72.48 Screen. (7248)

Justification:

1. 5059 screen required for this change.

2. Removal of the statement in FSAR section 9.1.3.3 does not require a 50.59 review since it Is covered by previous 50.59 screen, 5059screen-05-0080.

Page: 1 of 7 E

RIOCFR50.59 Evaluation ST EN ER Y W ESTControl No. and Revision No.

paeople. Vision. Solutions 5059-11-0003, Rev. 0 10CFR50.59 EVALUATION Activity Title and Applicable Document Numbers:

POC Review:

LDCN-FSAR-10-025 Signature I Date LDCN No:

POC Meeting No.:

LDCN-FSAR-10-025 Prepared By (Print)

Prepared By (Signature)

Date GREGORY, DE GREGORY, DE 11/28/11 14:48 Revied By (rnt) e~w y (gnatuq Me-MORALES, TP MORALES, TP 01/05/12

.1108 Based Upon The Results Of This Evaluation:

Implement the activity per plant procedures without obtaining alicense amendment O Request and receive a license amendment prior to implementation Brief Description Of Activity. (What is being changed and why)

What:

The spent fuel pool temperature limit for an anticipated operational occurrence, loss of a fuel pool cooling division, In FSAR Section 9.1.3 and FSAR Table 9.1-6 is being revised from 155F to 175F.

Why:

Columbia Generating Station was licensed with a 175F allowable temperature for the fuel pool for an anticipated operational occurrence, with the stated capability to maintain 155F with the design heat load and cooling water temperature. During the power uprate, the capability value of 155F was re-categorized as a limit instead of a capability....*Current NRC guidance, RS-001, Indicates that the proper value for fuel pool temperature during an anticipated operational occurrence is the design temperature, which for CGS Is 175F.

Sum mary of Evaluation:

(To be used in preparation for the NRC report pursuant to 10 CFR 50.59 (d)(2). Refer to the Resource Manual)

Summary The purpose-of this change is to restore the allowed temperature of the fuel pool during the anticipated operational occurrence (loss of onetrain of fuel pool cooling) to the original licensed Value of 175F, consistent with NUREG-0892 and RS-001.

Discussion:

In evaluating the change in temperature on Columbia Generating Station (CGS), It was determined that CGS was designed to operate with a fuel pool temperature of 175F in the abnormal condition of an anticipated operational occurrence of the loss of a fuel pool cooling division. All equipment that would be subjected to the Increased heat and humidity from a fuel pool with 175F water temperature is qualified for this environment. The fuel pool and its cooling system are qualified for this temperature. The fuel in the fuel pool is qualified for much higher temperatures than 175F. The sumps in the pump rooms have sufficient capacity to handle fuel pool surface evaporation. The water makeup system was designed to handle a boiling fuel pool. CGS was granted its original license based on an allowable temperature of 175F for the fuel pool, and evaluated against this temperature in NUREG-0892.

As this change is not adverse to a design bases function or a safety analysis, will not cause an accident or malfunction previously evaluated, will not create a new accident or malfunction not previously evaluated, and is not a departure from a method of evaluation described in the FSAR, NRC prior approval is not required.

References:

FSAR Section 9.1.3, Spent Fuel Pool Cooling and Cleanup System FSAR Table 9.1-6, Bounding Fuel Pool Cooling Events PER 205-0093 dated 3/18/2005 NUREG 0892 dated March 1982 FSAR Amendment 25, June 1982 FSAR Amendment 30 of June 1983 TS Amendment 137 of May 1995

Page: 20of7 ENERGY o1OCFR50.59 Evaluation NORTHWEST Control No. and Revision No.

• People. Vision-Solutions 5059-11-0003, Rev. 0 10CFR50.59 EVALUATION FSAR Amendment 44 of April 1992 License,Amendment G02-93-180 dated 71911993 with attachment. NEDC - 32141P Power Up-rate SER 1(G12-95-099)

MLEA Report No. 2009174001 of December 2009 RS-001, Revision 0, December 2003, Review Standard for ExtendedPower UpRates to Matrix 5 to RS-001 Fuel Pool Temperature History:

The October 7, 1981 NRC Auxiliary System Branch Meeting, draft SER review of open Items, section 9.1.1.3 stated for Energy Northwest to. verify that Spent Fuel Pool temperature does not exceed 175F when worst case :heat load is calculated with the loss of one Spent FuelPool Cooling System.

