ML20012E219
| ML20012E219 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/31/1989 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 90-137, NUDOCS 9004020064 | |
| Download: ML20012E219 (405) | |
Text
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>4 VIM 01NIA ELECTRIC AN1) POWEH COMPANY j HIC 11 MON!),Y1HU1NIA 2(12 61 J
i March 20, 1990 I U. S. Nuclear Regulatory Commission Serial No 90 137 Attn: Document Control Desk NAPS /JHL/TAH ; Washington, D. C. 20555 Docket No. 50 338 l 50-339 l License No. NPF 4 ; NPF 7 Gentlemen: l l VIRGINIA ELECTRIC AND POWER COMPANY l NORTH ANNA POWER STATION UNITS 1 AND 2 l l
SUMMARY
OF PROCEDURE CHANGES. FACILITY CHANGES
- j. AND SPECIAL TESTS PURSUANT TO 10 CFR 50.59 Pursuant to 10 CFR 50.59(b)(2), enclosed is a summary of procedure changes, facility changes and special tests and a summary of the safety evaluation for each, that were completed for North Anna Units 1 and 2 in 1989.
Very truly yours, i b W. L. STEWART Senior Vice President Nuclear l Attachments: i cc: U. S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900
- Atlanta, Georgia 30323 Mr. J. L. Caldwell NRC Senior Resident inspector North Anna Power Station 9004020064 891231 gi DR ADOCK0500g8 1 1
p -- o I o SUMMARIES OF 10 CFR 50.59 CHANGES NORTH ANNA POWER STATION L t ! i l l l i I t ! i
- l. ,
t I L L 1 VIRGINIA ELECTRIC AND POWER COMPANY I
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(, SPECIAL 1TBTS RIQUIRING A SAITIY EVA11RTIOi DLRING 1989 f 1-ST-79: Verification of Fall Safe Position of Safe Shutdown Air Operated ', I Valves i k
Purpose:
To verify air operated valves required for safe shutdown fall to their fall I safe positions upon loss af air. L. ' Dates: 1he test was performed from 05/25 to 07/12. b Unit conditions: p' The test was performed on Unit I with the unit is hbdes 5 and 6. Results.Sunnary:
.The - test verified the valves failed to their fall safe positions upon loss of air, f
Safety Evaluation Summary:
- 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the- safety analysis report was not increased. Doration and inventory addition flowpaths remained operable during testing. The unmonitored release of a waste gas decay tank is an analyzed event.
The valves and systems are designed for the conditions that were present during testing.
- 2. The possibility for an accident or malfunction c( a different type than was previously evaluated in the safety analysis report was not created. The valves are designed to be cycled and designed for the conditions that were present during testing.
- 3. The margin of safety as described in the bases of any technical specification was not reduced. A boration flow path remains operable at all times. A waste gas decay tank release will not occur due to .
4 normally closed isolation valves. ' i _ ,sJ
3-ST-80: Verification of Adequate Back Up Air Supply for Safeguards Area Exhaust Dampers
Purpose:
Tb verify _the adequacy of the Safeguards Area exhaust dampers back-up air supply. Dates: The test was performed on 05/15 and 08/01. Unit Conditions: The test was performed on Unit I with the unit in hbde S. Results Summary: The test verified the adequacy of the Safeguards Area exhaust dampers back-up air supply. Safety Evaluation Summary:
- 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated-in the safety analysis report was not increased. The dampers and fans are designed to be cycled. The accumulator is designed to cycle the dampers. Equipment -could be placed in service immediately if required.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. Operability of the safeguards exhaust fan and exhaust damper were not adversely affected during testing. The dampers and fans are designed to be cycled.
3.- The margin of safety as described in the bases of any technical specification was not reduced. The safeguards area ventilation system was capable of providing ventilation as described in the bases of the technical specifications.
aM_ G 4thl4 V LRf . fdj' M[ ) s &paa p,y' , (1-ST-81: Mitregon Reserye Tank Onpacity Verifieation (Pzr MRVs):
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Purpose:
'To evaluate,the pressurirer KRVs' nitrogen usage and to verify the adequacy of - . v the.KRV reserve tank. , s
' Da't es s ._ , .;
4' The' test was performed on 06/16 and 06/29.' :
.m ; *7,, Unit Conditions: ,j W The.' test was performed on Unit I with the unit in Modes 5 and 6.. i i
_ Resul t s Surenary: - l The test verified the adequacy of the' KEY reserve tank. j s.* ' Safety Evaluation Summary:
-)
j s A ,1. The probability of occurrence or the consequence of an accident or g malfunction of-equipment important to safety and previously evaluated. j slJ ' ~ln the safety analysis report was not increased. The nitrogen reserve , t u tanks - and -associated PORVs were isolated and tested individually to. j collect data for verification of tank capacity. When applicable, the limiting' condition or operation was comp! led with or the appropriate action statement was entered.--The overpressure protection system .was- .j D not affected during testing. [
- 2. The possibility of an accident or malfunction of a different type than !
was previously evaluated in the safety analysis ~ report was. not. 3
'[ ' created. The nitrogen reserve tanks and assoclated MEVs .werc )
< isolated and. tested individually to collect data for verification of ;
W , t'ank ' capacity. When applicable, the limiting condition for operation .i was' complied ~with or the appropriate action statement was entered. N
- 3. The margin of. safety as described in the bases of^ any' technical -{
N specification was not reduced. Overpressure protection was provided I when required during performance of the test. 'r
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'1-ST-82: Blowdown of Instrument Air System Main Headers and Branch Lines
Purpose:
To remove contaminants from the instrument air (IA) main headers and branch
!? lines..
i Ea.118' The test was performed from 06/06 to 07/05. t
)
L . Unit Conditions: h e test was performed on Unit I with the unit in Mode 5. ? [ Results Summery:
- After' the blowdowns, there was no evidence of particulates or- gross contamination in the 1A main headers and branch lines.
Safety Evaluation Summary: [ 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated i in the safety' analysis report was not increased. Loss of all IA has !- 'been previously evaluated with acceptable results. A loss of IA will be limited to a small portion of the system. The IA system was available during blowdown of the system because system pressure was l maintained.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. A loss of IA has been previously evaluated with acceptable I results. The loss of IA will be limited to a small portion of the h system. Maintaining system pressure ensured the IA system was available during blowdown of the system.
- 3. The margin of safety as described in the bases of any technical l specification was not reduced. %e ability of the IA system to supply equipment was not compromised during the test.
agg - 4 7 W' -..
.E' 11-87-83: Initial Startup, Qaeckout, and Head Ourve Verification.of 1-FW-P-2 ,
rr ,
Purpose:
To perform a post-maintenance checkout ~and a pump head curve -verification for ! 1 -IV-P-2 and.'to verify pump response time and the operability of specific j
. system valves.'
% , : Dates j
,h e: test was performed on 07/14.
l Unit Conditions: !
,, % e test was performed on Unit I with the unit in Modes 1 and 4. I 4 - . .,
Results Summary: ., The, test produced an acceptable pump head curve and verified post-maintenance pump- operability, satisfactory pump response time, and operability of specific .,
, . valves.-
+
Safety Evaluation Summary: - (. 1.. The probability of occurrence or the consequence of an accident or malfunction of equipment-important to safety and previously evaluated ; in the safety analysis report was noti increased. The auxiliary- i feedwater system was not modified or changed In' any. way. Administrative controls were in place and ensured the cross-tie is [ capable of being reclosed upon an auxi11ary'feedwater systom signal. ; The' test lineup-and procedure are bounded by the existing accident
-analyses on loss.of all AC, loss of Feedwater and Main Feedwater Line f
, break..
- 2. The possibility for an' accident or malfunction of a different type -
e k: , than was previously evaluated in the safety analysis report was not , created.. No system modifications were made during testing. -i
- 3. - The margin of safety as described in the bases of any technical specification was not reduced. Technical Specification Action Statements were adhered to during testing.
1 f. r 1
- f. - (
y 1-ST-84: Initial Stattup, Checkout, and Head Ourve Yerifleation of 1-FW-P-3A ). !l
Purpose:
To perform a post-maintenance checkout and a pump head curve verification for 1 -IV-P- 3A. , i F Dates: The test was performed from 07/06 to 07/07. i [ Unit Conditions: t { The test was performed on Unit I with the unit in Mode 4. r Results Summary: The test produced an acceptable pump head curve and verified post-maintenance ! pump operability.
. Safety' Evaluation Summary:
s.
- 1. The' probability of occurrence or the consequence of an accident or malfunction .of equipment important to safety and previously evaluated ,
t in the_ safety analysis report was not increased. No modifications ,[ were performed on the pumps. The test only runs the pumps after they f were rebuilt to ensure they meet their design function. ;
- 2. The possibility for an accident or malfunction of a different type
- than was previously evaluated in the safety analysis report was not' created. No modifications were performed on the pumps. The test only ;
se runs the pumps after they were rebuilt to ensure they meet their design function.
- 3. The margin of safety as described in the bases of any technical specification was not reduced. The special test only validates that the pumps meet their design specifications.
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I s 1-5T-85: lett tal Stertup, meckout, and Head Curve Verifleation of 1-PW-P-3B m Purposes-
'To perform a. post-maintenance checkout and a pump head curve verification for
. p 1-IW P-3B .
i e Dates s
, s The= test was performed,on 07/07.-
Unit' Conditions: L i'
'The(test was performed on Unit I with the unit in hbde 4. !
t Results Summary: ,
- The. test. produced an acceptable pump head curve and ev'rified post-maintenance [
pump operability. (- Safety Evaluation-Summary: l t cl.: 'The. probability of occurrence or the consequence of. an' accident or- 3 P s malfunction :of. equipment important to safety and previously evaluated j l
.inithe safety. analysis report was not increased. No modifications [
were performed on the pumps. The test only rugs the pumps af ter' they ; were rebuilt to ensure they meet their design function, i
- 2. The possibility:for an accident or. malfunction of a different type than-was previously evaluated in the safety analysis' report-was not ;
, , created. No modifications were performed on the pumps. The test only
-runs- the pumps after they were rebuilt to ensure they meet'their ;
, . designz. function.
- w. .
3.. The margin of safety as described in the bases of any technical l "( specification was not reduced. 1he special test only validates ' that the pumps meet their design specifications. l t r 4 s
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,( . 3-ST-42i nead Ostee Verificat ion '- t-PW-P-2.
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~p-kam -Purpose:
m . J s To obtain a'd n evaluate pump had curve data for 2-!W-P-2.
+ .t ,
V V Date U- The test was perforded on'02/19.
. ,9 k p <
, .. Unit Conditionst' sThe test was performed on Unit 2 with the unit in Mode 1.
, V '
- Results Summary:
a
+
e , The' test resulted in the accumulation and evaluation of pump head data.
.m ,
i Safety Evaluation Summary:
; / /An evaluation for'a potenttal Unreviewed Safety Question was performed, and'the' 1following was' determined, b,
^
f , . .
- 1. - ~The probability of occurrence or the consequence of an accident or
@ malfunction'.of. equipment important to safety and previously evaluated in' the. safety analysis report was not increased. The test proceddre !
n, , ensures only two pumps and flow paths are affected. 'At'least one pump' ! [, is available for ' discharge'to the steam generr. tors Administrative l
' controls ensured the cross-tie valve would be immediately closedL'if .
q required. ' Testing was performed within the bounds of the accident analysis. , f 2. The possibility for an accident or malfunction of a different l type -; 3' . .. . than ,was previously evaluated in the safety analysis report was not: m ; p$- created. 'The system was not-modified during testing. The system .was ~i hg 3
, . capable of performing its design function. !
[7 13. The margin of safety' as described in' the bases of any technical ,t m, , 1 gf . specification was not reduced. 'Ihe auxiliary feedwater system ;;a s ! capable'of-performing its design function during testing. {' .; y . 33 i f e f: ._ y , ' s.
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J 2-ST-76:'lBlowdownoffInsirumentAirSystemMain'HeadersandBranchLines, t . it n -i. .
Purpose:
J , i'Ib' remove contaminants from the instrument- air .(IA) main- headers and ' branch
, 'Llines.
W : Da tes: - .
- m. '
- The test was' performed f rorf 06/06 to' 07/05. .c l- " JUnit' Conditions:L e
v iThe test was performed on~ Unit 2 with'the unit in Mode 1. &L.l
? ,. ..
.Results Summary:
x' - After the: blowdowns, .there -was no' evidence of .particulates or gross 4
. contamination--in.the IA main headers and branch 1ines.
Safety' Evaluation Summary:'
.s
- 1. The probabill.ty.of occurrence or the consequence of an~ acc
- nt or-malfunction. - of equipment important to safety. and previously y. .uated
-in cthe3 safety analysis' report.was not. increased. Loss of,all IA has
'beenVpreviously evaluated with acceptable results. A loss of'IA will-bestimited to a small portion of 'the .systen. -The I A' system was.
1' avallable. during blowdown of the! system because system pressure was .1 maintained.
~ 2;t The possibility for an accident or malfunct'on of a different type than was previously evaluated in the safety analysis report was not-created.' A= toss of IA has been previously evaluated with -acceptable results. ~The loss of IA wiIl be limited to a small portion of the system. Maintaining system pressure ensured the IA system was available during blowdown of the system.
M y 3. The margin of safety as . described in the bases of any technical' 4l specification was not reduced. The ability of the IA system to supply equipment was not compromised during the test. i
) 4 i
( 'i
L (; h y+ O b 2-ST-77:' Verification ~of-l Fall' Safe Position' of-Safe Shutdown Air Operated t
-t- -Valves .
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Purpose:
p . . . f_ _ To verify air operated valves required for safe shutdown fall to their fall P safe positions upon-E of' air. O. . , ' Datest.
,p " ;
['< The test was performed from 04/17 to 05/03. p[ : Unit Conditions: ; n W^ n The test was performed on Unit 2 with the unit in Modes 5 and 6. l
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Results Summary: The,= test ' verified the valves failed to their fail safe positions upon loss of L . air. Safety Evaluation Summary:- g n J -1. The probability of occurrence or the consequence of an accident or
,m _
malfunction of equipment important to safety and previously evaluated in'the safety analysis report was not. increased.. Boration';and
,.v 4
' inventory addition-> flowpaths remained operable during testing. 'Ihe
+ unmonitored. release of a waste gas: decay tank'is an analyzed . event. ,
, < 'lhe valves. and systems are designed for-the conditlons that were
.present during testing,
~
y 2. The possibility for an accident or malfunction of a different type t than was: previously evaluated in the safety analysis report was not created. 'The valves are designed to be cycled and designed for the conditions that were present'during testing. 3., -The margin of safety as described in' the bases of any technical
! specification was not reduced. A boration flow path remains operable at ~all times. A waste gas decay tank release will not occur due to r normally. closed isolation valves.
[.- 11 N--' W
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* ! ' 2-ST-78 ,- Nitrogen Reserve Tank Capacity Verification. (Pr.r IMVs) h ,
V , , . 1urpose:' 5 ,.e To evaluate the pressurizer IG Vs' nitrogen usage and to verify.the adequacy of r the IGV reserve tank. .
~
Datest' K. W The"', butperformed'on05/01.
~
- n it~ Conditions:
- 'O
< The tes't was performed on Uni ( 2 with the unit in Modes 5 and 6.
Q Results~ Summary:
.The' test verified the adequacy of the IGV~ reserve tank. s Safety Evaluation Summary:.
~
The' probability of occurrence or the consequence of an accident or
~
1.~
. malfunction of equipment important to safety and previously evaluated ,
e in the safety analysis report was not increased.. The nitrogen' reserve
- tanks'and associated PORVs were isolated;and tested Individually to collect data for verification of tank capacity. hhen applicable,.the ;
limiting conditior. for operation was complied with or the appropriate o , action statement was entered. The overpressure protection system was s
- not affeeted during testing.
- 2. The possibility for an accident or malfunction of a different type 7
.than was previously evaluated in the safety analysis report was not created. The nitrogen reserve tanks and associated PmVs were isolated and tested individually to collect data for verification of tank capacity. hhen applicable, the limiting condition for operation
- . was complied with or. the appropriate action statement was entered.
The overpressure protection system was not affected during testing. 3.. The margin of safety as described in the bases of any technical I spectfIcation was not reduced. Overpressure protection was provided when required during performance of the test. f 4 . 3
, =- - - -
it i y r [ '2-ST-79: Verification of Adequate, Back Ups Air . Supply for Safeguards Area L , Exhaust Dampers - (,
Purpose:
-To: .. verify. the: adequacy- of.~ the Safeguards Area exhaust dampers back-up: air
. supply.
Da't es: The-test was performed on'04/29.
. Unit Conditions -
2 TheT t es t was per formed on Uni t 2 wi t h t he un! t i n f.bde 5. s - Results Summary:
'lhe' test: verified the adequacy of the Safeguards Area exhaust _ dampers back-up q-air supply. ,
p- Safety Evaluation Summary: l '. - The probability' of occurrence or the' consequence of an accident or
> malfunction ~of equipment important to safety and previously . evaluated E <
in the' safety _ analysis report was-not increased. The dampers and fans. are designed to be cycled. The accumulator is designed to cycle the dampers. Equipment' could be placed in service immediately if required. 2.- The possibility for an accident or malfunction of a different type t.
.than was previously evaluated in the_ safety analysis report was not created. Operability _of the safeguards exhaust fan and exhaust damper ;
were 'not adversely affected during' testing. The dampers and fans are designed _o t be cycled.-
- 3. 'The margin'of safetyLas described in the bases et any technical -
specification was not reduced. The safeguards area vent lation system l was capable of providing ventilation as described in tha bases of the technical specifications. I
Q & N , { ' i s2-ST-813- Initial-Startup, Q eckout','and Head Ourve YerifIcatIon of 2-PW-P-2 7 ,s ,
', Y f,n-
Purpose:
._o L :To perform a post-maintenance checkout and a pump head curve verification -for 4 2 -IM-P- 2 . :
- Datess.
1The testLwas' performed on 05/06.
~
c n
, 0 Unit Cbnditions: '
The test was performed on Unit 2 with the unit in Modes 3 and 4.
. . j Results Summary:
The , test produced an acceptable pump head curve and verified post-maintenance pump operability,-satisfactory pump response time, and operability of specific -
, valves.
Safety Evaluat' ion Summary: , l -1. The ' probability 'of occurrence- or the consequence of an accident ~or -; E malfunction of. equipment important to safety and previously evaluated 6 -- in the safety: analysis report 'was not increased. The auxiliary Of 'feedwater system was' not modified or chenged in any. way. $ Administrative controls; were in place and ensured the cross-tie is k
. capable of.being reclosed upon an auxiliary feedwater system signal.
The test lineup and . procedure are bounded by the existing; accident 0 analyses on loss of all AC, loss of Feedwater and Main Feedwater Line - P break. . [ '2. The possibility for an accident or malfunction of'a different type J ' than was previously evaluated in the safety analysis report was. not
- g. . created. No system modifications were made during testing.
^ g 3. The margin of safety as described in the bases of any technical specification was not reduced. Technical Specification Actton Statements were adhered to during testing. 1 4 lf L 3
e 3' e.M b _s B, p, 1-ST-86:l Charging.PumpSuctionHeader-NjInjectionTest. i e
- d. -!
J u Purposes; L, - - 4 h ' To determine 'the: minimum, liquid level from gas' accumulation-in-the charging' < p - ( , pump normal suction header. ; t - U Date: f7 The test was performed-on'06/24. O E , ..
, Unit Conditions:'
The test wasiperformed on Unit I with the unit in Mode 5.. i n j '
.Results Summary:- l gy
'lhe results of the test were inconclusive due 'to unreliableLultrasonic. testing t
readings when the header, indicated less than-50% level. 1.g. . Safety Evaluation Summary:- 9, 5
- 1. The probability of occurrence or the' consequence of an accident or-o
~ malfunction of equipmen' important.toisafety-and previously evaluated in'_the safety analysis report was.not increased. The injection of nitrogen:in'the charging system will not cause equipment or ~ component m .
damage. < Piping;was vented before the' charging pumps.were started. No.
. positive. reactivity addltions;were ..made- during -the test. The Technical Specification Action- Statement- for no operable charging .
4 pumps was complied with if required. A low head safety injection pump was available for makeup to the reactor coolant system if required. 2.-- The possibility for an accident or' malfunction of a different type- ; than was previously evaluated in the safety analysis report was not a created. The injection of-nitrogen in the charging system will not cause equipment or component damage. Piping was vented before the
< charging pumps were started. No positive reactivity additions were 1
o; , made during the test. The Technical. Specification Action Statement ] for no operable charging pumps was complied with if required. A low ] head safety injection pump was available for makeup to the reactor coolant system if required. j V' l 1 t i N' ;y
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i L1-ST-87: marging Pump Gas Accumulation During Normal Operation. r , f [,'
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Purposes y ,V H Tol evaluate gas accumulatlonLin the charging pump' normal-suction header during gJ normal'_ operating conditions, n.- fi s ' ' . Utr , O' ' Dates:
- The test.was porformed-from 08/02 to 08/03.
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-Unit Conditions:
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'.The' test was performed'on Unit 1 with the. unit'in Mode 1.
4 LResults Summary: n l~'A .The test demonstrated that during a 10 hour period, the water level 'in- the < . horizontal . charging pump suction header reduces to approximately 67% of the I b -
.w ater height when full.
1 ll) , Safety Evaluation Summary: _ y.
^
l
- 1. The' probability of occurrence or the consequence of an accident o r. H H malfunction of: equipment important to safety and proviously evaluated
- in the safety.. analysis.. report was not increased. .The test -was .1
- s F 4, performed to obt in data and did not affect normal unit operations, - j J > -2.. 'The.- possibility for an accident or malfunction of a different-type- ! than was previously evaluated in the safety analysis report was not' J ' {- , created. The test was performed to obtain data and did not affect M' normal' unit operations, j l
'3. The margin of' safety as described in the bases of any technical r
- specification was not reduced. The charging system and the unit were i not affected by this test. Swapping of charging pumps is a normal y operational evolution, j i y' .;
<Y i t' 4%
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'1-ST-87a } Different!ali Pressure Test of Service Water Bypaes IOVs -
s !? 4 6
Purpose:
R :To measure the torque requirements for the ' Service -Water Bypass h0Vs with
- r P simulated design basis. differential pressure / flow conditions. ,
- Dates:-
^
L % e' test'was performed from 11/21 to 11/22. Unit. Conditions: The tect was performed on both units with the units in hbde 1. , Results Summary: The test verifiedL the valves have sufficient torque available to operate-at design-basis con'ditions. I Safety Evaluation Summary: 1;- .The probability of occurrence or the consequence of an accident or malfunction of equipment-important to safety and previously evaluated
-in the safety analysis report was not increased. The test verifies
~
~the' ability =of,the service water system to operate as designed. The l
test does not' introduce additional failure modes. Components are designed to operate under design conditions. -An- operator was stationed at the. service water valve house to -manually reposition, valves,- if required, to prevent pump damage. Sufficient cooling capacity for safety related equipment is maintained.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not
> '~ created. The test lineup was designed to simulate design basis conditions. Precautions were established to prevent equipment aamage
-in the event of a component failure.
3 .- The margin of safety as described in the bases of any technical specification was not reduced. The test did not affect operability of the service water system. Service water loads have been realigned to
. prevent plant transients. Sufficient cooling capacity for safety related equipment was maintained.
940W ; n <m n- 1
/ . i ll .
y- " s1-STi B8:' LnGain Cbntrol and Relay Room Air. Co'ndit toning _ inittal Checkout - [9., ,
~
pt , L 4 Purposes
* , .To' provide reference performance data for the system.
( Datess' , 4: h4
.The test was/ performed from 12/27.to 12/29.
e p , Unit Conditions: l Thettest'was performed on Unit 1 with the unit in Mode 1. 3 h, g Results Summary: The. test'provided valve stroke times and pump performance data for the system. [ l Safefy Evaluation Summary: , G
. l. The' probability of. occurrence or the consequence of an accident or r ' malfunction of equipment important to safety and previously evaluated. ;
in the safety. analysis report'was not increased. The test did not I y affect the primary or secondary plant.or any power supplies. Control-and rack room. temperature; limits were ' observed.. -All applicable
- 1imiting conditionsLfor. operation were observed. ;
- 2. The. possibility for an accident ~or malfunction of a.different type than was previously evaluated in the safety analysis report was not created. The UFSAR evaluates; the' possibility of IIVAC maintenance' y, , problems.- The test did not challenge the llVAC chillers. Equipment was operated;within manufacturer's limits.
c~
. 3.. The margin .of safety as described in the bases of any technical N specification was not reduced. The Technica1 Specifications were u
adhered 'to during the test. The test did not challenge control and rack room temperature limits. 4-t '5 O s
l j. , ' , g 2-ST-823,InitialEStartup, Checkout, and Head Curve Verification of 2-FW-P-3B 3 Purposes
;. . To performi a post-maintenance checkout:and a' pump head curve verification-for:
7 2-RV-P- 3B. x s Datest.
- The. test - wa's per formed on - 05/01.-
' Unit' Conditions:1
? l The test was performed on-Unit 2 with the unit in hbde 4.
Results Summary:- 1The test produced an acceptable pump head curve-and verified post-maintenance _ pump operability. Safety Evaluation Summary:
- 1. The;-probability of occurrence or the consequence of an accident or malfunction of, equipment important to' safety,and previously evaluated-Ein=;the- safety analysis report was not increased. 'No modifications.
were performed on the pumps. The. test only runs the pumps af ter they
.were rebuilt to ensure they meet their design. function.-
- 2. The -possibility- for an ' accident or malfunction of'a1different type-
-than was previously evaluated in the safety analysis report' was not created. No modifications were performed on the pumps. The test only runs the pumps after they were rebuilt' to ensure .they meet their design' function.
- 3. The margin of safety as ' described' in the-bases of any technical
- o specification was not reduced. The special test only validates that the pumps meet their design specifications, s
E
- $'e
e c ' lh ~. ,.; p, '2-ST-83: Inittal Startupi. m eckout; and Head Ourve Verification'ofl2-PW-P-3A y lv
Purpose:
To:~ perform a post-maintenance checkout atid a pump head curve verification for 2-IWP.-35. / ' 1
= Dates ~
'Ihe. test. was per formed on 05/01..
, >UnitcConditions:
The: test.was performed on Unit 2 with the unit in Mode
^
4.- Results Summary: The test produced an acceptable pump head curve and verified post-maintenance ipump operabi1ity. n Safety Evaluation-Summary: e 3
.q 1., The probability of occurrence or the consequence-of an accident or j
= malfunction of. equipment important to safety and previously evaluated ini~the safety' analysis report was not increased. No modifications n were performed on-t'e h pumps. .The test only runs the pumps after they~
..)
.were rebuilt to ensure they meet:their. design function'.-
2.- :The= possibility lfor an accident or malfdnetion of a different-type than was~previously evaluated in the safety analysis report ~was not
. created.. No modifications were performed on the pumps. The test only. ..
- runs: the pumps af ter they were rebuilt to ensure they meet their .jl
'm -design. function. ;
- 3. The ' margin of -safety as described in the bases of any technical-specification was not reduced. 'Ihe special test only validates that i the pumps meet their design specifications.
m ,
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$1-ST-84: ' Main Control'and Relay Room Alh Oonditioning; Initial Checkout 5
, Pu r po s_e_: - ,
j'7 ;To lrovidei reference performance data for the system.-
. Dates::
The test was' performed.from 12/27-to 12/29.
,s h
- Uni t 'Co'ndi t ions :
The test was; performed on Unit 2 with the unit in Mode 1. Results Sumary:
The test'provided valve stroke times and ~ pump performance data for'.'the system.
.;> ~ Safety Evaluation Summary:
-1.. 'Ihe ? - p robabi li ty of occurrence or the consequence of an-accident or malfunction of equipment important to safety'and previously evaluated'
.i n : the safety analysis report was not increased. The test did not affect the primary or. secondary plant or any power supplies. Control
' 'and rack room- temperature limits were observed. All applicable limiting conditions for operation.were observed.
- 2. - The possibility for an accident or malfunction of a different type
'than was previously evaluated in the safety analysis-report was not created. 'The UFSAR evaluates the. possibility. of hVAC maintenance
~ problems. The test did not challenge the llVAC chillers. Equipment
~was operated within. manufacturer's limits.
- 3. >The margin of~ safety as. described in the bases of any technical
' specification was not reduced. The Technical Specifications were adhered to during the test. The test did not challenge control and h.' rack room temperature limits.
I fb.t N.
- rpn - + =,
rm h v IgnrW1RE DEVIATIGIS RBQUIRI!C A SAFEIY EVAIARTION I1R2NG 1989 e r f* b 1-OP-SA:- Valve Checkof f - Reactor Coolant p
Purpose:
The :, procedure. deviat ion was required to enhance pressurizer control due to spray valve leakby.- ~ t 4 Safety Evaluation Summary:. l'._ ~The probability of. occurrence or the consequence of an accident or malfunction of equipment'important to safety and previously evaluated in the safety analysis report was not increased. No changes were made to the Chemical- Volume and Control System. Pressurizer' spray valve- k 1leakby provides the same effeet as opening the manual bypass valves. Alarm conditions- for thermal stratification remain operable. All systems. required to r.itigate an accident are unaffected by the valve-lineup change. 2.- The possibillty -for an accident or malfunction'of.a different type. .
< than was previously evaluated in the safety analysis report was' not I created. .The only change-is the valve lineup. Equipment operation is i
.l
- . not altered.
3.: ;The: margin.of safety as, described in the bases of any technical: specification -was not reduced. The integrity of the RCS piping is j l maintained due to existing surveillance of the pressurizer spray line temperature. ! i f l 1 4
)
1 q j
M*Y n h, ' i Process Vent Particulate Radiation Monitor ICP-CW-1-HM-1012 [
- ,r g
Purposes: g . . The procedure deviation was. required to limit the amount of[ time containment - vacuum pumps-are tripped off while performance of~the= instrument calibration.= a P l Safety Evaluation Summary:
'l -?
[ 1. .The' probability of occurrence or the consequence of- an accident or malfunction of' equipment important to safety and previously evaluated l'
'in the safety analysis report was not increased- Additional radiation t
monitors remained operable on the process vent to detect increased , radiation levels and. initiate auto-closure of affected valves..
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. Additional radiation monitors remained: operable on . the l
_ process vent to detect increased radiation- levels and initiate Lauto-closure of affected valves. 3.. 'The: margin of safety.as described in the bases o f. any technical -;
. specification was not reduced. Removal of the radiation monitor from service for calibration is an acceptable practice. Additional radiation monitors provided auto-closure function to maintain the required isolation capability.
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t
' C: .
3.pp.71,3 g : Auxillary Feedwater Pump'(1-FW-P-3B)
Purpose:
' The procedure d,eviation was required to verify 'the flow limiting capability of
~
the auxiliary.feedwateripump (1-FW-P-3B)-recirculation flow orifice. Safety Evaluation Summary:
- 1. .The pro'bability'.of 1 occurrence or the consequence:of an accident or-
. malfunction of equipment important to safety and previously ' evaluated in the safety analysis report was not increased. The pump was capable
, of performing its intended function 'i.f required. Sko additional auxiliary,feedwater pumps were also available to provide cooling.
- 2. The- p'ossibility for 'an accident or malfunction of a different type
- than wasJpreviously' evaluated'in the safety analysis report.'was not 'f created. The pump was capable of performing its intended function if' required. TWo additional auxiliary feedwater pumps. were- also
"" b; available to provide cooling.
- 3. .The margin of safety as . described .in the baseslof any technical i
. specification was not reduced. Technical Specifications requirements-T were followed at all times during the--test.
- t
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J> w 1-OP-8.7: Boric Acid' Batching and Transfer System 0 t -- i- Purposes" , p The ' procedure deviation'was required to' clear unnecessary alarms generated by-equipment or systems not required to be operable and that place an additional burden for the operator to distinguish between insignificant and important alarms.- b Safety Evaluation Summary: 5
- 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important'to safety and previously evaluated-in the safety analysis report was not increased. The design function and operation of the boric acid storage system was unaffected with the
'B' boric acid storage tank was not inservice.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. The design function and operation of the boric acid storage-system was unaffected with the 'B' boric acid storage tank was not
' inservice.
i
- 3. The margin of safety as. described in the bases of any technical i specification was not reduced. The design function of the boric acid storage system was unchanged as a result of this procedure devintion.
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- IW-C-PROC-09:- RepIacement 'of a Narrow Range Reactor 0001 ant - Syatem If!D-p.
g
Purpose:
The procedure deviation was required to delete steps to allow replacement of bypass spanifold- R1Ds without isolating the affected R1D bypass manifolds with the Unit'in Mode 5'and pressurizer level and pressurizer relief tank pressure g - reduced to minimize water pressure at the.R1D penetration.- [ f{/
~
Safety Evaluation Summa.ry: i
- 1. The. probability o f- occurrence or the consequence of an= accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. Contingency actions-were in place to ensure that core cooling and shutdown margin were maintained in the event of excess leakage from the R'lD opening. The--
contingency actions ensured RHR operation was maintained.
- 2. The possibility for an accident or malfunction of a different type than was previously_ evaluated in the safety analysis report was not L created. Contingency actions were in place to ensure that core cooling- and shutdown margin were maintained in the event of excess leakage f rom the R'lV opening. .The contingency actions ensure RIIR
-operation was maintained.
- 3. The margin = of safety as described in the bases of any technical specification was not reduced. Applicable 1imiting conditions of
> operation and action statements were complied with during RTD 9, replacement.
IE i' s . 1 i Y I $ ,
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k . . ? . 2-PT-34. 5 : - .W rbine-Generator Overspeed Trip Test [ P[, , p p '
Purpose:
I .The-procedure deviation was required. to delete the requirement 'to. perform
' mechanical overspeed trip testing of the turbine generator for the surveillance
-interval.
i' Safety.Evaluat' ion Summary:
- 1. The probabi1ity.of occurrence or the consequence of an, aceident or
' malfunction of equipment important to safety and-previously evaluated in.the safety analysis report was not increased. A turbine trip' is evaluated. in UFSAR Section 15'.2.7. Deferring the mechanical overspeed
. trip. test does not make a. turbine: trip more likely. The overspeed protectionL' system is _ not affected. The ability of the reactor protection system to. sense a turbine trip is unaffected. The electrical overspeed protection instrumentation is fully operable.
- 2. The possibility.-for .an accident or malfunction'of a different. type thna was previously evaluated in the safety analysis report was not !
- created.- Main turbine operation and trip protection is unchanged by delaying the surveillance in,tervel..
- 3. The margin of safety as described -in the bases of any technical ~
spectfication was not reduced. The Technical Specificatlon, required. overspeed protection system is fully operable.
.]
.I 1
- -7 _
, w b ,
IHP-7.1.22: . Storage lof, Radioactive Material Outside the Protected Area.
, i ,
- l. 4 z ,.;
~ Purposes
' I'? The procedure deviation was required to allow the ' storage of a contaminated- .;
a ~
' reactor coolant pump motor in Warehouse 5. .
.. i Safety Evaluation Summary:-
- 1. The, probability of occurrence or the consequence of.an' accident or
' malfunction of equipment important to safety and previously' evaluated ,
in the safety analysis report was not increased. All systems required =r for accident mitigation were unaffected. Safety related equipment was not involved in the shipping and storage process. .;
, ) -2. The possibility for an accident _or malfunction of a different type .
l0 than was previously evaluated in the safety analysis report was not~ created.- The radiological hazards involved in the shipment and t l
. storage process are bounded by offsite dose calculations for design.
basis accidents. ,
- 3. The margin of safety as described' in the bases of any technical ;
specification was not reduced. The radiological hazards involved ~in ,
, the shipment. and storage process' are bounded by .offsite. dose ; ; calculations for design basis accidents.
g 5
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r qsj ' b . L {: 1-OP-4.5: , Operation of the New Fuel Storage Elevator i t [ -Purposes The.-procedure-devidtion was required to add an initial condition to allow' hi
> -transport.of new fuel:'from the spent fuel pool to the new fuel storage area.-
Safety? Evaluation Summary: 4 y ,. 1. : The probability of-occurrence or the consequence of an accident' or - malfunction of equipment important to safety and previously evaluated in.the safety analysis report was not-increased.' Operation of the new i fuel: elevator in -this manner--has no effect on previously analyzed accidents.: The probability of dropping a fuel assembly is not increased since movement of the fuel assembly to the elevator'is no-different:than described. A fuel handling accident has been described
- 1 in UFSAR Section 15.4.5. The evaluation bounds the movement of n'cw fuel.
~
- 2. The possibility for an accident or malfunction of a different type t q t
.than- was previously evaluated in the safety. analysis report was not created. No new accidents were created by the change. .1 i
- 3. LThe: margin of safety'as d? scribed in the bases of any technical
-.1, specification.was not reduced. The Technical Specification bases were not affected by the change.
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'> . 2-PI'-34; 3: . W rbine Valye Freedom Test
.i
^-
^
Purposes-(f '
% The procedure deviation was' required to delete the. requirement to stroke the (4 . governor.. valves .during turbine valve freedom testing. -This was to prevent turbine damage due.to reduced load during unit coastdown.
LSafety Evaluation Summary: 3-
'1. The probability of occurrence or the consequence of an accident or
& > malfunction'.of equipment important to safety and previously evaluated-
'in the safety analysis report was not increased. Adequate testing is
~
performed to ensure turbine trip capability is maintained.: Turbine
- trip capability via the throttle valves was tested to verify 1 stroke ,
and input to the reactor protection system. %e turbine manufacturer
. . recommended the test revision.
- 2. _ .The possibility for an accident or malfunction of a different type-than. was previously evaluated in the safety analysis, report was not created. . Adequate testing is performed to ensure turbine. trip j
. capability is maintained. Turbine trip capability via-the throttle l
valves was tested to verl.fy stroke and. input to the reactor protection { system.-The turbine manufacturer recommended the test revision, j
'l
- 3. The margin - of safety as described in the. bases of any technical l .
. specification was not reduced. Structural integrity of the turbine ;
was maintained and the throttle valve trip was verified. i
'l 1
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1 s D1 -OP-62.5 , . Flushing E SS-124 With Primary Grade Water:
Purpose:
The purpose of this' procedure deviation was to add steps to flush the Steam Generator Blowdown Radiation Monitors.
' Safety Evaluation Summary:
.l. The probability of occurrence or the consequence of' an accident or malfunction of equipment important to safety and previously-evaluated in the safety analysis report was not increased.. The blowdown radiation : monitors are designed to be flushed. The steam generators were continuously monitored for. radioactivity-using an N radiation.
. monitor during flushing operations. A steam generator tube leak would be detected while flushing the blowdown radiation monitors, m 2. The possibility for an accident or malfunction of a different type than was- previously evaluated in the safety analysis report was not-created. The blowdown radiation monitors are designed'to be flushed.
The _stcam generators were continuousiy monitored for radioactivity _s- using:an N-16 radiation monitor during flushing operations. A steam
-generator tube leak would be detected while flushing the blowdown radiation monitors.
- 3. The margin of safety as described in the bases of any technical specification was not reduced. An N-16 radiation monitor was operable during flushing of the blowdown radiation monitors. The blowdown radiation monitors are not required by Technical Spectfications.
, .m
pn r- . . gc' Ib . ; .- u - K M@-C-W-1.1 :1 Making New or Replacement Socket and' Fillet Welds on Pressure. . l . Retaining Piping, Yalves, Pumps and Vessels h Purpose s-The procedure deviation was required to revise the size-requirement for fillet 't welds.from 1.25t to'1.09t'as:specified in the Corporate Welding Manual. -i , , g Safety Evaluation Summary
- 1. The probability of occurrence or the consequence o f- an accident or -
. malfunction' of equipment important to safety and previously evaluated -
.in the safety analysis report was not increased. The change is in
-accordance: with AShE Seetion III and XI., Seismic qualification of-safety related systems is maintained.
2 .' The possibility for an accident or malfunction of a different type t than- was previously evaluated in the safety analysis report was not created. 'Ihe single active failure of _ a pipe are analyzed events. Compliance -with the ASME Code does not increase the probability of an , accident. 3.- The margin:of safety as described in the bases of any technical
' specification was not reduced. System integrity is being maintained
-with:the use of the AShE Code requirements.
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i L 190CHERIE RIQUIRING A'SAFEIY'EVALUATIOi DLRING 1989 1 e 5 i .. - Recirculation of. Refueling Water Storage . Tank Through Portable-2 '10P-7.11: Demineralizers
.i
Purpose:
4- ..The. ' procedure ;was developed to allow the installation of 'a temporary
'demineralizer rig on the refueling water storage tank (RWST) recirculation. -
= piping- to' remove sodium hydroxide- that was inadvertently injected into the JRWST.- ,
Safety Evaluation Summary
- 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. Operation of the-IBWST was not changed. Volume and temperature requirements were met.
Seismic' qualification on RWST recirculation piping was maintained.
.2. The possibility for an accident or malfunction of a 'different type than was previously evaluated in the safety analysis report was not created.- Operation of the RWST was not changed. Volume and temperature requirements were met. Seismic qualification on RWST recirculation piping was maintained.
- 3. The margin of safety as described in the bases of any- technical specification was not reduced. The RWST was capable of performing its intended function during this change. IB25T temperature, volume and
, boron concentration were maintained at all times.
W J!: 1
Yh a p i ~ 1-NDP-49.08: Removing.#1' Supply-- (A) and #4 Return (A) From Service and Returning to Service
~ Purposes.
. 'The- procedure was required to allow removal of the service water supply;and return headers from service to determine the rate of leakage from the header by.
-filling-the headers'and pressurizing with a hose equipped with a rotemeter.
b; Safety Evaluation Summary:
- 1. . The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. A leak check was ,
performed at a pressure below the design limit of -the system with overpressure protection provided. ,
- 2. The.-possibility- for an' accident or malfunction of a different type.
-than was previously' evaluated in the safety' analysis report was not created. The failure of a service water header is not a design basis accident. The redundant service water header will adequately provide ultimate heat sink capability.
- 3. The margin of safety as described in the bases of any technical !
specification was not reduced. Sufficient cooling capacity was available from the redundant service water header. e
) ci 2-PT-171.1 _ 10 , Year Hydrostatic / Pneumatic Test for the Low Head Safety.
^
Injection Pump Discharge Flow Paths
Purpose:
'Ihe procedure was required to hydrostatically test each of the low head safety
~
injection-(LilSI) pump discharge flow paths to the ICS hot and cold legs by - isolating the flow path while in Mode 3._ The discharge flow paths to the cold
- legs were _ tested simultaneously with hDV-2890C and D closed.
, Safety Evaluation Summary:
- 1. The probabillty of ' occurrence or th- consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The piping and test equipment was designed to tolerate the hydrostatic test pressure.
Overpressurization potential was minimal due to a relief valve that was installed on the test pump. Only one LilSi to hot leg flow path was tested at a time. Dedicated operators were stationed to open hDV-2890C and D if required while testing of the LilSI to cold leg-flow paths were tested.
- 2. The possibility for an accident or malfunction of a different-type than was previously evaluated in the safety analysis report was not created. Leakage from hydrostatic tested was capable of being.
Isolated and therefore minimized. The RCS is protected by check valves in the injection lines. 3.- The margin. of safety as described in the bases of any technical specification was not reduced. Appropriate Action Statements were entered during testing. 4
byM 9t W .lU l> s 9{}7.. - t h:. , ' $c ? ; 1-10P-33.1: iPurging;'C' S/G to the PRT . O
- l v.
f' D , %)w ~[Nrpose: , b The; procedure was 'to allow the-installation'of a sample / vent rig on the'"C" E! h Steam Generator' flow or level transmitter.and vent to the pressurizer relief g :. ? tunk. ?' p I ); L'? Safety Evaluation Summary: i t 1 '. -- .The probability of -occurrence or the consequence 1of an accident or-g malfunction of equipment important to safety and previously evaluated , g! In the safety analysis report was not increased. Equipment important to safety is not affected. The design function of the pressurizer ; relief. tank was not affected.
