Text

Annual Results & Data Rept for 1989, Including Description of Facility Changes,Tests & experiments.W/900228 Ltr
	ML20012A185
Person / Time
Site:	Point Beach
Issue date:	12/31/1989
From:	Fay C; WISCONSIN ELECTRIC POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	CON-NRC-90-016, CON-NRC-90-16 VPNPD-90-098, VPNPD-90-98, NUDOCS 9003080460
	Download: ML20012A185 (180)
	v • d • e

,l**

0

Wasconsin

' Electnc POWER COMPANY '

231 W Michgon, Roh 2046. Milwoukee W $3201 . (414)221-2345 VPNPD-90-098

- NRC-9 0-016

. February 28, 1990 U.S. NUCLEAR REGULATORY COMMISSION 10CFR50.59(b)

Document Control Desk Mail Station P1-137 Washington, D.C. 20555 -l Gentlemen:

DOCKET NOS. 50-266 AND 50-301 l

. ANNUAL RESULTS AND DATA REPORT 1989 i POINT BEACH NUCLEAR PLANT,-UNITS 1 AND 2 ;

Enclosed is the Annual Results and Data. Report for the Point Beach Nuclear, Plant, Units 1 and 2, for the year 1989. This report is submitted in accordance-with Technical Specification 15.6.9.1.B pursuant to the requirements of 10CFR50.59(b). The report contains information regarding operational highlights of the Point' -

Beach Nuclear Plant operations during 1989 and includes

-description of: facility. changes,; tests and experiments, personnel l occupational exposures, results of steam generator inservice inspections, and listings of reactor coolant system relief valve ,

challenges. Ten bound copies of this report are also being '

provided-to you under a separate cover.

Very truly yours,

, i g i

' C .- - W . Fay Vice President'

-Nuclear Power Copies to NRC Regional Administrator Region III NRC Resident Inspector i

w 1 i

OfU$0NK 0 @ r .M !

/

A subsMiwofliluvnsm DwqyGqwm&m

4 PREFACE This Annual Results & Data Report for 1989 is submitted in accordance with Point Beach Nuclear Plant, Unit Nos. 1 & 2, Technical Specification 15.6.9.1.E and filed under Docket Nos. 50-266 & 50-301 for Facility Operating License Nos. DPR-24 & DPR-27, respectively. ,

i:~

j,

?

~

?

s i .

f TABLE OF CONTENTS P3

1.0 INTRODUCTION

1 2.0 HIGHLIGHTS 2.1 ' Unit 1 1 2.2 Unit 2 1-3.0 PROCEDURE CHANGES, FACILITY CHANGES AND TESTS &

EXPERIMENTS REQUIRING PRIOR NRC AIPROVAL OR A 10 CFR 50.59 REVIEW

3.1 Amendments to Facility Operating Licenses 2.

3.2 Procedure changes 4-3.3 Tests or Experiments 69 3.4 Design Changes 71 3.5 Temporary Modifications 145 3.6 Core Reloads- 160 4.0 NUMBER OF PERSONNEL & PERSON-REM BY WORK GROUP AND JOB FUNCTION 162 5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION 5.1 Unit l' 163. '

5.2 Unit 2 164' 5.3 Errata - Steam Generator Plugging and ,

Sleeving (Unit 2) 176 ;

6.0 REACTOR COOLANT SYSTEM' RELIEF VALVE CHALLENGES

/ 6.1 Overpressure Protection During Normal Pressure &

Temperature operation 177 6.2 Overpressure Protection During Low Pressure &

Temperature Operation 177 7.0 REACTOR COOLANT ACTIVITY ANALYSIS 177 k

11

1 1

1 1.0 ' INTRODUCTION. l

The Point Beach Nuclear Plant, ' Units l' and 2, utilize identical pressurized water reactors rated at 1518 MWt each. Each turbine-

. generator is capable of producing 497 MWe net (524 MWe gross) of electrical power. The plant -is ' located ten miles north of Two =l Rivers, Wisconsin, on the west shore of Lake Michigan. .

2.0 HIGHLIGHTS 2.1 Unit 1 Highlights for the period January 1, 1989, through December 31, 1989, included a 43-day refueling / maintenance outage.