On Feb 10, 1982, John Redgley of the NRC voiced a concern with 175F. When told that actual expected peak temperature, based on design loads and temperatures, would be less than 155F, Mr. Redgley stated that this was acceptable, hewould close-the SER open item and he requested both a written confirmation and an amendment to place themanufacturers performance data in the FSAR.

On Feb 18, 1982, G02-82-216 was issued stating that an analysis had been pwrformed which determined that one pump and one heat exchanger was capable of maintaining the fuel pool below 175F.

In amendment 25,1dated June 1982, the following statements were added to the FSAR by SCN 81-506, which stated that the SGN was issuedto.cover, in part, -the ASB Meeting open issues of October 7th:

9.1.3.2.1 Normal Operation: An. evaluation of maximum spent fuel pool temperature was done using ASB TP 9-2, and; it was found that the temperature remained below 175F with one safety division of fuel pool cooling.

9.1*3.2.1 Normal Operation: The operating temperature of 125F Is permitted to rise to 150F when the circulating flow is Interrupted for draining the reactor well and dryer-separator pit, or to a maximum of 175F when one of the FPC pumps or heat exchangers is unavailable.

9.1.3.3 Safety Evaluation: The system maintains fuel pool water temperature below 175F in the event that only one pump and one heat exchanger are available.

In Amendment 30, June 1983, SCN 82-145 revised all temperatures to 155F. The stated purpose of the SCN was to properly reflect fuel pool cooling system capability. However, instead of differentiating between the design limit of 175F and the capability (with a set group of inputs), the SCN simply changed 175F everywhere to 155F.

NUREG-0892 was issued, stating that the ability to maintain the fuel pool at or less than 175F was acceptable.

In TAC M87076, Response to RAJ Power Uprate Review, the question was asked if the power uprate affected the capability of the FPC system to maintain 155F assuming a normal heat load and a maximum coolant temperature.

To which, the answer was that the resultant maximum temperature was less than 155F. Despite that the question recognized that 155F was a capability, the power uprate SER, amendment 137, characterized 155F as a limit. A subsequent FSAR change In 2005 added the word "limit" to be consistent with the SER.

Page: 3 of 7 ENERGY IOCFR50.59 Evaluation NORTH WEST Control No. and Revision No.

People.-Vision. Solutions 5059-11-0003, Rev. 0 10CFR50.59 EVALUATION Columbia was originally licensed under revision I of the Standard Review Plan (SRP). The original licensing SER, NUREG -0892, documents the degree of Columbia's compliance with the SRP. The SER noted that Columbia was designed to limit temperature to 125F fornormal heat loads, with two Fuel Pool Cooling (FPC) trains available, which was below the staff acceptance criteria of 140F. The SER noted that with larger-than.normal batches of spent fuel pool, the "abnormal" heat load temperature of 1.50F was within acceptable limits. Finally, the SER noted that during the maximum heat Ioad conditions and the single failure of one spent fuel pool cooling train, the maximum bulk fuel pool water temperature will not exceed 1750F. Based on the abnormal heat load, the single failure, andthe ability to maintain the pool temperatureat or below 175°F, the staff found this temperature acceptable.

As discussed above, the Power Uprate SER addressed the FPC capability by stating: "The licensee analyzed the failure of a singletcooling train under the same heat loading conditions and determined that the 155°FFSAR limit for a single failure would not be exceeded." The SER statement correctly Identified the. FPC capability, but erred in identifying the 155 *F as a limit. The true limit was 175F, consistent with the design of the plant and NUREG-0892.

Further, :RS-001 was issued by the NRC to supersede guidance in SRP Revision 1. RS-001's acceptance criteria are based upon an analysis of a bounding heat load, rather than normal heat loads. For bounding heat loads, the acceptance:criterion was the ability to maintain the spentfuelpool below the design temperature of the spent fuel pool structureaand liner following a single active failure or a design bases event. At Columbia this temperature is 212F. However, the Fuel Pool Cooling system design temperature is limiting at 175F.

In defining the limiting temperature back to its original licensed value of 175F, Energy Northwest is aligned with RS-001 forits bounding heat load case, as well as Columbia's original SER, NUREG-0892. As such, the change is consistent with NUREG-0892 and RS-001, both of which are NRC approved documents.

Page: 4 of 7E1 10CFR50.59 Evaluation ENO RTHT Control No. and Revision No.

ONORTHWEST People. Vision. Solutions 5059-11-0003, Rev. 0 10CFR50.59 EVALUATION iNOTES:

Provide a separate written response with the basis for the answer to.each question. The "10 CFR 50.59 Resource Manual" should be used to determine the contentof each response (see Section 6.2 for guidance).