- 2. The possibility 'for an -accident or malfunction of a different1 type than was previously evaluated in the safety analysis . report was- not ,
< created. The. design function -of the system is not affected and a radioactive release is unlikely.
o . 0 .' 3.- The margin ofLsafety as described in the bases of any ~ technical , specification was not- reduced. The Technical. Spucifications are b unaffected by this change. - L h o . a ., r s O 's N k'o i-x ? k-' h k (.
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V 1-OP-4.19: . Receipt and liandling of New Burnable Poison Hod Assemblies
Purpose:
The procedure was required to detail the guidelines for. receipt and handling of new burnable poison rod assemblies (BPRA). Safety Evaluation Summary: 1.. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the. safety analysis report was not increased. The accident analysis for the new fuel storage area and spent fuel pool conservatively bound the procedure. The probability of an inadvertent criticality was not created by the change.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. Current accident analysis were bounding for this change.
- 3. The margin of safety as described in the bases of any technical specification was not reduced. 14o Technical Specification bases were associated with the components or procedure. Design features of the Technical Speelfications are satisfied by bounding accident analyses.
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ff ra -.e. ,: iiL ;c z , , Ibergislag Procedure for Circult' Breaker 152 and 152 (Food to "A" "B" RSST Free Bus #5) , lh
- a
Purpose:
- The procedure was required to check the relays for circuit breaker 152 and 252. L b 4 Safety Evaluation Summary:
.1. . The probability of ' occurrence or the consequence of an accident or
,f ,
malfunction _of equipment important to safety and previously evaluated In!;the. safety analysis report was not increased. The flexibility of
'the'offsite:to onsite' power . supply ensures' power is available to required -onsite buses. All breaker protective relaying was operable to protect'from'a fault condition. The .chenge does not alter the loss
.i3 of offsite power. analyses. No~new failure modes were created.
2.. The posribility. for an accident or malfunction of a different-type than was previously evGluated in the safety analysis report was not created. The loss of all offsite power was;previously analyzed. 1, ,
- 3. .The margin of . safety .as described in the bases of any technical
. specification was not reduced. Electrical buses remained unaffected-and remained operable to ensure Technical Specification compliance..
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. \ i-Procedure For Installing a New 34.5 KV Feed From aus #5 to "C" Reserve Station Service Transformer
Purpose:
'Ihe procedure was required '.o allow energizing and load checking of the new feeder breaker and protective relays. v Safety Evaluation Summary:
- 1. %e probability, of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The independence of the required offsite power supply network is not reduced. Only one of three offsite power distribution systems is affected. The change does not invalidate the loss of offsite power analysis.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. The independence of the required offsite power supply network is not reduced. Only one of three offsite power distribution systems is affected. The change' does not invalidate the loss of offsite power analysis.
- 3. The margin of safety as described in the bases of any technical specification was not reduced. %e change does not reduce the independence of the required offsite power supplies, therefore, the ability to maintain the plant in shutdown or refueling for an extended-period of time with sufficient instrumentation and control is not reduced.
i I
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I. 1-10P-4.12 3 Removal of Damaged HPRA 11P99.From Fuel Assembly WO4 i y i;
Purpose:
The procedure was required to allow' removal of a damaged burnable poison rod assembly (BPIM) f rom a fuel assembly. ,
?
r-Safety Evaluation Summary: , ( ' j 1. 'The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated increased. 5 in the safety analysis report was not The change has [ instructions for removal and storage of a damaged (BPRA). Damage or rupture of'the'BPRA will not create an adverse radiological accident. Storage in a cell without an adapter is temporary until the BPla can , be disposed of. Dropping and complete rupture of a fuel assembly in , the spent fuel pit has been analyzed with acceptable results.
- 2. The-- possibility for an accident or nialfunction of a dif ferent type j than was previously evaluated in the safety analysis report tvas not ;
[- created. No new accidents are introduced with this change. Rupture of a fuel assembly in the spent fuel pit has been analyzed with acceptable results. I
- 3. The margin of safety as described in the bases of any technical specification was not reduced. No Technical Specifications were affected by the change.
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m [.+ Procedure For Disconnecting Transformer #5 From 575 laine and Reconnecting It to ! New Location, Verify Proper Protective Relay Installation and Verify Proper Breaker Indication
Purpose:
The procedure was required for disconnecting transformer #5 from 575 line and reconnecting it to new location, verify proper protective relay installation ) .and verify proper breaker indication. Safety Evaluation Summary: > 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The independence of the required offsite power supply network is not reduced. Only one of t three offsite power distribution systems is affected. The change does not invalidate the loss of offsite power analysis.
- 2. The possibility for an accident or Jialfunction of a dif ferent type than was previously evaluated in the safety analysis report was not created. The independence of the required offsite power supply network is not reduced. Only one of three offsite power distribution systems is- affected. The change does not invalidate the loss of ,
offsite power analysis.
- 3. The. margin of safety as described in the bases of any technical specification was not reduced. The change does not reduce the independence of-the required offsite power supplies, therefore, the ability to maintain the plant in shutdown or refueling for an extended period of time with sufficient instrumentation and control is not reduced.
y I Procedure For Installing Feed From Hus #51b Construction Power And Rework Of 3 Existing Feed From Dus 64 I r-
Purpose:
The procedure was required to install the feed from Bus 45 to construction
- power and rework of existing feed from Bus #4.
1: Safety Evaluation Summary: l '. The probability ofl occurrence or the consequence of an accident or F malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The independence of F the required offsite power supply network is not reduced. Only one of three offsite power distribution systems is affected. The change does-not invalidate the loss of offsite power analysis.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. -The independence of the required offsito power supply network is not reduced. Only one of three offsite power distribution systems is affected. The change does not invalidate the loss of offsite power analysis.
- 3. The margin of safety as described in the bases of any technical-specification was not reduced. The change does not reduce the independence of the required offsite power supplies, therefore, the ability to maintain the plant in shutdown or refueling for an extended period of time with sufficient instrumentation and control is not reduced.
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- h :n lterk-G-1.6: Adjusting LPacking of Safety-Related Motor Operated Valves in Gomeral.
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' Purposes' .
I The procedure was required to allow deferment of post maintenance testing .on the "D" Safety injection' Accumulator discharge motor operated isolation valve q g ( 2-SI-MOV-2865D)..
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Safety Esaluation Summary: - 1
- 1. The probability of occurrence or the consequence of an accident .or
' malfunction. of. equipment'important to. safety and previously evaluated I s
L
- in the safety analysis report was not increased. Deferment- of post j g >
maintenance testing did not affect the safety function of the' ! accumulators.- MJV-2865B was de-energized open. On low RCS pressure, ,
. t the accumulator would still inject _because the two system check valves ,-
.would open. ]
<, 2.- The possibility for-an' accident or malfunction of a different type' .!
p' than. was. previour,1y evaluated in 'the safety analysis report ~was not created;. The accumulator can perform its intended safety. function. j h !. 3 The margin of safety.as described in the' bases of any technical y specification was' not-reduced. The accumulator remained operable to perform its intended safety function.
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g , 'l-1tP-4.3As EIC-7 Puel Bod Receipt and Storage (i:1 j L. l
,Purposet The procedure was' required Lto receive, inspect and store Westinghouse- -
' demonstration. fuel rods.-
t- - e Safety' Evaluation Summary - l
- 1. -- 1The probability of occurrence or the consequence. of an ' accident or 1 o 4 ; malfunction of equipment important to. safety and previously_ evaluated ,
L 3pf < .in the safety analysis report was not increased. The procedure Lis'
-bounded by the ~ fuel handling accident described in the UFSAR. No i equipment operation is being changed as a result of this procedure..
~
} , 3 "21 The possibility for an aceldent or malfunction of- a different ' type-5 than :was. previously evaluated in the safety analysis' report was not , t L' } , created. Existing fuel handling accident scenarios are bounding, j 13.' The margin of safety as described in the bases of any ' technical' i 1 1 specification .was not reduced.- The- design bases of the new fuel
~
storage area and cspeat fuel pool' are- maintained.- .No- Technical [ Specifications are directly affected. t i i a: '
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1 l IWF-C-MISC-05: Troubleshooting, Repair and Replacement of Non-Safety , Related Equipment Purpose The procedure was required to repair 2-TW-PCV-2488 air line leak by manually overriding the main feednater regulating valve at 100% power with the feedwater bypass valve open. Safety Evaluation Summary:
- 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. Feedwater flow and plant conditions were monitored by operators. _If a feodwater isolation signal was generated, moln feedwater regulating valve and fast acting. feedwater isolation valve will be closed.- The feedwater mass added by the' limiting accident would be below the mass assumed in the accident analysis.
- 2. The possibility for an accident or malfunction of a different' type than was previously evaluated in the< safety analysis report was not created. Overfeed would be controlled by closing the valve, hhin feedwater pumps and fast acting motor operator valve get a close signal on a safety injection and 111 -11i steam generator level, hbnually overriding a valve does not increase the chance of the valve failing open. Feedwater flow mass was within the design limit.
- 3. lhe margin. of safety as described in the bases of any technical specification was not reduced. The ability of instrumentation to monitor parameters and initiate a signal was not affected.
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W4MW-1.4:- Adjusting Packing of Safety-Related' Motor Operated Valves' in' l
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t > Pu rpose,i _ .
' '!hel procedure was required to allow deferment of post maintenance testing on' {
( 3 the charging. pump seal water injection isolation ~ (1-Ol-hDV-1370).- [ k. b .
, Safety Evaluation Summary: j - 'i 0 fl.- The probability of occurrence or the consequence o f. an accident: ' or .;
m
' malfunction'_of equipment important to safety and p'reviously evaluated .
p , in tho' safety' analysis report was not increased. Deferment of , post. , p maintenance testing didt not affect the operation of the seal water j
' injection system., The valve' remained in its normally open position.to ;
supply seal water to the reactor coolant pumps. There are no j protective functions associated with 1-Oi hDV-1370. j
- 2. The possibility for an accident or malfunction 'of a different; . type thanj was' previously evaluated in the safety analysis report.was not.' 'I
.- created. Operation ~of.the seal water injection l system was: unaffected .I by this change. l
- 3. The margin of safety as ' described in 'the bases of any. technical- ,
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specification was not reduced.' The Technical-Specifications are not. } e, ' affeeted by.this change. [ , b, i j' L-tx .
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WW-C-N-1. 6 : Mjust ing Packing of Safety-Related Motor Operated Valves in General Purposes' The procedure was required to allow adjustment of the packing on 1-SI-hDV-1865B and deferment of post maintenance testing. . Safety Evaluation Summary: r
- 1. The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. Deferment of post ,
maintenance testing did not affect the safety function of the : accumulators. h0V-1865B was de-energized open. On low RCS pressure, the accumulator would still inject because the two system check valves t would open.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. The accumulator can perform its intended safety function.
- 3. The margin of safety as described in the bases of any technical ;
specification was not reduced. The accumulator remained operable to perform its intended safety function, t
4 i Procedure for checking the Third System Reserve TYansformer , Purposes-The procedure was required to allow energizing, load checking, and phase checking the newly installed Dus #5 Transformer #3. Safety Evaluation Summary:
- 1. The probability of occurrence or the consequence of an accident or
. malfunction of equipment important to safety and previously evaluated
'in the safety analysis report was not increased. The independence of the required'offsite power supply network is not reduced. Only one of three offsite power distribution systems is affected. The change does not invalidate the loss of offsite power analysis.
- 2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created. The independence of the required offsite power supply network is.not' reduced. Only one of three offsite power distribution systems is affected. The change does not invalidate the loss of offsite power analysis.
3.. The margin of safety as described in the bases of any technical specification was not reduced. The change does not reduce the independence of the required offsite power supplies, therefore, the ability to maintain the plant in shutdown or refueling for an extended period of time with sufficient instrumentation and control is not reduced. 1
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l c JIMPERS REQUIRING A SAFETY EVALUATION IN 1989 I L JUMPER NO. N2-861: I. Install blower over lube oil reservoir manway with discharge outside the turbine building. pURp0 set Maintain a negative pressure on the lube oil reservoir; Vapor extractor will not stay running. EVALUATION FOR UNREVIEWED SAFETY QUESTION: An evaluation for a potential unreviewed safety question was performed, and the I l. following was determinedt l
- 1. The blowar will perform the same purpose as the lube oil reservoir vapor extractor. Automatic turbine trip on low lube oil pressure is not affected nor is emergency oil pump adversely affected. Ability to trip the turbine or the reactor are unaffected.
- 2. Turbine trip and loss of offsite power transients bound the failure of this temporary modification. The turbine is designed to trip on a loss of lube oil pressure.
- 3. Ability 'to trip the turbine and trip the reactor on a turbine trip is not ,
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JUMPERS REQUIRING A SAFETY EVALUATION IN 1989 ir > JUKPER NOS. N1-1335 & N2-862: i Install gaftronics in the auxiliary building penetration area. , PURPOSE: l I i Provide communications via gaitronics to and from the penetration area. V
- EVALUATION FOR UNREVIEWED SAFETY QUESTION
- I s
p An evaluation for a potential unreviewed safety question was performed, and the following was determined. ' L , E, 1. Vital bus power will be unaffected. Temporary gaitronics will use station-power not vital bus power. Fuses will prevent significant damage in the event of failure. Presently installed fuses will protect the vital bus. ;
- 2. Equipment- needed to mitigate the consequences of an accident are not affected. Fuses and breaker from the vital bus to gaftronics are not i affected. t
- 3. The ability of the vital bus electrical distribution system to supply power has not been adversely affected.
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f, ~\ o d 11 h' JUMPERS REQUIRINC'A SAFETY EVALUATION IN 1989 ! l. L . 1 JUMPER NO. N2-864: l': Connect PDTT pump discharge (2-DG-91) to RP suction from cavity (2-RP-10) I PURPOSE: Collect and return RCS loop stop valve leakby during flooded up condition. EVALUATION POR UNREVIEWED SAFETY QUESTION: p An evaluation for a potential unreviewed safety question was performed, and the !~ following was determined.
- 1. The uss of the DG-RP jumper does not affect any previously evaluated accident. Safety related system performance is unaffected. Any previously evaluated equipment malfunction is not affected by the jumper. Jumper only provides a return path for RCS leakage past loop stop valves.
e 2. Containment sump alarm is available to alert the operator in the event of a loss of cavity level. Equipment performance is unaffected in tt.is jumper. p
- 3. Adequate surveillance exist to ensure cavity level is maintained.
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.4 4 JUMPERS REQUIRING A SAFETY EVALUATION IN 1989 I
JUMPER NO. 1336 , l . Air supply jumper to full open 1-DG-PCV-101, drain header to _ primary drain
;_ transfer tank.
k i PURPOSEt j To allow adequate flow from the drain header to the PDTT to be established. EVALUATION FOR UNREVIEWED SAFETY QUESTION: An evaluation for a potential unreviewed safety question was performed and the following was determined.
- 1. Failure of the PCV to operate either open or closed does not affect any accident analysis evaluated in the SAR. Jumper does not create an increase in the probability _ of a malfunction. Jumper only bypasses valve controller. Jumper has no impact on equipment important to safety. No changes in the consequences of valve position failure.
- 2. Jumper only bypasses valve controller. Failure either way will not create ;
an accident. Function of PCV is not required for proper operation of equipment important to safety.
- 3. Jumper does not affect any Tech Spec items.
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gr . . [ ' i T : e [ JUMPERS REQUIRING A SAFETY EVALUATION IN 1989 6 L JUMPER NO. N1-1337: l L Jumper out RCS hot and cold leg temperature interlock, as well as relief line L flow and hot leg stop valve nnd bypass volve position interlocks normally l; required for cold leg "B" loop stop valve (MOV-1593), f PURPOSE: To allow electrical stroke at the "B" Loop Cold Leg Stop Valve (MOV-1593) in an attempt to flush the valve seat of any material obstruction presently causing leakage between the disc and seat of the valve. EVALUATION FOR UNREVIEWED SAFETY QUESTION: An evaluation for a potential unreviewed safety question was performed, and the following was determined.
- 1. Though the " Cold Water Slug Interlocks" are being bypassed, the conditions of the RCS and core and the reason for the valve stroke in this condition precludes-the possibility of accidental startup of an inactive loop, which is an _analyred condition. Plant conditions are bounded by assumed conditions in accident analysis. Nature of jumper does not increase probability of malfunction of loop stop valve.
- 2. Accidental startup of inactive loop has been analyzed, but is not credible due to existing plant conditions.
- 3. Though the " Cold Water Slug Interlocks" are being bypassed, the conditions of the RCS und core and the reason for the valve stroke in this condition precludes the possibility of accidental startup of an inactive loop, which is an analyzed condition. Plant conditions are bounded by assumed conditions in accident analysis. Nature of umpre does not increase probability r umlfunction of loop stop valve.
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JUMPERS REQUIRING A SAFETY EVALUATION IN 1989 JUMPER NO. N1-1338: Route PDTT discharge (1-DG-81) to charging header downstream of 1-CH-FCV-1122 (1-CN-291). p PURPOSEt L+ j_ Recover RCS leaking by 1-RC-MOV-1593, o > k i EVALUATION FOR UNREVIEWED SAFETY QUESTION: LAn evaluation for a potential unreviewed safety question was performed, and the following was determined.
- 1. Previously evaluated . accidents are not more likely to occur due to the jumper. Basic system operation is unchanged. A boration flow path will ttill be operable. Jumper is adequately rated for service condition. The ability to deliver. borated water to the RCS core to maintain -shutdown margin is. unaffected.
- 2. No new accident scenarios are created. Jumper flow path will not degrade CHfDG systems in any way.
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'. 3. Boration flow path will be maintained, n
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JUMPERS REQUIRING A SAFETY EVALUATION IN 1989: b:. . JUMPER NC. N1-1340: <
, Connect PDTT pump discharge (1-DG-77) to RP suction from cavity (1-RP-ll6) ;
PURPOSEt F Collect RCS loop stop valve leakby and return it to the cavity during Mode. 6. L EVALUATION FOR UNREVIEWED SAFETY QUESTION: i An evaluation for a potential unreviewed safety question was performed, and the
'following was determined.
g 1. The use of the DG-RP jumper does not affect any previously evaluated accident. Safety-related system performance is unaffected. Any previously evaluated equipment malfunction is ne,t affected by the jumper. Jumper only provides a return path for RCS leakage past the loop stop valves. 2.- Containment sump alarm to unavailable to alert the operator in the event of a loss of cavity level. Equipment performance is unaffected by this , jumper. p 3. Adequate surveillances exist to ensure cavity IcVel is maintained. F i t_
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lJ JtBtPERS' REQUIRING A SAFETY EVALUATION IN 1989 h JUMPER NOS N2-866 .867 ,868 .869, and 870 u'
^
ng To provide temporary power from 480 VAC bus 2J1-1 loads to various loads on f 480 VAC bus 2H1-1. Specifically: RSST "C" cooling fan to H EDG MCC.. Swing charger to 2-II battery charger, RSST "A" & "B" Cooling fans to 2-I' charger, Welding receptical to 2A semi-vital bus, RPII cabinet to radiation monitoring cabinet.
- PURp0SE:
b To supply power to selected 480V H-bus loads while the H bus is de-energized.
.) EVALUATION FOR UNREVIEWED SAFETY QUESTION:
- An evaluation for a potential.unreviewed safety question was performed, and the following was determined.
- 1. Since adequate breaker protection exists, the probability of an accident
< has not increased. .The consequences of an accident have not- increased-
'- since one bus would remain operable.
- 2. Since. the design function and operation of the distribution has not changed..the possibility of a'different accident or malfunction has not increased.
- 3. ,Since the design function is unchanged, the msrgin of safety as described is unaffected, t
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GW l JtMPERS REQUIRING A SAFETY EVALUATION IN 1989 l t s JUMPER NOS. N2-871 ,872, 873 and 874: provide temporary power from 2H 480 volt buses to MCC 2J1-1A, 2J bus semi vital .; bus, 2-III battery charger and 2-IV battery charger. l' i [ PURPOSE: l [ [ , Supply power to selected 2J 480 volt loads during the 2J bus outage. EVALUATION FOR UNREVIEWED SAFETY QUESTION: p An evaluation for a potential unreviewed safety question was performed, and the following was determined.
- 1. Since breaker protection exists and one train of emergency power remains i operabic, the probability of previously evaluated accidents has not +
increased. ,
- 2. The possibility of a different type of accident has not been created
~because one emergency bus remains operable. Brraker protection will exist.
L 3. Electrical requirements for modes 5 and 6 from Technical Specifications are being met and satisfy the associated basis section. i
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t JIMPERS REQUIRING A SAFETY EVALUATION IN 1989 i [ JUMPER NO. N2-882: F Pull local alarm horn relays. I PURPOSEt. !, To silence the local horns while deluge is valved out to the main transformers.
- EVALUATION FOR UKREVIEWED SAFETY QUESTION t
An evaluation for a potential unreviewed safety question was performed, and the following was determined. p
- 1. The jumper.will merely disable the local alarm on the deluge system to the main transformers. Water to the de-energized main transformers is valved out. The consequences of previously evaluated accidents have not been increased by the performance of this jumpe . The main transformers are not safety related and do not supply power to any safety related equipment.
- 2. Water is valved out to the main transformers. Disabling of the horns is being done because they are a nuisance. The jumper will not affect safety related equipment.
- 3. Adequate water suppression systems are available to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located.
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- JUMPER NO. N2-883:
F Jumper IA to Unit 2 penetration (Aux. Bldg.) area. PURPOSE: The supply header to U2 Aux. Bldg. penetration area must be isolated for leak ! repair on the header. EVALUATION FOR UNREVIEWED SAFETY QUESTION f An evaluation for a potential unreviewed safety question was performed. and the L following was determined.
- 1. The installation of the jumper does not increase the probability of the
- occurrence of an accident. RHR systems and boration flow paths of both 1
units are not affected by the IA jumper. The consequences of the loss of IA are the came with or without the installation of. the IA jumper. AP-28 is available to take action in the event of the loss of IA. The installation and testing of.the jumper following installation and removal ensure that the probability of occurrence of a malfunction of safety ! equipment is not increased. The jumper will only be in place for a short period of time.
- 2. ~ The nature of the jumper precludes the creaction of a different type accident., RHR systems for Units 1 and 2 or designated boration flow paths are affected by the jumper.
! 3. Ecquired Tech. Spec. systems will be maintained in an operable condition during the jumper installation or the appropriate statement will be
- entered. .Neither unit RilR system or designated boration flow paths are affected by the jumper.
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?~'*;. . p. I e JUNPERS REQUIRING A SAFETY EVALUATION IN 1989 , JUMPER N2-886.687 and 888 Jumper loop stop valve open interlocks. PURPOSE: Contacts in circuit are not making up. EVALUATION FOR UNREVIEk'ED SAFETY QUESTION: E An evaluation for a potential unreviewed safety question was performed, and the !. following was determined.
- 1. The conditions required by the cold leg loop stop valve interlock will be ,
verified before opening the valve if the jumper is installed. The
. conditions specified in the UFSAR for opening a cold leg loop stop valve t
will be met. Loop stop valve is designed to be cycled when interlock is met. Valve and system integrity not affected.
- 2. 11ur changt af a cold water injection accident or any other accident is not I created' because interlock will be verified before opening cold leg valve.
The UFSAR assumes cold leg loop stop valves are cycled when interlock is met. ' i
- 3. The-cold water slug positive reactivity accident will not occur because the !
interlock will be verified before a cold leg loop stop valve is opened. ,. Tech' Spec requirements will be met. ' .o l
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- JUMPER NO. N2-893: '
i The jumper will allow 2-TV-CC-204C to remain in the open position, utilizing n, two pipe wrenches'and a come-along while repair / replacement of damaged. air -! tubing to the valve actuator is performed. I ,
!b PURp0 set
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f To maintain 'CC flow to 2-RC-P-lO coolers during air line repair for ! TV-CC-204C. t [ EVALUATION FOR UNREVIEWED SAFETY QUESTION: i An evaluation for a potential unreviewed safety question was performed, and the [
'following was determined. '
i
- 1. The jumper maintains TV-CC-204C in- the open position during air line replacement. The probability of an accident is not. increased by this. A' !
second isolation valve (check valve) in containment exists and will provide ,
>- penetration isolation when required during phase "B" isolation conditions.
This condition, is bounded by the Action Statement of T.S. 3.6.3.1 which will be entered when the jumper is installed. The jumper only affects ,
'2-TV-CC-204C and its ability to provide containment isolation during phase ;
-B conditions..
- 2. No new accident-is created by blocking open the trip valve.
statement of T.S. 3.6.3.1 will be followed. The action j
- 3. The action ' statement of T.S. 3.6.3.1 will be followed while the jumper is in service. ,
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pn-l > . - m 1 JIMPERS REQUIRING A SAFETY EVALUATION IN 1989 JUMPER NOS. N1-1347, 1348, and 1349 Provide temporary power from 480 VAC bus 1J1-1 to various loads on 480 VAC bus , 1H1-1. Specifically, MCC IH1-1A, H bus semi vital bus lA, 1-I battery charger and the 1-Il battery charger. i PURPOSE: 4 Supply power to selected 1H 480 volt load during the IH bus outage. EVALUATION FOR UNREVIEWED SAFETY QUESTION ' An evaluation for a potential unreviewed safety question was performed, and the following was determined.
- 1. Since breaker protection exists, the probability of previously evaluated accidents has not increased. One train of emergency power rettains, operable, therefore the consequences of previously evaluated accidents has not been increased.
- 2. The possibility of a different type of accident has not been created because one emergency bus remains operable breaker protection will exist.
- 3. Electrical requirements for Modes 5 & 6 from Technical Specifications are being met and satisfy the associated basis section.
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L s JUMPERS REQtf1 RING A SAFETY EVALUATION IN 1989 ! l r. [ JUMPER NO. N1-1351: Provih temporary power to 1-SW-C-1B during the "J" bus outage. (Service Water Pumphouse air compressor normally powered from MCC IJ1-3 Bkr A3, jumper will power compressor from MCC IH1-3 Bkr C2 which supplies 1-HV-UH-36A SWPH heater- - normally.)' - REASON: [' No SW reservoir level indication with the "J" bus outage. Need this compressor , , operable for indication in main Control Room.
. EVALUATION FOR UNREVIEWED SAFETY QUESTION: l An evaluation for a potential unreviewed safety question was performed, and the following was determined. j L 1. Breaker protection exists to protect the available emergency bus thereby t not increasing the probability of previously evaluated accidents. SW - ;
reservoir level ' indication will be available to- verify level so that > adequate inventory exists to provide normal cooldown of. Unit 2 and mitigate the effects of accident condition. One train of emergency power remains-available on Unit I and SW reservoir level indication remains available to verify minimum water level for Unit 2. SW reservoir level required to be ; maintained greater than 313 feet in Modes 1-4. . The jumper is being 1 installed- so that SW reservoir. level can be verified and the LCO of T.S. 3.7.5.1 satisfied. . t i 2. The possibility of creating a different type accident has not occurred for ! Unit 1. An emergency bus remains operable for Unit 2 Service Water level .. I can be verified satisfactory to provide normal cooldown and mitigate the effects of accident conditions. The malfunction of safety related - equipment'of a different type is not created by this jumper. The jumper , allows normal operations .to continue on Unit 2 without jeopardizing equipment or systems on Unit 1. 3 The jumper will allow the operator to verify adequate inventory exists to > provide normal cooldown and to mitigate the effects on accident conditions. s
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k> , .' JtBIPERS REQUIRINC A SAFETY EVALUATION IN 1989 h l JUMPER NO. N2-8981-Li- Installation of a steel flex hose from 2-MS-PS-218 to 2-MS-None. temporary PT u , connection on #3 gland steam supply to temporarily supply gland steam to the #3-F , ' gland on the Unit 2 #1 LP turbine to maintain condenser vacuum.-
' PURPOSE:
Gland supply line.to the #3 gland is plugged by furmanite. [ EVALUATION'FOR UNREVIEWED SAFEIT QUESTION: o-b An evaluation for a potential unreviewed safety question was performed, and the following was determined. 1
- 1. . The-jumper will provide gland steam supply at the proper pressure to the E affected gland. to prevent in or out leakage along the turbine shaft. No safety related systems or. equipment is affected by the jumper. Jumper materials are adequate.for the application. Chemistry limits on secondary L will'be maintained.
- 2. : Gland-Steam is not safety-related. The function of this jumper does not
, create any new accident.- Secondary chemistry continues to be monitored and will be maintained within limits or proper actions will be taken.
- 3. Gland Steam is not safety-related and is not Tech. Spec. No Tech. Spec.
required system or component will be adversely affected. L
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9 a l .? s JUNPERS REQUIRING A SAFETY EVALUATION IN 1989
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-JUMPER NO.-N2-897 -
"1 ' Install temporary carbon steel piping from 2-MS-381 (Gland Supply Strainer-Blowdown):into a flange and f'tting on valve body of~2-MS-PCV-221.
~
This will 1 1
-jumper gland ~ supply steam-to the'#3 gland on the Unit-2 #1 LP turbine.
g 'PURPOSEt l- Glan'd supply line to-the #J ,and is plugged by furmanite.
. EVALUATION'FOR UNREVIEWED SAFETY QUESTION:
-An: evaluation for a potential unreviewed safety question was performed, and the following was determined.
- 1. ' The jumper will provide gland steam supply at the' proper pressure -to the-affected_ gland to prevent in or out Icakage-along the turbine shaft.- No safety-related system or equipment. is affected by the jumper. Jumper materials- are adequate for the application. Chemistry limits on secondary will be maintained..
-2. GlaEd eteam is.not'cafety-related. The function of the jumper _does not
-create _any new' accident.
Secondary Chemistry continued to be monitored and willtbe maintained within limits or proper action will be taken; 3 3.- l Gland steam system is not safety-related and is not Tech.- Spec. No Tech. Spec. required system or component will be adversely affected. 7 1 p 1 1 l,
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j ,.! L< 1 i . 't JUMPERS REQUIRINC A SAFETY EVALUATION IN 1989 ( L- . JUMPER NO. N1-1354:
-Close 1-CC-201, 1-CC-202 and 1-CC-237. Install hose from 1-CC-231 to 1-CC-232
[ J and open valves. Open ~1-CC-TV-108A and 107A. Start 1-NS-P-1A, Open 1-CC-265. l t i F PURPOSE - l p Jumper'is to'be installed to allow transfer of CC from the Neutron Shield Surry Tank- (1-NS-TK) to the "B" RHR CC return line. This will accomplish two tasks 1 (1) Draining the Neutron Shield Surge Tank for Maintenance and~(2) Filling "B" RHR CC return 111ne fo11owing11-CC-MOV-100B maintenance. EVALUATION FOR UNREVIEWED SAFETY QUESTION An. evaluation for a potential unreviewed safety questions was performed, and the following was determined. [':
- 1. The jumper only transfers CC water between subsystems of the CC system.
The -sybsystems will not be adversely affected and the affected RHR CC i return line-is tagged out. . The Unit is defueled and the RHR system is .not ' in operation.- The neutron shield tank cooling water systems is not safety related. -
- 2. The' transfer ~of CC wtaer between the CC subsystems does not create' the possibility of a new accident.
- 3. The RHR system is not in operation (the unit is defueled). The expected .,
'RHR CC return line'is tagged out for the fill. No LCOs will be violated, n ,
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JR ., Y JUMPERS REQUIRING A SAFETY EVALUATION IN 1989
?
i U JUMPER NO. N2-901:= Pullpatchcardlfor2-RS-P-2BSealHeadTank11evel' alarm. ,
' PURPOSE:
~ Alarm does- not'prov'ide additional information at this time.- Problem has been
- identified-and a W.O. has been generated.
. EVALUATION FOR UNREVIEWED-SAFETY QUESTION: I
.\
An evaluation for a. potential unreviewed' safety question was performed, and the following was, determined.- 1., All . classes of accidents--are unaffected by-defeating the alarm. Pump. operability in the event of an actual reduced seal cooling inventory is unaffected. .Since, 2-RS-P-2B can still perform its intended function,. ,;
.previously evaluated offsite doses are still valid. . Pump vendor' has
-verified that- seal operability 11s not jeopardized by the possible reduced inventory'in;the sealccooling-system.
~
2L-A new scenario has not been created. RS system is fully operable.
- 13. RS system will still reduce containment pressure to subatmospheric following a'DBA.
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JUMPERS REQUIRING A SAFETY EVALUATION IN 1989
- JUMPER NO. N1-1355:
Install. a temporary push button on containment side fuel transfer control panel. PURPOSES. The " conveyor at Rx" limit switch is erratic and does not indicate that the conveyor has left the at home position. This condition would allow upender movement with the conveyor in the wrong location. The push button resets the interlocking relay by de-energir.ing it. EVALUATION FOR UNREVIEWED SAFETY QUESTION:
- An evaluation for a potential unreviewed safety question was performed, and the following was~ determined.
- 1. = Maintaining the interlock function operable ensures that the probability of or consequences of a malfunction or. a different malfunction is not increased.
- 2. Maintaining the interlock function operable ensures that the probability of or consequences of a malfunction or a different malfunction is not increased. :
- 3. No systems or_ components described in the T.S. or basis are affected. -
i
pql = ..Ji. 3:; , 'n , N [l ' ' 1 4 . JUMPERS REQUIRING A SAFETY EVALUATION IN 1989-t . E; b JUMPER'NO. N1-1364: l, g Air supply. jumpers ~to fully open 1-DG-PCV-101 (drain to PDTT) i! lt _ LPURPOSE - n: I'v; To allow adequate flow 1from' drain header-into the PDTT. ho 1 EVALUATION FOR UNREVIEWED SAFETY QUESTION: An evaluation for a' potential unreviewed safety question was performed, and the ' p following was determined. k. i
- 1. Failure ofl-the .PCV' to operate either open or closed does not affect any- '
anceident analysis evaluated in the SAR. . Jumper does not create an increase _the ' probability. of a1 malfunction. ' Jumper only -bypasses -valve e in '
' controller.,l Jumper.has no impact:on equipment important to safety.
2.- Jumper only: bypasses valve controller.- Failure'either way will'not create
- an ' accident. . Jumper only! bypasses valve controller. . Function of PCV is-not required for. proper operation of equipment important to. safety.
fL 3. Jumper does'not affect any Tech Spec: items. c i o f, i L I 9 1i 1 5
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-JUMPERS REQUIRING A SAFETY EVALUATION IN-1989 .;
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JUMPER NO. N1-1367: 4: k, LUpen 1-DG-PCV-101' fully to allow increased flow.into the PDTT.. PURPOSE:-
$ Tol allow adequate flow from the drain header.to the.PDTT to.be established..
l f EVALUATION FOR:UNREVIEWED SA'FETY QUESTION f An evaluation ~for a potential unreviewed safety question was. performed, and th'e , t following was determined. g , ;
- 1. Failure) of the. PCV. to operate either open or closed does not affect any ;
accidentianalysis evaluated in the SAR. Jumper does not create an increase- .!
.in the. probability of- a= malfunction. Jumper only bypasses valve
, controller. Jumper has no impact'on equipment important to safety. +
- 2. Jumper only bypasses valve controller. Failure either way will not create j an accident. '
- 3. 1 Jumper does not affect any Tech Spec items.
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-JUMPERS REQUIRING A SAFETY EVALUATION IN 1989-i
-JUMPER NO. N1-1370B:-
ir ,
< Defecting 1-SW-P-8_ input.into annunciator IK-36, Recire. Spray 1 A ,1 B ,10, ; and
[ -ID,' Rad Mon. Pump-Low Flow. c-g 4 PURPOSE: t: f
.1-SW-P-8 is tagged out. _ Loss of power 'to l relay (2-1SWSN12)'results in-annunciator being illuminated. This can be considered an invalid alarm.
l' EVALUATION'FOR UNREVIEWED SAFETY QUESTION: t An~ evaluation for a potential;unreviewed safety question was_ performed, and the
~
following was determined. i t-L1~. Jumper only defects.the low / low input to alarm window for 1-SW-P-8 which'is out of service for maintenance. Alarm function of 1-SW-P-5,6 and 7 are not affected. ,0peration of running-RSRX sample pumps is not affected.
- 2.. Jumper-only. defects 1the low / low input to' alarm window for 1-SW-P-8 which is out of service .for' maintenance. Alarm function of 1-SW-P-5,6 and 7 are i not affected. Operation of running RSHX sample pumps is not affected.
- 3. This' low-flow alarm is not described in Tech' Specs. !
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t , (c .- , LJUMPERLNO.'N1-1370E:' -f p DefeacLannunciator " Stripper Dia. Comp. Disch.;Nor. Hi/Lo pressure.-
~
'A . PURPOSE: n Up qAlarm' is locked in.due toJcompressors not being used; distraction to operators' i
- from' Human. Factors standpoint.
I, , EVALUATION FORfUNREVIEWED SAFETY QUESTION:~ ! F:, i G ;An evaluation for a potential;unreviewed safety question was performed, and the O following was determined.- t. E .1~.. No'isystem' operational. changis have been made. All systems required to mitigateidesign basis accident s are available and unchanged. There are no
+ adverse effects on BR system ;omponents. Present system operation does not
-alter; operation of BR syster. components.
2'.- .Present ' operation' of. the- gas stripper does not' require. compressor operation. Alarm on'Hi/Lo compressor discharge pressure is not needed.
- 3. 'None are affected.
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j'{; ' > o n PERS REQUIRING A SAFETY EVALUATION IN 1989' L. JUMPER NO. N1-1370H:' n Jumper out annunciator 21J-G4 "1RS PP-2B seal head tank High/ Low level. g b' PURPOSE - t p Support' black. board' program alarm does not provide useful information, f p -EVALUATION FOR UNREVIEWED SAFETY QUESTION:
~
.An-evaluation for a potential unreviewed safety question was performed, and the following was determined. '
.c
- 1. All classes of accidents are unaffected by defeating the alarm. Pump. .
operability -in the' event - of an actual reduced seal . cooling inventory is - unaffected. Since 1-RS-P-2B can still perform its intended function previously evaluated offsite doses are still valid. Pump vendor- has a verified'that seal operability is not. jeopardized by the- possible reduced inventory in the -seal cooling system. SFGG area ventilation is-fully ' operable-and is. designed to swap to iodine filter tank on CDA. Therefore, ' in case of a total seal failure, the SFGG exhuast will be properly' treated. 2.- A new scenario has not been created. RS system is fully operable. J 3. : RS' . system will still reduce containment pressure to. subatmospheric ! following a DBA. l r
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~ JUMPER NO. N1-1373 -
cInstall blower over lube oil reservoir, i
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. ' PURPOSE: c: b
~
' Maintain'a negative pressure on ^the lube' oil' reservoir-.to allow' maintenance on
~
j GM-F-6.- i b. EVALUATION-FO'R'UNREVIEWED SAFETY QUESTIONt il . -Blower- will' perform the-same purpose as-the lube oil. ~ Automatic turbine- ?! trip.'on low-lube-oil pressuresis not affected.s Emergency oil pump:-is-.not-adversely affected. Temporary blower will perform the same function as the'. lube'. oil. reservoir vapor extractor.
- 2 .- Turbine: trip and loss of offsite power transients bound the failure of..this temporary modification.- The-turbine is, designed to trip on a loss of' lube-N . oil pressure,
,3. Ability'to trip the-turbine-and trip the. reactor in a turbine trip is not
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Fh io , i ( ., , : JUMPER NO.~N1-1370E: , p [ Defeat-annun."StripperD'a. i . Comp.Disch.HeaderHi/LoPress" :s b U PURPOSE: -t i
'; 'A'larm in.usually locked'in'duetto infrequent use'of^ compressors. j g . EVALUATION FOR'UNREVIEWED SAFETY-QUESTION .;
f An' evaluation for a' potential.unreviewed safety question was performed. and'.the
-following was. determined.- ;g a
- 1. .No physical changes to system operation will occur. -All systems ; required' :i to mitigateLDesign Basis Accidents are a'vailableLand unchanged.. Failure of the_ gas stripper lis not more 'likely .due .to _ annunciator- defeat. No-components ~are--physically altered. 'Present system use ensures constant- g
' surge' tank bleed.to a PCV.
- 2. Use of BR' system is unchanged.
'3.t None are affected. ,
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p. y- JUMPER NO.:N1-1370J:- L Defeat' alarm from-1-RS-P-2A seal' head cauk Hi/Lo level, annunciator- C-4 on. " 6f EI-CB-21J.. .
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PURPOSE: 4; Defeat _ alarm which is locked in support bJr:ckboard program. EVALUATION FOR UNREVIEWED SAFETY QUESTION:
, An evaluation'for.a potential-unreviewed safety-question was performed, and the D following.was~ determined.-
, -1. -System integrity is'not affected. dnalysishas_determinedOSRS pumps are operable with. seal head tank and seal. problems. RS system is designed for a single failure, safeguards. ventilation is directed through. iodine filters !
during a CDA.' Single failure'and leakage of radioactive fluid are analyzed events. The OSRS pump;is operable as documented in JC0 88-26.
. RS. system 6 - will continue to depressurize' containment with seal tank level alarm defeated.
- 12. RCS system operability is not adversely affected by defeating alarm.
3.--RS system is operable, 4
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'DC-84-36-3
' SERVICE WATER' RESERVOIR IMPROVEMENTS VALVE HOUSE MECHANICAL / ELECTRICAL AND FINAL STRUCTURAL NORTH ANNA / UNITS 1 S 2 4
- DESCRIPTION This DCP' installed a new- reservoir spray and bypass system. The partially constructed North Anna Units 3 and 4 pump house was completed and serves as a i valve' house for the spray and winter bypass headers. -All piping exits the new '
valve house to serve the two bypass headers and the eight spray headers in the
- new- system. The final tie-in to the existing service water system are in the buried return headers approximately 150 feet outside of the protected area. 1 This. design concept. provides a system with more efficient heat transfer
- characteristics, improved resistance to corrosion, lower. maintenance ,
requirements, better access for periodic Inspection and improved operability.
~
SUMMARY
-OF SAFETY ANALYSIS This modification does not constitute an "unreviewed safety question" as defined In'10CFR50.59. This design does not; a.. Increase -the probability of occurrence or consequences of an accident or malfunction of equipment Important to safety as prevlously evaluated { in the ~UFSAR as part of the on-line service water system.- This design ' change does not. place the new service water system into operation;- however, the new design and equipment meet or exceed all of the , original safety-related requirements of the existing service water system. The new service water system will merely use a new safety related flow path and spray arrays to acccrnplish the cooling process. The new spray array will be'more efficient than the existing one and
'the new valve arrangement provides more flexibility. The new 480V MCC's that are energized in this design change are safety related, selsmic' Class 1.
- b. Create the possibility for an accident or malfunction of a different '
type than any evaluated previously in.the UFSAR. The new service water design uses the same basic technology as the present design. It consists of the reservoir, pumps, piping (some existing and some new) and a new, more efficient spray array. The winter bypass is a new feature, but It wi11 not be used durIng an accident. The same accident types will exist for the new system configuration. The new system will' be as good as, or better than, the existing system in the terms of preventing and/or mitigating the postulated accidents,
- c. Reduce the margin of safety in the basis for any Technical Specification. The margin of safety is increased with this new system since the new valve arrangement will minimize the effect of one valve falling closed, and the new spray array will be more efficient and more durable than the existing system. The margin of safety with the new system is not decreased when the system is operating in the winter bypass mode because ~the system automatically switches from bypass to spray upon receipt of a Safety Injection signal .
m DC-84-60 REG. GUIDE 1.97. PRESSURIZER LIQUID TEMPERATURE MODIFICATION NORTH ANNA / UNIT 2-DESCRIPTION'
. Regulatory Guide 1.97 Rev. 3 "Instrunentation for Light-Water-Cooled Nuclear Power Plant and Environs Condition During and Following an Accident," (R.G.