Unit 1 operated at an average capacity factor of 84.9 percent and a net electrical / thermal efficiency of 32.4 percent. The unit and reactor availability were 88.0 percent and 88.2 percent, respectively. Unit 1 generated its 60 billionth kilowatt hour on January 14, 1989; its 61 billionth kilowatt hour on May 5, 1989; its 62 billionth kilowatt hour on August 16, 1989, and its 63 billionth kilowatt hour on November 13, 1989.

y L 2.2 Unit 2 l'

o Highlights for the period January-1,~1989, through December 31, l 1989, included a 63-day refueling / maintenance outage and two I

unplanned maintenance outages. On March 29,1989, the unit '

tripped and remained off line for 98 hours0.00113 days 0.0272 hours 1.62037e-4 weeks 3.7289e-5 months when an inadvertent actuation of the fire deluge system caused a phase-to-ground fault on 2X01C main step-up transformer. On August 20, 1989, the unit tripped and remained off line for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months when an ,

actuation of the main step-up transformer 2X01B sudden pressure trip relay caused a generator lockout. This trip was classified as an unusual event. '

Unit 2 operated at an average capacity factor of 82.0 percent and a net electrical / thermal efficiency of 32.8 percent. The unit and reactor availability were 81.2 percent and 82.7 percent, 1 respectively. Unit 2 generated 1ts 60 billionth kilowatt hour-on January 26, 1989; its 61-billionth kilowatt hour on April 21, 1989; its 62 billionth kilowatt hour on July 11, 1989; and its ,

63 billionth kilowatt hour on December 5, 1989.

I '

1 I

$ '3.0 ~ FACILITY CHANGES, TESTS, AND EXPERIMENTS j u .3.1 .Amendnents to Facility Operating Licenses 1 ,

j. During the year 1989, there were nine license amenhents. issued - )

by the U.S.~ Nuclear Regulatory Commission to Facility Operating

-License DPR-24 for Point Beach Nuclear Plant Unit 1 and 10 l'

amen &nents issued to Facility Operating License DPR-27 for Point Beach Nuclear Plant Unit 2. These license amen &nents are listed

- by date of issuance and are summarized as follows: 1 3.1'.1 .02-08-89, Amendment 116 to DPR-24, Amendment 119 to DPR-27 These amen &nents modify the testing frequency specified -

in Technical Specification Table 15.4.1-1 for a number of instrumentation channels.

ll 3.1.2 04-14-89, Amendment 117 to DPR-24, Amendment 120 to DPR t

!! These amen &nents modify Technical Specification 15.5.4.2 which relates to the permissible enrichments i for storage of fuel assemblies in the new fuel storage L

vault and spent fuel storage pool.

l 3.1.3 04-17-89, Amendment 118 to DPR-24,, Amen &nent 121 l' -to DPR-27 These amen &nents modify various Technical ."

Specifications dealing with fire protection to include recently-installed systems.or to be e consistent with Standard Technical Specifications, p$.

, NUREG-0452. >

7 I 3.1.4' 04-25-89, Amen &nent 119 to DPR-24, Amendment 122

'>m to DPR-27

These amendments modify Table 15.3.5-3 to revise the L g. permissible bypass conditions for the " Trip of Both -

Main Feedpumps Starts Motor Driven Pumps." This was:

necessitated by the installation of the ATWS mitigating system actuation circuitry (AMSAC).

3.1.5 05-08-89, Amendnent 120 to DPR-24, Amen &nent 123 to DPR-27

'l These amendments modified various sections of the L Technical Specifications to reflect the use of upgraded l optimized fuel assemblies and low-low leakage core L

designs at Point Beach.

L 2

n: . - . .

3.1.6 ; 05-18-89, Amendment 121 to DPR-24, Amendment 124 '

to DPR These amendments modified Technical specification !

Table 15.4.1-1.to~ clarify the requirements for reactor coolant flow logic testing.

3.1.7= 06-09-89, Amendment 122 to DPR-24, Amendment 125- .

g to DPR-27 i These amen &nents modify Technical specification 15.3.8 to provide more specific.and precise requirements _;

regarding the containment purge and vent system.

Additional editorial changes' were made to Tables

-15.7.3-2, 15.7.4-2 and 15.7.6-2. j 3.1.8 07-31-89, Amendment 123 to DPR-24, Amendment 126 '

to DPR ,

These amendments modify Technical Specification i i

15.2.3.1.B(5) to eliminate the fAI function from the overpower AT setpoint to increase the flexibility of ,

operation at full power by allowing use of the full l flux difference operating envelope. l 3.1.9 08-24-89, Amen &nent 124 to DPR-24, Amendment 127 to DPR-27 These unendments modified Technical specification 15.6.9.2.c to clarify the reporting requirements for. .

operation of the overpressure mitigating system. They also deleted two schedular conmitments which had been implemented.