Identify references used to perform evaluation (either In a single list In the reference section or within the written responses)

If the answer to any of the 50.59 questions Is "YES", then the proposed activity may not be Implemented until a License Amendment has been obtained from the NRC.

Throughout this evaluation, UFSAR refers to the current FSAR AS UPDATED per 10 CFR 50.71, approved changes to the FSAR which have not been submitted to the NRC by amendment and documents incorporated Into the FSAR by reference.

r EFFECT ON ACCIDENTS AND MALFUNCTIONS PREVIOUSLY EVALUATED IN THE UFSAR

1. Does the proposed activity result in more than a minimal Increase'in the frequency. of occurrence of an ACCIDENT PREVIOUSLY EVALUATED IN THE UFSAR? (See Section 6.2.1 of the 10 CFR 50.59 Resource Manual)

DYes

]No None of the accidents previously evaluated in theFSAR were postulated to be caused by fuel pool temperature or from evaporation from the fuel pool surface. At the time that the accident evaluations were performed,,the allowable fuel pool temperature during an anticipated operational occurrence was 175F, as this was the allowable temperature at1the time of Columbia Generating Station obtaining'its license. All equipment'in an area whose

,relative humidity could be affected by the fuel pool have been evaluated for,100% relative humidity. Therefore, the changing of.the allowed fuel pool temperature during an anticipated operation occurrence (loss of.one train of

,cooling) from 155F to the design temperature of 175F will not result in more than aminimal increase in the

.frequency of occurrenceof an accident previously evaluated in the FSAR.

2. Does the.proposed activity result in more than a minimal increase in the likelihood of occurrence of a.MALFUNCTION OFAN SSC IMPORTANT TO SAFETY previously evaluated in the UFSAR? (See Section 6.2.2 Of the 10 CFR 50.59 Resource Manual)

QYes jiNo No malfunctions previously evaluated In the FSAR were postulated to be caused by fuel. pool temperature-or

.evaporation from the.fuel pool surface. At the time that the safety analyses were. performed, fuel pool had an allowable temperature of 175F. Therefore, the conditions created by the increase In temperature from 155F to 175F have already been taken.into consideration in whether the conditions would cause a malfunction of equipment. All equipment in the upper areas of the reactor building, the area that would be affected by Increased

• humidity from an increased fuel pool temperature, is qualified to operate in a 100% relative humidity environment.

Therefore, the changing of the allowed fuel pool temperature during an anticipated operation occurrence (loss of one train of cooling) from 155F to the design temperature of 175F, consistent with the stated temperature In the FSAR at the time of obtaining its license, will not result in more than a minimal increase in the likelihood of occurrence of a malfunction previously evaluated in the FSAR.

3. Does the proposed activity result in more than a minimal increase in the consequences of an ACCIDENT PREVIOUSLY EVALUATED IN THE UFSAR? (See Section 6.2.3 of the 10 CFR 50.59 Resource Manual)

QYes No As stated in the FSAR, the fuel pool cooling system is not credited for mitigating the consequences of a design basis event. During an accident, the purpose of the fuel pool cooling system and the fuel pool itself is to prevent fuel pool boiling. The proposed activity, increasing the allowable fuel pool temperature during the anticipated loss of a division of fuel pool cooling during normal operation from 155F to 175F, has no -relation to an accident, or to fuel pool boiling, as an anticipated operation occurrence of the loss of one division of fuel pool cooling and design bases accidents are -treated as separate events.

Accident analysis, such as the drawdown analysis, assumes the fuel pool is at Its maximum normal operating temperature of 125F. This change does not affect this temperature and therefore does not affect the consequences of an accident.

Therefore, the changing of the allowed fuel pool temperature during an anticipated operation occurrence (loss of one train of cooling) from 155F to the design temperature of 175F will not result in more than a minimal increase

Page: 5ENERGY 1

1 OCFR50.59 Evaluation NORTHWEST ControNo and Revision No.

People. Vision. Solutions 5059-11-0003, Rev. 0 1 OCFR50.59 EVALUATION in the consequences of an accident previously evaluated in the FSAR.