1.97) requires that variables required for planned nenual actions be classified-as_" Type A" variables.- Pressurizer Liquid Tanperature was identifled asi a
" Type A" variable and. as such nust neet the Category I requirenents of R.G.
1.97. R.G.1.97 requires that'the entire Instrunent channel have an cdequate range, be redundant and separated, have an adequate power supply, and be se l sni ca l l y . and - env i ronnenta l l y qua l l f l ed. , Pressurizer Liquid' Temperature RTD, TE-2453 was removed and replaced with a new Weed dual element RTD. The RTD ls seismically and environmentally quallfled to IEEE 344-1975 and IEEE 323-1974 and purchased to the requirenents specified in NAS-2033 -" Specification for Pressurizer Liquid Tenperature RTD". The new RTD ' retained the previous nark nuTber TE-2453. The existing non safety related temperature channel T-2453 was deleted and two-new redundant Pressurizer Liquid Temperature channels (T-2453-1 red _and T-2453-2 blue) were installed. These new channels neet the intent of the Category 1 requirements of R.G.1.97 except for the electrical separation and redundancy criteria. The exception to these criterla_ exists whern one dual elenent RTD will be used for both channels.
-One, RTD is used because there is only one Pressurizer Liquid Tenperature thermowell available and the installation of considered practical. a- new thennawel l was not Virginia Electric Power Company submitted to the NRC (Ref. letter dated January 31, 1984,: Serial'#054) the proposed nethod of providing a dual eleaent for redundancy.
The NRC acknowledged this as an acceptable nethod of complying with R.G.1.97 as noted in letter dated February 8, 1985, Serial H85-094 The Main Control Board was modified by this change. The indicator T1-2453 was relabeled as T1-2453-1 and a new Indicator T1-2453-2 was added to Vertical Board Section 2-1 next to T1-2453-1.
SUMMARY
OF SAFETY ANALYSIS The removal of the existing Pressurizer Liquid Temperature channel T-2453 and the addition of the new safety-related channels T-2453-1 and T-2453-2 does not constitute an "Unreviewed Safety Question" as defined in 10CFR50.59. .
- 1. The implementation of this nodification does not increase the probability of occurrence or the consequences of an accident or nelfunction of equipnent important to safety and previously evaluated in the Final Safety Analysis Report. The existing Pressurizer Liquid Tenperature channel T-2453 is a non safety-related channel, as such, no credit has been taken for it in any accident analyses previously
evaluated in the_ UFSAR. The replaconent of the existing RTD 'wlth an envi ronnental ly- and- sel unical ly quallfled RTD 'wlli reduce the' probability of failure 'during nornal and accident conditions.
- 2. The implecentat ion of this nodification does- not' il ncreasr. the possibility. of an ' accident or a nelfunction of 'a different t',po than evaluated previously_ in the- Final Safety Analysi s : Report . Since Pressurizer Liquid Temperature channel is a non safety-relat0d channel,.
no credit has been .taken for it in any accioent anesyses. The replacement of the existing _ RTD.- with an environmentally. and seismically quallflod dual elenent RTD will provide an. Increase in reliability' for; monitoring Lthe Pressurizer Liquid Temperature. This will enhance the-control room operator'sLabilityJto' accessEthel plant status' during and. after an accident. The use of a dual - elenent RTD (TE-2453) to provide redundant nonitoring is not regarded as an "Unreviewed Safety-Question" because a fault in the RTD or in.the. cannon cable to'the first Junction box would result in only_a loss of Pressurizeri Liquid' Temperature : Indication which does not affect any present safety analysis. A review has been nede to determine if a short- circuit in the connon cable (i.e. either a ground fault or hot short between channels) would-have any. impact on any other existing circuits in the Prinary Plant Process Rack. The investigation has determined that: this would have no inpact on existing circuits in ~the
~
Primary. Plant Process Rack.
- 3. .The Inplementation of this nodification does not reduce the margin of safety'as defined in the basis for any Technical Specification. The safety limits as defined in the -Technical Specification are not affectcd by this design change.
V F DC-84-69-1 and DC-84-70 ' PRESSURIZER SAFETY VALVE LOOP SEAL INSULATION OVENS NORTH ANNA / UNITS 1 6 2 DESCRIPTION The reactor coolant systen safety valves must be quallfled for design-basis transients and accidents. NUREG-0737 Section ll.D.1 provided qualification and docunentat ion requi renents and stated in part that " valve qualification shall include qualification of associated control circuitry, piping and supports, as well as the valves thenselves." There~-are three safety valves provided to protect the reactor coolant system from overpressure. Each safety valvo is connected to the steam space- of the pressurizer with six inch piping in a loop-seal arrangement to maintain subcooled water (rather than steam) at the valve seats. The original piping analyses did not consider the fluid transient loads associatec with a water slug passing through the systen. EPRI studies (Review of Pressurizer Safety Valve PerforFance as Observed in the EPRI Safety and-Relief Valve Test Program,. dated June 1982, WCAP-10105 - Westinghouse Corporation) Indicated significantly higher loaos than those postulated in earlier analyses.
. This DCP installed netal reflective thennal Insulation boxes which enclosed the safety valve loop seal piping, exposing the loop to an uninsulated portion of the pressurizer. The safety valves are located directly adjacent to the upper section of the pressurizer vessel . A snell area of the pressurizer adjacent to the' safety valves is exposed. Insulation boxes were Installed enclosing the exposed pressurizer areas and the loop seal piping. The Insulatlor, boxes are designed to 'nalntain the water in the loop seals at an elevated temperature C>4000F).
This tencerature will allow part of the water to flash into stean after passing' through the safety valves.
~
The reduced density of the water-stesn mixture will
.cause reduced fluid transient loads for the safety valve discharge piping and plpe supports.
S,UMMARY OF SAFETY ANALYSIS This design change will not create an "Unreviewed Safety Question" as defined in 10CFR50.59. The implementation of this modification: a Does not increase the probability of occurrence or the consequences of an accident or malfunction of equi pnent important to safety and previously evaluated in the Safety Analysis Report. The Installation of insulation boxes around the pressurizer safety valve loop-scal piping will not affect the operation of the pressurizer or the safety valves. The function of the Insulation boxes is to nelntain the water in the loop-seal piping at an elevated tenperature to minimize fluid transient forces on the valve discharge piping. b Does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Original piping analyses did not consider the fluid transient
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hacf.m,,Y Esupports;are qual.1 fled for the design basis transients.- f y -;
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W, [.ys , ; safety-related; system or component.? W;e ,'- 9 5 f'p : ' ji ; 2 q ,[f, . ,? t({ (! ._,- p- ' 's ' j
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DC-84-71 and 84-72 k VALVE DISCHARGE PIPE SUPPORT MODIFICATIONS NORTH ANNA / UNITS 1 & 2 DESCRIPTION
' The_ _ reactor coolant' system power-operated relief valves and safety valves must be quallflod for design-basis transients and accidents. NUREG-0737 Section ll.D 1- provided " qualification shall include 'aualification of associated
. control circuitry, piping and supports, as well as the valves themselves."
There are two power-operated relief valves (PORVs) and three safety valves provided to protect the reactor coolant system frcm over-pressure. Each safety valve is connected to the steam space of the pressurizer with six inch diameter p piping _ in a loop-seal arrangement to maintain subcooled water (rather than steam) at the'. valve seats. [ This DCP Installed additional pipe supports and modified existing supports I which are designed for the fluid translent loads (as a result of_ elevating the k ! loop seal water temperature) associated with tha water slug transient. The support tredi ficat ions ensure that the pressurizer PORV and safety valve discharge piping is c:ual Ifled for the expected operating conditions :for design-basis transients and accidents. The piping analyses included thermal, hydraulic, deadweight, and seismic loading. In addition, the analyses included a fatigue evaluation for Safety Class Q1 piping. ! SUt+%Ry OF SAFETY ANALYSIS This .DCP will not create an "Unreviewed Safety Question" as defined in 10CFR50.59.
- a. Does not increase the probability of occurrence-or the consequences of I an. accident or malfunction of equipment important to safety and previously evaluated in the Safety Analysis Report. The pressurizer safety and relief valve discharge piping and supports have previously been evaluated for transient and accident conditions. This design change will upgrade the piping supports to withstand the loads associated with the water slug transient as required by NUREG-0737 Section !!.D.1. See Appendix 8.2 for reference calculations.
b.- Does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. This design change nodifies the pressurizer safety and relief valve piping supports only. The operation of the power-operated relief and safety valves is not affected.
- c. Does not reduce the margin of safety as defined in the basis for any Technical Specification. The pipe supports which are hydraulle shock suppressors (snubbers) are addressed by Technical Specification 3/4.7.10. The new/ modified snubbers should be included in the ADM 11.8.
The overall design taking into account the recent EPRI pipe and valve test data and North Anna specific stress analysis, results in an improved margin of safety.
~"~
fA3 h' DC 82 PIPE PRESERVATION AUXILIARY SERVICE WATER SUPPLY AND RETURN' NORTH ANNA / UNITS 1 C 2 I DESCRIPTION
- The service water: system piping at . North Anna Power Station has undergone <
' severe corrosive attack due to the aggressive nature of the water- and the
. presence of sulfate reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting. ,
-In order to preserve the Integrity of the service water system piping,. and thereby- avert or postpone any requirement for= further pipe replacements, a corrosion inhibitor chemical treatnent progran has been. Initiated.
However, the treated service water cannot presently come in contact with the Internal surface of a majority of.the service water system piping due to the layer of corrosion product, silt and slime that has developed over years of plant operation. in order for the corrosion inhlbitor to be effective on service water pip ~1ng, the Internal surfaces must be cleaned to bare metal. , The desired pipe surface preparation was acconpilshed by chemical and nechanical cleaning nethods. The snount of corrosion product on the pipo Internal surfaces made it impractical to clean the piping using chenical methods alone. In order to reduce the amount of chenicals and time required ' for. final chemical cleaning, and to insure the effectiveness of the chenical cleaning process, the -piping was first-cleaned using a mechanical technique referred to as hydrolasing. , The mechanical and chemical cleaning operation performed on the service water piping . required that all valves in the system provide proper isolation. Those
' valves that did not seat properly were. repaired or replaced before many sections of. the service water piping was nechanically cleaned. .,
This design change covered the cleaning of portions of the Auxiliary Service Water piping using this technique; specifically the following: 24" -WS-25-151-Q3 24" -WS-58-151-Q3 Two butterfly va'Ives were repaired.as part of this design change. Both valves are for 24 inch nominal piping and are MOV-SW-115A and MOV-SW-2208. SUt4%RY -OF SAFETY ANALYSIS The nechanical cleaning of sections of the auxillary service water piping by this DCP (including the repair of butterfly valves) does not: 5
- 1. Increase the probability of occurrence or the consequences of an accident or nelfunction or equipnent important to safety and previously evaluated in the Updated Final Safety Analysis Report; because water supply w!11 be maintained to the charging pump coolers by an alternate
usupply in thefevent of-failure of the on-line service water loop, and
#1,ood i ng of the auxillary and turbine building basenents are prevented by adninistrative control by locking closed valves- 18 Inches and
-greater. In'the event of a LOCA, the off IIno system will be placed in a backup status which could be utilized to provide service water to critical equipnent . by installing b'llnd flanges In place of valves renoved for cleaning and repair.
- 2. ' Create a possibility for an accident or nalfunction of a different type than-any evaluated previously in the Final Safety Analysis Report. The only modification was the addition of snel l bore- piping ~which has' already .been evaluated in the UFSAR; therefore no possibility for an accident- or malfunct?on of a different type than any evaluated previously In the Updated Final Safety Analysis Report will-result.
- 3. Reduce the margin of safety as defined In the basis of any1 Technical Specification.
All naterials tused in the nodifications, resulting froa this design change, are in accordance with NAS-1009 and/or neets the requi ronent s - of the systan's original design criteria. The service water system consists of two-(2) 100% capacity piping loops shared by both Units 1 and 2. The Technical Specification (3/4.7.4) requires that both be operable. If one is . inoperable for ' a period greater than 72 hours, then .both units; must be shut - doan. The conttruction period will require that one loop be renoved from ' service in order to perform the necessary cleaning, but this-will be performed within the prescribed 72 hour tine period. _Therefore, based on- the-Technical Specification criteria, no threat of increased probability of accident will occur. -f
Fy> q ? L-F DC-84-83 1 PIPE PRESERVATION - AUXILIARY SERVICE WATER SUPPLY NORTH ANNA / UNITS 1 & 2
' DESCRIPTION The service water system piping at North Anna Power Station has undergone severe corronlve attack due to the aggressive riature of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general ,
wall thickness reduction and' wide spread pitting. ' in order l to preserve the integrity of the service water system piping, and thereby avert or postpone any requirement for further pipe replacements, a corrosion inhlbitor chemical treatment program has been initiated. i
. Mcwever, the treated service water cannot presently come in contact with the Internal surface of a majority of the service water system piping due to the layer of corrosion product,. slit and slime that has~ developed over the years of-plant operation.-
In order for the corrosion inhibitor to be effective on service water piping, the Internal surfaces must be cleaned to bare metal. The desired pipe surface preparation was accomplished by chemical and ! mechanical cleaning methods. The annunt of corrosion product on the pipe internal surfaces made it impractical to clean the piping using chemical methods alone.. In order to reduce the anount of chemicals required, reduce the i time required for final chemical cleaning, and Insure the effectiveness of .the chemical cleaning process, the piping was first cleaned using a mechanical i technique referred to as hydrolasing. The mechanical and chemical cleaning operations to be performed on the service water piping. required that all valves. In the systen provide proper isolation. : Many of the valves in the service water. system do not provide tight shutoff. Those valves that did not seat properly- aust be repaired or replaced before
- many sections of the service water piping could be meche,nically cleaned.
This design change covered the cleaning of porticm of the Auxi':ary Service li Water Punp .and Screen Wash Pump piping using this technique, s;:c ifically the following: 24"-WS-25 -151-Q3 24"-WS-425-151-Q3
-24"-WS-48 -151-Q3 L8"-WS-90 -151-Q3 ,
-8"-WS-93 -151-Q3 8"-WS-94 -151-Q3 '
i
)
J
/
r- m. a 4 r, p *
-e -
0" h V , . . h '
.The: size and identification'of.the. butterfly valves that were repaired-as part ,
+
j' 3 .of;thisLdesign change are: ' w 8" - MOV-SW-119
~8".--MOV-SW-219 24" - MOV-SW-115B:
+ 24" - MOV-SW-117 '
24" --MOV-SW-1181
- 24" - MOV-SW-215B '
24" - MOV-SW-217 l
-l I '
SUMMARY
OF SAFETY ANALYSIS. The mechanical ' cleaning of sections of the auxiliary service water piping ~-
~
~
(including the . repair of butterfly valves) and the modifications to ; be performed under this design-change have been carefully reviewed to insure that ! no unreviewed safety question is created.. The review Includes the systen's , final (configuration as well as the construction methods used to implement the work. Specifically, this design change does not: [
- a. Increase the probability of occurrence or the consequences of an accident. or nelfunction of equignent Important to safety and previously ;
evaluated'In the Updated Final Safety Analysis Report.
.The1 auxillary service water. Supply lines'are isolated fran the naln systen by double Isolation valves. Double Isolation will be maintained during~ nost of the conpletion of'this design change by the Installation of 1a blind flange in place of one of the isolation-valves. Since only ~
one of the two ' auxiliary service water. supply lines will be out of 1 service at.any one tine, the other line will be available to . provide make-up to the service water reservoir, should it be required. ,
- b. Create a possibility for an. accident or malfunction of a different-type y than any' evaluated previously in the Final. Safety Analysis Report.
The only modification created by this design change is the addition of T 'snell bore piping. - Failure of small bore piping has already been evaluated in the. Updated Final Safety Analysis Report. Failure of the small bore piping would not cause a loss of adequate service water-flow i to any safety related component. 6
- c. Reduce the nergin of safety as defined in the basis of any Technical Specification.
All naterials used in the modifications, resulting from this design change are in accordance with NAS-1009 which neets with the requirements of the system's original design criteria. The new circulating piping is also Class I seismically supported. Installation of the 3/4 inch circulating lines.ls an insignificant diversion of the service water from systen components. Diversion of this flow will not affect the operation of any safety related canponent, in addition, I this flow can be nanually isolated if deemed necessary for any reason. ? r i
g L p DC-85-01-3 and DC-85-02-3 t ' STEAM GENERATOR ON-LINE CHEMISTRY MONITORING SYSTEM ' NORTH ANNA / UNITS 1 & 2
+
DESCRIPTION = t
=In order to fulfill the terms of the Steam Generator Maintenance Agreement with Westinghouse, a comprehensive on-line steam generator secondary ~ side. water
._ chemistry ~ monitoring system was installed. The new - on-line chemistry Lt monitoring system was installed under two Design Change Packages, DC-85-01-3
'and DC-85-02-3. - Design Change 85-01-3 installed outage related piping / tubing tie-Ins to affected existing systems, portions of the new piping / tubing required for -the new system, the new onsite computer and associated disc and tape drives, peripheral' consoles, terminals, and printers, and an air-conditioned room to house the new computer and disc and tape drives.
Design Change 85-02-3 completed the Installation of the new on-line chemistry monitoring system. The system provides continuous analytical information from four main sample locations in the secondary system (final feedwater, steam generator blowdown - each steam generator, main steam - each steam generator, and condensate) and from two-supplemental sample locations (make-up effluent and heater / moisture separator reheater drains) to both an onsite computer and an offsite Westinghouse (W) computer system for logging and analysis. Information will form the basis of corrective and/or mitigating actions toThis be taken -to' reduce the frequency and duration of outages caused by-out-of-specification steam generator steam purity and water chemistry. Samples from various sample locations are cooled by primary coolers which use either bearing cooling water or component cooling water. The reactor percent power is provided vla isolated-output of the Nuclear Instrtmentation System. SUMr%RY OF SAFETY ANALYSIS These design changes did not create an "unreviewed safety question" as defined in. 10CFR50.59. The addltlon of the steam generator on-1Ine chemistry monitoring system does not;
- a. Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Updated Final Safety Analysis Report. The majority of-the modifications made by Design Change 85-01-3 is QA Non Safety (NS)
The safety related component cooling modifications were performed in accordance with station's procedures for QA Category SR seismic piping.
. Modifications will not affect the safety function of the component cooling water system or create any new accidents by the addition of the new piping.
The modifications made by Design Change 85-02-3 is QA Non Safety (NS) or Non-Nuclear Plant (NNP). These nodifications will not affect the safety functions of any equipment important to safety, in addition, the reactor percent pcwer signal is a fully isolated signal output from the Nuclear Instrtmentation System, therefore it is not safety related or capable of increasing the probability of an accident or malfunction of equipment important to safety.
m e[ b.' Create -the possibility of 'an accident or nalfunction of a different ' type that evaluated previously in the Updated Final Safety-. Analysis [ Report. The modifications installed by this design change are designed b consistent-- with the affected system's. designed bases. The new-piping / tubing tie-Ins Installed by this design change are installed in a manner such that the operations of any of the affected existing systems are not changed. The operation of the Nuclear Instrumentation System is not changed by Design Change 85-02-3 by utilizing an existing isolated : reactor ' percent power signal for chemistry sonitoring alarm Initiation. 'The reactor signal is: fully' Independent of the NIS.
'; c., Reduce the.nargin of safety as defined in the basis for any ' Technical Specification. These design changes do'not-change the bases for any Technical Specifications. . Also, during implenentation of -this design
' change, with reautrenents of the Technical Specifications will not be i vlotated. The tie-Ins to the camponent cooling- water system (DC-85-01-3) which are- done during plant operating modes 5 or 6 will
-not affect system operation. Penetrations and tie-ins are being nede '
.to the containnent penetration coolers.which are not required during these operating nodes. Therefore, the nodifications do not affect' the requi rmnents of Technical Specification 3/4.7.3 for the camponent cooling systam which addresses system operability and survelliance .in operating nodes 1, 2, 3 and 4.
P L
DC-85-29-1 REPLACEMENT OF STATION BATTERIES l NORTH. /MM / UNIT 1 DESCRIPTION i. ' The existing 125V.DC Station Batteries were approaching the end of service life and required replaconent. New calculations indicated that larger capacity i batteries were required to provide an adequate safety nergin. Installing larger-batteries into the existing Battery Roons also required the use of two two-tier racks in each Battery Roan to utilize the minimum space available in the roans. SUtt%RY OF SAFETY ANALYSIS L The proposed modification was reviewed to detenmine if an "unreviewed safety question" as defined in 10CFR50.59, exists. This proposed nodification will replace the existing Station Batteries, with new and larger capacity batteries. It has been detennined that an "unreviewed safety question," as defined in 10CFR50.59, does not exist because:
- a. The implementation of this modification does not increase the probability of occurrence of an accident or nm1 function of. equipment important- to safety and previously evaluated in the Final Safety Analysis Report. This modification replaces existing Station Batterles in Battery Roons 1-I,1-II,1-111 and 1-IV with new and larger capacity batteries which will coordinate with the balance of the present 125V DC systen. The new batteries will be designed, fabricated, and installed j
to neet or exceed the requirenent sections of the original Design Basis
. Document , ;
h
- b. The implementation of this modification does not create a possibility for an accident or malfunction of a different type than- any evaluated previously in the Final Safety Analysis Report. The naterials used in the modifications are totally conpatible (relative to design, quality, and functional requirements) with original equipnent. Therefore, the possibility for an accident or nelfunction of a different type than previously evaluated does not exist.
- c. The laplementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification. New j larger. capacity batteries will be installed in Battery Roons 1-I,1-II, '
1-III and 1-IV enhancing the present 125 V DC system. Safety limits as ; defined in the Technical Specifications are not changed. ; i
DC-8546-3 SERVICE WATER CORROSION RATE MONITORING NORTH Aft 4A / UNITS 1 C 2 The service water. system at North Anna Power Station has undergone severe corrosion attack-due to the aggressive nature of the water and presence of sulfate reducing bacteria. This is in the form of general wall thickness reduction and wide' spread pitting. In order to preserve the Integrity of the service water system piping, and thereby avert or postpone any requirement for further piping replacements, a
- corrosion inhibitor chemical treatment program ha.:,been Initiated.
in order to determine the adequacy of the corrosion control program and to ensure that the high corrosion rate of the service water system piping has been mitigated, a permanent corrosion rate monitoring system was Installed. The corrosion rate monitoring system consists of the following Petrollte Equipment t and Instrunents Group components: one (1) M-6000 nulti-station Instrument chassis, ten (10) Model M-6010 corrosion rate Instrunents, one (1) 12 channel strip chart recorder, ten (10) corrosion rate sensing probes, and six (6) corrosion rate coupons. In addition, apprcximately 3,500' shleided cable with connectors to interface an Inst runent cable between each corrosion rate Instrunent and sensing probe is required. To provido a broad picture of overall service water corrrosiveness, the sensing probes and coupons are Installed in diverse locations in the service water system piping. Each probe and coupon is connected to the service water piping through an existing 2" instrunent tap and Isolatlon valve without affectIng the operability of the service water system at any time. When in service, each probe / coupon wi11 be inserted through Its opened Isolation valve Into the service water flow. To inspect the coupons and perform routine maintenance, each probe / coupon can be withdrawn and the Isolation valve closed to maintain uninterrupted operation of the service water system. SUbNARy~0F SAFETY ANALYSIS The modifications performed under this Design Change have been carefully evaluated to ensure that no unreviewed safety question as defined by 10CFR50.59 is created. The evaluation included the system's final configuration as well as the construction methods used to implement the work. Specifically this Design Change:
- a. Does not increase the probability of occurrence or the consequences of
_ an accident or malfunction of equipment important to safety and previously evaluated in the Final Safety Analysis Report. At no time during the construction, implementation or maintenance of this system will any part of the service water system be made Inoperable. Sensing probes and coupons will be isolated frcm the service water system during construction with an isolation valve. After installation, probes / coupons will be inserted into the flow path without adversely affect ing the system's functionality. Prior to l 1 E . . .
4 3 [p- .! h placing the corrosion rate -probes and coupons in service, thelr b insta11atlons' shall be hydrostatIcally' tested to ensure the service , f water systom pressure boundary Is - in no way compromi sed. - The ' possibility of a probe or coupon failure inside the Service Water -
" piping has also been reviewed. As the retractable probe assemblies are
. greater than }" diameter stainless steel, it is highly unilkely that an assembly could break off, although it could bend and become difficult h
to- retract. If -it were to break or bend it would nost Ilkely result from being struck by a large loose object noving in the Service Water flow path.- In that case, the broken probe assembly would be a minor concern In comparison to the loose object. F lt is more probable that instead of a gross probe failure, an electrode or coupon (coupon being worst case due to larger size) could detach-from its probe assembly and flow downstream., if this is the case, the flow blockage from a piece of steel approximately 2" x .8"'x .125" (for coupon) would be negligible and have no impact on service water' system operability. To minimize the possibility of an electrode or coupon detaching, they will be perlodically inspected to ensure they are in satisfactory condition as part of each scheduled weight . loss / corrosion effect examination. The electrodes, coupons and fasteners will be replaced as necessary,
- b. This modification does not create a possibility for an accident or malfunction _of a different type than any evabated previously in the Final Safety Analysis Report.
The Installation of the corrosion rate monitoring system introduces no new unanalyzed failure modes to the service water system or plant. The_ only significant modification created by this design change is the addition of probes or coupons to existing 2" instrument taps. A + packing leak or rupture of any probe or coupon would constitute a small bore piping failure which has already been evaluated in the Updated , Final Safety Analysis Report. Failure of the small bore piping would not cause a loss of adequate service water flow to any safety related component, j: c. This modification does not reduce the margin of safety as defined in i the bas'is of any Technical Specification. .l e No service water operational parameters shall be affected by the installation of the probes or coupons, in addition, the service water system will remain completely unaffected and operational during all _ phases of project Installation and implementation. I e 4
C JE .i I
+
P > c DC-86-01-3 AUXILIARY' BOILER CONTROLS NORTH AtNA / UNITS 1 & 2
' DESCRIPTION Controls for the auxiliary boiler are pneunatic and located on two independent control panels. In addition there were two Flane Safeguard Control Panels and a_ Deaerator Control Panel which are part of the Auxillary Boller System. The Auxillary Bollers are shutdown for extended periods and the controls are subject- to condensation and dirt buildup due to lack of use. Maintenance is-time consuning as most nelntenance training is perfonned on electronic controls and the Auxiliary Boller- Controls are outdated pneumatic devices.
The Auxillary Boller Controls were replaced with a microprocessor-based control ' systen. The new system incorporated conbustion, boiler feodwater and deaerator
- controls. _ The controls and equipnent located in the flame safeguard cabinets
.was replaced.with a new microprocessor based flane safeguards system with first 3 out annunciation for boiler trip identification.
The existing pneumatic transmitters were replaced with electronic transmitters for process control and Indication. Oxygen analyzer probes and transmitters were installed in_ the boiler outiot ducts to control excess air for proper combustion. Air flow probes and transmitters were installed at the discharge of forced draft fans to neasure-air floa. SUtHARy OF SAFETY ANALYSIS" The Auxiliary Boller Control Systen nodification has been reviewed to detennine if any unreviewed safety questions exist as defined by 10CFR50.59. -The resulting determination is that this modification involves no unresolved safety questions.
- 1. The laplementation of this modification does not' increase the probability of occurrence or - the consequences of an accident or malfunction of equipnent important to safety previously evaluated in the Final Safety Analysis Report.- The modification does not contain equi pnent important to safety that has been previously analyzed.
2~'. The implementation of this modification does not create a possibility for an accident or a malfunction of a different type than any evaluated previously in the Final Safety Analysis Report. The modification involves replacing the pneumatic control system with a state of the art electronic control system. On this basis, no new accident situation is created during the construction phase or post construction operation.
- 3. The Inplementation of this nodification does not reduce the nargin of safety as defined in the basis of any Technical Specification. As this modification involves no equipnent important to safoty and has not increased the potential for radioactive release, no Technical Specifications limits will be exceeded during construction or operational phases of the modification.
5 DC-86-09-1 and 86-10-2 LNtGE BORE SNUBBER LEAK-BEFORE-BREAK MODIFICATIONS NORTH AfNA / UNITS 1 & 2 i l l DESCRIPTION All snubbers in the reactor coolant system are required to be operable to i ensure that the structural Integrity of this system is maintained during and following a seismic event or other < vent which initiates dynamic loads. , The support arrangement for the steam generator and reactor coolant ptano of I each primary loop Included twelve snubbers per unit: four at the upper steam generator restraint, two from the steam generator lower support frame to the j i primary shleid wall, two from the reactor coolant ptarp support framo to the primary shleid wall, and four between the lower steam generator support and reactor coolant ptinp support frames. One reservoir supplied hydraulle fiuld to l the four upper steam generator snubbers, and one reservoir supplied hydraulle fluid to the eight lower snubbers. Virginia Power requested a license amendment to the plant design bahls via , letter No. 86-477A dated November 6, 1986. Based on the technical details of < fracture mechanics evaluation, leakage detection system evaluation, and loading evaluation for reactor coolant system piping, equipment and supports, i Supplemental inforTnation was provided in letters 86-4778, dated February 25, 1987, 86-477C, dated March 12, 1987, 86-477D, dated March 8, 1988, and 86-477E, dated June 10, 1988 to provide additional technical detalls requested by the NRC. r in accordance with this design basis change, both snubbers between the steam generator support frame and the primary shleid wall, both snubbers between the reactor coolant ptsup support frame and primary shleid wall, and two of the four , snubbers between the RCP and SG support frames were eliminated, in addition, two snubbers at the upper steam generator restraint were replaced with rigid struts. The net result is the elimination of Pie of twelve large bore snubbers in each of the three reactor coolant loops and substitution of two snubbers per loop by rlgld struts. - In order to further resolve potential problems associated with snubbers on the RCS, the remaining four snubbers per loop were replaced with new snubbers p having long Ilfe seals, in-situ testing capability, and snubber valves in the piston that eliminates the' need for fluid reservoi rr.. The new snubber cylinders are pressurized to assure effective sealing. All comTon reservoirs for the existing large bore snubbers and associated tubing have been eliminated. SUf4%RY OF SAFETY ANALYSIS The proposed m3dification was reviewed to determine If "ur >rev i ewed safety questions", as defined in 10CFR50.59, exist, it has bee' detarmined that implementation of this madification does not involve an unreviewed safety question. The proposed modification has been reviewed and approved by the NRC.
4
- 1. The implementation of this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of- equipment important to safety and previously evaluated in the Final Safety Analysis Report other than the dynamic effects of a primary coolant piping rupture, which per NRC approval will no longer be required as a design basis.
Generic Letter 84-04 provided the NRC safety evaluation concluding that an acceptable technical basis exists so that the blowdown loads
- j. resulting from double-ended pipe breaks in the RCS primary loop need not be considered as a' design basis. Subsequently, based on their evaluation,- the ' ccanission revised GDC-4 (effective May 12, 1986).to allow the use of leak-before-break to eliminate dynamic effects of the primary loop pipe ruptures from the design basis of PWR's.
The modification associated with the reactor coolant system equipnent
, support snubbers does not increase the probability of occurrence or the t-consequences of an ~ accident or malfunction of equipnent important 'to safety and previously evaluated in the Final Safety Analysis Report (except as noted above). The modifications made to the roactor coolant i:
system equipment supports do not change the design basis (except as " previously noted), functional requirements, or operation of the reactor coolant. system and _Its associated equipment supports. In fact, possibility of accident due to probability of malfunction of snubbers is reduced due to reduction of a total nwber of snubbers.
- 2. The. Implenentation of this nodification does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the Final Safet' Analysis Report. Analysis has shown that regulatory requirements apt cable to this design change have been met:
- a. Nonnal operating plus selsmic loads are sufficiently small to ensure that an existing through wall crack in the primary piping would remain stable during a selsmic event,
- b. The reactor coolant system equipment, piping, and supports continue to have acceptable margins of safety under licensed loading conditions other that the now-eliminated RCS main loop rupture,
- c. The leakage detection system has been found adequate to detect a throughwall crack, with margin, before it could become unstable during or after a selsmic event.
- 3. This modification of supports associated with the reactor coolant system does not reduce the margin of safety as defined in the basis for any Technical Specification. This design change does not change the basis of any Technical Specification. During implementation of this change, the requirements of the Technical Specifications shall be met.
The sequence of snubber renoval and replacement will ensure operability l' of all three reactor coolant loops should a seismic event occur during cold shutdown (Mode 5) or refueling (Mode 6). The NRC has reviewed and approved the factors of safety associated with the component supports. I
e L
'DC 87-31-1 and 87-12-2 l
ANTICIPATED TRANSIENT WITHOUT SCRAM l NORTH ANNA / UNITS 1 & 2 l r DESCRIPTIO4 An Anticipated Transient Without Scram (ATWS) 1s a postulated accident sequence lL initiated by a loss of feedwater, loss of condenser vacuum, loss of offsite E
- ^
power, or other design basis expected translent coincident with a failure of the Reactor Protection System (RPS) to shut down or scram the reactor. 1 i Anticipated Translent Without Scram (ATWS) Mitigation System Actuation j Circultry (AMSAC) is required per 10CFR50.59.62 to provide a backup turbine [= trip and auxillary feedwater Initiation which would operate if a failure of the existing Reactor Protection System (RPS) occurs. The RPS Includes the 7300 ; Process Instrunentation and Control System (7300 System), and the Solid State Protection System (SSPS). The AMSAC nust be Independent of the RPS and be 1 diverse from sensor output to final actuation device.
'?
This modification uses three progranmable logic controllers (PLCs) (one per o steam generator) to implement an AMSAC which neets and exceeds the Westinghouse ! Owners Group (WOG) design, approved by the NRC per WCAP-10858. This resolution ' enhances the WOG design by using all steam generator narrov range level sensors 4 to input to the logic circuits. The WOG design uses one narrow range level i sensor from each steam generator (different channel on each steam generator). Both designs use two turbine impulse chamber pressure sensors,.one from each of two channels.- The coincidence of two out of three level sensors for two or three stean generators detecting 13 percent or less of level while two out of T two turbine impulse chanber pressure sensors detecting a pressure (load) greater than 38 percent Initiates an AMSAC. This modification uses all nine ; available steam generator narrow range level sensors (three per stesn . generator). This enhancenent allows one sensor to be placed in bypass for i maintenance. while still allowing a failure of another sensor without causing AMSAC. The two out of three coincidence logics taken twice, reduces the > probability of inappropriate (spurious) turbine trips, c When the AMSAC logic for any PLC is satisfied, the PLC Initiates a pre-trip 27 second tiro delay to allow the RPS to operate, trip the turbine, and initiate D auxillary feedwater flow to amlntain short tenn reactor coolant pressure below 3200 psig, if two out of the three PLCs time out (27 seconds), then isolating output relays are energized. These safety-related output relays trip the turbine and initiate auxiliary feedwater. Availability of heat sink is additionally assured by using the safety-related output relays to close all 1 steam generator blowdown and sanole isolation valves. The control rod t notor-generator sets are tripped by the nonsafety-related output relays to e provide a redundant and diverse reactor trip.For trips when the unit is above 38 percent load, it is anticipated that the steam generator levels will drop below and stay below 13% of the narrow range level even though the RPS has functioned properly. l This activation of AMSAC witi be due to post trip steam generator Inventory shrink and not an ATWS. The activation of AMSAC in this situation is expected.
t i. L l Ln 9 St)9%RY OF SAFETY AtALYSIS -
; The Insta11atlon of the AMSAC modifIcatlon does not constitute an "unreviewed i
safety question" as defined in 10CFR$0.59. -
< i
- a. The inplanentation of this nodif1 cation does not increase the {
( probability of occurrence or the consecuence of an accident or malfunction of equipnent important to safety and previously evaluated - : In the Final Safety Analysis Report. l This. equipment will be used to provide additional backup to the existing reactor protection system (RPS), therefore, reducing the probability of and reducing the consequences.of RPS failure. : i
- b. The ~ lmplanentation of this nodlfication does not create a poss'bility [
h for'an accident or a nelfunction of a different type than any evaluated : previously in the Final. Safety Analysis Report. 1 This design change- does not !ffect the operation of safety-related h equipnent or systens, the availabi;ity of safety-related power sources, and does not change existing instrumentation in the Control Roon. This i design change adds new Interlocks to auxillary feedwater- pumps . start, ;
- c. turbine trip and reactor trip circuits. However, this design change 1 g- enhances the safety-related systens by providing backup protection to ;
[ the reactor coolant system safety limits to prevent overpressurization ; by initiating auxillary feedwater, turbine trip, and a reactor trip on * -r complete loss of feedwater in the existing RPS falls. 4
- c. The Inpl enentat ion of this modification does not reduce the nergin of j safety as defined in the basis for any Technical Specification.
This nodification provides backup protection to prevent overpressurization of the RCS If the existing RPS falls and there is a ' complete loss of feedwater. Therefore, this nodlfication-provides an improvement In the' ability to provide a- nergin- of safety for RCS i overpressurization, i r t h i
- V
'I
DC-87-22-1 and DC-87-23-2 CRDR - SW AND RS ICC RELOCATION NORTH ANNA / UNITS 1 & 2 DESCRIPT104 To comply with NUREG-0700, NUREG-0801, Supplement I to NUREG-0737 and STD-GN-0005 (Hunan Factors), instrunents and controls were relocated. These modifications correspond to Corrective Actions delineated in the NRC letter (Serial No. 85-26BC) and are as follows;
- a. CORRECTIVE ACTION 31E:
The Outside Recirculation Spray Puup 1-RS-P-2A (2-RS-P-2A) is controlled fran the " Orange" section of the Main Cantrol Board, 1-El-CB- 05 (2-El-CB-05), and discharges to the Recirculation Spray Heat Exchanger 1-RS-E-1D (2-RS-E-1D). Before the Installation of Recirculation Spray indicator (T1-RS156B, TI-RS256B) and the Service Water Indicators FI-SW1000 (F1-SW2000) and TI-SW1000 (TI-SW200D) associated with this Heat Exchanger, 1-RS-E-ID (2-RS-E-ID) were located on the " Purple" section of the Main Control Board, 1-El-CB-05 (2-El-CB-05). The sane is true for Recirculation Spray Punp 1-RS-P-2B (2-RS-P-2B), In reverse. This pump is controlled f ran the " Purple" section of the Main Control Board, 1-El-CB-05 (2-El-CB-05) and discharges to Recirculation Spray Heat Exchanger 1-RS-E-1C (2-RS-E-1C). The Recirculation Spray and Service Water Indicators TI-RS156A (T1-RS256A), F1-SW100C (F1-SW200C) and TI-SW100C (TI-SW200C) associated with this heat exchanger were located on the " Orange" section of the Maln Control Board, 1-El-CB-05.
- b. CORRECTIVE ACTION 13E:
The following control switches, Indicating and systen status light to the Vertical Board Section 1-3 (2-3) were relocated. New control switches and indicating light were included in the modification. 1-RS-P-3A (2-RS-P-3A) 1-RS-P-3B (2-RS-P-3B) MOV-RS100A (MOV-RS200A) MOV-RS100B (MOV-RS200B) MOV-RS101B (MOV-RS201B) MOV-RS101A (MOVRS201A) The following Casing Cooling Tank Level and Temperature indicators to the Vertical Board Section 1-3 (2-3): LI-RS-103A (LI-RS-203A) L1-RS-103L (LI-RS-2038) TI-RS-100A (TI-RS-200A) TI-RS-100B (L1-RS-200B)
StJ4MRY OF SAFETY ANALYSIS The CRDR - SW and RS 1&C Relocation nodlficotton does not constitute an "unreviewed safety question" as defined in 10CFR50.59. This modification relocates Instrunentation and controls on existing panels. The relocation of instrumentation within this design change does not alter the function or operation or existing safety related equipment or systems, will not create a possibility for an accident or a malfunction different from those evaluated in the Final Safety Analysis Report and will not alter any setpoints, function, or operation of any safety systems. Safety limits as defined in the Technical Specifications are not changed nor the margin of safety as defined in the basis of any Technical Specification.
I L l DC-87-24-1 and 87-25-2 l PRESSURIZER SPRAY VALVE POSITION INDICATION NORTH Aff4A / UNITS 1 & 2 DESCRlPTION To comply with NUREG-0700, NUREG-0801, Supplenent 1 to NUREG-0737, and STD-GN-0005 CHuman Factors), Pressurized Spray Valve Position Indication was added. Red and green Indicating lights were Installed on the neln control board above the respective auto-manual controller station. These nodifications correspond to Corrective Action 1.2E delineated in NRC Letter Serial No. 85-768C. l $UMMARY OF SAFETY ANALYSIS The Pressurizer Spray Valve Position Indication nodification does not constitute an"unreviewed safety question" as defined in 10CFR$0.59,
- a. The Inpl ementat ion of this nodification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipnent important to safety and previously evaluated in the Final Safety Analysis Report.
This nodification installed Indicator lights for the Pressurizer Spray Valves. The Installation of this DCP will increase the chance of operator error / delay, therefore decreasing the consequences or the probability of occurrence of an accident,
- b. The implementation of this modification does not create a possibility for an accident or nelfunction of a different type than any evaluated previously in the Final Safety Analysis Report.
The Installation of Indicating lights provides a nore reliable neans of tronttoring station operation from a "hurnan factors" viewpoint. during and after a design basis event and will not create a possibility for an accident or a malfunction different from those evaluated in the Final Safety Analysis Report,
- c. The Inplementation of this nodification does not reduce the nargin of safety as defined in the basis of any Technical Specification.
The addition of valve position Indicating Ilghts does not alter any setpoints, function, or operation of any safety system. Safety limits as defined in the Technical Specifications are not changed nor the nergin of safety as defined in the basis of any Technical Specification. l
l4 l DC 87-28-1 and 87-29-2 CRDR INSTRUMENTATION RANGE CHANGES NORTH AtNA / UNITS 1 & 2 DESCRIPTION During the perfonnance of the Control Roan Design Review, the Irdicators for the parameters listed belav were Identified as not reeting the guidelines of NUREG 0700, "Guldelines for Centrol Roan Design Reviews". These Indicators did not have the appropriate range for reading the values required by the Energency Operat ing Procedures. Operators had experienced difficulty in reading, monitoring, and verifying Indications which nay have resulted in delay or error in the operator's response. The identifled paraneters are: CContalnnent Pressure oPrimary Component Cooling Water Punp Discharge Pressure oPrimary Canponent Cooling Water Heat Exchanger Flav Charging Flow oCharging Letdavn Flow The ranges and scales of the pressure transmitters were increased and the flav
. loop indicators were expanded or linearized to enable the operator to better read then.