1 3.1.10 10-24-89, Amendment 128 to DPR-27 This amendment modified parts of Technical specification 15.3.1 to. revise the Unit 2 reactor-vessel surveillance capsule removal schedule.

Il 3

n p. 8 m i l

c 3.2 ' Procedure Changes

- There were no procedure changes made during 1989, beyond those

' authorized with license amendments as noted above, which rugdf red ,

Nuclear Regulatory Commission approval. Following is a list of .I procedure changes made at Point Beach Nuclear Plant during 1989, j that required a 10 CFR 50.59 review. In each case, the safety ;

evaluation determined that the procedure change ~did not pose an- ]

unreviewed safety question. The probability of occurrence or the. ;

consequences of an accident or malfunction of equipment important i to safety was not increased. The new or revised procedure did not create the possibility.for an accident or malfunction which has not been previously evaluated, and the margin of safety as defined .

in the Technical Specifications was not reduced.

3.2.1 Critical Safety Procedures (CSPs), Emergency Contingency Actions (ECAs), Emergency Operating Procedures (EOPs) and Status Trees (STs)

A common safety evaluation summary applies to changes in the CSPs, ECAs, EOPs and STs. That summary is as follows:

The CSPs, ECAs, EOPs and STs deal primarily with actions to be taken in response to emergencies which might occur l .: in the nuclear portion of the plant. They also include L secondary plant emergencies that'could affect the overall plant. The E0Ps are discussed in very general terms in FSAR Section 12.4. The description in the FSAR is not affected by the procedure changes which follow.

These changes do not alter the manner in which equipment is operated and do not change the intent of the procedure. The original 10 CFR 50.59 evaluation for each procedure remains applicable for these revisions. Human factors engineering was taken into. consideration. These procedures still conform to.the NSSS generic procedure,

except where reference to plant-specific ~ equipment is used I .. due'to differences between PBNP and the model plant used

} ,

for the generic EOPs.

a. Critical Safety Procedures (CSPs)

1. CSP-C.1: Response to Inadequate Core Cooling This revision changes Step 6 to read, " Check fewer than 5 core exit thermocouples >2200'F" to remove the ambiguity of the current wording, "a i- majority of" and meet the intent of the ERGS.

2. CSP-C.1: Response to Inadequate Core Cooling This revision added narrow range to reactor vessel level in Step 7a for consistency with other procedures. In Steps 16, 17, and 19, the 4

, ._ i i a

m

, exact number of thermocouples used were identi-fled. 'These numbers had nct been specified in, previous revisions. Steps 8 and-10 were revised ;

'from previous substeps to be high level steps _ ,

o because of the EOP practice to move transitions to high level ~ steps. -.

~

3. CSP-C.2: Response to Degraded Core Cooling l-In this revision, " increasing" was changed to ,

" decreasing" in Step 8 Substeps a and b in order to comply with ERG documents.

4. CSP-C.3: Response to Saturated Core Cooling I 1

+

This-revision adds a step-at the beginning_

of the procedure (Step 1) to check the RHRL <

system is not in service. This was done to be . j consistent with the WOG generic procedure. .PBNP-

~

omitted this_ step until we had a degraded RHR .it procedure (AOP-9C). ~Other steps in this proce-dure were renumbered as-a result of the addition. 'l of Step 1.

5. CSP-H.1: Response to Loss of Secondary Heat-Sink q This revision moved Step 14 (from Revision 0) i before Step 13 as a result of operator comments-received during training to reduce the number of- >

times the operator must go behind control board C01. The operator now only has to go behind that control board once. A note was added priori to Step 14 (Step 13'of Revision 0) to be consis-tent with standard step sequence in other '

procedures. In Step'15 Substep b, the words

" button depressed" were changed to " switch turned to block".because it is'a control switch rather than a button. Step 26 was added to: skip _

Step 27 after the last safety injection pump is ;'

stopped. -This was done following observations during simulator training of a seemingly unde-sirable transition to E0P-1 after all the safety

' injection pumps were stopped. The ERG guidelines showed we are more consistent with the generic procedure by inserting this step. Transitions in Steps 24 and-25 and all steps following this step were renumbered because of the addition of this step. <

6. CSP-H.1: Response to Loss of Secondary Heat Sink' This revision adds Step 4.c to reset feed regulating bypass valve lockout from SI if necessary. In order to operate the valves one must reset them prior to use after an SI signal. _

, 5 I

- 1 ._, . _ _

w -

' i Step 6.d.3 was modified to add the statement, j

" Hold switch in the open position which prevents l the switch from return to auto which will shut, the valve." Step 6.d.4 was modified to add the statement, "when main feedwater pump dischargei -;

valves.are open, open breakers to MOVs. This- _i maintains main feedwater discharge valves-open to feed generators:from the condensate pumps." ,

Steps 7.b and 7.c were added to restore the feed ~ J system lineup to its original state before returning to the procedure in effect.