4. Does the proposed activity result in more than a minimal increase in the consequences of a MALFUNCTION OF AN SSC IMPORTANT TO SAFETY previously evaluated in the UFSAR? (See Section 6.2.4 of the 10 CFR 50.59 Resource Manual) rYes No As stated in the FSAR, the fuel pool cooling system is not credited for mitigating the consequences of a-design basis-event. During a previously evaluated malfunction, the purpose of the fuel pool cooling system and the fuel

.pool itself is to prevent fuel pool boiling. The proposed activity, increasing the allowable fuel pool temperature during the anticipated loss of a division of fuel pool cooling during normal operation from 155F to 175F,has no relation to a malfunction event,. or to fuel pool boiling, as an anticipated operation occurrence of the loss of one division of fuel pool cooling and malfunctions are treated as separate events.

Accident analysis, such as.the drawdown analysis, assumes the-fuel pool is at its maximum normal operating temperature of 125F. Foruthat analysis, a concurrent single failure (e.g., loss of a division and, therefore, loss ;of a

• FPC train) does not affect the temperature during drawdown. The post-accident fuel pool temperature of 175°F occurs after drawdown. This change does not affect the initial-temperature and therefore does not affect the consequences of a malfunction.

Therefore,:the changing of the allowed fuel pool temperature during an anticipated operation occurrence (loss-of one train of cooling) from 155F to the design temperature of 175F will not-result in more than a minimal:increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the FSAR.

_POTENTIAL :FOR CREATION OF A NEW TYPE OF EVENT NOT PREVIOUSLY EVALUATED IN THE UFSAR

5. Does the proposed activity create a possibility for an accident of a different type than any previously evaluated in the UFSAR? (See Section 6.2.5 of the 10 CFR 50.59 Resource Manual)

"Yes

-'No The proposed activity-is to-increase the allowable temperature for the anticipated operation occurrence (AOO) of the loss of a fuel pool cooling division from 155F to the original licensed value of 175F. During the AOO, the increase in temperature would result in an increase in vapor generation from the fuel pool surface. The resulting increase in relative humidity would result in an increase In effluent to the ECCS pump room sumps, since condensing vapor would end up in the floor drains.

The difference In evaporation from an increase in water temperature from 155F to 175F is minor. The ECCS sump pumps operate for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> combined a day, indicating that normal flow of water into the sumps Is negligible. A boiling fuel pool would release less than 24 gpm into the sumps, based on the 10.1 MBTU/hr fuel pool heat load in FSAR section 9.1.3.2.1.1; a fuel pool at 175F would release significantly less than this value. Each sump pump is capable of 50 gpm. Therefore, the system is designed to prevent Water buildup in a pump room during the AOO. Equipment that are required to respond to a DBA, located in the areas of the reactor building that would be affected by Increased vapor release, have been evaluated for 100% relative humidity. Therefore, the plant has been. designed to handle the increased humidity levels from the AOO.

The minor increase in evaporation is well within the makeup capabilities of the fuel pool makeup systems, which were designed to provide makeup to a boiling fuel pool.

The temperature effects on plant equipment are bounded by accident analysis, which has an allowable fuel pool temperature of 175F and assumes loss of the normal HVAC system. During an AOO the normal HVAC is still operating, taking suction off the areas just above the fuel pool. Therefore, area temperatures will be less during an AOO than during an accident. All equipment that is required to respond to a DBA, and is located in the area, is designed for accident temperatures.

Therefore, the proposed activity does not create the possibility for an accident of a different type than any previously evaluated in the FSAR.

Page: 6 of T.,

10CFR50.59 Evaluation E

NORTHWEST ControlNo. and Revision No.

People. Vision, Solutions I

5059-11-0003, Rev. 0 10 CFR50.59 EVALUATION

6. Does the proposed activity create a possibility for a MALFUNCTION OF AN SSC IMPORTANT TO SAFETY with a different result than any previously evaluated in the UFSAR? (See Section 6.2.6 of the 10 CFR 50.59 Resource Manual' QYes RNo The proposed activity is to'increase the allowable temperature for the anticipated operation occurrence (AOO) of the'loss of a fuel pool cooling division from 155F to the original licensed value of 175F. During the AOO, -the Increase in temperature would result in an increase In vapor generation from the fuel pool surface. The resulting increase in relative humidity would result in an -increase in effluent to the ECCS pump room sumps, since condensing vapor wouldend up in the floor drains. The difference in evaporation from an increase in water temperature from 155F to 175F is minor. The ECCS sump pumps operate for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> combined a day, indicating that normal flow of water into the sumps is negligible. A.boiling fuel pool would release less than 24 gpm into the sumps based on the 10.1 MBTU/hr heat load in FSAR section 9.1.3.2.1.1; a fuel pool at 175F would release significantly less than this value. Each sump pump is capable of 50 gpm. Therefore, the system is designed to prevent water buildup in a pump:room during the AOO. Critical equipment in the areas of the reactor building that would be affected by increased vapor release have been evaluated for 100% relative humidity.