SUMMARY
OF SAFETY ANALYSIS The proposed modification was reviewed to determine if an "unreviewed safety question" as defined in 10CFR50.59, exists. This nodification changed the ranges of the containment pressure, primary component cooling water pump discharge pressure, primary canponent cooling water heat exchanger flow, charging flow, and charging Ictdown flav transmitters and the scales of their respective Indicator faces, it has been detennined that an "unreviewed safety question", as defined in 10CFR50.59, does not exist because: A. The implementation of this modification does not increase the probability of occurrence or the consequences of an accident or nelfunction of equipment inportant to safety and previously evaluated in the Final Safety Analysis Report. Any actuation setpoints associated with the subject systens will renaln the sane. The reason for the range and scale changes is to make the Indicators rore easily read by the operator. B. The laplenentation of this nodification does not create a possibility for an accident or neifunction of a different type than any evaluated previously in the Final Safety Analysis Report. The Indicator scale changes are being implenented to aid the operator during the time he refers to the Energency Operating Procedures. C. The lapl ementat ion of this modification does not reduce the nergin of safety as defined in the basis of any Technical Specification. With laproved infornat ion available to the operator, any nargin to safety will be more easily identified. l
I g DC-88-01-1 k o [ I BENCHBOARD Et#iANCEMENTS NORTH Aft % / UNIT 1 i L To comply with NUREG-0700, NUREG-0801, SUPPLEMENT 1 to NUREG-0737 and ! STD-GN-0005 CHunan Factors) -Instruments and Controls were relocated. These
~ nodifications correspond to corrective actions delineated in NRC Letter Serial
- 16. 85-268C and are as follows:
- 1. !
Corrective Action SE'
- Tte reactor coolant system wide -and narrow range pressure Indicators' !
F P1-1403A and B, located on the top row of Section 1-1 of the Vertical l Control Board were rescaled and relocated.. The Indicators were mounted too ' high on the vertical board for the 5th percentile fenale operators to road-
" accurately due to the curvature of the Westinghouse VX-252 cover plate. 'I Additionally, the- scaling of both narrow range Indicators did not lend v Itself to accurate reading of reactor . pressure, and wide range reactor pressure. Indicator PI-1402A contains a x100 scale multiplier while the opposite train reads pressure directly without a multiplier.
This ; corrective action required the relocation of one train of the wide and ; narrow range RCS pressure Indicators, PI-1403A and B, respectively, h - rescaling of wide range pressure Indicator PI-1402A and both trains of narraw. range Indicators PI-1402B, and PI-1403B.
- 2. Corrective Action 40E - Annunciator.Pushbutton Relocations A. ~ In Section 1-1 of the benchboard, three annunciator pushbuttons on the bottom row of the vertical portion were relocated.. The three !
pushbuttons are, " Annunciator Slience", " Annunciator Test" and !
" Annunciator Acknowledge".
B. In Section 1-2 of the benchboard, fcur annunciator pushbuttons on the vertical section were relocated. 3.. Corrective Action 56E e ' A portion of the plant permissive status lights which were located in annunciator panel IP, as well as a portion of the trip status lights which , were located in annunciator panels IL, and IM, were contrary to the other - Indicators in these panels and the other annunciator panels in that they i were on during nornal plant operations. The pennissive status lights are ' presently spilt approxinately 50/50 on and off and were scattered over the annunciator panel. Lights which were on at normal plant 100 percent power i operation were rewired at their respective annunciator output relay to be off under nornal conditions. The appropriate windows were re-engraved to o indicate the change in status and to standardize the annunciator acronyns used. Additionally, the permissives and trip functions were distributed about the annunciator panels in a random nenner. This DCP rearranged and regrouped the status lights such that they change state in a nore
e secuential nenner during a nonnal plant startup or shutdown, and like functions were grouped together inside their respective panels by function and channel.* 4 Corrective Action 68E t The Residual Heat Renoval Bypass Flow Controller FC-1605C was located on the bottan left hand corner of Section 1-2 of the Vertical Control Board. It is used in conjunction with HCV-1758 through flow controller HIC-1758, to control the cooldown rate of the plant during a nornal plant shutdown belav 3500F. In It's fonner location, it was necessary for the operator to leave the area in front of the benchboard in order to nentpulate the nanual ; controller. Additionally, because it was located on the bottan row of the vertical board, it was difficult to operate and monitor accurately since the flav Indicator, F1-1605, is located in the top rov of Indicators on Section 1-2 of the vertical board. The flow controller was relocated. '
- 5. Corrective Action 74E The' Safeguards Insert located on the vertical portion of Section 1-1 of the benchboard contains the manual safeguard initiation control switches as well as block and reset switches for various safety injection, containment spray, and containnent isolation functions. Four of these block and reset switches have a dual function, that is, they are three position, spring return to conter, switches with both a block and reset function on the sane switch. These switches are the Lo-Lo TAVG and Lo-Lo Pressurizer Pressure Safety injection, block and reset switches, which either block or reset a Safety injection Signal to the Reactor Protection System during plant startup and shutdown. These switches could change state if allowed to snap back to the center position from either the block or reset positions which could have resulted in an unwanted safety injection initiation or the block of a safety injection signal. The four switches were replaced with eight, two oosition spring return to left position switches with either a block or a reset function. Additionally, the ramalning switches on the Safeguards insert were arranged in a similar manner for equipnent in Trains A and B which perfonn similar functions. This increased operator reliance on memory for switch locations. Therefore, the renelning switches on the Safeguards Insert were rearranged in a nere logical manner with switches perfonning similar functions in both trains arranged such that the Train A switch is directly above the Train B switch, s
SUMr%Ry OF SAFETY ANALYSIS The Benchboard Enhancement modifications do not constitute an "Unreviewed Safety Question," as defined in 10CFR50.59.
- a. The incl enentat ion of these nodifications does not increase the probability of occurrences or the consequences of an accident or nelfunction of equipnent important to safety and previously evaluated in the Final Safety Analysis Report.
r These modifications will relocate existing canponent s to different locations inside the control roan, . and for the Flash Evaporator equipment fram the control room to the Flash Evaporator. -The relocation of these components within this design change package does not impact the operation of these conponents or systems. The modifications performed to the Main Control Boards has been seismically analyzed and the components are seismically installed so as not to degrade the selsmic qualification of the Vertical Control Boards and Benchboards. The Benchboard Enhanconents nodification Interfaces with safety-related equipnent and was performed during an outage. Testing of the affected components was complete prior to station operation.
- b. The Implementation of this modification does not create a possibility for an accident or a nelfunction of a different type than any evaluated previously in the Final Safety Analysis Report.
This design change does not affect the operation of safety-related equipnent or systens or the availability of safety-related power sources. The design change does affect the location of instronentation and components inside the control room. However, the. relocation of instrunentation is impi snented to sat!sfy the Human Engineering Deficiencies identified by Virginia Power's Control Room Design Review Program, and thus provides a more functional arrangenent of controls and Indicators on the neln control boards. The modification does not add or delete any safety-related or important to safety components from the control roan. The Benchboard Enhancenents modification was Installed during cold shutdown or refueling modes. Testing of the components relocated was completed prior to station operation,
- c. The implanentation of the modifications addressed by this design change package does not reduce the nergin of safety as defined in the basis of any Technical Specification.
The relocation of the camponents addressed by this design change provides the operators with a cicarer understanding of the plants condition and reduces the potential for inadvertent operation of incorrect cont rol devices. This change did not add or delete any safety-related Indicators or control devices fran the control roan. The Benchboard Enhancenents modification was Installed during cold shutdown or refueling modes. Testing of the canponents relocated was conplete prior to station operation. l
y DC-88-05-3 I
. INSTALLATION OF TH]RD SYSTEM RESERVE TRANSFORMER l
NORTH AtNA / UNITS 1 & 2
. DESCRIPTION Prior to the Implanentation of this modification there were only two 500-34.5 l
KV system reserve transformers (SRTs) which served as the two electrically and physically separate sources
' offsite for the reserve station service transforrers (RSSTs) at North Anna. SRT #1 feeds RSST C which serves energency buses 1H and 2J. SRT h2 feeds RSSTs A ar.d B which serve emergency buses 1J and L
2H, respectively. The existing design with two SRT's minimized the chances of a single failure removing offsite power to both energency buses of a single 1 . unit and satisfles the-original design criteria of 'having two independent offsite power sources. The existing Unit 1 energency bus ties add to the nunber of offsite sources available to the Unit 1 energency buses. ~ Flexibility
- of the - ' energency bus ties is increased further by feeding bus 1H from SST 1B L and bus 1J from SST 28 Instead of from the sane unit's buses, thereby increasing the probability of the availabl11ty of at least one source.
The loss of the offsite power could prevent operation or cause a trip of these units. The intent of relocating the 500/230KV transforner and replacing the 230KV breaker was to reduce the risks to the plant from transnission system faults. .The Third SRT provides backup capability for the 34.5 KV reserve station service buses. i SUtt%Ry OF SAFETY ANALYSIS
-The Third SRT nmy be utillzed as either an Installed spare or a permanent feed to RSST A and a backup supply to RSST B and C. It is concluded that these changes ido not result In the previously acceptable safety questions as defined in 10CFR50.59 exist. Specifically, with regard to 10CFR50.59:
- 1. The installation of the equipment specified above does not increase the probability of occurrence or the consequences of an accident as previously evaluated in the Updated Final Safety Analysis Report (UFSAR). This nodification nelntains the electrical and physical independence of the offsite sources, utilizes equipnent of sufficient capacity and capability, and improves the voltage proflies for certain GDC-17 scenarlos ehlle imposing specific operating restrictions. The Installation of a' third SRT provides an additional source of offsite I
power and thereby reduces the probability of a loss of offsite power.
- 2. The Installation of the equipment specified above does not create a possibility for an accident not previously evaluated in the UFSAR.
This nodification provides an additional source of offsite power to the Reserve Station Service Transforners (RSSTs). However, this can not j totally eliminate the possibility of a loss of offsite power which has been evaluated in Section 125.2.9 of the UFSAR. H li E
t i f
- 3. .The Installation of the equipnent specified above does not reduce the l
f nergin ' of safety as defined in the basis of any . Technical Specification. 'This nodification provides an additional source of offsite power and increases the ability to maintain two reautred
-independent offsite sources of power. The construction outages required by this modification must adhere to specific requirements.
Section 8.0 of this ER & SA delineates the operating restrletions imposed by the new switchyard arrangement. o. 1
DC-88-13-3 CONTAltNENT RADIATION MONITORING SYSTEM MODIFICATIONS NORTH ANNA / UNITS 1 & 2 DESCRIPTION There were three (3) containment high range radiation monitors installed in each unit at North Anna Power Station; RM-RMS161, 165 and 166 for Unit I and RM-RMS261, 265 and 266 for Unit 2. Monitors RM-RMS161 and 261 had becono unreliable-and difficult to nnintain. RM-RMS161 and 261 were in need of repair, replacement, nodification or renoval. These two radiation monitors are located outside the personnel hatches. They were Installed as original equipment with the objective of monitoring the containnent atmosphere post-LOCA. Subsequently, NUREG-0578 required the Installation of contalmnent high range radiation monitors inside contalnnent for the same purpose. To conply with NUREG-0578, radiation nonitors RM-RMS165, 166, 265, and 266 were Installed inside of the containnent polar crane wall approxinately 125 degrees apart to provide physical separation. . There is no requi renent for nonitors RM-RMS161 and 261, as their original design function is now perfonned by RM-RMS165,166, 265, and 266, which also fulfills the requirenents of Regulatory Guide 1.97. However, it is desirable to retain these nonitors outside the personnel hatch (RM-RMS161 and 261), but nodified to function at radiation levels expected during nonnal operation. This is particularly useful during outages, for ALARA concerns. The two nonitors, RM-RMS161 and 261, were nodified in order that they function at a new range of 10*4 R/hr. to 10 R/hr. This was acconpilshed using Westinghouse canponents to change the detector, and the rate noters at the ; local readout nodule and in the control roan. Also, a new printed circuit board (Rate Board Assembly) for the control roan electronics drawer was Installed and a modification to the power supply shunt resistor was nede. SUbMARY OF SAFETY ANALYSIS The modification of radiation monitors RM-RMS161 and 261 did not constitute an *
"unreviewed safety question," as defined in 10CFR50.59.
- a. The nodlfication did not increase the probability of occurrence or the consequences of an accident or nalfunction of equipnent inportant to safety previously evaluated in the Final Safety Analysis Report.
The modifications to the instrunentation within this design change did not impact the operation of existing safety-related equipnent or systens. The nodified instrumentation provides indication to the operator to allow him to assess radiation levels outside of the personnel hatch during nornal plant operation. Monitoring of the containment atmosphere post-accident is acconplished per NUREG-0578 and R.G. 1.97 by rediation nonitors RM-RMS165, 166, 265 and 266 which are inside containment, n
- wy <
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.. \ a s T .b. 'Th'Is modificat'lon'did not create a possibility for an-accident or a -
'f malfunctlon of -al different type.than any evaluated previously in the .!
Final Safety' Analysis Report.
> -l b JThefmodified Instrunentat ion is'-not' required to satisfy R.G. l.97 la {
requirements.- They provide a reliable means; of monitoring' radiation; ) levels at a spectfIc'locatlon durIng normal operatlon.' .; I ; p' " . cl 'Thismodificationi does not' reduce'the margin of safety as defined.In' t the basis of any Technical Specification.- J i [ The. modified Instrunentation provides continuous Indication in the' l
- control room of the dose rate in the area outside of_ the personnel- if hatch. .for 'each! unit.
Technical
~
Specification 3/N 3.3- Includes l c ' contalrvnent high, ' range' monitors inside- containment- but not- those- t:
'outside. of contalrvnent. Safety limits as defined in the Technical ~ !
Specification are not changed. t
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DC-89-02-2 CONTAltfENT ELECTRICAL PENETRATION SECONDARY PROTECTION NORTH At#4A / UNIT 2 DESCRIPTION To comply with position C.1 of Regulatory Guide 1.63, Rev. 2, secondary protection for circuits utilizing containnent electrical penetration is required to preclude the possibility of penetration seal danage which ney result from penetration feed-through conductor overheating due to circuit t overcurrents coincident with single failure of the circuit protection device, i The following is a list of equipment that was added or revised concurrent with DCP 81-521B which first implanented secondary overload protection but was not in conpilance prior to this design change: 2-RC-TV-2018-1 2-RC-TV-2018-2 t 2-RC-TV-202B-1 2-RC-TV-202B-2 2-RC-TV 201A-1 2-RC-TV-201A-2 2-RC-TV-202A-1 2-RC-TV-202A-2 2-HR-P-214 sum ARY OF SAFETY ANALYSIS The Installation of the secondary penetration protection modification does not constitute an "unreviewed safety question" as defined in 10CFR50.59.
- a. The impl ementat ion of this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipnent important to safety and previously evaluated in the Final Safety Analysis Report.
This modification adds secondary contalnnent electrical penetration protection. The time-current characteristics of the existing primary protective device and the new secondary device being added to each circuit coordinate with the t ine-current characterist ics of the associated electrical penetration. Proper coordination of both devices reduces the probablilty of an electrical penetration seal failure which could result from penetration conductor overcurrents coincident with failure of a single circuit protective device. Reducing the likelihood of a penetration seal failure similarly reduces the probability of a breach of containment integrity under those conditions when such Integrity is required. ' No systen design functions have changed.
- b. The impl enentat ion of this modification does not create a possibility for an accident or a nalfunction of a different type than any evaluated previously in the Final Safety Analysis Report.
l
~
- 0. ~
1
'This design change does rvyt offect the operation of safety-related
'i
'equipnent' or systeers, : the availability of-: safety-related power sources ~, ~}
and' does not change existing Instrwentation In the~. Control Room. In ,
. fact, the addition of properly coordinated electrical -penetration l l , secondary.< protection enhar.ces the reliability of_the penetration seal a to provide an' adequate containment- pressure boundary ; under- accident . ;
conditions coincident- with a fault or overcurrent on a penetration i rg circuit. ' 1: c.- The Irmlementation of. this nodification does not reduce the margin of
. safety as defined in_the basis for any Technical Specification, b This nodlfication-increases the nergin of. safety for .the ' contalnnent"
-electrical penetrations by adding secondary protection..
U ' As noted above, 'the addition of secondary penetration protection b enhances .the reliability of the penetration seal naterial' by_ reducing h, the Ilkelihood !of seal the rnal danage resulting from penetration L conductor overcurrents coincident with a~ single failure of a circuit ~ protective device, t-I i L q
DC-89-07-1 AIR. EJECTOR AFTER CONDENSER CHECK VALVE INSTALLATION NORTH AN'4A / UNIT 1 DESCRIPTION As a result of the February 1989 Unit I reactor trip and erratic air ejector radiation nonitor Indication, the air ejector after condenser piping was identified as requiring an enhancement. An actual high radiation condition did exist due to a steam generator tube plug failure. The air ejector radiation ronitor Indication becane erratic when the air ejector discharge was diverted fran atnosphere to containment. This caused air to be drawn into containment from the turbine building via the after condenser loop seal. This Design Change Installed two 4' swing check valves, one just upstream of each after condenser loop seal drain. The check valves will not interfere with normal system operation,-however they will change system operation when the air ejector exhaust is diverted to containment. The check valves will prevent reverse flav, and thus will not allow air from the turbine building to be drawn into containment. With this Design Change Installed,.the air ejector radiation nonitor Indication should be stable under all conditions. SUt4%Ry OF SAFETY ANALYSIS This nodlfication does not coastitute an unreviewed safety question as defined in 10CFR50.59 since It does not:
- 1. Increase the probability of occurrence or the consequences of an accident or malfunction of equipment inportant to safety and previously evaluated in the updated Final Safety Analysis Report. The modifications nede do not change the operation or ability of equipment important to safety to perform their safety functions.
The new check valves have a pressure /tsnperature rating quallfled for this systen. Pipe stresses and supports were evaluated and found to be acceptable for this nodification. The check valves will cause the system to operate as originally Intended.
- 2. Create a possibility for an accident or malfunction of a different type l
than any evaluated previously in the Updated Final Safety Analysis Report. The check valves will be installed in accordance with construction code ANSI B31.1, and they will allow the systan to function as originally Intended. The drain traps are redundant to the loop seals and are not required.
- 3. Reduce the nargin of safety as defined in the basis of any Technical Specification. The nodification does not affect safety related equipnent or equipnent identified in the Technical Specifications.
f-l EWt SLM4ARY
'EWR 81-316A DESCRIPTICN i
This EWR updates the tbrth Anna Setpoint Docment to correctly reflect the
, reset valves of the following switches:
1-EG-PS-610H,J / 611H,J 1-EG-TS-607H,J / 610H,0 / 611H/d EWR 81-316 specified the correct reset valves but did not provide for the updating of the Setpoint Docment, i-SAFETY ANALYSIS SLM44RY This EWR updates the ibrth Anna Setpoint Docment to reflect correct reset values EWR. of the affected switches. There is no nodification involved in this 6 i 1
i ! EMt SW9Wty EWR 82-272C DESCRIPTIGd Modifications in the Electrical Shop upper area fmezzanine) and Break Rotm hcve 7 '~ affected the existing sprinkler system coverage. Modifications to the sprinkler system are required to provido proper coverage. The mezzanine area i sprinklers in the area with a new ceiling are required to be reworked to ' provide protection below.the ceiling. The two walls of the room are required to be extended above the ceiling to the bottcm of the roof deck to isolate the space above and allow the sprinkler protection to be relocated below the ce l l irig. [ ! The existing sprinkler system In the Electrical Shop lower area is being extended into the Break Rocrn. Two heads are being added below the ceiling in the Break Room. 1 SAFETY MMLYSIS SW9%RY The modifications described in this EWR are being completed in the Electrical Shop of the Service Bullding. The modifications will require temporary isolation of the Service Building Warehouse fire sprinkler system which includes the Electric Shop. The trodifications to the existing sprinkler system sprinklers and the implementation of the modifications do not impact any Station system important to safety.
Eh8t$@9%RY EWR 82-483A , DESCRIMIOf This EWR was used for the Installation of a reach rod for valve 1-55-572, also known as valve "A", which is located in the Auxiliary Building Sample Room. This was in 11eu of relocating the valve itself which would increase radiation levels in the sample room'after an accident. SAFETY ANALYSIS SWe%RY This is not a safety rotated trodification. The Technical Specifications and the design basis as described in the UFSAR are not affected by this rrodi ficat ion. The system operation is not adversely affected, the addition of the reach rod is only for the aid of the user.
O
-+---- m4 Det SWMety Et.R 83-078A DESQtiPTI(M During disassembly of 10-SW-MOV-103C it was noted that the Disc was corroded.
The corrosion was in the seating area and would hinder obtaining a proper seal. The original' disc was carbon steel, however use of a stainless disc would eliminate the corrosion problems and improve the sealing characteristics of the valve. the new disc weighs the same as the original disc, therefore seismic Integrity is not affected. SAFETY ANALYSIS SGMUtY The ability of the valve to open on CDA Signal in the required time to provide design flow and provide containment integrity has not been adversely affected. Valve is designed for the conditions in SW System. Selsmic integority of valve and system are maintained. e
m EWL SL99%RY EWR 83-190F _DESCRIPT!CN This EWR Justifies leaving pressurizer heaters 29, 45, 46, 77, and 78 disconnected. in addition, heaters 46, 77, end 78_will be moggered by this EWR to evaluate insulation Integrity. SAFETY ANALYSIS SLM %RY There are no unreviewed safety questions. No Technical Specifications are violated as a result of this EWR, The design basis as defined in the Technical-Specifications will be maintained.
f t-EWt SIMMRY EWR 83-1901 v i DESCRIPTION j. This EWR Justifies disconnecting pressurizer heater tJo. 73. The conclusion was based on a review of IJorth Anna Technical Specifications. Heater 14 o , 13 will a;so be terporarily disconnected. ' s l
- i :
SMTTY ANN.YSIS SLMWtY i As~ a result of this modification, there will be five ($) heaters that are disconnected, one temporarily This will not pose any safety concern and falls within the bounds of the Technical Specifications, , 1 1-W
L M My EWR 83-190J I DESCRIPTICN This nodification reconnects pressurizer heater ib. 13 that had been terrporarily disconnected, it was disconnected because the connection between !' the heater and cabic was bad. Once the damaged portion of the cable was recoved and upon satisfactory resistance and megger tests, the heater was able to be l reconnected. SAFETY ANALYSIS SLM %RY This EWR returns equipment to its original configuration, which has already been analyzed for sefety. This EWR only improves upon the safety aspects of the system as outlined in the Technical Specifications.
i M MY ' !, EWR 83-276A DESCRIPTICN i This EWR provides for the docunentation of certain breaker Inspections. The breakers that were looked at were Westinghouse JA breakers with Al-Cu ; teminals. The reason for the inspection was to ensure that the proper type of - Al-Cu tenninal was Installed on the breaker, and that ' the Westinghouse guidelines for tenninating aluminun pcwer cables to Al-Cu terminals have been ' followed. SAFETY M MLYSIS S M Y This EWR will improve the reliability of secondary circuit protection. The ' margin of safety of the breakers will be increased by this EWR,
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l h l- - l. EWR 84-275 EWR Sut WtY i DESCRIPil(N This EWR provides the installation Instruction for a temporary corrosion mnitoring tystem. A corrosion inhibitor system supplied by Calgon will be permanently installed on the Service Water System. The reason for this EWR is to assess the effectiveness of the corrosion inhibitor system that will ultirnstely bo Installed. I i-L N ANALYSIS s m y The Installation of the temporary nonitoring system in the Service Water Pumphouse will not affect the safety aspects of the system. The terrporary system can be isolated from the Service Water System. The temporary Installation is not safety related. O l
p- " l i i EWR 84-307 . L EWR SLM%RY t DESCRIPTION This EWR Service ls for Water the initial reservoir. In addition addition, of corrosion this inhibitor _ chemicals to the EWR installs temporary equiprtent ! Including storage tanks, puros, tubing, etc. for the addition and maintenance j of proper chemical concentration, t I' SAFETY ANALYSIS SLM MRY Testing by the Chemistry Department has determined that corrosion inhibitor treat, ment will have no harmful effect on the Service Water piping. There are no unreviewed safety questions. The margin of safety will not be decreased by this modification Chemical addition is not addressed in the Technical Specifications.
(q k 9-EWR SW9%Ry V; EWR~85-320 & A,B,C,D,E-r DESCRIPTim i;
-The. rod control cabinets (Units IC2) and pressurizer heater control cabinets (Unit 1 only) have expertenced heat related failures as a result l of ' extremely
-high ambient temperature during the surmer months. As a short term fix to-
. alleviate the problem scrne of the ventilation ducts in the area will' be rerouted to the area beneath the. cabinets. This will reduce the Internal cabinet temperatures to c. more acceptable' level' and decrease the incidents of failure within the cabinets, q.
SAFETY ANALYSIS SW9%RY The failure of the ductwork, spin collars, and supports, with the exception cf the seismic support installed by this modification, will have no effect on surrounding equipment. All work with respect to duct modification and rerouting has been ana'lyzed for dafe Installation, and. where required, has been l seismically designed. 3 l. i p i i .. i It 85-3206ABCDE t p i
?y I :* 2 h EWR - SIA9Wty EWR 85-467 C A- ' (;I- 3 DESCRI.*TICH - The reason for this modificatton is to. trodify .the control circuit ~y of .
, MOVs-1535, 1536,.2535, and 2536. This is being done to ensure that vil ve direction' reversal will. not occur until the valve has fully stroked. The modification will be inplemented by adding a Junper in ~ the control circuitry of the affected valve.
SAFETY ANALYSIS StMWty This modification will not affect the valves ability to perform isolation. The design basis as described in the UFSAR remains unaffected by the addition of
- the Junpers. The margin of safety of the valves is not reduced, i
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?EWR'85-486B DESCRIPTION
' 'There'is a concern that the leakage frm the SW line could flood the Auxillary Bu11 ding basement. To prevent flooding of the critical. equipment, the charging punps, the renovable blocks will be sealed with RTV sealent and the- drains in
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the cubicles will also be' sealed, i SAFETY ANALYSIS SLM SRy Operation of the charging punps is unaffected by this rrodification. Sealant is considered moisture and weather resistant and:will-not be affected by a seismic
. event. Sealant does not alter previous seismic analysis.
The probability of a malfunction of charging pump is not increased by this modification. Flooding of charging pums is prevented. A different type of accident has not been created. A line break inside a cubical would render the charging pum inoperable with or without the cubicle sealed.- The margin of safety has not been reduced. Sealant will prevent flooding of
, charging pumps which increases safety, i
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- n. , . IMt SL3HARY EWR 85-494A DESCRIPTION,
~This EWR was used to revise a UFSAR figure. The figure needs to be revised to reflect the construction of a building near the Security Emergency Diesel _ house in which RVs will be tested. This structure is a non-safety related building.
i i b .' SAFETY ANALYSIS SLM4ARY - This EWR does not create or increase the probability of an accident or malfunction of equipment important to safety. The Technical Specif: cations are unaffected by th!s modification. e
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. ag my EWR 85-507A
_ DESCRIPTION This EWR specifies the correct Isolation valve type to be used in the service water piping for flow switches 2-HV-FS-2215A, B, & C. EWR 81-377 was written to Install stainless steel piping and Grinnell diaphram valves in place of the carbon. steel piping which continually clogged up. EWR 85-507 was written to Install additional; Isolation valves in order to work the flow meter (PM and calibrate, etc.); however, the wrong type valves were specifled and mark
- nmbers were not assigned. -e addendun corrects the valve type and identifies the valve mark nmbers.
SAFETY ANALYSIS SLM4ARY The Safety Analysis (ADM-3.9) reviews and evaluates the installation of 1/2" Grinnell diaphram valves in the seal water piping supply to the Control Room chill er ci rculat ion punps 2-HV-P-22A, _ B, & C. The Safety Analysis does not consider this modification to constitute an unreviewed safety question or required change to Technical Specifications. This modification will not create'any different type of accidents or equipment malfunction than any previously evaluated in the UFSAR.
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,1 EWR 85-570, 8, C,' D DESCRIPTI(pd f This EWR replaces the diesel driven fire ptrnp batteries (NAPA)'wlth batterles that are seismically quallfled. The EWR also verifles that the existing, battery rack and pan is selsmically quallfled (per CE-0615) . The new batteries are oxide COM-8D-P and meet or exceed specification requirements for the pumps.
SAFETY ANALYSIS SLMMRY The batteries, battery racks, and battery pans are seismically quallfled per
- calculation CE-0615 The starting capability of the diesel driven fire ptrno is unaffected as the battery capacity exceeds original specification requirements.
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, EWR 85'-7014
__o- _DESCRIPTIGN This . EWR 'was for the replacement of aux 111ary air tanks for various valves- In the maln' steam sytem. The original tanks _were not designed for use' In -air systenu, thus they needed to be replaced, w-4 SAFETY ANALYSIS SLpqMARY Valvei operation will be checked after tank installation. The air tanks will be
. seismically mounted, and the system operation does not change. .The original safety features of.the tanks will not be affected.
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J EWR 86-130C-f. _ DESCRIPTION-This EWR allows reduction in enbedmentlof Hilt! Bolts from 6" to 3-7/8", based on calculation, for the tool chest to be-Installed in the Unit-' 2 containment. The . tool chest will remaln' seismically: quallfled, .as had been previously_
-specified.
. ;:n-SAFETY ANALYSIS SLDNARY
, The failure of the tool chest would not in any way Jeopardize safe shutdown, of the plant.
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_DESCRIPTI(W i This' EWR- ' addendan installs a toolbox in the Unit I containment to store spool-pieces and tools used by the Mechanical Maintenance Department. The box hat been seismically des!oned. In addition, a calculation has been prepared and ' s flied to qualify the "as-built" Hilti bolt enbednents that secure the tool. box.' , t SAFETY ANALYSIS StM SRY-The box is seismically. designed and placed in the annulus of conta i rrnent . It has no affect on any safety related systems. s L
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- DESQt!PTI(N -
The liquid waste radiation monitor RM-LW-115 will be removed; This tronitor has !' never been used at North Anna. To save man hours (8 MAN REM) the equipment will iLi
.be removed from service. The radiation ny;nitors 'do not enhance the present irode of operation.=
i-L SAFETY ANALYSIS SLMWtY L There are no unreviewed safety questions. The consequences of the radiation monitor's removal are nonexistent since the system is not used Since the-
. system is not used the possibility of a different type of1 accident has not'been-
. created. The design basis of Technical Specifications remains unaffected,by
'the removal of the radiation monitors.
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- f. EWR.86-303D-1
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Containment Pressure Indicators are being replaced with. Newport. panel meters,.
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a iThe new panel meters are acceptable. The operability of the'. system .(LML has: j' i :. not. been affected, ;There Is no unreviewed safety question. The modification does not impact.: Technical Spectficatlons,.The;holessIn the[H :S' d. Safeguards. panel will have:to be enlarged in order to accomodate the new meters
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'EWR 86-494C & F DESCRIPTION:
This 'EWR: installs a fabricated sheet metal anti-rotation device:for Masoneilan valves This prevents stem rotation which could cause-limit switches to'make-up improperly. iThe affected valves: are 2-SS-TV-200B, 212B, and- 1-SI-HCV-1850F. l o r SAFETY ANALYSIS SIANARY q TheLweight of the ; device is very smallJand does not require a seismic re-'
. evaluation'and the device will increase- the reliability of limit switch-operation. Installation will .not increase the possibility of any failures not presently evaluated in the UFSAR, and will not affect the margin of safety of.
any= equipment system or components. l '
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. This modifIcatlon is for the Insta11atlon of a fabricated sheet metal, d anti-rotation device for masonellan valves to prevent stem-rotation which could M cause limit swltches to make-up Improperly.
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- - E seismic re-evaluation.. The device will increase the reliability of Ilmit Installation will not increase the-probability.of an
' switch operatloc.
accidentTas presently evaluated in the UFSAR; will not increase the possibility
- 'of any fallures not presently evaluated in tne UFSAR, and-wlli not: adversely
- .g 'affact the margin of safety of any-equipment, system, or components.
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EtSt SIM MRY-- EWR 86-498C y DESCRIPTION The purpose of this EWR was to perform modifications to-each ASCO solenold operated valve (SOV) located inside the Unit 2 Reactor Containment Bullding (excluding those that are in " radiation harsh onl' y" areas) that appears on the
!:r.s ircrtxntally Cunli t led 14 ster List. The rodification entailed the
- r.! t i.i l tt Ir t' ( f r - icu .v.i r (: e i rect ec' CC' < 100. c.r. tre exhaust port ,
u SAFETY ANALYSIS SL90%RY-The modifications performed in this EWR were done' to reduce / eliminate nulsture entering the SOVs In accordance with the manufacturer's reconmendations. The nodification will only enhance the reliability of the SOVs, t' s
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EtSt SLM4ARY EWR 86-498H. DESCRIPTION- ,
'The purpose of this EWR was to perform a walkdown'for each ASCO solenold
-operated; valve (SOV) located inside the Unit 1 Reactor Containment Building which appears on the Envircrvnentally Quallfled Master List. Modifications were performed to ensure that each SOV exhaust port is fitted with a street elbow or 90o tube bend oriented downward. In addition, the walkdown and resultant-modifications were performed to ensure that the electrical leads for each SOV are fed to a suitably vented electrical junction box or sealed by a Conax connector.
SAFETY ANALYSIS SL9 NARY The above modifications have been performed to reduce / eliminate moisture entering the SOVs in accordance with the manufacturer's recemnendations. The design modification will enhance' reliability of the SOVs.
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>3 EWR"86-615 u ,
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'& This modification reduces the length of pipe:for valve 1-SW-355 by 3"..-This!
change will allow for the insta11ation of the regulred pipe can', s t (C. ,
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The changeIn pipe length will have no impact on the seismic integrity, The ]" l,1 s roduction'.'Is-less than 10W.. This modification-will not affect system operatlon.- o O j 's ;r ' o . .
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' " , EWR 86-616A, & C '
[.) [ , DESCRIPTICN The. purpose of.this EWR'Is to provide Instructions for the Installation of a tit ' 9 . ' crossconnect, in the instrunent air lines to allow continuous pressurization of the receivers. f r N:' < 0;+ l- SAFETY ANALYSIS SLM MRY-This-modification does not affect the Design Basis of the Instrument Air
- system,: nor does 'lt affect -any Technical Specifications. This EWR does not
' . increase the probability of occurrence nor the consequences of an accident or malfunction of equipment important to safety. The margin of safety.as'
, described 11n the basis of any-Technical Specifications is not reduced.
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- EWR 86-621 DESutIPTIGd.
' The purpose of this EWR.was to modify a small portion of the ll' quid waste system. .The nodification will involve.the removal of a spool piece and the Installation of_a_ valved wye connection so that resin flushes to the shipping l i ne r .- .
SAFETY ANALYSIS SLMSRY Thl5 Is a non-safety related'insta11atlon. The only part of the UFSAR that Is affected is,flgure 11.2-4 to illustrate this Installation. This EWR does not affect _any system component or structure as defined in the Technical Specification basis and, therefore, does not affect the margin of safety. b i 1
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[ , , EWR 87-025B r _ DESCRIPTION The - Robert - Shaw FMS~ 1000 will have a point reloader nodule (PRM) Instal' led so ' that when power is lost, the FMS 1000 memory will not be destroyed. Fire
- . protection is unaffected, ,
' Thic modification provides for protection of fire zone information, allowing reload following loss of power.
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SAFETY ANALYSIS SLMMRY
' Fire detection Is! unaffected by the installation of the PRM-Into the FMS 1000.
- If the PRMs . falls, zone information can be loaded into the COPS -mencry. There are no unreviewed safety questions. This modification does not impact the UFSAR; 'nor does it ' impact the margin of safety of Technical Specifications.
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- EWR 87'-062A, B,-C, d
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[~ DESCRIPTION ~ L v -r The - purpose. L of this EWR was' to provide additional- office space for the Health-i;. Physics department.- This will be~ accomplished by renovating the clean change shower area Into an office space. F; t s SAFETY ANM.YSIS SLM MRY The . work involved in this EWR Is non-safety related. .No nodificatlons are
- beIng mado.to safety related structures, components, or systems. The portion.
of. the service building in which the work is taking place is non-seismic. This
'~ modification Is of the cosmetic type only, therefore, there are no safety. 3 1
concerns as the result of this EWR. i
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-- EWR 87-162H 0ESCRIPTICN U
The'EWR provides the Installation Instructions for replacing valve 1-SW-330.
.The valve.ls a 3/4" globe valve that will be replaced by;a 3/4"' gate valve.
9 , 3 r J SAFETY ANALYSIS Swe%RY The replacement valve perfonns the same function as the existing valve, thus the safety aspects of the Service Water System'are not affected. There are no
- unreviewed safety questions. The design basis as'outilned in the Technical SpectfIcatlons remains unaffected.
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' ' -EWR 87-162L' h:' ,
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, DESCRIPTION ~ .
h This EWR provides the installation Instructions for' replacing valve 1-SW-424 l
~ The existing valve,.Is a 3/4" globe valve and it will.be replaced with a 3/4" -
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t" i .! 9 SAFETY ANALYSIS ^SIM4ARY. nThe replacement valve performs the same function as the existing valve, thus the safety aspects of the Service Water system are not affected. There are no
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.unreviewed safety questions. - The design basis as outilned in the Technical Specifications' remains unaffected.
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- EWR~87-162Mz ,
,DESQtIPTIGdi This_EWR provides the Installation Instructions for replacing valve 2-SW-258.
~ This existing. valve Is a 3/4" globe valve and'It will be replaced by a 3/4" J
-gate valve.- '
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'questlons._ The design basis as outilned in the Technical-Spectficatlons
. remains unaffected. '
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EWR 87-1620-
- e, 3 4 DESCRIPTION' This;EWR is being used_for the replacement o#-valve 1-SW-331. This valve is presently-'a 3/4".globo: valve (Henry Vogt series SW12141) and it will be-replaced with a gate valve (Henry Vogt-series SW2801).
s. t i 5 SAFETY ANALYSIS SLANARY 1The replacement of a globe valve with a gate valve does not affect the safety taspects of the-system. There are no unreviewed safety questions. The design
. basis as outilned in the Technical Specification.s remain unaffected.
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' EWR 87-162P, S._Q,'S R '
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t This EWR was used for the replacement of valves 2-SW-190 and 2-SW-249. Valve ^ 2-SW-190 1s a 1" Henry Vogt globe valve whereas 2-SW-249 is a 3/4" Henry Vogt
; globe valve. Both valves will be replaced with Henry Vogt gate valves of the same: size. '
L SAFETY ANALYSIS StMWty 1'
;The installation of a gate valve in these appilcations will allow for future
. clean outs. ...The replacement will not affect system operation nor reduce the- >
margin of.-' safety as defined in any Technical Specification. The modification
.will not ' increase the probability of occurrence or the consequences of an accident previously evaluated nor will it affect any equipment Imnortant to safety..
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- EWR 87-199J, & K, ~ C . L, & M, & N
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The notor for 2-SW-MOV-231B was replaced with a similar notor. The thermal-
. ,c overload will also be replaced by this EWR.
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y 3' The. replacement notor ls compatible and Interchangeable with the original' equipment, and installation of the notor will not adversely affect any existing qualification of the actuators involved. ,The additional weight of the new. , motor. does nct have 'any adverse lapact on the operator, y 5
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-EWR'87-269A S'B~'
b DESCRIPfl(N A supply of fire protection water aust be provided to the sprinklers in the trallers.that are used for clean change rooms.during outages. This nust be done because it is an insurance requirement and a good safety practice. During the 1987 outages, fire protection was supplied from the yard fire loop. This was tine consuming and expensive as it required digging up'the yard south of the aux boiler roam before and after the outage.
- The proposed solution is to nake a connection to the sprinkler piping in the aux boiler room. This connection will have an Isolation valve where it tees off the existing system. The piping will run through the south wall of the aux boiler room and end at a flanged connection. Normally a blank flange wlll be Installed on this connection, but during outages tenporary piping will be heat traced and Insulated if In use when freezing is a possibility.,
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This modification does not effect Nuclear Safety Related or Environmentally. .! Qual i fied equipnent. The piping added by this nodification will not change the .' design basis or operation of the fire protection system. 1 l l. '.
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7 -. i'- a b EWR SLM %RY l i. EUR 87-3/3B i , b _DESCRIPflCN
'This, EWR describes the history of operating problems with the pressure control valves for the High Pressure Drain purps gland seal supply. This EWR doctments the ' review that was conducted by Maintenance Engineering, and the corrective
' actions required to correct the gland seal supply system. An Installation procedure is supplied-In the text of the EWR to ensure proper operation of the system. The EWR Includes all docunent changes and updates required.. The EWR i
installs 0-600 PSI gauges and a 400 PSI relief valve to protect the PCV and F packing gland of the high pressure heater drain purps, k SAFETY ANALYSIS SlM %RY A Safety' Analysis is attached due to the UFSAR drawing changes required. Figure No . - 10.4-18 (11715-FM-24A) and 10.4-20 (12050-FM-24A) Bearing Cooling Water required updating to show the new isolation valve for the' pressure Indicators
- and the pressure indicators which are installed by this EWR.
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EMtSWMMty EWR 87-326 CA,B,C DESQtlPTI@d EWR 87-326 thru 87-326C Involves replacement of four Safety Supports Mark fos. WCMV-P-004,-007,-001,-002 on the WCMV System in the Auxiliary Feedwater pum
' house (AFPH). The purpose'of the replacements was to prevent corrosion of the new supports by renoving them f rom a trench in the AFPH and thus preventing them frcm contact with water.
SAFETY ## LYSIS SG MUty The four pipe supports are Safety Related replacements for the existing pipe supports mark tos. WCMV-R-004,-007,-001,C-002 on the WCMV system in the
' Auxillary Feedwater pu mhouse. The replacements are Identical in design function, fit, etc. and do not affect the design basis of the system.
Ebst SLMSRy s EWR 87-331 DESCRIPTION
- The purpose of the EWR was to evaluate and approve the modification to upgrade the Reactor Coolant Pwo seals. This nodification will be done in order to Inprove seal performance and reliability, b .
SAFETY ANALYSIS SLM mRy The Roactor Coolant Pump seals will continue to function as originally designed; therefore, the probability of occurrence of a malfunction will not be increased as a result of this nodification. The margin of safety is not _ reduced, and no Technical Specifications are affected. I
v ,- LI DR SLMMRY EWR 87-395 OESQt!PTI@d g The purpose of this EWR 15 to lengthen the retaining pin on the service water radInt ion tronitoring purm,1-SW-P-10. This modification will prevent rotation between the channel rings and the casing thereby eliminating the flowpath blockage. I f f-SAFETY AN4 LYSIS SLMMRY The probability of occurrence or the consequences of a malfunction will not be ,_ Increased by this modification. Release Ilmits are unchanged and can be i monitored to prevent exceeding allowable releases, and therefore, the margin of ; safety as described in the Basis section of the Technical Specifications is not reduced, 3 i
-a y my EWR 87-396D DEscitIPflGd The purpose of this EWR is to provide direction through a procedure for the wiring of parallel reactor breaker inputs to the Dranetz Sequence of Events Recorder.- This will allow a backup recorder in the event the plant P-250 recorder is experiencing problems.
SAFETY ANALYSIS Swe%RY This modification does not affect a Safety Related system, the Techni ~1
' Specifications, or the UFSAR. The margin of safety as described in the Technical. Specifications is not reduced by this nodlfication. There are no unreviewed safety questions as a result of this EWR.
EWR SMY EWR 87-4148, C, D DESCRIPTION: This rrodificat ion determinated the grounded pressurizer heaters and replaced those heaters with heaters that have been repaired and are ready to return to service. The heaters that were de-terininated are 1HS-EHR-62, 63, and
- 41. The heaters re-terminated are 1-HS-EHR-9, 18, 53, 59 and 61.
SAFETY ANALYSIS StM*ARY: This rodification will increase the total KW of pressurizer heaters available and will maintain the system above the Technical Specifications mintntm for the H and d bus. The margin of safety will not be reduced for H and J bus. 87-414BSCSD i s
' :n- -(; q .y l ,h' ' EWR !RM MRY EWR 87-576B DESCRIPTI(N This EWR adjusted torque switches to the max trrun for butterfly MOVs. This is a
- conservative approach to be irrplemented as traintenance is being performed.