7. -CSP-H.1: Response to Loss.of Secondary. Heat Sink This revision deletes the transition to ECA-2.1~

in the first caution prior to Step 1 because-there are'other guidelines which this caution q

could be associated with. The caution was also- -

-changed to a note because it better fits the ;

definition of a note as delineated in the EOP' Writers' Guide. -In Steps 19 and 26 the word #

" vessel" was added to reflect the nameplate ;

information in the control room. A caution was $

. added to Step 23 to verify containment spray for I.

the possibility of when the containment spray setpoint' may be exceeded during a feed and y bleed. . A standard caution and step for Step:33.

were added to stop;the RHR pumps so'a. transition l i to E0P-1.1.can be made. The RHR pumps need'to be secured prior;to SI termination. All pages- ;

. . were changed to the new page. structure and all cautions and notes were changed to the structure specified in the EOP Writers Guide.

8. CSP-H.2: Response to Steam Generator Overpressure

'This revision changed steam generator pressure '

-from 1125 psig to 1120 psig in the symptom and entry conditions note prior to Step 1, Step la, '

Step 5b, and Step Sc. The change was made in

the conservative direction to improve read- '

L- ability of the instrument.

p i 9. CSP-H.2: Response to Steam Generator Overpressure l

This revision makes all pages changes to a new page structure and changes all cautions and notes to the structure specified in the EOP writer's guide.

u 10. CSP-H.3: Response to Steam Generator High Level l

l This revision transforms Substep d of Step 4 into a high level step because of PBNP policy to have all transitions in the AER to be high level 6

l

4 \

' steps. All pages were-changed to the'new page

. structure and all cautions and notes were. '

changed to the structure'specified in the 80P writer's guide. .

11. CSP-H.4: Response to Loss of Normal Steam.

Release Capabilities' ,

This revision'makes all pages:into a new page.- I structure and changes all cautions and notes to; a new structure specified in the E0P writer's guide.

12. CSP-H.5: Response'to Steam Generator ~ Low Level' ;

This revision' changed steam generator' narrow-range level from <28% to <8% [28%) in the .

symptom and entry conditions prior ~to Step 1 to

J maintain consistency throughout the procedure l- because the status trees had the entry conditioni a and the procedure did not.

' 1

13. CSP-H.5: Response to Steam Generator Low Level- t This-revision changes all pages to a new page.

L structure and changes all cautions and notes to the structure specified in the EOP writer's.

L guide. .

g

14. CSP-I.1: Response to High Pressurizer Level This revision change to Substep 8.b RNO adds l.< direction to isolate the auxiliary spray line for the case when the auxiliary spray line..

cannot be isolated normally. It also changes "

all pages to a new page structure and changes ;

all cautions and notes to-the' structure specified in the EOP writer's guide, i- 15. , CSP-I.2: Response to Low Pressurizer Level "

L This revision changes the page structure of all !

pages in the EOP.

16. CSP-I.3: Response to Voids in Reactor Vessel ~

This revision corrected the nomenclature for.

valve SOV-570A, B in Step 19 to coincide with ,

the control room nameplate data.

17. CSP-I.3: Response to Voids in Reactor Vessel This revision changes existing Step 14 substeps to the substep format specified in the E0P writer's guide. The symptom and entry condition from L 7

f g: [ y

')

'EOP-1 was deleted because of a change _to.E0P-1 __ l to conform to HED M24. Step 20 RNO changed the ]

-transition from Step 15 to Step 12 because the 3 operators are not directed to obtain,a hydrogen -1 concentration for recalculating reactor vessel venting times in Appendix A before this revision.

All'pages were changed to the new page structure ,

~

and all cautions and notes were changed to'the.

structure specified in the E0P writer's guide.

18. -CSP-P.1: Response to Imminent PTS Condition- ?

This revision moved Step 9 (from Revision 0). 1 e; before Step 8 as a result. of operator comuments '(

received during training to minimize the number of times the operator must go behind main ;

control board C01. The operator now only_must- ;

go behind the control board once. T 7

19. CSP-P.1: Response to Imminent PTS Condition-This revision changed. reactor coolant system ~

l-(RCS) cold leg temperature in the symptom and entry conditions prior to Step.1 from 283'F to -

u 285'F. This:is a conservative change._ The word

'"nonfaulted" was changed to "non-faulted" throughout the procedure for consistency. The note prior to Step 7 was changed to be a caution note prior to Step 9. - This is a standard caution used in the EOPs. l 1.