Therefore, the plant has been designedto handle the increased humidityllevels:from the AOO.

The mino1r increase in evaporation is well within the makeup capabilities of the fuel pool makeup systems, which were designed to provide makeup to a boiling fuel pool.

The temperature effects on plant equipmentlare bounded by accident analysis,which has an allowable fuel pool temperature of 175F and assumes loss of the normal HVAC system. During an AOO the normal HVAC is still operating, taking suction off the areas just above the fuel pool. Therefore, areastemperatures will be less during an AOO than during an accident. All critical equipment in the area is designed for accident temperatures.

Therefore, the changing of the allowed fuel pool temperature during an anticipated operation occurrence (loss of one train of cooling) from 155Fto the design temperature of 175F, as allowed by RS-001, will not result in more than a minimal increase in the likelihood of occurrence of a malfunction previously evaluated in the FSAR.

IMPACT ON FISSION PRODUCT BARRIERS ASDESCRIBED IN THE UFSAR

7. Does the proposed activity result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered? (See Section 6.2.7 of the-10 CFR 50.59 Resource Manual) r,] Yes

..NNo The new allowable temperature of 175F is well below the temperature limits for fuel, and is the design temperature for the fuel pool cooling system. The fuel pool structure and liner were designed for a boiling fuel pool (212F);

however the design temperature of the fuel pool cooling system is the limiting temperature. Therefore, increasing the allowable temperature for an anticipated operation occurrence of the loss'of a fuel pool cooling division from 155F to the Fuel Pool Cooling system design temperature of 175F Will not result in a design basis limit for a fission product barrier being exceeded or altered.

IMPACT ON EVALUATION METHODOLOGIES DESCRIBED IN THE UFSAR

8. Does the proposed activity result in a DEPARTURE FROM A METHOD OF EVALUATION DESCRIBED IN THE UFSAR used in establishing the DESIGN BASES or in the SAFETY ANALYSIS? (See Section 6.2.8 of the 10 CFR 50.59 Resource Manual)

[] Yes 1@ No The proposed change does not affect an analytical method of evaluation used in establishing the design bases for the fuel pool cooling system or in a safety analysis. The design basis is to maintain 125F with two heat exchangers and to prevent boiling during an accident. The proposed change is a change in the allowable result that is specified in the FSAR for an anticipated operational occurrence of the loss of one division of fuel pool cooling. Before every outage, Energy Northwest determines the maximum fuel pool temperature that could exist based upon operating the entire outage with only one heat exchanger available, to bound the temporary loss of one heat exchanger. The acceptance criterion used for this analysis is the 155F stated in the power uprate SER.

Page: 7 of 7

4.

ENERGY 1

IOCFR50.59 Evaluation

• NORTHWEST Control No. and Revision No.

People -Vision-SolutionsI 5059-11-0003, Rev. 0 1 OCFR50.59 EVALUATION I

This LDCN is changingthe acceptance criterion to the 175F allowed by Columbia's-original SER (NUREG-0892) and which is less than the 212F allowed by RS-001, Review Standard for Extended Power Uprates. Both NUREG-0892 and RS-001 are NRC approved documents. Since the use of lower than RS-001 recommended limits is not a change to an analytical method, and the method of analyzing the fuel pool is not changing, this change does.not result in a departure from a method of evaluation described in the FSAR used in establishing the design

.bases or in the safety analysis.

Engineering Change Print Date:

03/09/2012 EC.Number Status/Date Facility Type/Sub-type:

0000010787 000 ACTIVE 03/09/2012 02 MALT ENERGY ONORTHWEST

.Page.

2.&

EC

Title:

REMOVE A SECTION.OF THE BIRD SCREEN ON THE ROA SYSTEM INTAKE Mod Nbr :

25736603 KWI:

KW2:

KW3 KW4:

KW5:

Master EC Outage WO Required Adv WkAppvd:

Auto-Advance:

Caveat.Outst:

Resp Engr Location N

N N

Y

.PAUL Work Group :

Alert Group:

Image Addr Alt. Ref.