SAFETY APMLYSIS SLM%Ry
, Torque switch settings will be Increased to reduce the likelihood of an inadvertent torque switch trip resulting in valve failure. .This will improve valve reliability without adversely affecting it's operability, e
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[- , i EWL StMMitY ? - L ' EWR 87-598C i !~ DEScitIPTI(N Incore detector thirrble tube LOS showed a wall loss of 49% in one locatlon. This was an increase of approximately 9% frcm the 1987 Unit W2 outage. Because of this increase this tube will be retracted an additional two (2) Inches. i s' SAFETY ANALYSIS SLM WtY This rrodification will provide a new wear surface for the tube being retracted. This will not affect the operation of the incore detector thirrble tubes. Modification will not increase the probability of an accident, t i
IMt SGMMtY EWR 87-61$ OESQtIPTIG4 No podification is required by this EWR. The EWR assigns a mark nwber to i various fire dampers and requires that they be identified by stenciling the
' ductwork, access door or area adjacent to the fire dampers. The Station drawings are to be revised to show the mark nmbers assigned to the fire darrpers.
I I SAFETY ANE.YSIS SG MNtY Identification of the fire dampers does not affect system operation. Periodic testing and maintenance will be more easily conpleted once the fire dampers have been identlfled In the plant. I l
,i
- a. EWt SLMWtY EWR 87-620C . -
_DESCRIPTICN The EWR allows the nodifIcatIon of the Unit I feedwater regulating valves, 1-FW-FCV-1478, 1488, and 1498, cage trim and valves stem. A balance drtm is Installed on the top of_the cage and baffles installed on the bottom of the cage. A 1/4' diameter valve stem is Installed. The velve bonnet is machined to accept .the 1/4" diameter valve stem and balance drun. The purpose of the nodification is to eliminate the history of valve stem failures and vibration experienced which causes the valve stem failures. The 1 1/4" diameter valve stem increases valve reliability. The balance drun and baffles reduce flow Induced vibration within the valves and system piping. 7
~
SAFETY ANALYSIS SLM WtY The operation and system Integrity remains unaffected. Valve operation is unaffected. Technical Specification requirements remain unchanged. The selsmic structural Integrity of valve and piping system is not affected, Appilcable UFSAR sections, 68.3.3, 10.4.3 and 15.2.10. appilcable Tech. Spec. 3.3.2.1 Table 3.3-3 Item SA. I-
L Oft SLM%RY EWR 87-671A V DESCRIPTICH In the original EWR 87-671, three Hilti Bolt holes in the baseplate of Plate B, for 2-MSS-WGCB-PRR-3 were enlarged frcm 1-1/8" to 1-1/4" to permit reinstallation of the baseplate. The baseplate was removed temporarily to facilitate valve maintenance. The holes were enlarged to ensure that Hilti Bolt threads would not be damaged during plate Installation. An Engineering analysis was provided to allow the enlargement of the three holes. EWR 87-671A provides a trore descriptive' design basis and method in support of L the engineering judgement made In the original EWR. L SAFETY #% LYSIS SLM%RY Enlarging three holes in a worst case scena r lo, may cause a slight redistribution of shear to the remaining five (5) Hilt! Bolts; however, all anchors will remain fully effective in resisting tenslie forces due to the plate washers provided. The redistribution of shear will be more than offset when considering the anchor's actual entedment depth and the in place 4000 psi strength concrete. The resulting margin of safety will be increased when
-taking advantage of the deeper embedment and higher strength concrete.
Reference Addendtm No.1 to Stone & Webster Calculation 12050-2-1020, Rev. O for details.
E , _) r EWtSGM4tY It EWR 87-677 i DEsatIPTIGd ' The purpose of this EWR is to provide Instructions for the retroval of the i emergency condensate supply to High Pressure Heater Drain ptrnps. These ptrms ; are non-catagory I nor safety related. The Ernergency Condensate Supply is used : to prevent the High Pressure Heater Drain purros from cavitating, but the ptsms already have other protection for this, therefore, this system is not required.- 4 L l i. l' SAFETY ANALYSIS SW M4RY The retroval of the Emergency Condensate Supply to the High Pressure Heater Drain ptrrps does not affect any Technical Specification, so the margin of safety as described within the Technical Specifications is not reduced. The renoval of this system does affect an UFSAR figure, however, no design basis within the UFSAR is affected. There are no unreviewed safety questions. e l l l
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'EWR 87-706, & B,'S D '
pESQtIPTIGd 1 The purpose of this EWR is to provide a method for the upgrade of the existing oil coolers for the Auxiliary Feedwater Ptnps. The upgrade is necessary in , order to reflect the current pressura conditions of the system. The steam driven oil coolers will be upgraded by replacing the bonnet, whereas the motor driven oil coolers will be completely replaced. , F SAFETY AN4 LYSIS StMMftY This EWR will upgrade the oil coolers to a higher pressure rating, however, the operation of the oil coolers remains the same as described in the UFSAR. This mod!fication will not increase the probability or consequences of a malfunction > of equipment important to safety. The margin of safety is not reduced by the Auxillary Feedwater Ptmp oil cooler modification.
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N. Hz y my i; o EWR 87-714 + p l i i ( DESQtiPTIG6 This EWR is providing the necessary Instructions for the replacement of the .; bearing retainer plate for the motor of 2-QS-P-2A, This rrodification consists i , of utilizing the existing catponent as a template for machining a replacement j, part from ASTM-36 plate material. p, i-i , 4 F t SAFETY ANU YSIS SLMERY .c i The design basis of the puTo motor will not be affected. The probability of ) occurrence or the consequences of a malfunction of equipment important to safety will not be Increased. The margin of safety remains unaffected as long as RWST temperature remains within it's limits, which is being maintained with the "B" ptrip. + 9 e
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IMt SWMMtY EWR 88-080 l o DESQt!PTIGd l The purpose of this EWR was to evaluate and provide engineering guidance in locating a replacement outer piston guide gasket for 01-SI-SOV-1008. The conclusion is that it is acceptable to use a gasket from a previously purchased repair kit that is environmentally quallfled for use at the station. ; l!. SMTTY ANN.YSIS SWNMtY The operation and design basis of the SOV remains unchanged. No Technical Specifications will be affected, therefore, the margin of safety remains unaffected. The replacement gasket is equal to, or better than, the original . equipment; therefore, the probability of an accident or malfunction is not increased. l~ l
s+ b ~ EWL SG9%RY EWR 88-086A C B & D & E & F C G DESCRIPTIGd l The existing Kerotest manifolds and Hoke isolation valves have leakby problems. j_ This EWR will provide the evaluation and concurrence for replacing these components with Anderson Greenwood equivalents. The affected manifolds pertain to 01-MS-FT-1474, 75, 84, 85, 94 and 95. g V L l-SAFETY #MLYSIS SLMMRY The replacement equipment treets or exceeds the specifications of the existing equipment which is being replaced, therefore the probability of an accident is not increased. The margin of safety as defined in the basis section of the Technical Specifications will not be reduced. a
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i r EWR 88-099 & A DESCRIPTIG4 The top two rows of vital bus indicating Ilghts irrmediately belcav annunciator IG are not arranged in proper top-to-bottcm and left-to-right sequence. Rearrange and reliable the nameplates and rewire five (5) Indicating lights such that when viewed frce the vertical board 1-2 from left-to-right the top row will be "Res Sta Ser A Leads W", and "Res Sta Ser B Leads UB", and "Ros Sta Ser C leads UV". The second rcw lef t light will remain the same, the middle [' light will be "Turb Sol Trio DC Pn1 A Ckt 11", and the right will be "Back Turb
! Sol Trlp DC Pn1 Ckt 11".
l' i t SAFETY # MLYSIS SLM %RY t This nodification does not involve an unreviewed safety question as defined by 10CFR50.59 because it is a rearrangement of existing ccmponents for Human Engineering aspects only. Also Appendix R and the UFSAR are not affected by the modification, e i
l lL EMt SL89WtY EWR 88-100 & A S B C C _DESCitIPTICH The top two rcws of vital bus indicating Ilghts imediately belcw Annunciator 2G are not arranged in proper top-to-bottom and left-to+right sequence. Rearrange and relabel the nameplates and rewire five ($) Indicating Ilghts such as when viewed from the vertical board 2-2 f rom lef t-to-right the top row will be "Res Sta Ser A Leads W", and "Res Sta ser B Leads W", and Res Sta Ser C 1 cads W". The second row left 11ght will remain the same, the middle light will be "Back Turb Sol Trlp DC Pn1 Cktll", and the right Ilght will be "Back Turb DC Pn1 Ckt 11". SAFETY MMLYSIS SL99WtY This modificat'.on does not involve an unreviewed safety question as defined by 10CFR50.59 because it is a rearrangement of existing components for Hanan Engineering aspects only. Also Appendix R and the UFSAR are not affected by this nodification. i
4 gs , o' Elst SLM4NtY f j ,. fi EWR 88-108 & B i- ' DE$CRIPT!(N l This EWR will cause nonna11y lit amunciator B-7 "CH PP IC Auto Trip 1$J7 LO" J .to be enabled only when breaker l$d7 15 racked in. I 4 l; ? i r j. j,' b F :
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SAFETY ANM.YSIS St#MNtY
- g. This change enables the "C" charging pum J Bus Feeder breaker lockout' -
annunciator to be enabled only if the J Bus breaker is racked in. The
!. operability and function of the "C" charging pum is not adversely affected. '
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, This nodification will ensure better control room operator reliability. ,
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If I I EWt Supt %RY l EWR 88-271 DESCRIPTim This EWR nodified the Grinnell Valve Reach Rods. The problem with the original reach rods is described, le; poor maintenance ability to reach rod and valve, reach rod does not allow for misalignment, radiation dose rates are high due to design during maintenance, no easy way of opening / closing valve upon fallure of reach rod.
.The nodified reach rod corrects the above problems. The EWR is used as an evaluatlon to allow use of the nodifIcatlon through an approved statlon n.alntenance procedure which incorporates this EWR. All work would then be controlled by work order and Ff4P. (Mechanical Malntenance Procedure)
The reach rods will be nodified during maintenance of a Grinnell Diaphragm Valve if required as stated in the MMP or by the preventive maintenance program to change the valve diaphragms. SAFETY #MLYSIS SLD9%RY Reach rods will be nodified on a case by case basis by the maintenance group while performing preventive maintenance. Each reach rod modification will be completed using an approved procedure. Each reach rod will be evaluated by Design Engineering on a case-by-case basis to ensure that the seismic Integrity of the piping system or any other safety equipment will not be effected. The operation of the valves does not change. The margin of safety is not changed or degraded by the nodification. O
e , , p i:" Det SU9%RY i' EWR 88-296 DESCRIPTICN 1
, This EWR replaces the secondary potential fuses in the ground detector circults {
on the Unit 1 omorgency buses. The new fuses will still protect secondary 4 loads while provide coordinated protection. The new fu?ies are sized equal to
, those on the same circuits of Unit 2, thus would el!minate discrepancy and l maintain uniformity.
l l ! 1 SAFETY ANALYSIS SLM SRY The change out of fuses will not result In adverse conditions nor allow the plant to be exposed to different conditions from which it was previously protected. l l
EhlR SLM%RY 88-312A DESCRIPTION: Substitution (replacement) of Valve Ib. 1-MS-457, a 1" VOGT VOS-60C, was reQulred. A suitable valve replacement in design and material was available. This valve, Conval Figure 11G2, is an acceptable substitution. SAFETY ANALYSIS SLM %RY: The replacenent vale performs the same function and reets the design reculrements of the existing valvo. Fallure may include leakage past the Isolation valve. Since the valve meets the original design criteria, probability of failure is not increased. 88-312A
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EWR 88-312B DdR St#9%RY DESCRIPTION Various Henry Vogt valves are experlencing problems such as body to bonnet leaks, seat leakby and steam cuts. The Conval Y-globe valve is a better design and is easier to perfom maintenance on the Conval valve. Both valves are constructed of campatible materials and have pressure / temperature ratings which exceed that of the piping. The Conval valve is an acceptable replacement for (Vogt) valve 2-MS-330. SAFETY ANALYSIS St#9%RY The replacement valves will provide the same function and meet the same design requirements as the original valvo. All original design criteria are met. Since the valves meet the original design criteria, the probability of failure is not increased, t l
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/ 3 l-EWR 88-312C EWR StM Wty i
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- _DESCRIP11(N Various Henry Vogt valves are experiencing problems such as body to bonnet
' leaks, seat leakby and steam cuts. The Conval Y-globe valve is a better design and is easier to perform maintenance on the Conval valve. Both valves are o ' constructed of compatible. materials and have pressure / temperature ratings which i exceed that of the piping. Conval valve is an acceptable replacement for (Vogt) valve 2-BD-106. '
r SAFETY #MLYSIS SLMWtY The replacement valves provide the sane function and meet the same design requirements as the original valve. All original design criteria are met. Since the valves meet the original design criteria, the probability of failure is not increased. 1
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EWR 88-312D y Sg,ggy DESCRIPTION Various Henry Vogt valves are experiencing problems such as body to bonnet F-leaks, seat leakby and steam cuts. The Conval Y-globe valve is a better design and Is easier to perform maintenance on the Conval valve. Both valves are constructed of ecmpatible materials and have pressure / temperature ratings telch exceed that of the piping. The Conval valve is an acceptable replacement for L (Vogt) valve (2-MS-422), i SN"ETY #MLYSIS SWe%Ry i. The replacement valves provide the same function and meet the same design requirements as the original valve. All original design critoria are met. Since ' the valves meet the original design criteria the probability of failure is not increased. e
I EWt SLM%RY , ,f ! 1 l EWR 88-312F ! DESCRIPTim i i l L 1 Substitution (replacement) of valve No. 1-MS-360, a 1" Vogt VOS-60C, is reculred. A suitable replacement in design and materials is available. This valve, Conval figure 11G2, is an acceptable substitution. ' r I s l- ! l SAFETY ANALYSIS SLMmRY l I The replacement valve performs the same function and meets the design requirements of the existing valve. Since the valve meets the original design criteria, probability of failure is not increased.
w EWt SlM%RY i r l DdR 88-312G 1 i DESCRIPTION: 1 Substitution (replacement) of Valve No. 1-MS-425, a 1" Vogt VOS-60C, was reaulred. A suitable valve replacement in design and material was available. This vaive, Conval Figure llG2, Is an acceptable substItutlon. P SAFETY ANALYSIS SlM %RY: The replacement valve performs the same function and meets the design requirements of the existing valve. Failure may include leakage past the isolation valve. The valve neets the original design criteria, probability of failure is not increased. l l I l 88-312G
i EWt SlM%RY l' l EWR 88-312H i r DESCRIPTICN: Substitution (replacement) of Valve No. 1-MS-64, a 1" Vogt VOS-60C, was required. A suitable valve replacement in design and material was available. This valve, Conval Figure 11G2, is an acceptable substitution. L SAFETY ANALYSIS SLM MRY: i The replacement valve performs the same function and meets the design requirements of th3 existing valve. Fallure may include leakage past the Isolation valve. Since the valve meets the original design criteria, probablilty of failure is not increased. 1 i 88-312H
7_ - EWt SLM%RY EWR 88-312N DESCRIPTION: Substitution (replacement) of valve No. 1-FW 523, A 1/2" Vogt V05-60C was ! ~ 7 required. A suitable valve replacenent in design and noterial was available. ! This valve, Conval Figure 11G2 was an acceptable substitution.
!- SAFETY ANALYSIS SlM %RY: :
i The replacement valve perfonned the same function and met the design 1 requirenents of the existing valve. Failure nay include leakage past the
, solation valve. Since the valve met the original design criteria, probability of failure is not increased. ;
2 F i l l l 88-312N
um f.- EWt SLM4ARY EWR 88-3120 DESCRIPTION:
. Substitution (replacement) of . valve No. 2-MS-333, a 1" OVS-600, was
- required due to leakage. .A suitable valve replacement In design and material was available. This' valve, Conval Figure 11G2, was an acceptable substitution.
SAFETY ANALYSIS SLM4ARY: The replacement valve performed ti m same function and met the design requirements of the existing valve. Failure may. Include leakage past the Isolation valve. Since the valve met the original des.lgn criteria, probability of failure was not increased. 4 4 I 88'-3120 a., I ,1' .' ic
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EWR SLM%RY EWR 88-312P OESCRIPTiCN Varlous -Henry Vogt valves a. e expeelencing problems such'as body to bonnet leaks, seat;leakby and steam cuts. The Conval ~Y-globe valve is a better
-design,-and it_Is easier to perform maintenance on this valve. Both valves are constructed of compatible materials and have pressure / temperature ratings which exceed that of the piping. The Conval valve is an acceptable replacement for the Vogt valve. The affected valve was 01-BD-120.
SAFETY ANALYSIS SLM eRY The replacement valves will provide the same function and meet the same design requirements as the original valve. All original design criteria are met. Since the valves meet the original design criteria, the probability of failure is not inc reased.
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b EWR SLD99Ry EWR 88-312R DESCRIPTION Substitution (replacement) of valve No. 2-MS-68, a 1".VOS-60C, was-required due to leakage. A sultable valve replacement in design and: materials was available. This-valve, Conval figure 11G2, is an acceptable substitution. SAFETY ANALYSIS SLD9%RY The replacement valve perfonns the same function a_nd meets the design requirements of the existing valve. Failure may include leakage past the
-Isolation' . valve. Since the valve meets the original design criteria, probability.of failure is not increased.
'\l DdR SG9%Ry l
EWR 88-312S DESCRIPT10N Substitution (rep 1acement) of valve No. 2-MS-106, a 1" VOS-60C, was required due to leakage. A .sultable valve replacement In design and material was available. This valve, Conval figure 11G2, is an acceptable substitution. SAFETY ANALYSIS StM SRy The replacement valve performs the same function and meets the design' requirements of the existing valve. Since the valve meets the original design
, criteria, probability of failure is not increased.
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- b. . EWR'88-312T EWR SLD9%RY DESCRIPTION' Substitution (replacement) of valve No. 2-MS-408, a 3/4" VOS-60C, was required due to leakage. A suitable replacement in design.and material was available.
This valve,.Conval figure 11G3, Is an acceptable substitution. SAFETY ANALYSIS SLM4ARY The replacement valve performs the sme function and meets the design requirements of the existing valve. All original design criteria are met. Since the- valve meets the original criteria, probability of failure is not increased.
l l. I EWR 88-312U EWSMY DESCRIPTION-Substitution (replacement) of valve No. 2-MS-412, a 3/4" V05-60C, was reculred due to leakage. A suitable replacement in design and material was available. This valve, Conval figure 11G3, is an acceptable-substitution. SAFETY ANALYSIS SUP94ARY
.-The. replacement valve performs the- same function and meets the design requirements of the existing' valve. Since the valve meets the original design criteria, probability of failure is not increased, s
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Y c . ; E j ; Ii -EWR 88-312X EWR SLM%RY ' n - DESCRIPTION Substitution (replacement) of. valve No. 1-8D-115, a 3/4" Vogt VOS-60C was required. A suitable valve replacement in design and material was available. This valve, Conval figure llG2 Is an acceptable substItutlon. F SAFETY ANALYSIS SLM %RY The_ replacement valve performs the same function and meets the design requirements of the existing valve.- Fallure may include leakage past the isolation valve. Since the valve meets the original design criteria, probability of failure is not increased.
y~ - l + ) e i e E EWR 88-312Y- yggy t x [-
. DESCRIPTION
~ Various Henry Vogt valves are experiencing: problems such as body to bonnet leaks, seat leak by, and steam cuts. The Conval Y-globe valve is a better-design' and is easier to perform maintenance on the Conval valve. Both valves-c are constructed of compatible materials and have pressure / temperature ratings which exceed that of the piping. The Conval valve is acceptable replacement for (Vogt) valve 1-MS-1018.
1 l-4 SAFETY ANALYSIS stpMARy ' The replacement valves provide the same function and meet the same design requirements as the original valve. All original design criteria are _ met. Since-the valve meets the original design criteria, probability of failure .is not increased. I I I i
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- p. 7 iEWR 88-312Z EWR ' StPNARY E
r DESCRIPTION Various Henry Vogt valves are experiencing problems such as body to bonnet leaks, seat leak by, and steam cuts. The Conval Y-globe valve design and~. Is easier to perform maintenance on the Conval valvo.is a better Both valves
. are constructed of compatibio materials and have pressure / temperature ratings. -
l which exceed that of the piping. The Conval valve is an acceptable replacement for (Vogt) valve 1-MS<1017. -- I e SAFETY ANALYSIS SlDNARY The replacement valves provide the same function- and meet the same. design requirements as the original valve. All original design criteria are met. Since the valve meets the original design criteria, probability of. failure is not increased. l '. 1
EWt StM4ARY EWR 88-312AA DESCRIPTION Various Henry ~Vogt-' valves are- experiencing problems such as body to bonnet leaks, seat leakby and steam cuts.- The Conval Y-globe valve is a better design and- In easier' to perform maintenance on. Both valves are constructed of compatible materials and'have press / temp. ratings which exceed that of the piping.- The Conval is an acceptable replacement for Vogt valve which-is a }" VOS-60C. The affected valve is 2-MS-236. 1 SAFETY ANALYSIS SLM1ARY' i f The replacement valves will provide the same function and meet the same design ' requirements as the original valve. All original-design requirements are met. l Since the valves meet the original design criteria, the probability of failure Is not increased. i l l a
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.EWR SLM SRy EWR 88-312AB-DESCRIPTI@l.
Substitution (replacement)'of valve No. 1-BD-21, a 1" Vogt VOS-60C, was. required. A suitable valve replacement in design and material-was available.
- This valve, Conval figure-11G2, Is an acceptable substitution.
e. 1 ( I-SAFETY ANALYSIS SLM MRY' The replacement valve' perfonns the same function and meets the design requirements of the existing valve. Failure may include leakage past the isolation valve. Since the valve meets the original design criteria, probability of failure is.not-increased.
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- EWR 88-312AC W DESCRIPTION L
Substitution (replacement) of' valve No. .1-MS-400, a 3/4" VOS-60C, was required due to' leakage. A suitable valve replacement-in design and material was
^
- avallable.-ThIs valve..Conval figure 11G3. Is an acceptable substItutlon.' - .
I t SAFETY ANALYSIS SLM MRY The- replacement valve performs the sane: function and meets the design requirements for the-existing valve. Failure may Include leakage past the isolation-valve. Since the valve meets the original design criteria,.
. probability of fallure--is not increased..
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. 3 .- EWR 88 312AD .
e i DESCRIPTION .
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halve 1-FW-514 Is:aL1/2". Henry Vogt welded bonnet globe-valve. It--.Is'a vent ~- (valve' for- the LT-1475 condensing piece. It was. replaced with a Conval~ Y-Globe: valve. .!
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The; replacement. valve meets the design requirements for the-system. The material is compatible with the system. The probability of accident 'or the'
. consequences of a.fallure are not Increased.
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, DESCRIPTION:
Substitution ~(replacement) of. Valve No. 1-MS-1011, a 1/2" VOS-60C, was required due to leakage. .A suitable valve replacement in design and material was-' available. This valve, Conval Figure 11G2, is an acceptable substitution. 7 SAFETY ANALYSIS REVIEW: The. replacement valve performs the same function and meets the design: requirements of the existing' valve. Failure' may. Include -leakage 'past the F isolation- valve. Since. the valve meets the original design criteria, probability of failure is not increased. .!
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EWR 88-312AF i-DESCRIPTION:
. Substitution (replacement) of Valve No. 1-MS-348, a 1" VOS-60C, was required due to leakage. A suitable valve replacenent in design and material was available. This valve, Conval Figure 11G2, is an acceptable substitution.
SAFETY ANALYSIS SUNNARY:' j The .replacenent valve performs the sane function and -meets the design-requirenents of the existing valve. Failure ney include leakage past the isolation valve. Since the valve meets the original design criteria, probability of fallure is not increased. r I u i 88-312AFL m i
i S{k r i A6 EWR SlM%RY j j.; , , f 3 ',j.5 R -EWR 88-312AG
> t DESCRIPTION:
Substitution- (replacement) of' Valve No. 1-MS-324, a 1" VOS-60C, was required l' due to leakage. A suitable valve replacement = in- design and . material was available. This valve, Conval Figure 11D2, was an acceptable substitution. i SAFETY ANALYSIS SLM MRY: , The replacement valve performs the same function and meets the requiremont's of:
.the existing. valve. Fa11ure may include leakage past the Isolation- valve.. >
Since the valve neets the original design criteria, probability of failure is - not-Increased, t h r a g
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Is EWR 88-312Al ySpy DESCRIPTION Substitution (replacement).. of valve No.1-BD-40, a 3/4" VOS-60C, was required due to leakage. A suitable valve replacement in design and materials was-available.This valve, Conval figure 1103, is an acceptable substitution. SAFETY ANALYSIS SLDNARY The . replacement valve will provide the same-function and meet the same design requirements as the original valve. All original design criteria are met. Since the valve neets the orig!nal design critoria, the probability of failure is not increased. I t
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p b p EWR 88-312A0 EWR SL) NARY i-DESCRIPTION
. ; Various1 Henry Vogt valves are experiencing problems such as body to bonnet leaks, seat leak by and steam cuts. The Conval Y-globe valve is a better design and easier to perform maintenance on
.Both valves are constructed _ of i compatlabic materials and have pressure and temperture ratings which exceed
' that ~of the piping. The Conval is an acceptable replacement for the 3/4"vogt gate valve. The affected valve is 1-FW-74 -
4
-t SAFETY ANALYSIS SLDNARY The replacement valves provide the same function and meet th' esame. design requirements as the original vlave. All original _ design requirements are met. '
Since the valves meet the original design criteria, the probabillity of failure is not increased. 1 A
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. i
'EWR 88-3238 l DESCRIPTION The changes ,'of material used in the electrical penetration test assembly from brass to stainless steel will provide the conmonality betweeen both Unit'l and.
2 and simplify purchasing the material. Also the stainless steel material.may reduce the maintenance on penetratlon leak or the test assently.
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SAFETY' ANALYSIS SLM gRy I This~ change provides cenmonality of materials for test assent 11es on Unit I and'
- 2. . The original Specification No. 21.for Conax penetration provided for Unit'l and 2 were Cat I Specifications which supplied Unit I with stainless steel-parts and Unit.-2 with-brass material. Functional testing will be performed on
'each modification-to verify the test assembly operability. .
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p , x h L p EWR SW9%RY-EWR 88-328, A, B, C' L ; p DESCRIPTION The Nuclear Regulatory Conmission (NRC) has issued a bulletin (NRCB 88-08) that identifies a' potential for thermal stress Induced failures on sections of unisoluble piping connected to the Reactor Coolant System. This bulletin requires'.11conse holders to analyze their piping systems connected to the RCS to -identify potential locations where cold water could leak past Isolation valves into the RCS causing thennal stress to pipe welds near the RCS piping and provide means to measure any temperature differentials in the piping. Six [." piping sections in the Safety Injection System (SIS) have been identified between the RCS loops and the last SIS check valve where such thermal i stratIfIcatlon is possible. Plant engineering has requested- that- temporary thermocouples be Installed ~on the top and bottom of.each affected piping section to nonitor temperature differences across the pipe dianeter. Stone S e Webster Engineering Corporation Identified a possible similar concern in the pressurizer surge line, based on IE Notice 38-80 'and reconmends the addition of i six thermocouples on } the pipe dianeter between the top and bottom of the ' surge line in the location shown on the sketches attached to the EWR, The thermocouples will provide input to a' digital Data Logger for the purpose of documenting any thermal gradients or oscillations in each of the identified piping sections. A positioner monitor will nonttor pressurizer surge line transducer / motion caused by thermal expansion.
. SAFETY ANALYSIS StM MRY This modification does not involve an unreviewed safety question as defined by 10CFR50.59 because this change including all required components, are.
designated non-safety related, nonclass 1E and-have no connections for safe shutdown. . The thermocouples, cables, and. recorders, being classifled as temporary, will be seismically installed in a manner which will prevent them. ; from. affecting safety related components during a seismic event. Also the Technical Specification and the UFSAR are not affected by this modifIcatlon. I 1 l l
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F b ' R p EWR SLM%RY
'EWR 88-329, A, B, C, D, E,'F, G f
DESCRIPTION: The nuclear Regulatory (NRC) Issued a bulletin (NRCB 88-08) that-Identifled a potential. for thernal stress Induced failure on sections of Insoluble piping connected to; the Reactor: Coolant Systan. This bulletin reaulred license > holders to' analyze their piping systems connected to the RCS and identify potential locations where cold water could leak past isolation valves into the-RCS causing thernal stress to pipe welds near the RCS piping. This txal let in also required -license holders to provide' neans to neasure any tanperature differentials in the piping. Piping sections in the Safety. Injection System (SIS) was identifled between the RCA loops and the SIS check valve where such thornal stratification. ls possible. Plant engineering requested that temporary thernocouples be Installed on the top and bottan of each affected piping section to monitor tenperature differences across the pipe dianeter. Stone and Webster Engineering Corporation Identified a possible similar-concern in the pressurizer surge line, based on IE Notice 88-80 and reconnended the addition of thernocouples on 1/2 the pipe dianeter between the top and bottom of the-surge line. The thernocupies would provide input to a digital data logger for the purpose of documenting any thennal gradients or oscillations in each of the Identified piping sections. A positioner nonitor would nonitor pressurizer l
'llne transducer / notion caused by thennal expansion.
SAFETY ANALYSIS SLM MRy: This modification did not involve an unreviewed safety question as-defined by 10CFR50.59 because this change, including' all required canponents, ' arc designated non-safety related, nonclass lE and have no connections for safe shutdown. The thermocouples, cables, and recorders, were classified as temporary, would be seismically Installed in a manner which would prevent them fran affecting safety related components during a seismic event. Also, the Technical Specification and the UFSAR were not affected by this nodification. f 88-329&A-G
V w EWR SLM MRY b 88-330,'A,-B, C, E, F, G, H, 1, J, K
- j. : DESCRIPTION in order to ensure that the High Head Safety Injection (HHSI) pumps have a.
. flooded nuction, additional vents will be installed by this EWR on the high" points of the HHSI pump suction IInes. New vents to be Installed are as follows:' '
(a) One. vent downstream of MOV-2863A (8"-SI-440-153A-Q2) . (b) _ (c) Onevont
.One vent upstream downstrean of MOV-28638 (8"-CH-417-153A-Q2) of MOV-2863B'(8"-SI-449-153A-Q2)
(d) One vent downstrean of MOV-1863A (8"-SI-40 -153A-Q2) (e) One vent downstream of MOV-1863B (8"-CH-17 -153A-Q2) (f)' One vent upstream of Refueling Water Storage Tank (10"-SI-8-153A-Q3). , SAFETY ANALYSIS StM MRY This nodification does not involve an unreviewed safety question as defined by 10CRF50.59 because the operation and design function of the HHSicand LHSI are unaffected. Installation of the vent valves will reduce the possibility of an accident or malfunction caused by air binding. All design basis for the vents are consistent with the piping that will be Installed.
'L i
y EWR Sl99%RY EWR 88-333 DESCRIPTION This EWR Installs Swagelok Compression fittings in the seal' head tank piping of
~2-RS-P-2A as the final connection point as allcwed by NAS-1009 tube specification.
The modification allows for the retroval of the head tank for maintenance or repairs. Th'Is.rrodification will also eliminate the possibility of damage- to the seal head tank piping. This'same trodification has.been L previously ccrnpleted on the_ Unit -l's outside recirculation spray ptmps and Unit - i:- 2 28 ptmps. I h- _ 1 SAFETY ANALYSIS St99%RY i The safety analysis approves the. Installation of the Swagelok unions. -The .: analysis valldates that.the systems f unctlon does not change, and ~ls not affected by the Installation. The seismic integrity Is not affected. The { materials are suitable and adequate to maintain system Integrity. The Technical
. Specifications and UFSAR remain unchanged. l l
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;- EWR-88-335A'
(- , DESCRIPTION-01-MS-TD-420D is defective;and requires replacement. The only'replacementu 7 available has 1/2" NPS connections whereas the' existing-trap has 1" NPS connections. The two traps are equivalent in construction, capacity, function, and-pressure / temperature ratings. Therefore the trap will be installed with 1" k X 1/2" reducers. E:: n i .. SAFETY ANALYSIS SLM %RY
~
The replacement trap perfonns' the same function and meets the Design 3, , requirements'of the existing trap. Since the trap meets the original design criteria, .probabilty of failure is not increased. A
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m . c b g EWR SLMMRY E EWR 88-340 h DESCRIPTI(W The existing valve, a 3/4" bolted bonnet valve, has been Inspected. This valve : shall be replaced with a 3/4" welded bonnet valve. Both valves are supplied by
, the same manufacturer, constructed of the same material, have the samo CV and end to end dimensions. Use of the wolded bonnet valve will eliminate a potentini leak path, the body to bonnet joint, and is acceptable to use.
SAFETY ANALYSIS SLM4ARY
.The replacement valve perfonns the same function and meets the design requirements of the existing valve. Failure may include leakage past the Isolation valve. Since the valve meets the original design criteria, probability of~fallure is not increased.
EWR SlMMRY I EWR 88-340A DESCRIPTION: Replacement of 1-MS-110 was necessary due to inadequate function. Tf'e existing valve was a 1-1/2" Henry Vogt bolted bonnet gate valve (Series SW12111). suitable ~ valve replacement in design and material was a 1-1/2" Henry VogtA i welded bonnet gate valve (Series SW2801). The use of a welded bonnet valve eliminated a potential body to bonnet leak for the valve. SAFETY ANALYSIS SlMMRY;- The replacement valve performs the same function and meets the design requirements of the existing-valve.- Failure may include leakage past the 1solatlon- valve. Since the valve meets the original design criteria, the; probability for failure is not Increased. li:
.r 88-340A
V: c: I? EWt SLM %RY s EWR 88-340B .; DESCRIPTION: Replacement of 1-MS-367.was necessary due to seat leakby. The existing valve was a 1" Henry Vogt bolted bonnet gate valve (Series. SW12111).- A-..sultable valve replacement in design and material was a 1" Henry Vogt welded bonnet gate-valve (Series SW2801). The use of a welded bonnet valve eliminated a potentisi j body to bonnet leak for the valve. ' SAFETY ANALYSIS StM %RY: The replacement valve performs the same function :and meets the- design requirements of the existing valve. Fallure may Include leakage past the isolation valve. Since tho' valve meets the original design criteria, the probabi1Ity.for fallure Is not increased. 9 88-340B i
k EbR StM4ARY EWR 88-340C DESCRIPTION: Replacement of 1-MS-104 was necessary due to seat leakby. The existing valve-was a 1" Henry Vogt bolted bonnet gate valve (Series SW12111). A sultable
-valve replacement-In design and material was a 1" :enry Vogt welded bonnet gate valve (Series-SW28011. The use of a welded bonnet valve eliminated a potential body to bonnet leak for the valve.
SAFETY #% LYSIS SlM4ARY:
-The. replacement valve performs the same function- and meets the design requirements of the existing valve. Since the valve meets the original design criteria, the probability for failure is not-Increased.
9 88-340C
m ,., . . . . s Elst S!M94tY EWR Bl.-3400 DESCRIPTION Substitution (replacement) of valve No. 2-MS-394, a 3/4" VGS-60B, is required because the valve has been injected. A suitcole valve replacement in design and mat erials is available. This valve, Vogt figure SW2801, !s an acceptable substitution. SAFETY ANN.YSIS StM WtY The replacement valve performs the same function and meets the design requirements of the existing valve. Fallure may include leakage past the isolation valve. Since the valve meets the original design criteria,
. probability of failure is not increased.
'l
r ?* EWR SGMWtY: F EWR 88-340E DESCRIPT10N: l Replacenent of 1-MS-108 is necessary due to seal leakby. The existing valve was a 1-1/2" Henry Vogt bolted bonnet gate valve (Series SW12111). A suitable ; valve replacerrent in design and material was a 1-1/2" Henry Vogt wolded bonnet gate valve (SW 2801). The use of a weldo1 bonnet valve eliminated a potential body to bonnet leak for the valve. SAFETY M MLYSIS SlM MRY: The replacerrent valve performs the same function and meets the design requirements of the existing valve. Since the valve meets original design criteria, the probability for failure is not increased. i P k b 88-340E
EWt SLM%RY {. EWR 88-340F DESCRIPTICN Replacement of 1-MS-106 is necessary due to inadequate function. The existing valve is a 1" Henry Vogt bolted bonnet gate valve (serics SW;211 0 . A avitable valve replacement in design and material is a 1" Henry Vogt welded bonnet gate i valve (series SW2801). The use of a wolded bonnet valve will eliminate potential body to bonnet leak of the valve. I SAFETY ANALYSIS SLM MRY The replacement valve perfonns the sme function and neets the design requirements of the existing valve. Since the valve meets the original design ' criteria, the probability for failure is not increased, i 88-340F
F f I-EWR SLM SRY EWR 88-3401 i DESCRIPTION: Substitution (replacenent) of Valve No.1-FW-76 was required due to leakage. A i' suitable valve replacement in design and neterial was available. This valve,
.Vogt Figure SW2801 was an acceptable substitution.
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SAFETY ANALYSIS SlM MRY: The replacenent valve perforns the sane function and neets the design requirements of the existing valve. Since the valve neets the original design criteria, probability of failure is not increased. l 88-3401
m I Oft SLMWtY r EWR 88-340J DESCRIPTICN 1-MS-398 is a 3/4" Henry Vogt bolted bonnet gate valve and is the downstream isolation valve for steam trap T0-423C. There is a nistory or body to bonnet leaks on bolted bonnet MS valves. The trap assembly is to be retroved to allow for maintenance of the upstream isolation valve. 1-MS-398 Is to be replaced with a welded bonnet gate valve while the trap assembly is rerreved to prevent ; the occurrence of a body to bonnet leak. This type of valve replacement is acceptable with regard to design and materlats. SAFETY ANALYSIS SLMWtY The replacement valve performs the same function and reets the same design requirements as the existing valve. Since the replacement valve meets the original design criteria, the probability of failure is not increased. l l 9 I
I [ , L EWt SLM%Ry l EWR 88-340K _DESCRIPTf(W This'EWR provides the guidance for the replacement of valve 1-MS-396. Since it is an exact one for one replacement, and engineering evaluation is unnecessary. This EWR was used sirply to correct inaccuracles about the valve on station doctments, such as the Q-List, UFSAR, and Station Drawings. ! i i t:
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i SAFE"TY #MLYSis StMgRy I Since this is an exact one for one replacement, there are no unreviewed safety questions. The original design basis remains unaltered. The margin of safety t as described in the Technical Specifications is not reduced. > L 1 l f
EWt SLM%RY EWR 88-340L DESCRIPTION: Substitution (replacement) of Valve No. 1-MS-34, A 1-1/2" VGS-608, was required due to leakage. A suitable valve replacement in design and rnaterial was available. This valve, Vogt Figure SW 2801, was an acceptable substitution. SAFETY MMLYSIS SLM%RY: The replacement valve performs the same function and meets the design requirements of the existing valve. Since the valve meets the original design criteria, probability of failure is not increased. 88-340L
gn= - 4 EWR SlMWtY
. EWt 88-340M DESCRIPTICN: ! Substitution (replacement) of Valve No. 1-MS-32, an 1-1/2" VGS-69B, was required due to leakage. A suitable valve replacerent in design and material I
was available. This valve, Vogt Figure SW 2801, was an acceptable substitution, i SAFETY M MLYSIS SLM WtY: i The replacenent valve performs the same function and meets the design requirements of the existing valve. Since the valve meets the original design criteria, probability of failure is not increased. i a b T I f i e 88-340M
I EWR StM %RY EWR 88-340N i i DESCRIPTICN: Replacenent of 1-MS-359 is necessary due to seat leakby. The existing valve Is a 1" Henry Vogt bolted bonnet gate valve (Series SW12111). A suitable valve l replacenent in design and materials is a 1" Henry Vogt welded bonnet gate valve (Series SW2801). The use of a welded bonnet valve will eliminate a ! potential body to bonnet leak for the valve. SAFETY ANALYSIS SLM%JR : The replacenent valve performed the s wne function and neets the design requirements of the existing valve. Since the valve neets the design criteria, the probability for failure is not increased. i 88-340N
F t-EWR SLSOERY i EWR 88-3400
. p DESCRIPTION:
Replacement of.1-MS-355 is necessary due to seat leakby. The existing valve is a 1" Henry Vogt bolted bonnet gate valve (Series SW12111). A suitable valve , replacement in design and material is a 1" Henry Vogt welded bonnet gate valve (Series SW2801). The use of a welded bonnet valve will eliminate a potential j body to bonnet leak for the valve. SAFETY ANALYSIS SLNORRY: The replacement valve performs the same function and meets the design requirements of the existing valve. Since the valve meets the original design criteria, the probability for failure is not increased. ; l l l l L L L 88-3400 L
E EWR SLM %RY EWR 88-340Q i DESCRIPTICN: Replacement of 1-MS-356 Is necessary due to body washout. The existing valve Is a Henry Vogt bolted bonnet valve (Series SW12111). The Henry Vogt welded { bonnet gate valve (Series SW2801) is exactly the same as the bolted in terms of j' material, design, and flow coefficient ;therefore, it is an accepted i replacement. t i l SAFETY ANALYSIS SLM %RY: The replacement valve performs the same function and meets design requirements of the existing valve. Since the valve meets original design criteria, the probability of failure is not increased. I 88-340Q
l i L EWR SLM4ARY
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EWR 88-38+0P I i DESCRIPTION: Replacement of 01-MS-358 is necessary due to body washout. The existing valve is-a Henry Vogt bolted bonnet gate valve (Series SW12111). The Henry Vogt j welded bonnet gate valve (Series SW2801) is exactly the same as the bolted design in tenns of materials, design and flow coefficient ;therefore, it is an ' acceptable replacement. SAFETY ANALYSIS StM4ARY: The replacement valve perfonns the same function and meets design requirements I of the existing valve. Since the valve meets original design criteria, the probability of failure is not increased. 88-340P
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EWt SLM%RY EWR 88-341. ! DESCRIPTION - i This EWR concerns the replacement' of a motor starter for a num in H2 Analyzer (COMSIP, Inc.). The Installed starter is obsolete. The new starter and-the Installed starter have the same electrical characteristics, thus the new l u, starter is an acceptable replacement. il 4 e SAFETY ANALYSIS SLM SRY The change is a part replacement only. There ls no change to'the design basis of the equipment / system. Qualification of the H2 analyzer remains intact. t t
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c k EMt SLM %Ry EWR 88-344 & A i _DESCRIPTI(N This EWR provides instruction for the Inspection and replacerrent of parts (reed switch wires and terminal blocks) Insido quallfled target rock SOVs. A cracking problem was found on reed switch wires and terminal blocks at other plants. This work was done in response to target rock service bulletin SB 88-01 and an INPO Information release (dated 10/31/88). a SAFETY ANM.YSIS SLM %RY , The part replacement enhances valve reliability and decreases the potential spurious operation of valves. 1 l l Y
M SLM %RY
; EWR 88-346 c
DESCRIPT!G( The existing globe valve (02-SW-1027) is clogged and cannot be rodded out. I This valve will be replaced with a gate valve that can be rodded out. [ seismic evaluation was performed and concluded that the seismic Integrity of A the line is maintained, p l l:. i t t ? SAFETY ANALYSIS SLM ERY The replacement valve is identical except for the type. The performance characteristles of the system or components are not affected. The new valve
- will not affect system operability. The replacement enhances system availability by allowing in-place cleaning Instead of renoval or_ replacement of clogged valves.;
9 L L 1
T 4 ; ,[. i .- EMt SLM%RY t: t EWR 88-346A & B & C , e DESCRIPTim I The existing globe valve is clogged up and cannot be rodded out. This valve
- will be replaced with a gate valve that can be rodded out. An evaluation was :
performed and' the valve is acceptable. The affected valves to be replaced are , 2-SW-332, 1-SW-229, and 2-SW-301. l- o
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i i l \, ' SAFETY # Mt.YSIS SLM %Ry : The replacenent valves are identical except for the type. The performance characteristics of the system and components are not affected. Replacement !