20. CSP-P.1:~ Response to Imminent- Pressurized Thermal Shock Condition-This revision adds a note prior to Step-1 to-remind the operator of the definition of a:

D 's faulted steam generator. The Step 1 RNO was 4 l reorganized so the substeps may be more clearly. t presented. Step 3 was divided into Steps 3,.4, ;

5 and 6 because of the PBNP policy of transi-tions in the AER to be high 3evel steps. 'In - t Steps 8 and 17 a subcooling requirement was :!

added-to the RNO. This was intended to permit

, the operator time to detect the loss of-sub-

cooling and prepare for initiation of safety injection if subcooling decreases to the reini-L, 1 tiation value and to assure that no potential P RCS inyt,ttory aggravation will' occur due to RCP restart Substep a to Step 19 was added to check !! termination criteria prior to isolating accumu]; tors. An RNO was added to Step 27 'to use auxiliary. spray if normal spray is not available because the possibility exists of reaching this step without going through the previous steps that cover spray. The step u 8 l

N <

i reference in the caution prior to Step 3 was .

revised because of the divisiot. of Step 3 into.-

~

' multiple steps. 'All.pages were changed to the' new page structure and all cactions and notes were changed to the. structure specified in the EOP .Wr:1,ters Guide.

-]

j. '21. CSP-P.2: Response to Anticipated PTS ConditionT This revision changed the wording in Steps.3l l and 4 from "within limits" to'"within the- .

acceptable region" to more clearly state the intent of ths steps.

22. CSP-P.2: Response'to Anticipated PTS Condition.

I This1 revision changed reactor coolant' system'

'(RCS) cold leg reference temperature.in the: i symptom and entry conditions prior to Step 1 ,

from 283'F to:285'F to coincide.with the read- .t ability of the control room instrument. -This-change was accomplished in the conservative -

direction, n ,

23. CSP-P.2: Response to Anticipated Pressurized- 4 '

Thermal Shock condition This revision' adds a note prior to' Step 1 to remind the operator of the definition of a. i faulted steam generator. Step 1 RNO was.

reorganized so the' steps requiring feedwater control are more clearly' represented. All pages were changed to the new page structure and-allf cautions and notes were changed to.the structure 7 specified in'the:EOP Writers Guide.

L 24. ' CSP-S.1: Response to Nuclear Power Generation-lA ,

ATWS

)

1 This revision changes Step 9 RNO to ensure b positive reactivity.is not being inserted due to l: a controlled cooldown during the rest c2 the procedure. This change was in response to a WoG 4; ERG maintenance program suggestion,'DW 85-042.

g 25. CSP S.1:-Response to Nuclear Power Generation- ,

L ATWS L

n This revision changes the wording in Step 6-L- Substep a RNO to more accurately describes how to locally open the reactor trip breakers from .

the rod drive room if the trip breakers do not ,

open per discrepancy sheet #16.

9 l1 h

~

i: ,

26. CSP-S.1: Response to Nuclear Power Generation-

C ADfS l

r The revision in Step 7 ensures auxiliary feedwater-flow to remove decay heat during an ATWS event -

or a return to. criticality. .For other transients, 1 a feed flow >450 gpm could be excessive; so as L long as feed flow can be maintained, the higher ,

l, feed flow does'not need to be maintained.' ~The' ;

Step 9 RNO was refonnatted to remove substeps.

A caution was added prior- to Step 12 to alert the operator that at least one steam generator ,

must be available as a heat sink and for decay- '

heat removal and RCS cooldown. -A.second caution ,

was added prior to Step.12 to alert the_ operator that a minimum feed flow should be maintained to minimize any subsequent thermal shock-to steam :

generator components. The note prior to Step 14-L.

l' was changed into a caution.because it falls- '

i3 under the definition of a caution for PBNP EOPs.

l The words Hand bypass" were added to a substep l1 of Step 1 because PBNP now has indication of the bypass breakers. All pages were changes-to a '

,, new page structure ~and all cautions and notes were also restructured as specified inlthe E0P! ;

writer's guide. ;

27. CSP-S.2: Response to Loss of Core Shutdown ,

Tliis revision makes the note prior to Step 3 into a caution because the statement more '

h amplifies a caution rather than a note as defined-H in the ERGS. This revision:also changes all, H pages to the new page structure and changes all

[l crutions and notes to the structure specified I in the EOP writer's guide.

l r

28. CSP-Z.1:_ Response to High Containment Pressure This revision deletes valves CV-2042 and

l. CV-2045 from Appendix A, Page 3, Section B. The ,

"T" signal and the high radiation iauto-close-signal were removed from the valve control -

circuits. These valves are no longer considered 3 to be containment' isolation valves.