Priority Department T HAND AR-EFIN Temporary Aprd Reqd Date:

Exp Insvc Date:

Expires On Auto-Asbuild:

Discipline N

03/09/2014 N

Milestone 1.10-PREPARE EC 130-'SDI & RVW 155ýDES RVW EC 170-SUPV MMOD 550-TURNOVER CI 580-OPERABILITY 805-.CLOSURE RVW Date 03/01/2012 03/06/2012 03/06/20.12 0.3/09/2012 PassPort PHAND PHAND JLEGAR2 DBRAND2 Name HAND HAND LEGARE BRANDON PAUL PAUL JASON DENISE Reql By H/APPR APPROVED APPROVED APPROVED MODIFIED MODIFIED CLOSED Units Fac Uni-t 02 00 Description COLUMBIA GENERATING STATION -

POWER BLOCK Systems Fac Syst 02 ROA Description REACTOR (BUILDING)

OUTSIDE A Affected Documents List Sub -

Fac IU Type Document 02 AED ARC A503 Minor Rev:

Major Rev:

Title:

W T AND REACTOR BUILDING SOUTH EL 02 AED MEC M810 Minor Rev:

Major Rev:

Title:

REACTOR BUILDING EL 572FT HVAC PLAN Sheet Ops Rvw Pri Inc N

Y I

Y Updt Due:

04/08/2012 N

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04/08/2012

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• OR WEST~~

MINOR PLANT DESIGN EC 103*5 NO'If "WEST CHANGE Pmuiae-ViNion* SolutionC Equipment Status:

Control panels E-CP-H 13/P680, E-CP-H I3/P60 1, E-CP-H 13/P683, and E-CP-C61/POO will still beoperating and have live circuits. Take extra cautionary measure when working in these control panels, Interfacing System Impact:

" Valve RHR-V-6B supports Low Pressure Coolant Injection (LPCI), suppression pool cooling, containment spray cooling, and maintains primary containment integrity.

" Valve RHR-V-I 6B maintains pressure integrity to support containment spray cooling.

" Valve RHR-V-49 supports Low Pressure Coolant Injection (LPCI), suppression pool cooling, containment spray cooling, and maintains pressure integrity-to support primary containment.

Valve RHR-V-74B maintains pressure boundary integrity.

" Valve RHR-Výl 15 is used to direct SSW into the reactor vessel and primary containment fbr containment flooding and is used only in the emergency operating procedure. The valve is used to maintain Pressure Integrity to support secondary containment integrity.

Technical Specification Requirements:

One RHR loop has to be in operation to meet Tech Spec 3.4.9 requirements.

Special Considerations:

Due to live circuits-in panels E-CP-H 13/P680, E-CP-H13/P601, E-CP-H13/P683, and E-CP-C61/P00l, ensure cautionary measures are taken to avoid personal injury and equipment damage.

" Under normal operating condition power for the RHR-V-6B is fed from E-MC-8BA compartment 2B, When working on this valve ensure that power is disconnected to avoid any safety hazards.

" Under normal operating condition power for the RHR-V-16B is t-d from E-MC-8BA compartment 5C.

When working on this valve ensure that power is disconnected to avoid any safety hazards.

" Under normal operating condition power for the RHR-V-49 is fed from E-MC-8BB compartment 7A.

When working on this valve ensure that power is disconnected to avoid any safety hazards.

Under normal operating condition power for the RHR-V-74B is fed from E-MC-8BB compartment 5C.

When working on this valve ensure that power is disconnected to avoid any safety hazards.

• Under normal operating condition power for the RHR-V-l 15 is fed from E-MC-8BB compartment 7D.

When working on this valve ensure that power is disconnected to avoid any safety hazards.

Specific Construction Activities:

All work activities in the MCR and the Remote Shutdown Room shall be performed with caution with respect to the precautions listed in PPM 10.25.46 section 4.6, revision 18. Extreme caution must be exercised to avoid hazards.

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Vilan-UaluIon MINOR PLANT DESIGN CHANGE MPDC Pg.No.f 2-77V 7n-pk

,/;/a EC 10365 The functional test for the-RHR system cabling requires all the components to be installed at their permanent locations with required utilities to be in place (i.e. power). Thepurpose of these tests is to ensure full functionality of the system.

I Continuity for cables Ai I rfktv rn'i PPM 10.25.19 0 10,

• ..2.

Test Continuity Test

'DO NOT

STROKE,

'RfR-V416B.

Stroke MOV Test Valve Full PPM 10.25.74 See additional Open/Full Closed steps below to validate wires are correctly

_installed.