. enhances system availability by allowing inplace cleaning, ,
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Elst SLM%RY
- I EWR 88-347A, W ,
i (4SCRIPTi(N j. This EWR is used for the replacement of valves 1-AS-304 (13" Bolted Bonnet Gate > L Valve) and 2-WT-375 (3/4" Bolted Bonnet Gate Velve). Both valves will be ' replaced by a Henry Vogt Series SW2801 Welded Bonnet Gate Valve of the same ; size as the axisting valve'. I j. f 1 SAFETY A>W yst$ S W %RY This nodification does not affect system operation, nor does it affect the + design' basis ~ as described in the UFSAR. The only change to the UFSAR is to a figure in order to Indicate the new valve identification nmber. These valves , are not specifically addressed in the Technical Specifications. 6
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! EWR 88-350 DESCRIPTIGd 4
i= The electro-manual pliot valve for the cardox master delp valve has been redesigned by the manufacturer. The valvo and trcunting plate are a direct replacement with limited fitting connections which are supplied by the vendor. The valve has been redesigned by the vendor to elliminate the linkage- ~ 1evers
, which had proven to be a trouble spot in operatlon, i
F l _SNTTY ANALYSIS SW9WtY The replacement valve is the vendor's redesigned tredel and is intended for this appilcation. Fire protection to the 2JEDG will not be adversely affected. Technical specifications will be ccrmlied with. I l
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EWR 88-355, & A DESCRIPTI(N r The ribbon connector cable for Westinghouse Radiation it>nitors is no longer made by SIGNAFLO. New cable made by Thomas & Betts will be used for the J105 and J106 connectors In Rad Monitors 1-VG-RM-105, 1-VG-RM-106, and 1-VG-RM-108. i m e h-. SAFETY ANALYSIS SLMmRY The replacement cable has the same function as the existing cable. No unreviewed safety questions are involved. The modification does not impact the
-UFSAR nor does it change Technical Specifications. No Protective circuitry is involvod. Margin of safety has not been, impacted by this rx>dification.
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L , EMt SLM%ftY ' EWR 88 361 C A .
" -i DESCRIPTIGd i
This EWR allows the use of new NUPRO valves, Model Ib. SS-4C-1/3 to replace the ex1 sting valves, NUPRO Model No. SS-4CA-5 for valve 1-HC-H2A-101. The valves .{ are used as relief / check valves on the callbration gas line, reagent gas line, [ and the reagent gas relief'Ilne. i. h L, t-t b l' ; l ! 1 i SAFETY # % LYSIS SlM %RY The use of new valves does not change the design basis of any conponent/ system. The new valves are quallfled (environmentally and selsmically) for their application. b 4
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, EWR 88-363 EWR SLM %RY p
I I _ DESCRIPTION [ This EWR- provides the requirements necessary for replacement of a rubber air j ,, line to the spent fuel pit canal gate seals, with a bralded stainless steel jL hose.
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4 r SAFETY #MLYSIS SLMMRY The replacement component fulfills the exact srvne form, fit, and function as the existing rubber air hose but provides additional abraslon resistance and decreases likollhood of hose failure, I i 1
l Ebst SW9%fty D1R BP-364 6 DEScit!PTICN ' Currently, there is no means of hydrostatic testing the Charging Punp mintrun ; flov bypass lines. A 3/4" test connection and valve will be added to allow for ; this testing. ' 1 i SAFETY ANALYSIS SLMufty The new pipe .and valve meet the system design requirements and do not impact system operation. Since the pipe and valve meet the system design criteria, potential of failure is not increased. The modification will be performed in mode 6 with the reactor de-fueled since all three charging punps must be _ secured in. order to perfonn this work. e 4 k 6 I 9
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EWR SLM MRy
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EWR 88-367 1 I DESCRIPTI(N j i: This EWR allows the use of safety related and direct replacement parts on i breaker 2-EP-BKR-24J1-4 (for 2-QS-P-1B) which is installed on a bus required to be' environmentally ouallfled. This EWR also allows the drilling of two holes to ' trount the switch plate . . J (, F-i SAFETY ANALYSIS SLM MRy i, The design basis or function of the breaker does not change and is not
- i. affected. The breaker is quallfled for its application.
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N l . / (.: f i l EWR 88-369 EWR SIM4ARY DESCRIPTION Substitution (replacement) of 3/4" T-88"s is required due to leakage. A suitable valve replacement in design and materials is available. This valve, Conval figure 12G2 or Yanvey 56178, is an acceptable substitution. (: SAFETY ANALYSIS SLM4ARY The Replacement valve perfonns the same function and ' meets the design requiroments of the existing valve. Since the valve meets the original design criteria, probability of failure is not increased. , L
t Oft ' SLMWty ' a EWR 88-369E DESCRIPTim Substitution (replacement) of valve No. 2-RC-90, a 3/4" T-88, is'recuired due [ to leakage. A suitable valve replacement in desic" and materials it. evallable. This valve, Conval figure 12G2, is an accepteble subst!tution. Y 1 SAFETY # # LYSIS SLM %RY i The replacement valve perfonns the same function and meets the design requirements of !.he existing valve. Since the valve meets the original design criteria, probability of failure Is not increased. 9 i I I
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L . p [ EWR SlM %RY EWR 88-369H p DESCRIPTION i-F Substitution (replacunent) of valve No.2-RC-80, a 3/4" T-88, is reovired due to L leakage. A suitable valve replacement in design and veterials is available. This valve, Conval figure 12G2, Is an acceptable substitution. l- . l. f' SAFETY'#MLYSIS SLM MRY 1-The replacement valve performs the same function and meets the design requirements of the existing valve. Since the valve meets the original design criteria, probability of failure is not increased, a
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EWR 88-369J twR stM%Ry DESCRIPTI(N Substitution (replacenent) of valve No 2-SI-243,a 3/4" T-88, is reautred due to leakage. A suitable valve replacement in design and materials Is available. This valve, Conval figure 12G2, is an acceptable substitution. I SAFETY # MLYSIS StM %RY The replacement valve perfonns the same function and neets the design requirements of the existing valve. Since the valve meets the original design criteria, probability of failure is not increased. i l _.1
Det SLM4Mty EWR 88-369K j DESCRIPTI(N Substitution (replacement) of 2-RC-20, a 3/4" T-88 valve, is required due to leakage. A suitable valve replacement in design and materials is available. This valve, Conval figure 12G2, Is an acceptable substItutton, I SAFETY ANN.YSIS SLM %RY The replacement valve performs the same function and meets the design requiroments of_the existing valve. Since the valve meets the original design criterla, probability of failure is not increased, 6. l l
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,j EWR SLM %Ry:
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EWR 88-369L DESCRIPTICN 4
- Substitution (replacement) of 2-51-230 a 3/4"'T-88 valve is required due to S'
leakage. A suitable valve replacement in design and materials is available.- This valve,-Conval figure 12G2, is an acceptable substitution. SAFETY ANALYSIS StM %RY. s fI I.' The replacement . valve performs the same functlon and meets the design requirements of the existing valve. Since the valve meets the original design-
, . criteria, probability of failure is not increased.
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W n: t, W ', . l I. 0-~' ' EWt SUP9%RY l C.. 7 EWR 88-369M 't . DESCRIPT10N " xx Substitution (replacement) of valve.2-SI-214, a 3/4" T-88Y, Is required due to
' leakage. A.sultable valve replacement in design and materials is available, This> valve, Conval figure 12G2, Is an acceptable-substitution,
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t t a SAFETY ANALYSIS SLM SRY-The- replacement valve performs the same function and meets the' design
-requirements of the existing valve. Since the valve meets the original design criteria,~ probability of failure is not increased.
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EWR SlM %RY
.EWR 88-371-DESCRIPTION:
1-CH-278 is a 3/4" Rockwell Edwards vertical stem globe valve that " developed a leak In its body. The Yarway Welbond valve is constructed of " compatible materials and has dimensions similar to the Rockwell valve. The pressure / temperature rating of the Yarway valve exceeds the design conditions-of the CVCS system. Therefore, the Yarway valve is acceptable. SAFETY ANALYSIS SLM MRY The replacement valve will perform the same function as the original valve. The new valve's pressure / temperature requirements exceed the design conditions of the CVCS system. The new valve :neets the design requirements of the original valve. Evaluation concludes that seismic Integrity of~ the 1Ine. Is-maintained. i I t 88-371
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', 4 EWR SLD9lARY.:
;EWR 88-371C DESCRIPTION-
~
c, Replacement of'1-CH-265, a 3/4" Rockwell Edwards 1X58-N valve-Is required'due to-
. seat-erosion. A suitable valve replacement' in design and materials is the Yarway figure 5617B valve. '
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l:The replacement valve serves the same function as the existing valve.'Since the j replacement valve meets.the original eriteria, the probabillty of failure 1 s'
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'not' increased.
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( y-n F ,.. oj , - EMt- $4MMRY - h i . ? EWR 88-373 !? DESCRIPTIGd-n P' This EWR provided for a modification to be performed on valve 1-QS-4 which is a
. swing check valve. The modification, which'Is intended to stop seat leakage,.
t. simply. entails the welding of the disc nut to the disc stud. 'This will ensure 7 . proper seating of the valv'e. f SAFETY ANALYSIS SLMMRY
.j
-l This modification will only: Improve the present operation of the valve, but it
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.will not affect the design basis of the component or the system. This- L modification will not increase the probability of occurrence or the consequences of an accident previously evaluated, nor adversely affect any equipment:Important to safety. The margin-of safety as defined in the basis for:any.-technical spectfIcatlon wi11 not be reduced.
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ej n . o L. . EWR 89-001 , i
, DESCRIPTION n--
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' -This EWR replaced clapper ann retalning block studs on Anchor Darling swing check valves. These valves are' located on the Si accmulator .^ discharge lines..
- 'The~. studs are being replaced because' Industry experience has shown these studs to be susceptible to stress corrosion cracking..
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! adversely affected. Replacement retalning studs are of an acceptable material-that is less susceptible to stress corrosion cracking.- Replacement' studs are recmmended by - the valve manufacturer and .are being supplied by same.
Operability tests required by the' Technical Specifications -- will be perfonned *
, before'the check valves are declared operable.
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g. H EWR SLM MRY : EWR 89-002 l DESCRIPTICN A cross connect IIne between the ccrnponent cooling water (CC) and liquid waste (LW) systems was installed by EWR 85-622. The purpose of the cross connect line ; 1s to allow' the chromated component cooling water to drain to the Evaporator ! Test Tanks (1-LW-TK-5A,B) when maintenance must be performed on the CC system. 1 When maintenance has been completed the Evaporator Test Tank Punps (1-LW-P-5A,B) were supposed to punp the chromatd water back through the. cross connect line to the CC system. This has not worked properly. The cross connect ilne is connected to the CC system in the supply header to several coolers.
- Because 'of- the high operating pressure In the header, the Evaporator Test Tank Punp cannot punp to the CC system while it is in operation.
This EWR will' cut the cross connect line near where it enters- the component cooling water system, and reroute it to the return header from the coolers. The operating pressure in the return header is 50 PSI lower. This should allow the evaporator test tank punps to pump to the CC system when it I;. operating. > SAFETY ANALYSIS SLM MRY Operation of the CC and LW systems is not affected by this change. This
- modification does not create the possibility of an accident of a different type than previously identified in the UFSAR, nor does it increase the probability of an occurrence of an accident previously' evaluated in the UFSAR.
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k! s h EWR SW9%RY q a -. . EWR 89-007 C A - H e ll: g _ DESCRIPTI(N m
- The. Emergency Operating Procedures require the operator to take post.LOCA action to prevent the nitrogen blanket above the borated water In the Safety injection Accurulators- from entering the reactor coolant. system (RCS). This.Is acecripilshed 'by either cloalng the isolation valves (MOVs) between the y" accurulators and the RCS, or by opening the vent valves in the accunulator vent lines. Since the isolation MOVs are located: below the flood plane, qualification is not-feasible. Therefore, the solenold operator valves (SOVs) which provide actuating air to the vent valves were selected for. upg rade. The L
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.11mit swltches associated with the vent valves, which provide position Indication, are upgraded as well because of conmon electrical circuitry with p; the SOVs.
SAFETY ANALYSIS SW94ARY Operation of accunulator vent valves has not been adversely affected. The replacement SOVs are the environmentally quallfled version of the' exiting SOVs. The new limit switches are reconmended by the- valve manufacturer. The modification does not prevent the accumulators from discharging borated water ill to RCS as-described in the UFSAR. h l lc. 1k .s !
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F'!L _EWR'89J 010 0' o _DESCRIPTIGd
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.This Instructlon 'for valve operator modlfIcatIon for EWR ~provides.
1-CH-MOV-2287C 1n accordance with IEB85-03 program. The '1Imlt. swltch wiring W nodifIcatlon. set up the operator for the four space 1Imit swltch rotor operation. This allowed the torque bypass switch to be set more' conservatively
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. -i This - modification does not- decrease the integrity of the motor. operated valve (MOV). The MOVs rollability will be increased.
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W{'O[ P yy H l u: ' EWt StMWty ' y 3 4 EWR 89-011, A' - [:.7 / P _ DESCRIPTION d Thi" :- EWR i- provides - detail s . and procedures for performing Internal vlsual i t' inspection of the buried and encased service water piping 24"-WS-26-151-Q3,
, 24"WS-34-151-Q3, 4"-WS-67-151-Q3 and 4"-WS-68-151-Q3 by rerrote nobile closed
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circuit' television system. This piping Is located _In the . area of detected leakage. ~This inspection will help to identify the source of the leakage, so i corrective measures may be developed and implemented. 1 4 /e d 4 , 1it ,v. ;
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a SAFETY ANALYSIS StM44RY y' This inspection does not . increase probability of occurrence or the consequences
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l of' an accident or malfunction of equipment important to Safety:and previously . evaluated-In Section 9.2.1 of the UFSARj and does not create possibility for an accident or malfunction of a different type than'any evaluated.previously in the~UFSAR. The Inspection also does not reduce margin of safety as defined' in-
' the basis of 3/4.7.4. and 3/4.7.7'of the-. Technical Spectfications.
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, ' _DESCRIPTICN Leakage -- In the service water piping was discovered in the alleyway between the
' Service Bullding and the Malr. Valve House. Groundwater observation wells have been installed. and! are monitored- to = determine the water level. This EWR addresses' icworing of the water table- to allow excavation to within six(6) feet -
of the service water
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. lines 4"-WS-67-151-Q3 and 4"-WS-68 151-Q3. The six (6).
icct cf fill is rectu!reci above the f.ervice water line of missile- prntection. L a s t er i r.(; . rtz i.cccr pl i t,hed ut11171rp nrr.s = aro 'v.cllicirt s . Drnation crd l storing was acccrvilshed by Installing Armco Tunnel. Liners. P SAFETY ANALYSIS SLMtARY The Irnr>lcmentation of this ELR cces not incrotn the prebtbility cf cr cte t t,r.c.c ' or the consequences of an accident.- Installat len of the Sm11 points arc vert ical ' shnf t i le,er ar'c: e.=ce.vatirr ti ,hin the liner to e re i r:t e l i. rect r t ri c the cerv!ce water l 'rie cces not affect any safety related er cceroonert oc ces ter
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- The Implcn tr. tat icr. c f .this EVR c'en s r ot cr u.t r t Fr rc s s it l i i t y (.t ci i cc ic'crt r
.p rr nol forct. i cr 'ef a c lfferert type ther. any evalueted previously, ncr r'ert. It Mt ect any safety systerrt cr ccr pcrents.
1 The IrT.lementat ien of this EWR does -not reduce the rergin of safety as defirec in the basis of any technict1 specification. tb functicnal changes are belrig -l
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made to any safety related systems or ccmponents. f
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EWR 89-028, A-H DESCRIPTION This EER provides the detalls and procedures for the temporary repair of the
. portion-of. buried 4"'dlameter service water pipes 4"-WS-66, 67, 68,- 69-151-Q3.
After, the Installation of new 4" diameter pipes per DCP-89-01, all-the above V pipes will be disconnected frcm the 24" diameter headers, filled up with : grout and abandoned In' place. The temporary repair will decrease service-water-
' leakage into the basements of surrounding buildings. It wl11 also prevent short D. term service water' piping leakage rate increase.
(- SAFETY ANALYSIS SLM ERY The Implementation of this EWR does not increase the probabl.11ty of occurrence or the consequence of an accident or malfunction of equipment important: to safety and 'previously evaluated .I n Sectlpn 9.2.1 of the UFSAR; nor does it create the possibility for an accident or malfunction of a different type than - any evaluated previously in the UFSAR. This repair also does not reduce the margin of safety as defined
-Technical Specifications.
In the- basis of' 3/4.7.4 and 3/4.7.7 'of- the No functional changes are being made to any safety ' related system or ccmponents. r l
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EWR SLM @RY
'l EWRj89-036, A,.B,-C, D, E, F, G, I,K,L,M,P 1
DESCRIPTION i A1 Safety System Functional Inspection (SSFI) was perfonned on the instrtrnent
' ' AircSystem at North Anna In 1988. The inspection report included a sttrmary ' of ifindings and recomnendations.
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I One near-tenn reconmendat ion was 'that a diesel or gasoline powered ai r- l compressor be connected to the compressed air' system. This reconmendat ion j stems from the unreliability of the Instrument: Air System compressors (1/2-IA-C-1) which are powered from the emergency busses. Since instrtrnent air l 1s Important -to continued unit operation, the unreliability of the system is - j compounded by the fact that the normal instrument air supply ccmes from the Redundant (Sullair) Instrunent Air Compressors (1/2-IA-C5), which are powered from non-safety related sources. Therefore, loss of offsite power Jeopardizes operation of. the Instrtment Air System.
!? This' EWR, including addenda through and. Including -036P, installs a temporary i diesel-driven instrtrnent air compressor to be used for emergency purposes; and an electrically-driven Instrtrnent air compressor to provide the primary source.
of Instrument air while maintenance is being performed on the -Sullair l Ccmpressors.- These compressors are Installed in parallel to utilize all piping ] and components down stream of the tie-In, such 'as the desiccant air dryer which was also added as a result of this EWR, j SAFETY ANALYSIS ~StM MRy i 4 This EWR does not modify any safety related equipment, nor reduce the margin'of r safety of.the systems defined in the Technical Specifications; .therefore, 'no ) unreviewed safety question is created. q
' This EWR' does not change- the design basis or nonnal operation of the Instrtment !
Air System. It is being implemented to maintain system reliability during a- ! loss of offsite power scenarlo; therefore, the probability of occurrence or the -{ consequences of an accident or malfunction or equipment Important to safety is ~! not Increased. The implementation of this EWR provides additional assurance that continuous f Instrunent air will be available in accordance with the system design as i
. ,.- described in the UFSAR. The possibility for an-accident or malfunction of a T '
different type than previously evaluated does not exist. The Instrunent Alr System is not addressed in the Technical Specifications; however, f' s the ccmponents to be installed under this EWR provide ccrnpressed air which meets or exceeds the requirements of the system's original design criteria. Therefore, based on the Technical Specification crlteria, no increased probability of an accident will occur, a
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p.j 4- + p: J EWR SW9%RY 3 ju EWR 89-037 < _ DESCRIPTION During the disassenbly of the fuel handling tool, a few of the spiral pins were distorted and could not be reused-, Since no replacement spiral pins were available,. .it was decided to manufacture solid pins, similar to those on the
-other fuel handling tools: at 14APS. The material that the. new pin is constructed of is compatible with the' with the other materials and the
, environment (borated water). Westinghouse, the manufacturer of the fuel tool,.
\ concurs that this Is an acceptable modifIcatlon. '
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SAFETY ANN.ySIS SLM4ARY
..t Use of the new type of pin will not alter the function of the tool.-The new pins'are equal to or exceed the original pins in terms of strength. ]
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A fuel handling accident :outside of contalnnent is addressed in- UFSAR Section 4 15.45. o Since operation of the tool is not . changed, no new type of an unanalyzed -j accident is introduced. . 7 i ('
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-EWR 89-045 /
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-DESCRIPTIGN;
- This EWR provided Instruct'lons for valve operator modification for-L 2-CH-MOV-2370 In accordance with the station's IEB 85-03 program. The Ilmit switch wiring nudification' set up the operator for four Ilmit -switch rotor operation.' This allowed the torque bypass switch to be set more conservatively without affecting the. light Indication.or limit stop.
L l l SAFETY ANALYSIS SLM4ARY This modification does not decrease the integrity of the motor operated valve (MOV). The MOV reliability will be increased.
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c 1,. DESCRIPTIG4 3 1 v- Thisl EWR- provided- Instructlon for valve operator _ nodifIcation ' for y/ 02-SI-MOV-2885C in accordance with IEB 85-03 program. The limit switch wiring. rrodification set'. up the operator for four limit switch rotor operation. This allowed the torque bypas's ' switch to be set nere conservatively without' ' affecting the light Indicator or limit stop. , i t b I I r r , SAFETY AtMLYSIS StM %RY
. This ~ modification does' not decrease the Integrity of the motor operated valve (MOV).:The MOV reliabillty-will be increased.
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1 DESCRIPTIGd
'This EWR~ provided Instructions for valve operator nodi fIcat lon for
'2-SI-MOV-2885C & 2885D in accordance with-the station's IEB 85-03 program. The q 1Imlt swltch . wiring mod I f.Icat ion. set up the operator for four 1Imlt swltch rotor operation. This allowed the' toroue bypass switch to be set nore '
conservatively without affecting- the light Indication or limit stop. I 1 I 1 i i SAFETY ANALYSIS SLM44RY 1 V l This modification does not decrease the integrity of the notor operated valve OOV). The MOV reliabl11ty will be increased. 6
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, _ .:V EWR 89-049 & A' >
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- This. EWRL provided instructions. for valve operator _ modification for 02-SI-M0V-2862B in accordance with the station's IEB 85-03 program. The limit' -
, . switch wiring modification set ~ up the operator for four limit switch rotor m operation. This' allowed the torque bypass switch to be set more conservatively G without affecting the light indication or limit stop, t
h L l SAFETY ANALYSIS SLM MRy This modification does not decrease the integrity of the motor operated valve <, -(MOV),and the MOV reliability will be increased; a [ 1' o Ik l l, c , I i
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y s 4 EWR.89-050 g/ [/ j . , DESCRIPT10N This EWR provided Instruction for valve. operator modification for 02-SI-MOV-2860B In accordance with the station's IEB 85-03 program..The Ilmit switch wiring modification set.up the operator for four limit switch rotor ; operation. This allowed the torque bypass switch to be set more conservatively wlthout affeetIng th'e 1Ight Indicatlon or 1Imlt stop..
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. This modification does not decrease the Integrity of the motor operated valve (MOV) and the MOV reliability will be increased.
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.- M g-E' =EWR 89-053, A s
DESCRIPTION
-This EWR provided instructions for valve operator modification for
.02-QS-MOV-201B in accordance wlth IEB 85-03 program. The 1Imlt swltch wlrIng modification ' set up the operator for four limit switch rotor operation. This allowed the torque bypass switch to be set more conservatively without affecting light Indication or limit stop.,
SAFETY ANALYSIS SL794ARY This modifIcatlon does not decrease the integrity of the motor operated valve ] (MOV). The MOV-reliability will be increased. l l l
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EWR SLM %RY EWR 89-053 & B,-C, D E DESCRIPTION This EWR provided instructions for valve operator modification for 02-QS-MOV-201A in accordance with the station's IEB 85-03; program. The limit switch wiring modification set up the operator for four limit switch rotor operation. This allowed the torque bypass switch to be set more conservatively
.without affecting the light indication or limit stop.
1 l 1 SAFETY ANALYSIS SLM MRY
.This modification does not decrease the integrity of the motor operated valve (MOV) and the M0V reliability will be increased.
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c 7 y , EWR SW9%Ry EWR 89-054 & A & B. _ DESCRIPTION This EWR provided instructions for valve operator modification for 02-RS-M0V-200A and B in accordance.with the station's IEB 85-03 program. The limit switch ' wiring modification set up the operator.for four limit switch rotor operation. This allowed the torque switch bypass to be set more conservatively without affecting the light indication or limit stop, i SAFETY ANALYSIS SW9%RY This modification does not decrease the integrity of the motor operated valve. (MOV) and the MOV reliability will be increased. e l l l
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EWR 89-055 ,
,w DESCRIPflG8 r.iy ,
s 6 , This.EWR provided Instruction for valve operator modification for .
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02-RS-MOV-201A in accordance with IEB 85-03 program. The Ilmit switch wiring i
;.. . modification set up the operator for four 11mit switch rotor operation. This- -
- allowed the torque bypass switch to be set more conservatively without L;, affecting the light- Indication or limit stop.
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- This modification does not' decrease the integrity of the motor operated. motor
-valve (MOV). The MOV reliability will be increased.
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.EWR 89-055A & B '
DESCRIPTim l This EWR provided instructions for valve operator modification for 02-RS-MOV 201B In accordan' c e with IEB 85-03 program. The 1Imlt swltch wiring modification set up the operator ~ for four limit switch rotor operation. This allowed the torque bypass switch to be set more conservatively without affecting the light _ indication or limit stop.
- SAFETY ANALYSIS SLDNARy 'i
( This modification does not decrease the Integrity of the motor operated valve (MOV). The MOV reliability will be increased. 4 i
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$ t EWR 89-061'& AL p
_DESCRIPTLON
- This EWR provided instructions for valve operator modification for 02-RS-MOV-256A in accordance with the station's IEB 85-03 program. The limit switch wiring modification set up the operator for four limit switch rotor ,
operation. This allowed the torque bypass switch to be set more conservatively without affecting the light indication or ' limit stop. SAFETY ANM.YSIS SLM WtY The modification does not decrease the integrity of the motor operated valve (MOV). The M0V reliability will be increased. i L
r-; I 'r 't , . e. "- , , 6 EWR SLMMRy , EWR 89-063 S B DESCRIPTI(N e When- 1-SW-295, and 296 and_297 were cut _out, fit was discovered that the header , isolation valves were not holding therefore new valves could not be - Installed due' tol the ; leakage problem. As a temporary fix, the nipples will be threaded and caps installed. This is acceptable per ASME section !!! for. class three (3) Ilnes. i c. t SAFETY ANALYSIS St99%RY Installation of the threaded cap will not affect system _ operability or the.
, Integrity of the SW piping as a pressure boundary. The tradi ficat ion is acceptable per ASME Section 11! Subsection ND. New valves will be installed ~ln the-1989 refueling outage. The cap is compatible with the existing pipe and Is, acceptable to use on this class piping per NAS-1009.
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EMt StMWtY EWR 89-084 &: A- & 'B
,{ DESCRIPTIG4
.New bearings were needed.for 02-HV-P-20A motor. This EWR evaluated and dedicated non-safety related bearings for use in this safety related motor.
If .r 'k. + ,.. l SAFETY ANALYSIS SLMWty - J l The replacement bearings-will provide the same function and meet'the same
. design requirements as the original bearings. Since the bearings meet the original, design criteria, the probability of failure is not increased.
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{ DESCRIPTI(DI The exposure of spring hanger 1-QS-SH-125 to the outside environment has' caused' the rusting and degradation of this component. This component will be. repl aced
;, . with an identical component except that.it will be galvanized per ASTM A-153-and the spring will be coated with neoprene. , w 4
1.! SA EfY ANALYSIS SLM MRY 3, Operatlon of the.-RWST, and Its support -systems; will .not- be = altered by '
. replacement of the pipe hanger. Modification to the hanger, which is an exact
; replacement, will preclude outside amblent conditions degrading'the replacement-hanger.. .
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EWR 89-113 DESutiPTlWd t.- h This EWR providcs the justification and machining requirements for replacing
}. /, the valve cap of 2-FW-134, which is currently fabricated from ASTM A216,GR WCB material, with ASTM A 105 material, jy
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SAFETY M MLYSIS SlM MRY The operation and design furction of the AFW system is unchanged. The material is consistent with this application therefore the Integrity of the component is maintained. Failure of the bonnet would be consistent with accidents previous'ly y . evaluated.
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Det SLMtifty I EWR 88-109- _DESCRIPTig
- This EWR'.wlli cause normally lit annunciator B-7 "CH PP 2C Auto Trip 25J7 LO" ;
to be enabled only when breaker 2$J7 is racked in. i l i s SAFETY AleLYSIS SLMSRY
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This change enables the "C" charging pum J Bus . feeder breaker lockout annunciator to be enabled only if If the J bus breaker is racked in. The
. operability and function of the "C" charging pum is not adversely affected. i This rrodification will ensure better control room operator reliability, ;
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L< EWt SLMSRY jl DlR 88-112,& A & B & C & D & E & F p DESCRIPTICN The Emergency Operating Procedures require that the operator take post LOCA action to prevent the nitrogen blanket above the borated water in the Safety injection Accmulators from entering the Reactor Coolant System (RCS). This is acccrrpl i shed by either closing the Isolation valves (MOVL) between the [1 accmulators and the RCS or by opening the vent valves in the accmulator vent l lines. Since the Isolation MOVs are located below the floor plane, qualification is not feasible. Therefore, the solenold operated valves (SOVs) which provide actuating air to the vent valves were selected for upgrade. The limit switches associated with the vent valves which provide position Indication are upgraded as well because of conTron electrical circultry with the SUVs. SAFETY # MLYSIS SLM MRY Operation of the accmulator vent valve has not been adversely affected. The replacement SOVs are the environmentally quallfled version of the existing SOVs. The new limit switches are recomnended by the valve manufacturer. The modification does not prevent the accmulators from discharging borated water to RCS as described in the UFSAR.
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f~" OdR SLMWty EWR 88-119 C B DESCRIPTICN t- i The renedeling of the Labor Department Offices and breakroom in Warehouse h2 ,
; includes the addition of a dropped ceiling. The addition of the ceiling requires the existing sprinkler system to be modified to add sprinkler heads in the new office areas and breakrocm below the ceiling. The existing sprinkler ,
piping will be tapped for two new branch lines, and twelve new sprinkler heads ; i are being added. f I SAFETY ANALYSIS SLM MRY The modifications described In this EWR are being completed in Warehouse #2 which is outside the protected area. The nodification will require the termorary isolation of the Warehouse #2 fire sprinkler system. The nodifications to the existing sprinklers and the implementation of the n nodifications do not impact any station system important to safety. ' i l l
g, 3 1 EMt SLMWtY 2 f 1 . [ , EWR 88-120 i _DESQtlPTIGd i L This EWR provides the instructions fortrodifying the limitorque actuators for 2-FW-f0V-200A and C. These nodifications are for four (limit switch) rotor ( operation and will be done in accordance with the station's IEB 85-03 valve program. F i L , SAFETY ANALYSIS SLM WtY s This nodification will Improve the valve reliability without adversely affecting its operation. Neither the nodification nor testing will affect safety function of the affected valves. J l
W:c L ,. i > [' EWt SLMMRy [ - I EWR 88-137C & D p DESCRIPTICW Drain valves spray (RS) have been Installed in the SW lines to and from the recirculation exchangers. This was done so that water leaking by 01/02-SW-MOV-101/102 and 105/205 may be drained before reaching the RS heat-exchangers. k
- SAFETY ANALYSIS SW9%RY While the cross tie trotor operated valves (MOVs) are shut, the heat exchanger Inlet and outlet MOV'wl11 be administratively controlled. The valvo and line Installation are selsmically quallfled, and post maintenance . testing requirements conply with appilcable codes and standards. Drain valves will remain closed after Installation except during leak testing. No potential mode fo. cenmon failure exists. The continued operability of the RS heat exchangers and service water headers is ensured during Installation by administratively controlling the inlet and outlet MOV's to the heat exchangers. With the cross tie valves closed, an " Info only" action statement will be entered.
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- n. EMt SG9Wty EWR 88-146'.
_DESCRIPTICN l This EWR provides instruction for valve operator modifications in accordance 4 [ .with the station IEB 85-03 MOV program. This will allcw the torque bypass switch on the valves (1-QS-MOV-100A,B) to be set more conservatively without t affecting the Ilght Indication or the limit stop. ! I p i. i l i i i SAFETY ANAL.YSIS SW9Wty This modification will not adversely affect the operation of the MOV or q associated Interlocks. The operation end reliability of the valves will be ' enhanced. The probability of occurrence or consequences of a malfunction of equipment Important to safety is not increased. The margin of safety as described'in the Technical Specifications is not decreased, l~ 1 I c l 6 5
g$ EWt SLMmRy n . EWR 88-155 & A, B, C, D, E, F (( _DESCRIPTI(N F. II
't.
t Additional SCBA Units are to be placed on the walls of the computer rooms of '
' the Control Room. Engineering has reviewed drawings and concluded that there are no embedded conduit or pipe in there locations. '
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;. . i SAFETY AN4 LYSIS SLM SRY Neither the SCBA bottler or the installation of the mounting brackets-Is safety related. However, the installation will be seismic since there is safety o' related-equipment in the area.
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b EWt SLMWtY ~i EWR 88-156 & A DESCRIPTICN I I i This EWR provides alternative methods for repairing stripped holes in the l assenbly f rame for trounting of the circuit breaker closing spring charging i motor. Repair may be performed by either using through bolts with special locknuts or repairing the stripped holes by welding and retapping to the original configuration. 1 1 SAFETY ANALYSIS SlMWty The work is considered a standard repair practice. This repair will restore the componem to its original condition. This repair will not introduce any new failure andes that have not already been evaluated in the UFSAR, this repair will not' Increase the probability of the failure of any equipment.important to safety already evaluated in the UFSAR and will not affect the margin of safety
- of any components, systems or structures. >
I t e
1. p. Oft NY u EWR 88-157 & A & B _DESCRIPT!G( The Llould Waste Discharge Proportional Sampilng PLmp (1-l.W-P-28) has required frequent maintenance. Much of this maintenance is a result of foreign matter getting into the ptrm. The suction piping for 1-LW-P-28 cones off the bottom of a low point in the 11guld waste effluent piping. When effluent flow stops, 4 particulate matter settles in the low point of the piping and is drawn into the - sampling pump when flow starts again. The EWR gives requirements and Instructions for relocating the suction for : 1-LW-P-28 to a vertical run of pipe downstream from its current location. This point in the system is always flooded and there are no nearby fittings to disrupt flow. The modificatlon will meet the same requirenents as the original Installation. Operation of 1-LW-P-28 will not change. Required testing consists
; of a leak check and functional check. .
1
,S_F #MLYSIS SLMSRY The relocation of the suction of 1-LW-P-28 to a vertical run of pipe, downstream from its current location, will not change the function or operation of the system. The new design meets the original design criterla and the probability of failure is not increased.
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p --~. 4 1 EWt SLM%lty , s li-t EWR 88-161A, B I DESCRIPT1(N ' l t a' This EWR' provides the Instructions necessary for modifying valves r 2-HV-MOV-211A,B / 213A,B and 1-HV-+0V-113A,B,C in accordance with the station's p IEB 85-03 valve program. These modifications are done in order to allow for i i increased toroue switch bypass and overall operator reliability. l (- . [,- I-p i SAFETY ANALYSIS SLMWtY
- This modificatlon will only inprove the operation of the affected MOVs. The design basis of either the system or corror nent is not affected. The margin of safety as described in any Technical Specification will not be reduced. tb
- Safety Related equipment will be adversely affected as a result of this :
I nodification.- % 4 4 i' e t 9
p, I s4 i EWR SLM mRY L EWR 88-178 C A C B & D ; DESCRIPTICH I The present piping configuration for the Auxillary Feedwater pump lube oil coolers allows for reverse flow through the coolers of donna at pumps while a punp running. This [' is is making it difficult to meet the Technical Specification flow (Recirc) of 53GPM for the motor driven punps. Installation of check valves on the oil cooler line will direct all finv through the rectre , line ano eliminate any problems with meeting the flow reculroments of the
- t Technical Specifications. This modification will be done for both units,
[ SAFETY ANALYSIS StMMRY The operation of the AFW system as described in the UFSAR will not be adversely affected. The modification will allcw better indication of recirc flav for a running punp during testing by preventing recirc backflow through the non - running punps oli rooler. The reduction of cooling water flow through the lobe
-oil cooler line of the running punp due to installation of the check valve is small in magnitude, and calculations have shown that the manufactures rococmended cooling water flow for the oil cooler will be maintained.
The check valves are to be Installed on each AFW punps lube oil cooler line. Any failure of a single check valve would not Impair operabilliv of the other two pumps. Therefore, the potential for carnion mode falltm docs not exist. Should the Installed check valve fall to open dur!% operation of the pump, cooling water flow to'the cooler would be prevented. This could cause damage to the punp as lube oli temperature would increase due to lack of cooling. This type of failure would be highly unlikely given the piston / low tension spring design of the Conval check valve and the periodic testing of the valves per the IW program. Any degradation of the valve, preventing full opening of the valve would be identified by the proposed IW testing, and corrective action would be taken to restore integrity of the valve.
Wg f g my f t r I 88-187 (FOR APPLICABLE ADDENDUMS, SEE ATTACHED) i _DESCRIPT!Gd Various Henry Vogt. bolted bonnet valves are experiencing problems such as body to bonnet Icaks, seat leakby and steam cuts. The Conval Y-globe valve is a better design, and It is easier to perfom maintenance on the Conval valve. l Both valves are censtructed of corroat ible materials and have pressure / temperature ratings which exceed that of the piping.. The Conval valve is an acceptable replaconent for the Vogt valve. , I l' t SAFETY MMLYSIS SLM MRY The replacement valves will provide the same function and meet the same design requirements as the original valve. All original design criteria are met. Since the valves meet the original design criteria, the probability of failure is not increased. l l
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, EWR-#. . VALVE # VALVE-REPLACED REPLACEbe4T VALVE
-187G 1-FW-052 3/4 Henry Vogt Series SW2821 3/4".Conval Fig. v11G3' 1/2" Henry Vogt Series SW12141
-187W. 1-MS-337 1/2" Conval'F1g.:#11G2
-187X 1-FW-208 3/4" Henry Vogt Series SW15821 3/4" Conval: Fig. 112G2
. :-187BC. 1-FW-234' 3/4" Henry Vogt.Serles SW66723 3/4" Conval Fig. #12G2 l -- -187BD 1-AS-321 1/2" Henry Vogt Series SW2821 1/2" Conval Fig. #11G2
? -187BF 1-FW-092 3/4" Henry Vogt Series SW12141 3/4" Conval Fig. $11G3 z -187BJ 1-MS-174 3/4" Henry Vogt Series SW2821 3/4" Conval Fig.-411G3
[
-187CK: 1-FW-240 3/4" Henry Vogt Series SW66723 3/4" Conval Fig. #12G2-
-187CM- 1-CN-124 3/4" Henry Vogt Serles "" 3/4" Conval Fig. #12G2 l
--187AZ' 1-MS-309 1/2" Henry Vogt Series SW12141 1/2" Conval Fig. #11G2 tj -187BN 1-MS-313 1/2"' Henry Vogt Serles SW12141 -1/2" Conval Fig. #11G2-
""EWR does not give series W, however it does state that it is one of the' series that had previously been evaluated, y
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EMt SLMW4y EWR 88-206 DESCRIPTION
.This EWR provides reccrnnendations for troubleshooting heat trace circuits with amperage readings outside 21% of nameplate current.
SAFETY ANALYSIS SLM mRy The troubleshooting checks of this EWR will ensure heat tracing ci rcuits are operable and performing as per manufacturer's requirements. 1
( ' i I EWR SLM%RY
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l l EhR 88-220, A ' DESCRIPTION: j tbdifications performed per this EWR Involve enhancements to the grounding system at the Primary Meteorological Tower to improve performance under lighting storm conditions. Modifications consist of the Installation of six (6) new ground rods, two (2) new electrolytic type ground rods, ground mat conductions, new tower ground wire and ground bus at the Instrument rack. i SAFETY #MLYSIS SW9%RY: Operation of the Met Tower is not affected by this change. The Met Tower ; 1s not required to be seismically quallfled. Instninent channels or power ' supplies are not being changed by .this design, nor being cross-tied. The grounding nodifIcatlon wl11 be performed in accordance with NAS-3014 and - STD-EEN-004 No Appendix "R" equipment is affected by this EWR. Modification only affects the primary Met Tower. Existing surveillance testing is not affected and is adequate. The function of the Met Tower Is unchanged. Failure of the grounding network could expose the Met Tower electronics to further lighting strike induced failures. The back up Met Tower cold then be used to rronitor untl1 the primary Met Tower electronics were repaired. 9 88-220SA
EW SLMWty 9 EWR'88-223 _DESCRIPT10N in order to Irnprove and enhance performance of the meteorological twer e equipment during Ilghtening storms this EWR perfortis the following trodi ficat ions; Delta "T" rrodule will be replaced with similar new enhanced L' version, Surge protector boards will be added and wired to provide protection c !- for electronic modules during transients and lighting events, and existing power strips feeding electronic devices will be replaced with new units L equipped with surge protection circuitry to prevent spread of transients and spikes and avoid damages to equipment. s i SAFETY ANALYSIS StMWtY No adverse effects will be caused by replacing the Delta "T" rredules and performing lightening protection enhancements. Operation as defined by UFSAR will not change. Original equipment and tower are non-selsmic. Modification will not cause loss of separation of channels or power supplies, and has no impact on protective circuitry, tb increased risk of fire or explosion exist. There is no ccmnon mode failure scenario created per the attached procedure. The appilcable Technical Specification Action will be entered, and the requirements of emergency plan will be satisfied. No safety related electrical load will-be affected. Work will be performed outside protected area and non-radiological area, therefore there will be no impact on radiation levels. L P
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EWR 88-225 & B C C DESCRIPTI(Bl
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[ ' The old Bendix 01-GW-H2/02-102 analyzer was replaced with a new Servamex unit that consists of. $40A 02 analyzer and T3-MSA Hp analyzer. The old analyzer l was worn out and needed to be replaced. (* i p i SAFETY ANALYSIS SU NARY The -installation of the new; analyzer did not change the UFSAR nor does it affect Technical Speelfications or LCO's. The margin of safety is not Impacted by this . modification. This modification will not increase the probability of an. accident as descr! bed in the UFSAR. No unreviewed safety questions are created by this modification. s t l. l
- , EWR SLM%Ry p
p . L; EWR 88-244, A, B, C t ! DESCRIPTION: I Cracks reported on QCAR No. AR-N-88-255 were inspected and analysed for structural cause. The conclusion reached is that the cracks are dormant and require only surface sealing with an epoxy compound. Periodic inspections will be performed after repairs are irrplemented to ensure that cracks do not ' i reappear Indicating that movement Is still occurring. h SAFETY ANN.YSIS SLM MRy:
- The implementation of this EWR does not increase the probability of occurrence or the consequences of an accident. The wall will perform after repair exactly as it was originally designed to perform. -
The implenentat ion of this EWR does not create the possibility for an accident or reifunction of a different type than any evaluated previously. Functionally and dimensionally the chiller room wall will remain unchanged. The implementation of this EWR does not reduce the margin of safety as defined in the basis of any Technical Specification. No functional changes are being rude, only repairs to the existing structure. 88-244&A&B&C
t E Mt S4 M m RY EWR 88-255 DESCRIPTICN i i This EWR provided Instructions for valve operator rodification for t 01-RC-MOV-1535 in accordance with IEB 85-03 program. The limit switch wiring rodification set up the~ operator for four limit switch rotor operation. This ' allowed the torque bypass switch to be set more conservatively without affecting Ilght Indication ur Ilmit stop. - SAFETY M MLYSIS SLM MRY This modification does not decrease the Integrity of the motor operated valve (MOV). The MOV reliability will be Increased, t k i l
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. t EWR 88-255A
[ p:. ' OESCRIPTI(W This EWR provided instructions for valve operator nodification for 01-RC-MOV-1536-in accordance with IEB 85-03 program. The 1Imlt swltch wlrIng ' modlfIcatton set up the operator for four 1im't swltch rotor operatlon. This allcwed the torque bypass saltch to be set ' nore conservatively without
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affecting the light Indication or limit stop.