29. CSP-Z.1: Response to High. Containment Pressure l

This revision changed the order in which ,

u . Appendix A is presented to improve operator ,

access.

l b

10 l t

- .t 2

30. : CSP-Z.1: Response to High Containment Pressure This revision adds a caution prior to Step 5 to' :

alert the operator to maintain a minimum feed '

flow to minimize any subsequent thermal: shock to the steam generator components. The revision also changes all pages to.a new page structure and changes'all' cautions and notes to the -- ;

structure specified in the EOP writer's guide.

31. CSP-Z.2: Response to-Containment Flooding- ,

This revision changes all pages to the new page structure and changes all cautions and notes to. ,

the structure specified in the'EOP writer's guide. f 9

32. CSP-Z.3: Response ~to High Containment. Radiation' H Level ,

This revision deleted valves CV-2042 and CV-2045 from Appendix A, Page 3, Section B. The "T" signal and the high radiation auto-close signal' were removed from the valve control circuits.

These valves are no-longer considered to be containment isolation valves.

~

33. CSP-Z.3: Response,to.High Containment' Radiation Level 3 This revision changed.the order in which-Appendix A is presented to improve operator access.

34. CSP-Z.3: Response to High Containment Radiation Level-I This revision changes all pages to the new page l structure and changes all cautions and notes to- !

L the structure specified.in the EOP writer's guide. ,

b. Emergency Contingency Actions (ECAs) p

1. ECA-0.0: Loss of All AC Power I This revision deletes valves CV-2042 and .l CV-2045 from Appendix C, Section B,-Page'3. The "T" signal and high radiation auto-close signal were removed because the valves are no longer considered to be containment isolation valves.

' Numerous typographical and numerical errors were '

corrected in Steps 8 and 16. The change of the l

l 11

S ,

, g _

' 1 I

' transition in Step 25'from Step'19=to. Step 15 j allows additional checks to.be performed while !

power is being restored per the WOG ERG- i

< maintenance program (ERO DW 85-048).

, 2. .ECA-0.0: Loss of All AC Power i This revision places Appendix A in the appro- .;

D priate place within the procedure.(after Page 13, not as the second page of the. procedure). ,

This was an error which occurred'in Revision.l.

3. ECA-0.0: Loss of All AC Power f This revision changed steam generator pressures ,

l in Step 20.in the conservative direction from- -

245 psig to 250 psig. The note prior to Step 25 '

j., was revised from " change RCP seal water return valves" to."RCP seal water return flow transmitter, f inlet valve" because these valves are downstream-y' 'of the return valves and outside of the biological j l

shield. .In Step 9, the order ~of the indicated =

L equipment was revised to enhance control manipu-l !

.lations, A clarification.to the second part of Appendix A:

was made to state, "IF the bus lockout actuation is due to power not present, THEN ...."'

Appendix A was also revised to add possible feeder breakers to X13 and X14 for verification by operators that the proper breakers are e

placed in the pullout position. Appendix B was-

revised to add two more radio' communicators. '

i The order of Appendix C was rearranged for.

control operator manipulations. ,

4. ECA-0.0: Loss of All AC Power q

~

This revision makes Step 6 a high level step?

from Substep Sc. This is to stay in line with PBNP policy to maintain transitions'in the AER l as high-level steps. A caution was added prior-to Step 21 to maintain steam generator.>8% [28%). i and stop the depressuria.ation until level is- l restored. This is-a continuous action item that-should be used throughout the procedure and in '

support'of Step 21. -All pages were changed to the new page structure and all cautions and' notes were changed to the structure specified in the E0P writer's guide.

12 4

_s_ _m - + - _ _ _ . -- , +

m 'S E_CA-0.1: Loss of All AC Power Recovery Without' <

SI Required This revision changed steam generator pressures in Step 20 in the conservative direction from- .