Control Logic and Test Circuit logic PPM 10.25.74

-interlock check functionality Onarahilib, TaBtlnn nf RHR.V.16R fnlleiwina imnlamantntlnn nf Ft~ IO2SS riarabilitv Testinn of RHR-V-ISB f6iloWina imn mentation of.E(! 10365 The following steps were developed in discussions with Systems Engineering and the field and serve to demonstrate that RHR-V-16B circuit wiring disturbed in the course of this EC work being performed is electrically restored to operable conditions:

1. At H13-P601, verify the green valve position indication light is illuminated for RHR-V-16B.

2. At E-MC-8BA compartment 5C remove the 3 amp control power fuse for RHR-42-8BA5C.

3. At H13-P601, verify the green valve position indication light is extinguished for RHR-V-16B.

4. At the remote shutdown panel E-CP-C61/P001, locate terminals BB-8 and BB-9.

5. At the remote shutdown panel E-CP-C61/P001, using a multi meter, measure voltage from terminals BB-8 ýto BB-9 and verify there is less than 1 VAC (approx. zero VAC).

6. At the remote shutdown panel E-CP-C61/P001, using a multi meter, read resistance voltage from terminals BB-8 to BB-9 and verify is an.open condition for the circuit from BB-8 to the control room and back to terminal BB-9 (more than 50 ohms resistance). 50 ohms resistance was chosen as there are a couple of parallel paths for circuit continuity; one of which is through the open contactor coil and green valve position light and another through the open contactor coil and the coil for relay RHR-RLY-80NV16B. Both of these paths are expected to be greater than 50 ohms.

7. Request Operations to place and hold the control switch RHR-RMS-V/1 6B to the OPEN position.

MPOC Page No. 382-S

-r#-tP

/12.

ENORT E

MINOR. PLANT DESIGN EC 10365

',NORTHWEST CHANCE NOTE: The following step confirms the modification made under EC 10365 supports that when the control

,switch for RHR-RMS-V/1 6B is placed in the open position the circuit continuity is made from the remote shutdown panel terminal BB-8 to the control room panel H13-P601 and back to.the remote shutdown panel BB49. This step satisfies the Operability testing for the Open control circuit for RHR-V-168.

NOTE: The wiring associated with relays RHR-RLY-K109BX and :RHR-RLY-K58B within panel)H13-P618 was

ýnot disturbed as part Of EC 10365 and this testing of these relays or their logic is not required. Confirming that continuity exist when the control switch is placed to the Open position proves the valve will open when demanded based on the previously performed two year valve operability surveillance OSP-RHR/IST-Q706 performed under 01178077 during R20 and the next scheduled surveillance :WO 02008334 scheduled in R21.

8. While Operations is holding the control switch RHR-RMS-V/1 6B to the OPEN position, using a multi meter, read resistance from the following terminals:

8,a. At H13iP601 switch RHRR-RMS-,Vl16B terminals 3 and 4, verify there is~less than 1 ohm resistance, This check tests theRHR-RMS-V/16B 3-4.

8;b. At H13 P6318 terminals AAA7-8 and. AAA-20, verify there is less than 1 ohm resistance.

This check tests the RHR-RLY-K1098X contact 1-7.

8.c.At -the Remote Shutdown panel E-CP-C61/P001, read resistance from terminals BB,8 to BB-g.

9. Inform Operations to return control switch RHR-RMS-V/1 6B back to normal position (spring

,return).

10. Confirm!by adding the resistances measured in steps 8.a.and 8.b and subtracting this value from the resistance measured in step 8.c:that the cable resistance is lessthan 1.5 ohms.

11.,At the remote shutdown panel E-CP-C61/P001, using a multi meter. read resistance voltage from terminals BB-8 to BB-9 and verify there is an open condition. For the circuit from BB-8 to the control room and back to terminal BB-9 (more than 50 ohms-resistance). 50 ohms resistance was chosen as there are a couple of parallel paths for circuit continuity; one of which his through the open contactorcoil and green valve position light and another through the open contactor coil and the coil for relay RHR-RLY-80N 1 6B. Both of these paths are expected to be greater than 50 ohms. This value should match closely to the value measured earlier.

12. At E-MC-8BA compartment 5C re-install the 3 amp control power fuse for RHR-42-8BA5C.

(needs to have second person verification that the fuse is properly installed)

NOTE: The following step confirms the proper installation of the control power fuse for RHR-V-16B.