- l SAFETY ANALYSIS SLM44RY This ~ nodification does not decrease the integrity of the motor operated valve (MOV). The MOV reliability will be increased.
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-EWR 88-264 and ADDENDA 1
[ ~The. folicwing valves: are required to be replaced due to leakage. A sditable ; valve replacement both.In materials and design is available for each affected j valve. The;following is a table that lists the EWR.nmt>er, the mark nmt>er of i the valve being replaced,-and the replacement valve type. _In !all cases the l valve being replaced is a Rockwell-Edwards 3/4" T-58 valve. j 3 f EWR 4 1%RR # REPLACEMENT
- I A 1-SI-321 2 l D 2-S1-260- 2 l E 2-51-309- 2 ,
O b F 2-SI-46 2 i G- 2-SI-304 2 i [ H 2-CH-417 2 ]
! 2-SI-300 2-K !
1-51-137 3 ! L 1-SI-63 3
.M 1-SI-276 3
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M P- 1-SI-13 3 - i .Q 1-CH-472 2 i i R I- S 1-CH-465 2 i 1-CH-464- 2- i T 1-SI-220- -3 'i U- 1-S1-254 2 ; V 1-CH-463 2 W. 1-CH-311
'Z 3 'i 2-S1-303 2
' AA . d 2-S1-39 2 AB ,
2-S1-213 2 AC 1 1-RC-15 3. ; AD- 2-SI-215 2 : AE 2-51-233 '2 ' AF. 2-CH-396 2 ; AG 2-S1-69 AH 2 j 2-SI-237 2 ' Al- 2-SI-252 2 AL 2-SI-120 2
.AN: 2-SI-48 2 AO- 2-S1-82 2 AP' 2-S1-129 2 AQ 1-CH-488 3 T AS 1-RC-53 3-AV 1-51-198 3 AW 1-CH-148 2 AX 1-CH-147 3 AY 1-CH-544 2
- AZ 1-51-137 3 BA. 1-CH-491 3 BC 1-RC-9 3 BG 1-CH-449 2
""(2) Conval Figure 12G2 ""(3) Yarway Figure 56178
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p EWR 88-323C DESCRIPTIGd The change of material used in the Electrical Penetration Test AsJc ibly from brass to stainless steel will provide the unification consistency between Unit 1 and 2 and simpilfy purchasing the material from available suppliers. Also the stainless steel material 'wlli reduce the maintenance work reoutred on a penetration leak in the test assembly. s-j'; d SAFETY #MLYSIS SLD9%RY This change will ensure that test assentiles on Unit 1 & 2 will be of similar metal. Originally, it was required for Conax penetrations that Unit I use stainless steel parts while Unit 2 use brass. Therefore, it is acceptable to use stainless steel parts for both units, without adversely affecting the margin of safety.
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l y o G ; A nrequirements of the' t existing valve. Since the" replacement ~. valve' meets.>the. *.i Pf 4 fM J: originali? design'f eriteria, . _ the' probability'of fallure is not increased. The: - e,
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. ope.rabili,ty.of.the affected: system will not.be~ altered..
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p , , L EW SLM %RY EW 88-266, A, B DESCRIPTION:- Service Water Putp Mechanical Seal Seal Installation (162-SW-P-1A&lB) I l This EWR installed a mechanical seal in a previously packed purro.- The putip's it shaft sleeve was machined from a 2 15/16" 0.D. to a 2 3/4" .0.D. to allow Installation of the mechanical seal. The manufacturer of the purip, Johnston Puno Co., was contacted, and agreed by telecon that the approach taken to
! modify the purp was acceptable. All four service water pulips were nodified for i a mechanical seal by this EWR.
SAFETY APMLYSIS SlM%RY: The Safety analysis approved the use of a mechanical seal. This type of seal was not addressed in the UFSAR. The mechanical seal will not adversely affect the operat ton of the service water pum. This nodIfIcatlon does not constitute j an unreviewed safety question and .does not change the license Technical - Specifications. 1 i i 88-266SASB
EWt SLM%RY
.EWR 88-267,'A , E ,
_DESCRIPT10N t The purpose of this EWR is to provide installation Instructions for the replacing of leakoff drain lines for various pumps. The purpose of this Installation is simply' to collect leakage from the ptrips and route it to the skid drains. The affected ptsrps are as follows: 1/2-FW-P-2 1/2-FW-P-3A 1/2-FW-P-38 ' I, ; f
;. SAFETY MeLYSIS SLM%RY The installation of a new drain line has no effect on the ptro design basis.
No sections of the Technical Specifications or the UFSAR are affected by this i EWR. The operation of the affected ptstos will not be deterred, and the margin of safety is not reduced by this installation. l l l l
cc h-b . ! r y my {s EWR 89-114, A, B, C, i lc DESCRIPTf(N i This EWR Involved the renoval of " Charlie Holsts" and all associated trolleys and ralls in the Emergency Switchgear Room, elevation 254'-0, the Control Rod Drive Room, elevation 280'-0, and in the normal Switchgear Room elevation t 307'-3. l-SAFETY ANALYSIS SLDNARY This modification did not change design basis docunents or any part of the Appendix "R". .The renoval of " Charlie Holsts" and all associated trolleys and supports places the cabinets and rooms in conellance with design basis seismic 1 > design. k
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! 't EWR SG9%RY l EWR 89-120 C A II i- _DESCRIPTICH This EWR provides the doctanentation and procedures necessary to Install fuel rod baffle .)etting clips on North Anna Unit I fuel assenti les designated for those core . locations identified as predisposed to baffle jetting. Advanced nuclear tucis (#4F) designed and ul11 Install the clips, g
s s SAFETY M MLys!S SWemRy The ellps are ccncatible with the fuel rcds frm a traterial s te nopr.,:t t . The clips are designed to remain in place during operation. If a clip were tc fell, Icoso parts, monitoring systom wculd cctect It. The prcbr.bIIIty er ccr : ect.t r.cc s nf er. ace k'ont ere not IrcreasecJ. l t 1 4
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, EWR SLM %Ry
, EWD 89-121, & A DESCRIPTIGd This EWR replaced 'the coils for certain radiation monitor SOVs. The replacement coils were rated for the area ambient temperature. The radiation monitors that were modified'are the following: RM-GW101 RM-RMS-159/160 RM-VG 103/104 RM-RMS-259/260
-RM-VG 105/106 RM-VG 112/113 SAFETY M MLYSIS'SLM SRy.
This EWR increased the reliability of the affected SnVs by increasing the' temperature rating- of. the coils. This modification did not change the
.. operation. of the coils.
The margin of safety as defined in the basis of any Technical Specification was not reduced. This modification Increased' the. .l
.j rellabi11ty cf the SOVs by increasing the antlent temperature rating of the' l 1, co l 1 s .' The modifIcat ton did not change operation of the affeeted SOVs.
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z ', j ' EWR 89-127A' U g_, , . .n.L. DESCRIPTI(N gy Y ' This EWR replaces the existing Bailey Positelonor with a Moore positioner. on the
.'01-FW-FCV-1478,1488 and 1498 Feed Water Floa Control Valves. -- The repl acement -
'model performs the same function as the _ existing and does not change the operation of the FCVs.' Loop operation Is unchanged.
_i-SAFETY ANAL.YSIS SLM%Ry
\ The "eplacement positioner will have no impact on equipment conditions, system operations or selsmic requirements, i
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c. y r , EWR 89-137 S-A i _DESCRIPTI(N
- This EWR provides instructions for valve operator modification for 01-SI-MOV-1860A In accordance with the station's IEB 85-03 program. The limit switch wiring modification set up the operator for four Ilmit switch rotor operation. This allcwed the torque bypass switch to be set nore conservatively
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without affecting Ilght Indication or limit stop. r L E er SAFETY ANALYSIS SLM4ARY This modlfIcatlon does not decrease the integrity of the motor operated valve (MOV7 and the MOV reliability will be increased. 4 ~ k i = b ii- ~
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L EWR StMWty p EWR'89-138,A DESCRIPTICN y This- EWR prov ided - Instruction for valve operator modification for 01-SI-MOV-1862A in accordance with the station's IEB 85-03 program. The limit switch.. wiring modification set up the operator for four Ilmit switch rotor operation.<This al. lowed the torque bypass switch to be set more conservatively without affecting the light Indication or Ilmit stop. SAFETY-ANALYSIS'SLM WtY i This , modification does not decrease the Integrity of the motor operated valve I (MOV). The MOV reliability will be Increased. I f g' 1 f i I
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- s. 'EWR 89-145 & A.S B C C ;
l o n , ig DESCRIPTION lThis EWR provide'd'for the replacement:of a' spring can and; associated hardware for. surge 1.Ine' pipe support No. 2FPH-RC-410-SH-01, f
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i x SAFETY' ANALYSIS SLMW.Y: _ Since th'Is modifIcat ton does not change- the support locatlon or system .i function, .no additional safety analysis concerns are introduced dlfferent frcyn - 6 the: original pipe' support design. . , O I g .t it 1. l-\' )
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- , DESCRIPTION
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' This EWR r,rovides direction for Installation-of tacking bolt's in the Auxillary !
r/ FeedwateriPunp (2-FW-P-38) Bearing Housing. to' alloro shaf t alignment. ;j Q F t; C s; SAFETY ANALYSIS SLM SRY
;This modification Is made to a-safety-related puTo with the concurrence'~of the Evendor. Design basis'of.the puno is unchanged as the modification does not affect.the structural / integrity or-pressure boundary.
. This modification' does~ not affect the failure analysis of this puno or;the
. failure analysls of any.other eaulpment.
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' This'.EWR provides direction for Installation of teck'ing bolts in the~ Auxillary
; Feedwater Ptsnp .(2-FW-P-2) bearing housing to allcw shaft alignment.
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~ vendor.~. ;. Design basis of the 'punp = ls unchanged, as the modification does not - -
. affect the. structural Integrity.or pressure boundary. This modification does not affeet the. failure analysis of this puno or.the failure analysis of any 1
- other? equipment, t
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.EWR 89-149 u:
DESCRIPTim - - This EWR provides directions for Installation of tacking bolts in the Auxillary
,7 Feedwater Ptanp (1-FW-P-38) Bearing Housing ' to allow shaft alignment.
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g L SAFETY ANALYSIS SLMWty - L This modification is made to a safety related ptrnp with the concurrence of the
. vendor. Design basis of the.punp is unchanged as'the modification does not i
affect the structural.Integri'v or pressure boundary. This modification does not: affect;the failure analysis of th!s ptinp or the failure analysls of any - other- equi pment . .. p , F E h i' 9' lyt ' . 'i ;..
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- DESCRIPTim h This EWR provides direction for Installation of tacking' bolts in the Auxillary Feedwater' Punp-(1-FW-P-3A) Bearing Housing to allow shaf t .alignnent.:
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c SAFETY ANALYSIS SW9%RY T g This modification is made to a safety related pump with the concurrence o'f the..
- vondor.' - Design basis of the puno is unchanged as the modificat ion does not :
; affect the structural Integrity or pressure boundary. This nodification does . '
not affect the failure analysis,of this punp or the failure analysis of any ,
. other) equipment.- '
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- This EWR provides direction for Insta11ationgof tacking bolts in the Auxl11ary
' Feedwater Punp f1-FW-P-2) Bearing Housing to allow shaft allgrment, at x
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. This. modification Is made to a safety elated ptrm with the concurrence of.the vendor . . Des ign - bas i s of the punp i s unchanged as ' t be nod i fi cat ion does not ' ~j affect tho- structural Integrity or pressuro boundary. This'modificatlon does- ~
not affect the failure analysis of this.ounp or the failure analysis of;any "other equipment,.
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!EWR 89-160 & B
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~ Fire protection spr inkl ers -. In the cable- vault and cable tunnel are n' dry,.
open-head system isolated by a manual' valve in the fan room. The fire protection. Supply header In ' the. fan rocm, which is full and pressurized, is non-seismic. If this piping failed, water could spray into the fan room and instrument rack room. To prevent this, an isolation valve-Is being added to the- 1 supply header. This valve will be located in the. Turbine Building, and will be norma 11y shut. y
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l SAFETY ANALYSIS SLD9%RY This modification does not- affect Nuclear Safety Related or Environmentally-Quallfled equipment. The valve added by this modification will not. change. the. design basis of the fire protection system. Operation of the fire protection 'j system will not be changed except that the fire-protection supply.for'the cable-- i
' vault and cable tunnel will'be Isolated in the Turbine Building Instead of the l fan' rocm, s
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/ =DESCRIPTI(N Fire . protection. ' sprinklers in the . cable vault and cable tunnel are.a dry, open-head system isolated by a manual valve in the Fan Room. . The fire protection. supply header in the Fan Room, which is full and pressurized, is non-seismic. If this piping' failed,. water could' spray into the fan room and-instrument rack room.- To prevent this', an isolation valve-is being added to the
. supply header. This' valve will be located in the Turbine Building, and will- be normally closed, t
SAFETY ANALYSIS SLM MRy C This modification does not affect Nuclear Safety Related or Environmentally Qualified equipment. The valve added by this mocification will not change the design- basis of the fire protection system. Operation of the fire protection will not be changed except for the following: (a) Fire protection supply for the cable vault and cable tunnel will be isolated in the Turbine Building instead of the Fan Room.
E r [1 EWR SLD9%RY , ,c EWt 89-166 o. DESCRIPTI(N . i Currently.Inside and outside containment five isolation. trip valves on Unit 2 containment. piping systems (SS & DG) are using the same ASCO model SOV as pilot n valves. This EWR will replace the five outside containment isolation trip
'" valves' SOVs with a different ASCO SOV model. The replacement SOV, performs the
-same function and does not affect operation of the trip valve. However, the replacement SOV does have a shorter continuous energized life and will alter.
the' trip vcives stroke time without affecting the valve's operability. This change will further enhance containment isolation reliability and reduco carmon mode fallure, since different model SOVs are being used in series..The SOVs mark No's are 2-DG-SOV-200A, 2-SS-SOV-2008, 2-SS-SOV-2018, ?-SS-SOV-2048,-and 2-SS-SOV-2128. a t I i SAFETY ANALYSIS SLM MRY. Replacement of SOV will have no impact on equipment exposure conditions. Enviromental qualification is maintained and this modification will enhance the
-re11 ability'of safety related equipment.
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[ _ DESCRIPTION *
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This EWR Incorporates- vendor reccmnedation for thy rredIfIcatlons.of check ' i.E
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' valves to' prevent rotation or separation of. the disk in the valve body.,
' Modifications include replacement of the cotter pin which locks the disc' retalning nut to the disc, welding of the nut to'the disc with an-1/8" seal '
- - j ;. weld and the,addltIon of an. anti-rotation pin.
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SAFETY ANALYSIS SLM44RY'
'While these are Safety Related containment isolation valves. This is a vendor directed nodification that will not affect the' valve design basis or the design p ,
basis of any other equipment. These minor changes will not affect the valvcs
.compIlence with ASME code, Selsmic or any other: design criteria.and will not l
' affect any failure potential or . failure analysis of this or any other
' equipmeric.
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. . EWR SLMMRY I t '.
EWR 89-169,.A,'.B;
- DESCRIPTION
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This EWR reccrmends incormating vent. reconmendation for seal welding and . pinning of the check valve dir.R to the tail link. This wl11. prevent' rotation oru.-separation of the disk in the valve body.- The affected valves are'02-QS-11 and.22.- 1 2 3 l l l l l SAFETY ANALYSIS SLM MRY- ' 1 While these.'are safety related containment icolation valves, this.'Is a. vendor directed modification that will not affect the, valve design basis or;the' design basis of any other equipment. These minor changes will not affect ASME . code, j.
-seismic or any ~other design criteria, and wil1~ not affect any failure. potential- ;-
.or: failure analysis of this or any other equipment'. -
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-EWR.89-171.C'B'.
n DESCRIPTION
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This EWR' reconmends incorporating vendor reconmendation for seal welding and
' pinning of check valve disk to tall link for valves 2-RS-20 and 2-RS-30. This will prevent. rotation,' loosening or separation of disk'in valve body.
q SAFETY ANALYSIS SG9Wty '! While these are safety related containment 1 solation valves, this is a vendor
-directed modification that_will not affect the valve design basis or the design Lbasis of any.other equipment. These minor changes will-not affect ASME code, -!
Selsmic or any other design criteria'and will not affect'any. failure potential or failure analysis of this or any other equipment. .< l 7 a. l r h ( (
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.EWR 89-1728- .
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h. DESCRIPTf(N V The new gland housings are the recomended replacement by the turbine manufacturer. The only difference is that the _new housing has two housing bolts -whereas the original has four housing bolts. The housing is kept In place by the turbine casing which is strong enough to hradle any loads. A New shaft had'an additional hole drilled in it which was not on the original shaft. The vendor stated that it was for the new design of overspeed trip'but' ' was acceptable for use due to low torsional loads at that end of the shaft. i SAFETY ANALYSIS SLM MRY lt was verified'that the turbine-driven Aux 111ary Feedwater Ptrnp meets the ' requirements stated in~the UFSAR. The pw1p will start and supply design flow at design pressure, as assumed. Justification for using the non-safety related parts 'was based. on the parts being constructed in accordance with the ,! manufacturers drawings. This included verification of dimensions by- Virginia
. Power -and . the manufacturers representative. In additlon, the_ parts
-manufacturer has supplied a certificate of compilance stating that' the ' parts <
havo -.been ! manufactured in accordance with their QA program,.and meet all
~ dimensional requirements.
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-EWR 89-175 >
. DESCRIPTION I
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'This ~ modification: removed the center phase overload heater frcm the redundant. )
-starters.(K12A and K12)r and ' Installed' a': Jtmper to.- eliminate overheating' nuisance trips for 1-RM-P-159A & B. i LV l
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q The' modification ~will not alter the abil ity of. the .radiat ion monitot* to ' "
- -automatically stop the discharge of gases frome the gaseoust waste d i e.posal '
j . system. Removal of the: center phase overload heater does not adverst.'.y. affect notor thermal; protection. - The. starters are1 adequately- protected by1 overload 1 heaters'in the remaining <twoiphases. e i' , \ . ' .'h'N
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- 2-SI-MOV-2836:cannot be MOVATS tested due to piping obstructlons. These obstructions also hinder removal of the actuator for grease changeout and valvee ;
maintenance..Therefore, the one' service water line that is causing the obstructton shall:be rerouted.
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SAFETY ANALYSIS SLM MRY 5
- Sleeved penetrations-w!11 function as described in UFSAR. Integrity of the
-service water system will'not be adversely affected and ability of .
SI-MOV-2836 to perform its function will not be affected by modification to. 1 cap.4 Now piping and fittings are the same as existing and seismic integrity is
~ maintained, ,
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s it< v h EWR SLMMRy > e r i EWR 89-180 I DESCRIPTION
.NuTerous areas in- the secondary side piping system have been identified as potentially having wall thinning due to' the erosion / corrosion phenomena .in- '
single phase flow conditions or steam erosion in twa phase flow conditions. , Inspection will be perfonsed on these areas during the Unit i refueling outage ' to determine if replacement of piping components is necessary. This EWR 89-180 will serve as the controlling docunent for replacement of piping components. identified';as requiring replacement due to erosion / corrosion or steam erosion. In addition,' this EWR wl11. serve as the controlling document for replacement of the first and second point extraction steam traps and associated drain line valves and piping components. Components identifled as requiring replacement will be replaced - per the controlling procedures contained In' this EWR.. Installations will be performed on these areas during the 1989 Unit I refueling outage. SAFETY ANALYSIS SLM MRY This EWR only' addresses non-safety related portions of the secondary side piping systems. Secondary side piping systems affected may-include :
. condensate,' high pressure vents and drains, low pressure drains, MSR drains, .
flash evaporator, auxillary steam,. gland steam and extraction steam systems. In-addition, this EWR addresses the first and second point extraction steam trap drain ' systems. This rrodi ficat ion does not~ affect EQ equipment, nuclear safety-related equipment or impact the UFSAR. < 4 (
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Ti s l EWt SLD9%RY lc EWR 89-181 S A DESCRIPTim ir- Upon- renoval' of the RV surveillance capsule "U", reinsertion of the capsule T plug In the core barrel flange was not possible. . The Westinghouse resolution to this problem is to machine as much as .01" from the diameter of the plug.to eliminate Interference. = Actual machining of the plug turned out to' be .02" which was still found to be acceptable. 1
... y SAFETY ## LYSIS SLD9%RY The -purpose of the plug is to control flow between lower and upper Internals.
Decreasing the_ Interfering diameter as instructed by Westinghouse will not 'd adversely affect the plug functlon or operabi1Ity. i u
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DESCRIPTICN; d k -Thi s- EWR provides justification for approval'of the continued use of a-ratchet j operat6r on 2-RH-37.. Continued use of the ratchet will also require use'of ' .a nullfled spacer and yoke nut.- i 9
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. The only safety ' functlon that this valve perfonns Is pressure retentlon. The Luse of a ratchet in place of the handwheel and adding a spacer will not: affect-the; pressure retalning abilities of .this valve. There will be no effect on .!
'/ah c.cerlen basis or c.ny other denign basis. , This nodification will' not ' reduce
- rrtety. r'erpin . ir roy ~way , .
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t 4 - EWR SLM MRy ! EWR 89-191 DESCRIPTION i This EWR provides the authorization, direction and materials list for the installation of heavier _ spring packs into.the SW spray array bypass MOVs. This is required so that the MOVs can operate as~ desired against the calculated l maximum differential pressures. Limitorque has stated that the. new spring packs 1 will enable the operators to deliver the 1500 Ft-Lbs. of torque required to the l butterfly valves, p. l l
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l l 1' SAFETY ANALYSIS SLM eRy This modification is a common upgrade to Limitorque operators which will provide the desired performance. .This improved performance will alleviate concern about _the ability of the valve to operate against -worst case- system differential pressure. As such, the elimination of this concern does not represent an unreviewed safety question. l ll l p-l' 3
[g,=, EWR SGNARY , L b EWR 89-193; DESCRIPTION
. j-1 This - EWR provides the Instructlon for- a Jtmper t_o be instelled on valve. ;
02-SI-MOV-28608. The purpose.of this is to-Jtsver out the torque switch.In the open direction since this particular valve cannot be MOVATS tested This is i considered sound practice and will not damage the ; valve since the station
" Limits Open" this MOV.
s i-i SAFETY ANALYSIS SWNARY < This modification does not increase the probability of occurrence nor the consequences of an accident previously' evaluated. The margin 'of safety as. l defined 'In .the basis for any Technical Specification is not reduced by this nodification. I? L i e 1 l '.
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- _This' dWR provides an'eveluation of exceeding the setoolnt doctment thrust band
-hlgh limit' for 2-CH-MOV-2373 durIng MOVATS testing of the-valve in:the opon - q direction. This. situation was found to be acceptable, .
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{- b P EWR SLD9%RY EWR-89-196 p _DESCRIPTICN This EWR nodified the test circuitry of alternate feeder breaker 15B11. the modification involved rewiring of interlocking devices between the 4160V emergency bus "1H" and its alternate feeder 4160V normal bus "1B"-(Breaker 15811), Incorporating the alternate feeder into various pts allows testing to reflect t'e' actual scheme of the "1H" supply breakers, it also allows the emergency bus "1H" to be fed from the alternate feeder while performing various pts without tripping of the alternate feeder breaker. SAFETY ANALYSIS SLM %RY
'This modification does not represent an unreviewed safety question'because system operability is unaffected. This rrodification is for testing purposes only. It will enhance the performance of various pts No future testing is required. Testing per this EWR, satisfies any testing that would be required by the appilcable PT.
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- EWR SW944Ry EWR 89-197 DESCRIPTIGN This~ FWR 'nodifles the test circuitry of alternate feeder breaker 25810. The nodification Involves rewiring of interlocking devices between the 4160V
, snergency bus "1J" and its . alternate feeder 4160V nornal bus 2B (Breaker 25-B10). Incorporating the alternate feeder into various pts allows testing to , reflect the actual tripping schone of the "1J" supply' breakers, it also allows [ the emergency bus "Id" to be fed fran the alternate feeder while preforming various pts without tripping of the alternate feeder breaker. . SAFETY ANALYSIS SW944Ry This nodification- does not represent an unreviewed safety question. Systen operability is unaffected. This nodification is for testing purposes only. It will enhance the performance of various pts. No further testing is required,
. Testing per this EWR satisfies any testing that would be required, per the appilcable PT.
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3 s r I. EWR StM %RY EWR 89-204, A, B i DESCRIPTION: This wiring modification allows and implements limit close control of 01-CH-MOV-1115C and-02-CH-MOV-2115C. This reduces the potential hazard to personnel form the handwheel. and lessens the potential for operator damage. New thrust values will be provided for the NASD. 'The torque switch will be adjusted to--deliver no less than the thrust to overcome system differential pressure, thus avoiding inadvertent torque switch trip.. While .these performance guarantees _ are provided, a Ilmit switch adjustment that minimizes total. seating thrust is implemented. This modification does not represent an unreviewed_ safety question because t component / system relIabi1Ity Is enhanced. The 1Imlt close method of valve ' i control is~ a conmon practice and does not present any failure modes or effects not previously analyzed.
. SAFETY ANALYSIS SLM %RY:
The MOVs are spring compensated which stores kinetic energy when the valve - disc reaches its end of travel. With torque close, a larger quantity of energy is stored due to motor contractor drop out time and nomentum. Limit stop reduces this stored energy without compromising full closure. When the stored energy plus contributed motor torque are released, the hypold gear is stressed,_ in addition to the handwheel " racing" which represents a personnel hazard. 89-204ASB
m i. EWR SLM SRy 4 t- EWR 89-213 ( l' i _DESCRIPTICH This EWR replaces the existir;g 2-CH-SOV-2201 ASCO Model ib. HT831654E with ASCO
-Model No.-NPKX831654E. The existing model number is no longer a stocking item at North- Anna Power Station. The replacement rrodel is an identical replacement in all mech nical aspects in addition to being quallfled for class IE service.
Electrical Lnd tubing connections w!11 remain the same. Electrically the replacement model has a 17.4 watt coil versus a 9.7' watt coil for the existing model, resulting in^a 7.7 watt increase in power consunption.- This increase is of no significance to the plant's.. electrical load. SAFETY ANALYSIS StM gRy
'The replacement SOV will have no impact on equipment', exposure conditions, system operations or seismic requirements.
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n- , n - o f,< . EWR SW9%RY: " r .4 1;; , 3 EWR^89-214,?A,.B, C,.D, E
'DESCRIPTIONI
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; This EWR- provides - the docunentatlon and procedures necessary to inspect, measure and replace. rods-on those assenb1les destgnated as_ part of the advanced
(- cladding denenstration program.- Westinghouse performed the work. F a a i 11 SAFETY ANALYSIS SW9%RY This program. does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important- to safety , previously evaluated in the -UFSAR. Fuel rod design criterla' is unchanged. Assembly design criteria is met even with seven inactive thimble tube Inserts. LOCA consequences are unchanged. Drop of rod storage basket is bounded by fuel assembly drop. This program does not create the possibility of an accident or malfunction.:
. Demo assent 11es satisfy current fuel criteria, even with seven inactive thimble tube Inserts. Handling of the fuel storage basket is bounded by the fuel assenbly handling accident. Calibration of the nozzle removal tool will Insure that Demo assenbiles will not be Ilfted during nozzio removal.
This program will not reduce the margin if safety as defined in the basis for any Technical Specification since the existing safety analyses remain unchanged and no new accidents are created. l l
m 1 EWR SLM%RY !~ EWR 89-215, A, B, Cl n DESCRIPTION: Valves 2-CC-276, 1-CC-559, 2-CC-552, and 1-CC-572 are 6" mission check valves that failed thelr. Type "C" test. Attempts to lap the valve to obtain a good -seating surface failed therefore it needs to be replaced. Mission was bought out by CSS. The CSS valve is equivalent or better than the mission valve. In terms of ' strength, ccrnpat ibility and design. It is an acceptable replacement. SAFETY ANALYSIS Sl99%RY:
'The replacement of the check valve with a different manufacturer wlll not compromise the function of the valve. Function of both valves Is the same.
Specifications for the new valve meet or exceed those for the original valve. A- calculation that was perfonned shcws that the _ seismic integrity is j maintained. Valve replacement will not cause malfunction of safety related i equipment, i l l i
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n ,, n EWR SlM%RY t m EWR 89-217,A-
.i t-DESCRIPTI@:'
Snubbers'2-SHP-HSS-217A, 2178, 218A, 2188, 219A, and 2198.In the Main Steam Valve House were requested to be changed out =from eight (8)- Inch Tcmpkins-Johnson : to eight (8) Inch Miller Snubbers by 151.. Since the snubbers were' safety related, an engineering evaluation, justification, and necessary- > guidelines were done.. With modifications to the rear snubber bracket of the Thompkin-Johnson to fit the new Miller, and an extension piece to the Miller, the nodifIcation'.was justIfled and was acceptable as a replacement. This EWR involves the change out.of snubber 2-SHP-HSS-217A and 2-SHP-HSS-218A. SAFETY ANALYSIS SLM MRY: The replacement of the eight (8) Inch Thompkins-Johnson Snubbers with eight (8) Inch Millers is acceptable and does not leave any unreviewed safety questions. The new MIIlers will provide a better-life for the snubbers, and thus help eliminate future rebuilds. 89-217A
i. EWR SlM %RY e [- !' cWR 89-219 DESCRIPT10N:, The ' downward thermal movement of the relief valve CH-RV-2203 ~Is restrained by'the west corner of the 1.5" thick plate of the lower support of
~
- the Recirc Spray Cooler 2-RS-E-1A.
To allow the~ free thermal trovement of the relief valve a notch is cut on the west corner of the 1.5" thick plate by 4.5",X 0.75". Since this notch- is m O cut on the excess portion of the.l.5" thick plate (cantilever part),Jit does. not affect the structural Integrity of the lower support of .the Rectre -Spray Cooler 2-RS-E-1A. SAFETY ANALYSIS SLM MRY: The -structural integrity of the support is maintained and the letdown line can thenna11y expand as designed. 89-219
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EWR 89-223, A, B,' C DESCRIPTION: Certain. original connection details for platforns inside containnent nay , not have been Installed as shown on the original design documents. To ' resolve ' this discrepancy,'certain critical joints (as specified by calculation CE-0657) will be' inspected and repairs to these Joints will be nede as required. SAFETY ANALYSIS
SUMMARY
The -inplementation of this' EWR, since it simply restores existing, i structures to. their -original design capacities, does not increase the-probability of occurrence or the-consequence of an accident; nor does it create the possibility for an accident or malfunction of a different' type than previously evaluated. The lmplenentation of this EWR does not reduce the margin of safety as defined in the basis of any Technical Specification, e l l l' 89-223,ASBSC
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($'M EWR S N Y' EWR 89-232' L
-DESCRIPTION:
a The! welds 'which ~ secure the ' stuffing box bushing to the stufflNg box'on ptros 2-FW-P-2, 3A, . 3B are cracked. The vendor reconinends Installing four. (4)
.1/4"-20 set screws.In the ctuffing box to prevent-rotation.
SAFETY ANALYSIS'SLM %RY: Auxillary FW Ptmps' wi l l continue to meet the requirements stated in the UFSAR.- New parts are the same size, material, and weight as. original' parts. Seismic integrity. Is maintained. Head curve tests and periodic tests will
. verify operability of purp prior to returning it to service.
4 l l 1 i 1 L' i 89-232 b !
. . ..y
, EWR S M Y
-.EWR 89-232A.
i' DESCRIPTION: During ~ disassenbly of 01-FW-P-2, it was noticed that the overspeed trip
' tappet was chipped. This problem had been brought up in' NRC IEIN Bulletin No.
88-67,. .Therefore, the tappet shall be replaced using the new design tappet-which shall improve reliability and durability. SAFETY ANALYSIS SLD9%RY: The upgrade of non safety related parts using the EWR process is considered appropriate, no design b& sis docunent, or calculations are changed. The Terry Turbine will perform as designed and its margin of safety has not been reduced. l
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L .I EWR 89-2328. DESCRIPTION: The . parts that are to be used frcm the CP&L punp are Identical to those in North Anna's punps in terms of form, fit, and function. The parts are manufactured-to the same drawings and have the same number.'The only difference-Is that the new parts are constructed of a slightly different material (ASTM A48 CL 308 -vs- ASTM A48 C L- 258). The difference between the two materials is that the CL 30B material is stronger (30,000 PSI tenslie strength-
-vs- 25,000 PSI). Therefore, it is acceptable to use.these parts, g SAFETY ANALYSIS SLM MRY:
Operation of the AFW punps as described in the UFSAR will not be altered. - Post modification testing will ensure operability.
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.EWR SLM4ARY L
EWR 89-232C-DESCRIPTION: n The' welds which secure the stuffing box busting to the stuffing box are I( cracked on 01-FW-P-2 and 01-FW-P-38. .The vendor recorrnends Installing four (4) 1/4"-20~ set screws In the stuffing box to prevent rotation. This is acceptable g since the set screws will serve the same purpose as the welds. The new parts (shaf t, balance drun, balancing sleeve) are as strong or stronger than the original parts.- Also, the new parts have' better erosion / corrosion characteristics. They are exactly the sane as the original parts in terns of fonn,. fit, and function. L i SAFETY ANALYSIS
SUMMARY
The -Auxillary feed water pumps will continue to neet the requirenents stated in the UFSAR. New parts are acceptable for this application and specified' changes- are acceptable to the pump nanufacturer. New parts are the-same as the originals In tenns of form, fit, and function. New naterials are as strong or stronger than the original naterials. Head curve tests and '
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i i i i 89-232C l' N[ i
? d EWR StM MRY L EWR 89-235A DESCRIPTION: The North Anna Secondary. Piping Erosion / Corrosion Inspection Program includes the computer stress analysis of piping systems. A potential code thennal . overstress was Identified in a segment of the steam generator blowdom piping 'Inside containnent. The thernal overstress occurred in- Line 1"-WGCB-412-601-Q3 'at the connection points of 2"-WGCB-411-601-Q3 and to Steam Generator 2-RC-E-1C. The overstress condition is a result of: (A) differential therral d isplacenents of the' stean generator at Nodes 245 and 285,.and (B) temperature differences in the two lines during systen operation. A computer analysis has been perfonned for the other two. piping arrangonents for Steam Generators 2-RC-E-1A and 2-RC-E-1B and concluded that this piping has acceptable stress level. SAFETY ANALYSIS StM %RY:
.lf the generator drain valves are opened, the steam generator drain Line 1"-WGCB-412-601-Q3 Will heat up to the sane temperature as the blowdown line.
This wl11: reduce the thernal stress in the pipe to an acceptable level. To ; ensure that- the steam generator drain pipes on 2-RC-E-1A and 2-RC-E-1B and ! Unit 1 steam generators are not thernal ly overstressed and to facilitate consistency of. valve line up, all isolation valves in the 1" drain lines-will , be opened. Blowdown enters two sparger pipes located on the top of the stesn ' generator. tube sheet. The sparger pipes are connected to two 2" lines on the opposite sides of- the stesn generator. These pipes are headered together to I fonn one blowdown pipe. An 1" drain line also originated from the top of the tube sheet'and connects to one of the blowdown lines. .The total length of the , drain -line is approximately nine (9) feet. The drain lines originate on the l top surface of the tube sheet as does the blowdown pipe. The conclusions of an evaluation of the effect of using the drain pipe as a blaadown is as follows: The sparger' pipe takes water from the steam generator across the diameter of the tube sheet. The drain is located near the e .l perimeter of the steam generator tube sheet and would drain water ; fran the dcwncomer. A conparison of blowdown -vs- drain pipe areas indicates that approximately 12% of the total flow would be drawn off by the stesn generator draln. This is consistent with i inspection of pressure drops in a 1" pipe and 2" pipe as given in : tables fran Caneron Hydraulic Data. The change in flow pattern ' resulting fran opening the 1" drain valve would be insignificant ; and therefore have no effect on operation of the steam generator i or the blowdown system. ' This EWR Involves valve realignment of the steam generator blowdown system. The system function is not changed and all original design criteria are met. Since the systen design is unchanged, the probability of failure is not increased. 89-235A 1
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i t r EWR SLM MRy ,, EWR 89-237 i r DESCRIPTION
.This EWR Initiates a Q List change which will list the motor operators for MOVs 2585, 2586, and 2587 separately and will identify them as QA. Category. tJSQ.
This is justified because the motor operators have no safety function. 'This-reclassification will eliminate the necessity of replacing the white torque
- switches in the motor operator.
SAFETY ANALYSIS SLM SRY The safety analysis concludes that the loop bypass valves provide an RCS boundary. This Is theIr sole reason for safety reIated desIgnatlon. The reclassification of .the motor operators to NSQ will not affect the valve
~ boundary, thus the integrity of the RCS boundary is not. compromised.
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EWR 89-237AL-b. y DESCRIPTION
- r-This' EWR initiates a Q-l.lst change which will list the motor operators for MOVs
'1585, 1586, and 1587 separately and will identify them as QA Category NSQ. This.
Is justifled because the motor operators have no safety function. No - clodi fication is involved. F . i.- l s .. SAFETY ANALYSIS SL7 NARY The safety analysis (89-SE-MOD-071) concludes that the loop bypass valves ' provide an RCS boundary. This is their. sole reason for safety related designation. The reclassification of the motor operators to NSQ will not affect the valve boundary. Thus the Integrity of the RCS boundary. Is not' compromised, r M l l 1 i l l l I 1
v n , p p: ; f' o a 1 t h' ' EWR 89-239, A 'd -- DESCRIPTIG4
.Certain original connection details for platforms inside containment may not
" have been. installed as shown on the original design documents.- To resolve this i discrepancy, certain critical joints (as speel? led by calculation CE-0662) wlll-be inspected and repairs t6 these Joints will be made as required. I" P SAFETY ANALYSIS SLMSRY 1 r The implementation of this EWR, since it simply restores existing structures to their original-design capacities, does not not increase the probability of occurrence or the consequences of an accident; nor does it create the
-possibility of an accident or malfunction of a different type than previously evaluated.
- The Implementation of this EWR does not reduce the margin of safety as defined in-the basis of any Technical Specif3 cation, t
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[ EWR 89-243 r _DESCRIPTIGd ' The UFSAR' states: "Dupilcate water level channels are provided for each acconulator. Both channels provide Indication to the control roan and actuate - high and low water level alarms". The failed level transmitter was unable to be repaired and an exact replacement was not available. The lead tine to procure an exact.replacenent was n'ot acceptable for plant operation. From past plant operating experience a suitable replacanent was determined based on operability, reliability, and availability. A new transmitter was drawn fran stock and installed with concurrence fran Design Engineering. l SAFETY ANALYSIS StM MRY ! Operation of the accunulator has not been adversely affected. The integrity of I the accumulator has not been affected. The new level transmitter has been I evaluated as an acceptable replacement. The dupilcate water level channels as described in the UFSAR ARE AVAILABLE. The modification does not prevent the i accumulator from discharging borated water into the RCS as described in the UFSAR. l l
4 EWt SIM%RY 89-247, B, ' C ' _DESCRIPTIG4 UNITS 1 & 2 Station. Instrumentation requires a safety eyebolt to be Installed on the steel' tube that supports the mineral Insulated thermocouple cable located adjacent to the control rod drive mechanisms above the reactor head shleid._ The purpose of the eyebolt'Is to clip workers safety belts to while performing connect / disconnect of-the transition cable connectors. Also, a portable work i platform that is supported by the reactor head shield was designed. .This ' platform shall be stored outside of contairrnent environment during' plant operation. SAFETY ANALYSIS SLM MRY 1here are no unreviewed safety questions. The platform is a temporary access platform while perfonning connect / disconnect of the transition cable / head lift rig cable connectors. The platform will not be stored in containment when not in use, therefore, there is no concern with aluninun in containment during LOCA. -The' eye bolts do not affect the structure integrity of the head lift cable support tube steel. The eye bolt will be used to tie off/ hook on a safety harness while utilizing the platform. e-
i r& ., . . !. i + h e-EWt St#9WtY - E , EWR 89-248,;B;
. DESCRIPTI(N This EWR replaced instrment tubing and fittings on- feedwater regulating valves with stainless steel tubing and fittings. ' The root cause analysis for a.recent trip identified the existing-fittings as being too heavy. . This, contined with
.'vlbration,. led to fatigue failure of a fitting.
i i SAFETY ANALYSIS SLM4ARY The margin.of safety is maintained as the ability of the valves to perform their design function is unchanged. Should the new tubing. configuration fai1, the consequences of failure-remain the same. The extreme accidents which could
= result such as full closure or full opening of- a feedwater regulating valve,-
-have previously been analyzed in the UFSAR. '
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- EWR 89-255 i,
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_DESCRIPTI(N This EWR incorporated a change to the North Anna Setpoint Docunent to show new mlnittm thrust values and " testing only" open thrust maxIntms for the S1 Accunulator Discharge Isolation MOVs. The new minimum values are taken from revised calculations provided by MOVATS Incorporated. The " testing only" open thrust ~ max imuns are based on evaluation The evaluation is based on the.. operator rating (less measurement accuracy) and that the valve stem degradation would- be detectable.- Based on these considerations the Signature Analysis Results for 2-SI-MOV-2865C are evaluated as being acceptable. SAFETY ANALYSIS SW944RY This EWR did not incorporata a physical modification or alter the operation of the SI/accunulator discharge isolation MOVs. The new NASD -values being incorporated did not lesson the margin of safety nor do they represent an unreviewed safety questlon. a 1 1 t 1
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; EWR 89-259 , I p > ;
Ep. . DESCRIPTIG4 A valve on-a drain line" for' the RCS RTD bypass line is being renoved and a ; pipe cap will be Installed. The pipe cap meets tho~ System design temperature ~ =
. and pressure. The nodification will have no effect on system operation.--
i li f r SAFETY ANALYSIS-Sl8 NARY i r The replacement of the isolation. valve with a pipe cap does not-Increase probability of an accident or increase the consecuences of an accident. .i 5 9 l' . l. l ^Y I 1' l f 4
W :. .. . (_ , s. EWL SW9%RY ijg.f- , b i; EWR:'89-260,'A- '"
, , DESCRIPTI(84 I -This.EWR installs notor Jacking bolts for- auxiliary feedwater purps by welding threaded lugs to the notor support bed' plate.