245 psig to-250 psig. -The note prior to Step 25= ?

was revised from," change:RCP seal water return L

- valves" to "RCP seal water return flow trans mitter inlet valve" because these valves are !

downstream of the return valves and outside of !

the biological shield. In Step 9, the order of :

the indicated equipment was revised to enhance.

control room manipulations. .l c A clarification to the second.part of Appendix A was made to state, "IF the bus lockout actuation-is due to power not present, THEN ..."; Appendix A E was also revised to add possible feeder breakers j P to X13 and X14 for verification by-operators that 1 L

the proper breakers are placed into the pullout

l. position. Appendix B was revised to add two more l: , radio communicators. The order of Appendix C was -

L rearranged for control operator manipulations.

6. ECA-0.1: Loss of All AC Power. Recovery Without SI Required This revision changes the symptom and entry-L1 conditions reference of ECA-0.0 from Step-30.to Step 31 because of the addition of a step to ..

.ECA-0.0. The caution on Step I was removed and . , !

l- placed as Step 2.of a step to ECA-0.0. Steps 8

? and 9 were combined as a new high level' step to ask if-letdown can be established due-to-the prerequisite of pressurizer level to initiate the letdown. The Step 1 note was changed-to-reference Step 13 instead of Step'-12 because a.

step was added in the last revision and this note was not updated. All pages were changed tot

i. the new page structure and all cautions and -

l notes were changed to the structure-specified:

in the E0P Writers Guide. ;

7. ECA-0.2: Loss of All AC Power Recovery.With Safety Injection Required This revision changed the word " throttle" to

" isolation" in Step 2 because A0V-761A, B and C are isolation valves and are not used for throttling.

13

~ .__ __

h_ 4 w

1 4

8.- ECA-0.2:-Loss of All AC Power Recovery With-

-Safety. Injection Required: q This revision corrected a period which was used instead'of a comma on the second note prior- 1

-to Step 1-(typographical error). In the REC column of Step 2, "CCW supply" was changed to

" coolant injection" to reflect _the control board nameplate. In the symptun and entry conditions, '

the reference to ECA-0.1 transition was corrected from Step 19 to Step 20._ The; note prior to o

. Step 4 was changed to be a step because auxiliary

feedwater pumps could be placed in pullout if:

the note remained as written. This would be an unnecessary hazard to the unaffected unit ~as the power supply is one per unit. As written, the !

step would have placed the plant in an LCO for i auxiliary feedwater.

9. ECA-0.2: Loss of.All AC Power Recovery With SI-4 Required This revision changed the symptom and entry conditions for procedures ECA-O'0 and ECA-0.1 because of additions'and subtractions of steps in each of the .tspective procedures. . The. .

caution in front of Step 3 was placed_in front of Step 3 because^this is the step to.which the caution applies. This revision makes a change !

to the step referenced in the note prior to Step 1 from~ Step 6 to Step 7 as this step change was missed in the last revision. This revision also changes all pages to the new page structure and changes all cautions and notes to the new structure specified'in the_EOP writer's guide.

10. ECA-1.1: Loss of All AC Power Recovery Without Safety Injection Required

. In this revision, the word " wide" was changed to i

" narrow" to have the operators use the correct reactor vessel level instrument for the case.

when all reactor. coolant pumps are secured in :

Step 18.

11. ECA-1.1: Loss of Containment Sump Recirculation In this revision, the change rewrites Step 1 !

because this procedure was originally written to explicitly address containment sump recirc although the guideline can be entered from ECA-1.2.

14 1

3 J A note was added prior to Step 3'to prevent main steam isolation valve closure on' low compensated s steam line pressure during a controlled RCS cooldown.

Steps 8, 11, 13 of Revision 2 were deleted from this revision. The replacement of Steps 8 through 27 will bring the plant to.a safe shutdown condition (<200'F) if.the SI termination criteria are satisfied.

^

Step 8.was added to remove the loc'kout signal so-

-containment valves may be realigned.

Step 9 is part'of the standard step sequence to reset SI and secure the'SI pumps.

Step 10, with its appropriate caution and note, restored compressed air to containment to allow control of air-operated equipment.

Step 11 was replaced with a modified Step 12 of-Revision 2 to delay the RWST depletion. Step 12 is Revision 2's Step 9 which was inserted here because of PBNP's sequence of steps that 'were inserted prior to this step.

Step 13 was:added to state that the preferred method of operation for sprays and cooldown is' forced coolant flow.

Step 14 checks whether the'SI-can be terminated-and return the plant to a safe operating condition or to still add makeup to remove the excess decay heat.

Step 15 allows-the continuation with'the-procedure'to cool down to a safe shutdown condition if the SI can be terminated.