13. At H13-P601, verify the green -valve position indication light is Illuminated for RHR-V-16B.

Drawing

References:

EWD 9E-098 EWD 9E097 EWD 9E-028 Scenario: 5k Scenanrio

Description:

Spurious motor-operated valve operation, AND Wire-to-wire short(s) bypass torque and limit switches.

DiscusdonlNotes:

General scenario is that fire damage to motor-operated valve circuitry causes spurious operation. If the same fire causes.wire;,to-wire short(s) such.,that the valve torque :and -limit switches are bypassed, then the valve motor may stall at the end of the valve cycle. This can cause excess current in the valve motor windings as well as valve mechanical damage. This mechanical damage may be sufficient to prevent manual operation of the valve. Scenario only applies to motor-operated valves. Note this generic issue may have already been addressed during disposition of NRC Information Notice 92-18. This disposition should be reviewed in the context of multiple spurious operations and multiple hot shorts.

An IN 92-18 analysis or mechanical analyses were completed as part Of PFSS analysis, NE 85-19. This disposition has been reviewed in the context of multiple spurious operations and multiple hot shorts. Ensure that the evaluation meets the circuit analysis guidelines of NEI 01 Rev. 2 1 RG 1.189. The IN 92-18 Analysis is documented in"Section lel of NE-02-85-19.

Is there any credit taken ABN-FIRE to manually open -a spuriously closed MOV / close a spuriously opened -one?

Review PFSS analysis.

Review ABN-FIRE for MOVs that are manually controlled.

There are also considerations for valve rupture, apparently, for these scenarios.

Final Disposition:

A Main Control Room Fire can cause all five of the normally closed valves RHR-V-6B, RHR-V-16B, RHR-V-49, RHR-V-115 and RHR-V-74B valves to fail open bypassing their torque switch. In accordance with IN 92-12 enforcement notice Weak link analyses have been provided for these valves as well, however following a Main Control Room fire, these valves may not be operable or in the desired position. A fire watch in the MCR is not necessary since it is continuously manned.

A :solution for a similar past problem protected the de-energized conductor up to the E-CP-H13/P601 panel "open" control switch, from hot shorts (using grounded flex as a 15 minute barrier) until Remote Shutdown transfer switches could be operated. RHR-V-74B and RHR-V-115B do not presently have Remote Shutdown transfer switches and will need to be added.

AR EVAL 226804 has been written to start the design change process. AR CR 226807 Option 1: IN 92-18 Main Control Room Modification:

RHR-V-74B

, Cable P/O 8001/El2A-006 in the main control room should be placed into grounded flex conduit.

C Icl>3&t5 fa At the remote shutdown panel term point BB-64 should be placed into RHR-RMS-RSTS13 {41-41C} (this RMShas spare contacts)

RHR-V-16B Cable P/O 8001/E21A-019 from E-CP-H13/P680 into E-CP-H13/P601 in the main control room should be placed into grounded flex conduit. (from term point 7 in E-CP-H 13/P680 to contact 3 on.RHR-RMS-V/16B in E-CP-H I.3/P601)

RHR-RLY-KIO9BX and RHR-RLY-K58B should be moved into the normally energized leg of the open contact for RHR-RMS-V/I:6B in E-CP-H13/P618.

RHR-V-6B From term 9 in E-CP-H13/P680 to contact 3 of RHR-RMS-V/6B.in E-CP-H13/P601 in the main control room should be placed into grounded flex conduit.

RHR-V-l 15

" From' term 5 in.E-CP-H.13/P680 to contact 3 of RHR-RMS-V/1 15in E-CP-H13/P601 in the main control room should be placed into grounded flex conduit.

At the remote shutdown panel term point BB-65 should be placed into RFHR-RMS-RSTS 13 {43-43C.) (this RMS 'has spare contacts)

RHR-V-49 From term 6 in E-CP-H13/P680 to contact 2T of RHR-RMS-V/49 in E&CP-H13/P60l in the main control room. should be placed into grounded flex conduit.

Option 2: Protect cable from MCC to Torque and Limit Switches:

An armored cable ran from the MCC to the valve's torque and limit switches.

Option 3: No Modification:

Risk violation in the 2012 NRC Triennial Fire Protection Inspection

References:

IN 92-18 ABN-FIRE ABN-CR-EVAC Actions:

CR 226807 written to capture issue. Complete design change.

Preparer Jim Civay s

Date I 10/5110 Reviewer Denise Brandon Date 10/5/10