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Y i i These Jacking bolts perfo'mr no operational function. They do not prevent the '
' operation of the AFW puTps or any other equipment. They-are to be used during maintenance. These are connected to: safety related equirrent through Uciding -
to puno support plate.~ Lug weight is so small as to not require seismic
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' Instal'1 motor Jacking Bolts for Unit 2' Auxillary Feed water Pumps ;
(02-FW-P-3A,B) by, welding threaded lugs .to the notor support plate. ; E i. 6 i i- i y- j
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SAFETY ANALYSIS SLMWtY l im These' lugs perform no operational function. They.do not prevent the operation -i
. of the Auxillary Feedwater ptinos or any ~other equipment. They are only used-during maintenance. The'only way that these are connected to safety related equipment -is .through welding 't.o puno support plate. Weight is so small as to
-not.rdquire seismic reanalysis, s
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y , -. - L r EWR SLM %RY , EWR 89-268A, C B \{ , i; DESCRIPT!CN p. -p This EWR revises the NASD mechanical stop and limit switch settings for the H recirculating spray heat exchangers Inlet MOVs. The new data was estabilshed-during performance of heat exchanger flow balancing test 2-PT-75.6. The PT instructs that the setpoints shall be changed to match the PT result. This EWR satisfled that requirement for a setpoint change. SAFETY ANALYSIS SLM SRY The PT (2-PT-75.6) balances recirculating spray heat exchangers flow to conform
-to the design basis. Having done that the NASD, by procedure, is changed. 'This method ~ assures system performance Is as designed. The probabi1Ity of equipment malfunction is unaffected, and valve operation is still as described in the UFSAR. As such, there are no unreviewed safety questions.
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[ ' y_ EWR SLM MRy: EWR89-268,B,'d,D,E _DESCRIPTICN-This -EWR revises the NASD mechanical stop and limit switch settings for the
' Recirculating Spray Heat Exchangers inlet MOVs. 'The new data was established ,
1 during ' performance of- heat exchangers flov balancing testing (1-PT-75.6). The-PT Instructs that the setpoint shall be changed to match the PT results, and-this:EWR conforms to that requirement. i SAFETY ANALYSIS SLM MRY-The PT (1-PT-75.6) Recircullating Spray Heat Exchangers flow conformed to the design basis. . The NASD, by procedure, is changed. This method was previously evaluated and approved by SNSOC. As such, there are no unreviewed safety questions. s S l l l l ' l l'
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EWR 89-274, A. DESCRIPTim - h The existing vent on the Auxillary Feedwater pump is not adequate to renove all
- of.the entrapped air in the-casing. Therefore, an additional vent shall be Installed at the ptinp casing high point. This will ensure proper venting of the
- ptanp and' increase reliabil Ity.
4 Y SAFETY ANALYSIS SLM44RY i-' The addition of a pump casing vent valve does not increase the potential for' pump fallure nor does It affeet the fallure analysis of any other equipment, v ,k Ia 1
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b , D8t SLM%RY , EWR 89-278A DESCRIPTIGd y Replacement of 1-CH-189, a Rockwell Edwards check va;ve (fig. 36174F316T3), is necessary due to leakage. A suitable replacement in design and materials is the Rockwell Echvards check valve (Fig. 36174F16T4). _ SAFETY # MLYSIS StM SRY The replacement valve has been provided to the same purchase specification as the original valve. A safety evaluation is not required per NRC inspection and Enforcement Manual part 9800 "CFR discussion.
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EMt 89-284A, C t (ESCRIPTIGd Work required by this EWR Includes fabrication of portable, tenporary steel d1kes to be used for 1-PT-64.8, ficav test of the Inside recirculation spray punps. There are minor codifications to existing dlke support steel such as plug welding of holes and ' redrilling the same nunber of holes in a different location. Also, a seam less than two inches needs to be welded to prevent leakage. I I SAFETY ANALYSIS SLM ERY The dlke panels will be renoved during plant operation and stored outside of i contairment . Modifications to existing support steel will not Jeopardize or compromise safety to the environment or pubile, nor put the plant in a less safe situation. 4 s
OdR SW9Wty EWR 89-287 DESCRIPTIG4 This EWR Incorporates the Safety injection signal back into the TSC Ventilation " System for the automatic start of Emergency Supply Fan 1-HV-F-150. Design criteria for the TSC Ventilation System, found in DC81-17, states that the Emergency Supply Fan will' automatically start through the Initiation of an Si signal. The $1 signal (cable 1HVZiNC341) is installed inside the TSC Ventilation Panel 1-EP-CB-330, but is not terminated. This EWR reconnects [ cable 1HVZfNC341 back into the Emergency Supply Fan control circuit, n h SAFETY ANN.YSIS SW9Wty This trodification does not represent an unreviewed safety question. The TSC Ventilation System, including the Safety injection signal was installed to the original design criteria found in DC81-17. The permanent removal of the 51 signal from the TSC Ventilation Panel was never perforTned. This EWR reinstates the requirement for the Si signal and provide the procedure for terminating the cable at the TSC Ventilation Panel and testing the signal. t
EWR SIM%RY 89-288 DESCRIPTim This EWR replaced the existing 02-RC-SOV-2456-1 Asco it> del tb. LB831654 with Asco Model nttrber NP831654E. The existing model nttrber is no longer a stocking item at North Anna Pcwer Station. The replacement rodel is an Identical replacement in all aspects in addition to being quallfled for class IE service. Electrical and tubing connections will remain the same. Electrically, the replacement model has a 17.4 watt coil verses a 5,7 watt coil for the existing model. This increase in power consurotion is ; no significance to the plants electrical load or circuit protection devices, r SAFETY ANALYSIS StM ERY Replacement 50V will have no impact on equipment exposure conditions, system operations or selsmic requirements.
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! EWR 89-289, A OESCRIPTIGd ;
During reasserbly of ptrip the proper alIgrment between pmp and notor could not > be obtained, in order to ob:.aln proper alignment the holes in the base of notor have to be clongated. This will Inprove realignment between the ptrp and notor. Also an evaluation has been perforTned and it is concluded that the selsnic Integrity of the notor has been retained. L i SAFETY ANN.YSIS SL89WtY Operation of the Auxiliary Feedwater System is not adversity affected alignment will enhance operation of system as described in UFSAR by increasing ptmo reliability. Modification does not affect conditions to which equipment will be exposed. Selsmic concerns have been addressed and codification determined acceptance. Allgnment of notor increases ptmp reallability and does not s increase the probability or consequences of an accident or ralfunction,
EWR SLM MRY g EWR 89-290 i DESCRIPTICH This EWR replaces the existing 02-RC-SOV-2455C-1 and 2 ASCO model ib. HT831654 with ASCO model ib. NP831654E. The existing trodel number is no longer a
- stocking item at North Anna power Station. The replacement tredel is an Identical replacement in all mechanical aspects in addition to being quallfled for class lE service. Electrical and Tubing connections will remain the same.
- Electrically the replacement model has a 17.4 watt coil verses a 9.7 watt coil for the existing model. The Increase in watts represents a .C6 amperage l~ Increase for the existing circuit protection devices for each SOV. This increase is of no significance to existing circuit protection devices.
SAFETY # MLYSIS SLM eRY Replacement SOV will have no impact to equipment or personnel exposure conditions, will not alter system operations or change seismic requirements.
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EWt 89-292 E DESCRIPTION: ' This EWR will revise torque switch setpoint based on calculations. As left, thrust values for RS-256A will not result in operation / valve degradation. SAFETY ANALYSIS SLM %RY: r The setpoints as revised are consistent with the NAPS safety-related value program. The values will continue to perform the same function. 89-292
n., .. I dst SLMWtY EWR 89-295 i DESCRIPTICH This EWR adjusts setpoints of nitrogen high pressure control valves (02-GN-PCV-225A-1, A-2, B-1, and B-2). The setpoints will be increased to p improve the stroke time of the pressurizer PORVs. The present setpoint of l i these valves is 180 psig which will be Increased 10 pounds to 190 psig. i e i SAFETY # MLYSIS SlM %RY The PORVs will continue to activate as designed. The new regulator setpoint will not increase the probability of an accident occurring. The margin of safety as defined in the basis section of any Technical Specification will not be reduced. m.
pv t-EtSt SW9WtY EWR 89-303 & A DESCRIPTI@i This EWR seismically. restrains a tool box in Unit 1 contairvrent that will be
- used for subattrospherle entries. The box will store tools used by the electrical and instnrentation shops.
i' SAFETY ANALYSIS SLM WtY The box will be seismically restrained such that the box cannot damage any safety related components. 9 4 i
EMt SLM%RY EWR 89-312 ?s DESCRIPTIGd IRC-187 is a 3/4" valve off of the loop stop valve. The valve was forTrerly used for disc pressurization but is no longer required for that purpose. The valve is being rerreved and the line capped because the valve has been injected due to a bonnet leak. 3 SAFETY ANALYSIS SlM MRY The new pipe cap will serve the same function (pressure boundary) as'1-RC-187 did at power. The pipe cap will maintain RCS Integrity and meets the design requirements of the RSC system. The failure of the pipe cap ls bounded by the double ended rupture of RCS piping in the UFSAR.
Det SW9%RY L EWR 89-316, C, D DESCRIPTION This EWR is'being used to provide the Installation Instructions tu replace valves 1-FW-237,243 and 249. These valves are feedwater recirculation line manual isolation valves. The existing valve is a Crane gate valve and the replacement will be a Powell gate valve of similar makeup. There is a ~ significant increase in weight (210 lbs.) but no additional supports will be required, as doctinented by Calculation CE-274 i SAFETY ANALYSIS SG9%RY The existing valves and their replacements are not safety related and will not change the affects on Safety Related hardware. Neither the Technical Specifications or UFSAR are effected by this nodification. The system operability will remain the same. The margin of safety as defined in the basis for any Technical Specification is not reduced. i r.
II t. EW SLDM4tY EWR 89-321 DESCRIPTICH tiodification will allow use of 1 & B 60000 series lugs on copper conductors on 1-RH-P-1A, IB. ? h-
. SAFETY ANALYSIS SLM %RY Use of 60000 series lugs on the RHR pwp motor leads will not alter operability of the ptsrp..
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w ' Oft SLM%RY . 1 EWR 69-323 DESCRIPTICN l Because of deterioration of the carbon steel J-nozzles on the steam generator feed water distribution ring by erosion / corrosion it was determined that . replacement of the J-nozzel was necessary. Inconel 600 was selected as the [. replacement material because of its superior corrosion resistance. l' 3 L i t. ?. SAFETY ANALYSIS SLM %RY Because the new J-nozzles are identical, except for material type, to the carbon steel J-nozzles, no pressure / temperature or performance parameters are ! changed. No increase in the failure probability results, no new failure modes- i exist, and no previously unanalyzed a safety concern is created. t I
EWt SLM %RY \ - i EWR 89-325, A, B - DESCRIPTI@
-In a recent inspection of the inside recirculation vortex breakers, it was observed that the 1/2" diameter concrete anchors have deteriorated, by corrosion, to a point that they need to be replaced.
The method of replacement will be to core drill around the existing bolt to '- extract it, The a larger diameter bolt shall be installed in its place. The new bolt will be stainless steel. [. SAFETY # % LYSIS SLM %RY Since a larger bolt Is to be Installed there will be more capacity than the original design. The vortex breakers will be adequately restralned during a selsmic event, , t S t
y-I EWt SLD9WtY b . [ EWR 89-328 b DESCRIPTI(N
. This EWR allavs the use of Westinghouse model ib. 771A523G06 sight glass in place of tredel No. 772A760H06 sight glass for 1-RH-P-1B rrotor on an Interim basis only. The correct tx> del number once procured will be installed during the next equipment PM or Unit Outage whichever occurs first. Work request No.
627590 has been initiated to ensure that this will be acconpilshed. l SAFETY ANALYSIS SLM %RY Replacement sight glass will perform the same function as the existing sight glass. The original operating parameters remain the same. There will be no increase in radiation exposure to personnel. The UFSAR and Tech. Spec. will remain unchanged. l
EMt SLDMNty EWR 89-329 DESCRIPTi(N Substitution (replacement) of valve No. 1-51-309, a 3/4" VGS-15A, Is reculred due to-leakage. Velan figure WO4-20$4B-13MS 1s an acceptable substitute and is currently available. l: I r SAFETY ANN.YSIS SLMWtY The replacement valve. performs the same function and meets the design requirements of the existing valve. Failure may include leakage past the isolation valve. Since the valve rrc-ts the original design criteria, probability of failure is not increased. 4 h . _ .
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l-I.' j EWR 89-330 DESCRIPTICH This EWR provided Instructions for valve operator trodification for 01-RH-MOV-1700 in accordance with IEB 85-03 progran. The Ilmit switch wiring !~ trodification set up the operator for four limit switch rotor operation. this i
- allowed the torque bypass switch to be set trore conservatively without affecting the light indicator or limit stop.
i v. I SAFETY AMLYS!$ S(ppgRy This modification does not decrease the Integrity of the trotor operated valves (MOV). The MOV reliability will be increased. I i
I? i g Det SW9%RY [ . EWR 89-341 DESCRIPTIGd A h " thick plate used in the inside recirculation spray strnp screens and support must be cut away to allcw removal of the inside recirculation punps. This EWR nodifles the plate to be rerovable by using fasteners. This will , allow the'ptinps to be renoved without the use of a cutting torch, thus Icworing dose exposure during puTp' renoval . The fasteners used are equivalent to the
- fasteners used throughout the sttro structure.
_ SAFETY ANALYSIS SW9%RY , Modifying the RS deck above the strro area will not change components as I I described in the UFSAR. The R$ System is not adversely affected by nodification of the trash rack support. The section to be modified is not part of any weight or load bearing menber of the ptrip support screens. b 6
t F { EWt SlM MRY EWR 89-342 DESCRIPTIGd Station requested to replace a failed alann bell on the Unit 2 Deluge System. The old bell is of the vibrating type,125VAC, the new bell purchased is of the same type and specifications. The work will be performed under Electrical Maintenance procedure EMP-SP-FP-2, which also includes the functional test of the replacement bell, SAFETY ANALYSIS St>9%RY Replacement of a failed alarm on the Deluge System does not compromise the safety function of the fire protection system, in fact it enhances the waring capability of the deluge system to inform fire protection personnel that the Deluge System is activated. I L 9
y = If 7 EWt SIM94tY 89-344 C A . DESCRIPTIGd These EWRs evaluated and accepted measured MOV thrust values for 01-CH-HOV-1275A & B. It was determined that the thrusts seen did not endanger or degrade the MOVs. i f I i SAFETY AMLYSIS SLM4MtY i The safety evaluation found that reliability. of the camponent/ system would not be adversely affected; therefore, the margin of safety was not impacted nor was . an unreviewed safety question created. i i
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0; { , r- y a my L ; , EWR 89-346 DESCRIPTI(N -
-Replacement valve plugs in valves 1-RC-HCV-1556A, B, C are required due to ,
leakage.-A valve plug replacement similar in design and materials is available. Part nuit>er 013434-107-780 is an acceptable replacement for part numer 013434-0950780. l r i l- SAFETY #MLYSI$ $(M SRy
- The replacement valve plug performs the same function and meets the' design ;
requirements of the existing valve plug. $1nce the valve still meets the original design criteria with valve function and operation unchanged, 3 probability of failure is not increased. f f f i f, 9 h
Det SLM%RY EWR 89-346A r. DESCRIPTI(N Replacement valve plugs in valves 1-CH-HCV-1200A, B and C are required due to leakage. A valve plug replacement similar in design and materials is available. Part nunter 013434-107-780 is an acceptable replacement part, l l SAFLTY ANALYSIS SLM%RY The replacement valve plug performs the same function and meets the design , requirements of the existing valve plug, Since the valve still meets the , original design criteria with valve function and operation unchanged, probabi1ity of fallure Is not increased. ' I t 1 P P
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L EWR SLM%RY Ic EWR 89-351A, B, C, D DESCRIPTION: L The original bearing bushing assorbly was constructed of ASTM A48 CL25B material. The replacecont bearing bushing assently is constructed of ASTM A48 i CL 30B material. Both assemblies are constructed to the same drawing and are Identical in form, fit and function. The materials are the styre except that 3 the new material is stronger (30,000 PSI tensile strength -vs- 25,000 PSI). L Since the new assembly will be stronger, it is acceptable for use, s SAFETY ANALYSIS SLM WtY: This EWR shows that, while these are "Conmercial Grade" items, they do meet the criterla for safety related items. The pump will perform as required by the performance of the listed pts and STs. These parts can be satisfactorily used as replacement parts in the Auxiliary Feedwater ptmps with no decrease in plant safety. 89-351ACBCCCD 1
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EWR 89-363 - i DESCRIPTION ' l The seat ring of valve 1-0C-MOV-100B, which is nede of natural rubber, must be replaced. Although a seat ring nede of natural rubber is not available, one that is dimensionally the sane, but rede of Nordel is available. Nordel is an
! Ethylene-Propylene synthetic rubber that has similar properties to natural rubber. This can be used as a replacement for the existing seat ring.
f b'. SAFETY MMIYSIS S(M MRy
- Use of the Nordel seat ring is acceptable as a replacenent, it's properties are similar to those of the existing seat ring, therefore, the probability of
- failure of the valve'Is not increased.
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n -. - [- r -- EWt CD9%RY n EWR 89-366 DESCRIPTICD4 t New bearings for the 1-t#-F-1C rrotor are needed. This EWR evaluated and dedicated non-safety related bearings for use in this safety related motor. j 0 l-l
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i SAFETY ANALYSIS SLM MRY r The new bearings will perform the same function and meet the same design ' requirements as the original bearings. Since the bearings meet the original
= design criterla, the probability of failure is not increased.
e, - 12 - , EMt StMWtv EWR 89-367, & A , 3 DESat! Prigs .p This EWR evaluates measured MOV thrusts for 01-SI-MOV-18658,C. It finds that the thrusts reported are acceptable based on operator capacity and MOVATS
' testing.
8 F SAFETY ANALYSIS StMWtY ' There are no unreviewed safety questions as the result of this EWR. There are no rrodifications to any systems or components, therefore, the UFSAR is unaffected. The margin of safety as defined in the basis for any Technical ? Specification is not reduced. 1
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[' i l ma mMwtY ' EWR 89-377, & A. , 1 OESCRIPTIGe ' A request was made to secure an " operational special test cabinet" In the Unit 2 piping penetration area. This EWR allows for permanent storage of the cabinet in this area. The attachrnents to a 20" x 2' concrete colunn are seismically quallfled per' Calculation CE-0494. An Appendix "R" review has boon made for small arrounts of ccxnbustibles in rags / rubber products. i : i i L SAFETY AMLYSIS SW9WtY
.There are no unreviewed safety questions. Attachment of the cabinet to a concrete colunn is selsmically quallfled by calculation CE-0696.
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EWR SlM%RY EWR 89-379A DESCRIPTION: This nodification will replace the tiner inside the RS ptmp 1A control circult. The existing Agastat tlner is a nodel 2412PF, .5-10 min,125VDC t iner, and the replacement will be a nodel 7012PF,1-10 min.,125VDC Agastat tiner. SAFETY ANALYSIS SlM %RY: The replacement of the tlner will not adversely effect the operation of the R5 ptmp control c1rcult, i I 89-379A
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.EWR 89-383 DESCRIPTION:
This' EWR evaluates and allows replac!ng.the existing SOVs 01-AS-SOV-100A
- and 01-AS-50V-100B with either ASCO Model Number NP8321ASE or NP831665E. Both are nuclear grade and rated for class 1E service.- Both have a MOPD rating above the existing instrument system pressure of 100PSIG. Both have a higher wattage rating for which existing electrical circuit protective devices are "
adequate. SAFETY ANALYSIS
SUMMARY
.Replacecent SOVs will perform the sane function as the existing SOVs, The original operating parameters remain the same. No unreviewed safety question ,
will be created by this modification. I l 3 9 89-383
o EWR 89-387 r DESCRIPTI(N A string, which is to initiate the power failure' alarm for card frame 1, 2,_3, and 6, is missing from instrwent rack cabinet 8. This annunciation alarm will ' alert the Control Room Operator of the problem in Cabinet 8 of the process ' rack. The operator will then notify the Instrtment Department for the necessary repair work. The string will be added to the string IIst to account for the missing wire. Procedure is also included in this EWR to direct'l & C - Department to add one internal wire to Cabinet 8 and provide a functional testing step to verify the operability of the annunciator. L s SAFETY ANALYSIS SLM MRy i This modification does not have any adverse impact on the safety related system,-nor will it create a new safety related concern to the station. -The modification will increase the survelliance on the' Ins'.rtment Rack, specifically, the power supply circuitry. -As such, it increases the-reliability of the Instrment Rack. l
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" SubstItutlon (replacement) of valve # 1-SI-270, a 3/4" T-38, Is required due toi-
' leakage.: A'sultable valve. replacement in design and materials is available. !
This' valve, Conval figure:12G2, is an acceptable substitution. ' w
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SAFETY ANALYSIS SLMWtY . f iThe replacement valve performs the same function and meets the design L. requirements of the existing valve. Failure may include ~1eakage past the V lsolation valve. Since' the valve meets the original design criteria,' , probability of failure !s not increased.- ~ 1 , iI U. ' i ,- s t [ li- U:
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~DESOtIPTIG6 Substitution.(replacement) of valve # 1-SI-234, which Is a-3/4" Rockwell u
~. Edwards (Figure 10484),Is required due to leakage. 'A suitable valve replacement in design and materials is the Conval (Figure'12G2) Y-globe valve.
SAFETY ANALYSIS SLMMftY The replacement valve performs the same function and meets the design requirements of the existing valve. Fallure may include leakage past the valve'. Since the replacement valve meets the~ original design criteria, the probability.of failure.is not increased, l i
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r- = ~3 p 7 EWt SLMWty 0 - EWR 89-393,A
' DESOtIPTIGO x This EWR w!Il serve as the controlling document for replacement of a service i
= water IIne 3"-WS-444-151-Q3 piping component identifled as requiring
-replacement: due to leakage' observed durIng system hydrotestIng. required by the ASME XI inservice ' Inspection Program. This is the normally-stagnant !
-containment pipe' penetration cooler return line located in the Unit 2 Quench Spray Pump house basement. In addition, a 3/4 connection will be installed on the subject component to facilitate post-rodification testing. A11 piping replacement activities will be covered -under Work Order #094931 and mechanical maintenance procedure MMP-C-W-1; separate frcm this EWR.
I SAFETY ANALYSIS SLMMRY r This EWR does not affect the design basis or normal operation of the service
-water system nor reduce the margin of safety of the service water system as defined in the Technical Specifications; therefore, no unreviewed safety question is created.
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C, > EWR 89-397-d i DESCitIPTIGO The. valve yoke on 1-MS-NRV-101C does not have guide blocks.as shown on vendor
$. drawings, Rockwell-PD-419662. A Sultable replacement material is required for- '
the guide blocks since the material Identified on the vendor drawings is not , available on site, g k i SAFETY ANALYSIS StMeftY- ' ASTM A36 carbon steel plate wl11 be used as replacement material for the material identifled on the vendor drawing. The guide. blocks are non-safety related'and do not perform a safety-related function. The substitution of material does not affect the valve from performing its intended safety related. function as described in the UFSAR. I' l 1-
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.EWR 89-399 i- t
.. DESCRIPTIG6 4
'This~EWR evaluates the s'ltuation and reason for damaged rubber boots on certain pressurizer heater connections. ' In addition, Justification'was provided.for.
the startup of. Unit 'I and the continued operation of Unit 2 with damaged' pressurizer heater connect'lon rubber boots. , SMETY ANALYSIS SLMMRY
' Operating the-Units with damaged protective boots on the pressurizer heater connections will not increase the probability of an accident or malfunction of.
equipment-Important to safety. -The margin of safety as defined-In the basis of any Technical Specifications will not be reduced.
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' s EWR 89-406 DESQtIPT!G6 '
- The new disc was supplied by the valve manufacturer as the-replacement for original disc. The new disc meets all of the requirements of.the original
. disc, and material is compatible with the original in tenns of strength and ccmpatibility with other parts in the valve. Seismic Integrity is still maintained due to negligible change in weight. -j 3
ii i s SAFETY ANALYSIS SLDMMtY No change in function of the valve will be caused by the EWR. Valve. f.
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- perfonnance and operation will be demonstrated to be acceptable by required-post-maintenance testing.- Selsmic Integrity is maintained due'to the negligible weight difference; therefore,. ECCS/ESF Systems are' not ndversely affected. The valve will function as designed to provide a Hot-Lt , injection "
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-EWR'89-412 :
L.. DESCRIPTION' This=EWR provided Instructions for valve operator modification for e: 01-SI-MOV-1864A in accordance with- IEB 85-03 program. .The Ilmit switch wiring modification set up'the operator'for four Ilmit swltch rotor operatlon. This allowed the torque bypass switch to be set trere conservatively. without - p affecting-the-light' Indication or Ilmit stop. '
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SAFETY ANAL.YSIS StM%RY This modification does not decrease- the Integrity of the riotor operated
. valve (MOV). The MOV reliability will be increased.
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m :- \L: ,, 3 , n .EWR SLM %RY w y EWR 89-412A DESCRIPTIGd h -i This EWR provided Instructions for valve operator modification for 01-SI-MOV-1864B In accordance with IES 85-03 program. The 1Imlt swltch wlrIng
.? modifIcatlon set up the operator for four 1Imlt swltch rotor operatlon. This.
allowed the torque bypass' switch to be set more conservatively without- .. $ affectIng the 11ght Indicatlon or.1imit stop. ' 'g: , SAFETY ANALYSIS SLM %RY e This modification does not decrease the integrity of the motor operated valve (MOV). The MOV reliability will be increased. .t
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f-EWR Sl9944RY g i EWR 89-416 _DESCRIPTICN . r When the turbine was disassenbied, it was discovered that the glano housings still had the shipping plugs In them. This prevented proper draining of the gland'and damaged it. The drain lines that will be Installed will route the leakoff to the suno. The supporting of the lines will be adequate to withstand a seismic. event'. L F F SAFETY ANALYSIS SIM4ARY
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-Length and. configuration prevents unacceptable seismic condition. The drain line is not required for puro operation, nor can it prevent pump operation.
-This -line is identical to the line used for Unit 2. Punp characteristics are ;
not altered. Failure of the line would only result-in overflow to the base of '
.the Turbine. Control and nonitoring capability are not affected. No new accident / malfunction condItlons are created. Auxi1Iary Feedwater System operation and availability are not affected. Also, the Auxillary Feedwater System capacity-is not affected, i
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- EWR 89-418,-A, B, C (Unit'2) r-DESCRIPTI(N It~ was discovered that the movable frame assembly In the seal table room was
'not bolted to the containment floor. Bolting of the frame is required by.
West inghouse/Teleflex Installation instructions and shown on Drawings 26641, 42748, and 42751. Since details of the bolt and other parts . associated with anchorage were not given, Stone & Webster was awarded the task of'selsmically-designing this restraint assembly.- h-SAFETY ANALYSIS StM4ARY Restraint of'the movable frame assently for the incore Flux Mapping System will .i
.be seismically quallfled with the _ implementation of this EWR, The movable frame assently will be restrained adequately during a seismic event such that there will be no damage to the Incore Flux Mapping Tubes protruding up, out of the seal. table, j
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k EWR SLM %RY 5 EWR 89-426 , L _ DESCRIPTION The 1H Emergency Diesel Generator's upper crankshaft and engine block was damaged when the No. 7 upper piston connecting rod bearing cap failed. The No.
-7 connecting rod crank pin was damaged and detennined not to be repat rable'by Fairbanks Morse Colt IND engineers. The block was damaged between No. 7 and 8 cy11nders on the control side of the engine, Just above the C.S. camshaft. The Inner block wall was hanmered out and- the weld was cracked under the top deck.
This area was repaired by Fairbanks Morse Welding S Block Engineers. A new crankshaft was purchased to replace the existing crankshaft. The new crankshaft is forged steel versus the original cast ductile Iron' crankshaft. This EWR evaluated and docunented its use in the engine. SAFETY ANALYSIS SGNARY The Safety Analysis shows that the crankshaft does not affect the operation, design or ' safety of the engine. The crankshaft is a acceptable replacement and interchangeable part. Technical Specifications are unaffected and the diesel will be proved operable by the appropriate pts for the 1H EDG.k The health and safety of the pubile is unaffected, and the station's margin of safety is also unaffected. l l l
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.EWR 89-432-F DESCRIPTION-t This EWR provided Instructions for valve operator modification for
.01-FW-MOV-100A'in accordance with IE8 85-03 program. The 1Imlt swltch wlrIng
- nodifIcatlon set'up the operator for four limit swltch rotor _ operatlon. This allowed the torque bypass switch to be set' more conservatively without affecting thel11ght'. indication or limit stop.
SAFETY ANALYSIS SLM4ARY r This modification does not decrease.the integrity of the motor operated valve '
.(MOV). .The MOV reliability will-be increased, r i
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{=, b EWL SLMWty [ EWR 89-432A DESCRIPTION This EWR provided instructlons for valve operator modifIcatlon for-01-FW-MOV-100C in accordance with IEB 85-03 program. The Ilmit switch wiring
- modification set up the operator for four '11mit switch rotor operation. This allowed the torque. bypass 'swltch to be set nere conservatively without
- affecting the light Indication or limit stop.
x 1 SAFETY ANALYSIS SLM MRY -e
- This modification does not decrease the Integrity of the motor operated valve
. (MOV). The MOV reliability will be increased.
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B -EWR 89-4328 . 4 DESCRIPTI(W . I When 01-FW-MOV-100A was Movats tested, the "as-left" total thrust was-above the ; maxtrrui value in= the setpoint docment. The upper value was due to the stem's limit. The manufacturer stated that the stem would bow as a result of the.
. overthrust and the value was conservative..Since the Movats test did not show r ' any abnormal readings, the "as-left" thrust is acceptable.
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'n SAFETY ANALYSIS SlDNARY Although the "as-left" thrust value is above the maximtm value stated in the <
setpoint. doc ment, the valve activator rating is above the "as-left" thrust. The Ilmiting component of the valve (the stem) has been shownjto be in good condition 1(i.e. no. bowing due to high thrust).by the Movats test, therefore the ,
"as-left" value is acceptable and the seismic Integrity'of the valve-is '
unchanged. This evaluation is for overthrust conditions only;:no physical changes have been made. , t l l l
p p . [ fp EWR SLM%RY bx v IJ EWR 89-439
, DESCRIPTION:
This EWR provides the . Instructions to fabricate a rock shaft for
' installation in 1-RS-27 'to meet- the vendor's specifications. The minor 3
modification for using slightly smaller Woodruff Keys to hold the tall. link arm in position on'the rock shaft is not considered a change to the design basis. l SAFETY ANALYSIS SlM MRY: l 1-RS-27 will continue to close with a small differential pressure and will continue to function as a containment isolation valve. In the event of valve i failure the MOV could Isolate the line (1-RS-27). In the event 1-RS-27 failed closed, the remaining one outside and the two inside recirculation spray lines would be sufficient to cool containment. .
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I E. ' EWR SlM%RY EWR 89-443A S D h; DESCRIPTION:
-This nodification involved' structural changes to Pipe Support 2.AJA.A 641 on Line 6".AJA.404.121.- The support was an anchor on an air ejector line, but will be converted into a two-way lateral restraint. This was' accomplished by cutting the baseplate free of the pipe wlthout disturbing the Integral baseplate / pipe weld that exists. Bunper plates were then welded to the portion of the baseplate that remained anchored to the. floor.
SAFETY ANALYSIS SlM %RY: This modification is consistent with the original design basis for pipe supports and does not affect any other component, structure or system. The modification provides greater line flexibility to acconnodate further differential building settlement. The line has been modeled with pipe support 2.AJA.A.641 as an anchor and as a two-way lateral restraint. Both models are currently acceptable (Ref. Calculation CE-0695, Rev 0).- As a two-way lateral restraint, 2. AJA. A 641 will pennit- l ine 6". AJA 404.121 to acconmodate greater differential settlement between Unit 2 main steam . valve and Reactor Containment. 89-443ASD
. - = . -
EWR SlM%RY i EWR 89-4438 & C' 1 i DESCRIPTION: This EWR revised the two (2) feedwater spring hanger nodifications. previously discussed in EWR No. 89-433. The original Intent of EWR 89-443 was carried out (i.e. Increase the free travel for supports 2-WFPD-SH-15 & 22). To accompiish this change additional nodifIcatIons were required to the existing baseplate and support framework. Modifications involved replacing the existing spring hangers with Grinnell Figure 98 spring hangers. To accomodate the .now , springs, the attachment point, on both supports' framework needed to be raised. Additionally, the existing baseplate for 2-WFPD-SH-15 needed to be reinforced to carry the seismic and settlement loads. The spring hangers were-non safety ' related but' the trodifications required drilling Hilti Bolts into a safety related concrete structure (Unit 2 MSVH). Therefore, the EWR became safety related to contro) work on the safety related MSVH wall. L SAFETY ANALYSIS SlM%Ry: The consequences for failure of-these two (2) spring hangers were reviewed by Calculation No.CE-0699, Rev. O and It was found that there would be no adverse effects upon safety related equipment. The supports were classified as-non safety related, seismic. Spring hangers are not considered when calculating seismic inertial loads. However, the failure of these two (2) , spring hangers would permit tributary dead load to be transferred to other adjacent supports as well as the safety related portion of the feedwater lines at the containment penetrations. Calculation CE-0699, Rev.0 reviewed these components for the additional tributary load and found them to be acceptable. 89-443BSC
E)dR SLM4ARY EWR 89 '460 _DESCRIPTI(N This EWR provides Engineering justification for not removing.Junper #711 and
- M03. - The Junpers were ' originally temporarily Installed and created to identify the temporary Installation of the supplemental loose parts tronitoring system, MIMS. At the time' of installation, a Junper was the only means of tracking this temporary modification. In the interim, the UFSAR was revised to address the temporary Installation.for which the Junpers were used,' thus providing an alternative method of tracking the MIMS.
SAFETY ANALYSIS SLM4ARY The margin of safety as described in the basis section of any part of the Technical. Specifications is not affected. The operability of the affected systems are not deterred by this EWR. This EWR was an engineering evaluation only, no modification was performed.
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EWR 89-470 , h, iDESCRIPT10N ' This EWR provides. instructions to rectify cable separation problems. The f i-cables are ?ocated in the Unit 2 safeguards area and do not meet cable
' separation criteria due to excessive looping of the cables. This EWR will-modify the cables in order'that they moet the criteria laid out in NAS-3012 and the-UFSAR. i I e SAFETY ANALYSIS SLM4ARY r
9
.This EWR d'oes not increase the probability of occurrence or the consequences of an' accident. The margin of safety as defined in the basis for any Technical Specification Is.not reduced'by this EWR, Cables involved in this EWR will simply.be tie-wrapped and as such no system operation will be affected.
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Ii-L J r EWR SLM%RY 'E+_. EWR'.89-516B l: ,
;. .DESCRIPTI(N During renoval of radiation nonttor sample punps 2-SW-P-6 and 2-SW-P-7 for maintenance, one of four A" dia. embedded anchor bolts broke off. This EWR.
provides instructions for replacement of the broken anchor bolts with 3/8" dia. HILTI-KWIK bolts II.- All 'four bolts for 2-SW-P-6 will be replaced.- t i i
.I SAFETY 4% LYSIS SLMMRY T$e replacement anchor bolt has the same or higher load capacity than the original boIt. . The margin of safety as described in any Technical
-Specification will not be reduced.
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DESCRIPTI(W Modification will replace control power transformer XF6 on 2J EDG which will
. require moving nounting locations slightly to facilitate Installation, e '[
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New transformer will be attached to the panel in a manner identical to that of the existing transformer.- *
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' Cable:ISISARX001 was damaged when conduit ICX9 or'RG was broken. This EWR is lssued to perfonn repair on the cable via a Raychem spIlce. The repair consists ,
p of spilcing a-new cable ISISARX003 to the existing cable-ISISARX001, and repalring the damaged condult'ICX900RG.
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i SAFETY ANALYSIS SLA9%RY This EWR only performs spilce repair. The damaged portion of' the cable wlll- be replaced and'new cable will be Raychem spilced onto the existing cable. Mounting circuitry will remain the same as well as location, configuration and operation. Therefore, no adverse effect to the station from safety point'of-view, and there are no unreviewed safety questions.. ' r 1 l< i l' t
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;,j EWR SW9%RY lc EWR 89-538 C A s
DESCRIPTf(N During a cable ' separation walk down per a NRC ccnmitnent, a color separation problem was noted between the= safety color coded-raceways of orange and purple.- The existing separation does not meet the reoulrements of_the UFSAR section u -8.3.1.1.2.3.or of NAS 3012' section G 1.2. This modification will add a~marinite
' barrier.to the top and bottom of the purple cable tray.
h,:.- F r SAFETY ANALYSIS SLM4ARY This modification-Is being made to meet the raceway separation requirements of the UFSAR and NAS 3012. There are no safety related systems being comprcmised as a result of the nedtfIcatlon.. No UFSAR changes are required due to this EWR
-and no Technical Specification will be required, l..
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g ,,; i p _ EWR SW9%RY i I: L < EWR 89-539 DESCRIPTION: The- purpose of this EWR was to provide Instructions for rebuilding one of L the centrifugal charging punps using replacement parts obtainod from Marble Hill; Power Station. The pump casings are completely interchangeable by design parameters, provided a statlon specific rotating element Is used.. This EWR provided all of the directions necessary for casing fit up such as alignment pinning, and punp casing feet hole enlargement. SAFETY ANALYSIS StM %RY: The replacement of the pump casing with one of a different casing material does not increase the probability of an accident as previously described in the
;UFSAR, create a new type of accident not previously described in the UFSAR, nor does'It reduce-the margin of nuclear safety. The punp fully retains its design head and-flow rate characteristics for normal system charging as well as those needed 'to fulfill design basis accident analyses. The material change in the replacement puno is' considered to be an upgrade by the manufacturer-provided in later model punps.
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EWR SLM%RY e .. p: . p EWR 89-540 DESCRIPTION The purpose of this EWR was to give engineering approval to a Unlt 2 Junper < j'
#892, and also to' alter the present monitoring of the level switch contacts used.to. indicate the high level: In the lower bearing oil. reservoir for Reactor Coolant punps lA and 18. This action will clear annunciator lights 112 and.
j 113. . r, i SAFETY ANALYSIS SlM%RY
-There is no change. to a safety related component, or to any section in the '
Technical Specifications. No new unreviewed safety questions will be created as a result of.this modification. The margin of safety is not reduced. ,
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i-.- o EWR SLDNARY EWR 89-556 DESCRIPTION The ' keep warm heater for the-dlesel driven-fire punn, 1-FP-P-2, is improperly connected.to the engine's coolant system. This causes the engine temperature to; be - nelntained higher than the specified 1200 F of the.NFPA code. -The high coolant temperature also causes the coolant to evaporate. resultIng in a low
' _ coolant- level.. To correct the problem the coolant piping from the heater to the engine' block was rerouted. The piping configuration is in accordance with-the manufacturer's recommendation and technical manual. The rerouting also
=a11ews the coolant to flow on natural circulation without restricticns, i
.g SAFETY ANALYSIS Sl9 NARY The Technical Specifications /LCO are not affected by the : modification. The diesel engine driven' fire pump does not. affect reactivity of the core. The ,
margin of safety for plant operation is not affected. The station's compliance with 10SFR50 Appendix "R" is not affected. The seismic integrity of the fire pump is not affected. There is no increased probability of failure to the , engine driven fire punp due to this modification. i i l'
I I EWR SLD9%AY EWR 89-561 _ DESCRIPTION i . When 02-SI-FE-2934 was inspected it was discovered that the orifico did not have a beveled edge, investigation revealed that the orifice was fabricated per E-1 DCR 6704-1 which required that the orifice be beveled. The original orifice was also fabricated for a 300 No. ANSI flange. The ficw element is in a 1500 pound line therefore a proper fit may not be obtained. New ortflee plates will be fabricated to the correct dimensions. _SN"ETY ANALYSIS SLD9%RY lsolation of recirculation flow is mentioned in UFSAR, this flow Indication will remain avallable and be enhanced by the nodtfIcatIon. Element provides local Indication only and will continue to be isolated upon an Si signal, Therefore no safety system characteristics will be affected. Physical characteristics of orifice and material of construction are equal to or better than existing element. I w i
L-U EMt Sl#$%RY EWR 89-570 i DESCRIPTICN 01-FW-7A is the upper Isolation for 01-FW-LT-1474. The valve has a body to L bonnet leak at'the seal weld. To stop the leak, an enclosure shall be i installed. The noterials of the enclosure are safety related and it is designed for the condition's it will be exposed to. In addition, Calculation CE-0731 concludes selsmic integrity will be maintained. This is only a temporary repair. The valve will be replaced during the next outage. I i' SAFETY ANALYSIS SLM%RY i An unreviewed safety questions does not exist since the ability of the S/G L level channel to sense level will not be adversely affected. In the event that Channel 1 level is inaccurate following box construction, Technical Specification conpliance is met by placing the channel in " Trip". Fallure of the Icvel channel will result in the channel being left in the " Trip" position. This is a fall-safe condition that ensures core protection if regulred. Y n
EWR 54M %RY EWR 89-588 DESCRIPTIM Replace 1-WT-484 with a Conval Globe valve. This is an acceptable replacement having eaulvalent materials, similar weight, and higher pressure / temperature ratings. I s m Analysis stM mRy Valve changeout of this nature (non safety related) will have no Impact on plant safety as analyzed in UFSAR or reduce margin of safety as described in the basis section of any Technical Specification.
- EWt SLM %Ry i-EWR 89-606
_DESCRIPTim The modification will replace the overload relay heater element for the IJ bus Emergency Diesel Generator standby Jacket coolant recirculation ptmp 1-EG-P-3J. The replacement overload relay heater coil will be a tredel H1018 which will provide better protection, f-b i -- ; i SMTTY MMLYSIS S439%Ry The trodifications provide adequate protection of motors as well we limit nuisance types. The UFSAR and Technical Specifications remain unaffected. The ; system will continue to operate as designed. r-
V l EWR SW9%RY i EWR 89-621 DESCRIPTICH i When current readings were taken on 01-HV-MOV-100D lt was noted that the value for C phase in the close direction was greater than 110% of the nameplate. Discussion with Limitorque in 1986 revealed that values below 120% were acceptable. Also, all of the other phase readings in both directions were good which shom the valve / actuator Is operating properly. In addition, bridge readings of the trotor show it is In good condition. Therefore, the overcurrent condition is acceptable. i SAFETY ANN.YSIS SWe%RY High current reading will not affect the MOV from perforTning It's function, if the valve falls to close, the redundant valve Inside containment will also receive a signal and close. Also, operators will still be able to cycle the valve manually If required. The valve has been Type "C" tested and passed. The valve also passed it's stroke time PT. l i l l u_
Em SW9%Ry 1 EWR 89-622 - DESCRIPT!Gd The existing thrust bearings for the Camponent Cooling pump need to be replaced. The new bearings are the same as the original ones in terms of form, fit and function. They shall be ver! fled to be fabricated to the correct ,
- dimensions. Also, the Impeller wear rings shall be fixed to the irtpeller by set screws which is acceptable per the pttro manufacturer. These trodifications will not affect ptmp performance.
r SAFETY ANALYSIS SWe%RY New bearings are equivalent to the original bearings in terms of form, fit and . function. The bearings will be dimensionally checked. Testing will verify , operability of the ptrro.
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EWR 89-624 r DESCRIPT!0N The trodification involves the partial retroval of pipe support 1-WS-A-177 fran line 3"-WS-44-151-Q3. The support, as Installed, is a one-way lateral restraint; however, the pipe will Impact the support under seismic conditions. The trodifIcatIon will uncouple the support from the pipe, leaving a vertleal section of 10x10 tube steel attached to building structural steel. The remaining portion of the support will no longer be able to come in contact with the pipe, precluding any potential damage due to impact at this location.' i SAFETY ANALYSIS StM %RY The tredlfication was required to conform to the latest pipe stress analysis for line 3"-WS-44-151-Q3. The modification prohibits the pipe from impacting support 1-WS-A-177. Calculation #14258.85-NPCB)-003-XE, Rev. O, Addendun I shows that seismic Integrity of the line can be guaranteed without this support. This analysis is consistent with the original design basis for pipe stress and Indicates that the margins of safety are not reduced, below code allowables, as a result of this modification, t i. !}}