Step 16 is Revision 2's Step.10 and is placed here to remove decay heat with the possibility of using the charging pumps. In' Step 16,-

substep b, the word "the" was removed to make'it similar to substep a.

Step 17 is a transition to Step 22 to verify adequate injection flow and skip the standard step sequence to secure the SI pumps.

Step 10 reduces flow into the RCS by stopping the RHR pumps.

Step 19 flushes the SI lines to prevent boron precipitation.

15 l

{ ;. ' o .

g', t' Step 20 secures the SI' pumps following the flush

}

.". because the conditions exist which enable'the securing of the-SI pumps and puts the SI lineup J back to normal- .

Step 21-ensures the RCS makeup flow is adequate.

before continuing. >

E , LStep 23 decreases RCS pressure to the lowest possible value without losing subcooling.

3

-Step 24 checks the conditions for RHR operation..

If RHR is not possible, the procedure directs' -

setting up the conditions, j Step 25 was added to prevent the accumulator.

nitrogen from being injected into the RCS.

. Step 26 was added to follow the logical L progression'of steps to cool'down the RCS to *

[ -less than 200*F. :y 1

I Step-27 checks if the cold shutdown conditions q have been achieved.

Step 31 changed from "depressurize the intact steam generator";to " Check if all intact steam generator _should be depressurized." If the g steam generators are already be below the 825 psig, this step will not be required.

. Step 32 adds reactor vessel level wide range

.because of the possibility of.having one RCP running with a 25% void fraction in the core. ;

Step 33 was changed to become a standard step as the ERGS view it as~a standard step.

In Figure 1,_ the solid line is returned to its '

1 position in' Revision 1 as Revision 2 was done in error-in-the conservative direction. The dashed line is maintained in its original. Revision 2 position.

Throughout the procedure, the pages are all [

4 changed to the new page structure and all cautions and notes take on the new structure specified in the E0P Writer's Guide.

L<

L 12. ECA-1.2: LOCA Outside Containment

' This revision deletes valves CV-2042 and )

CV-2045 from Appendix A, Page 3, Section B. The "T" signal and high radiation auto-close signal 16 l'

l-l5 s

s l f\)' .-

were removed from the control circuitry for these valves. The valves are no longer considered to be containment isolation valves..

13. ECA-1.2: LOCA Outside Containment This revision added isolation valves to and from the RCP for isolation of component cooling water .

to clarify exactly what needs to be done. i The entry condition from E0P-0 was changed because three steps were added to 50P-0. The order of Appendix & was revised in order to more easily recognire the valves in the control room.

Step 5b was made a high level step because of the general practice of making EOP transitions b in the " action / expected response" column into

, high level steps.

14. ECA-1.2: LOCA Dutside Containment This revision changes all pages to the new page 8

structure and changes all cautions and notes to- -l the structure specified in the E0P writer's i' guide.

15. ECA-2.1: Uncontrolled Depressuritation of Both !

Steam Generators

This revision moves Step 10 (from Revision 0) !

before Step 9 in response to operator training l comments that will only require one trip behind C01 during the evolution. The note before Step :

8 was moved to before Step 10 (Step 9 of !

Revision 0) to be consistent with the standard step sequence of other E0Ps. Step 13 changes

" saturated" to "subcooled" to correct an error in the previous revision. The procedure title- ;

is now correct with the procedure identifier. .

(ECA-3.1).. In Step 36, the last two substeps were deleted as the subject transmitters are now -

in service when at power operation. There is no '

need to make a containment entry to energize the transmitters. A substep was added to Step 35 to energize the low range pressurizer l pressure instrument because of the omission of 3 substeps in Step 36. In Step 41 "RCS" was added '

to better describe what is being cooled down in the high level step, nais was done based upon WOG ERG maintenance program DW 85-099. ,

16. ECA-2.1: Uncontrolled Depressurization of Both Steam Generators This revision corrected the valve identifiers in Step 6 (MOV-836A, B to A0V-836A, B).

17 t

' i f -- ,-., w e ., ,_m , -

% ,..e.,. ..r, ,.s.,.rm. . ,,_v._, ..v., m.... ,e.. . __,,..w.,-,__m. . . . , , . - , . , -

17. Sca-2.1: Uncontrolled Depressurization of Both Steam Generators L This revision changed the procedure designator in Step 40 because the procedure was recently revised. The word " locally" was placed in front of test valves that were operated locally in order to delineate operator action.

18. ECA-2.1: Uncontrolled Depressurization of Both -!

Steam Generators t This revision deletes the third caution prior to i Step 1 because the caution is covered in the foldout page. This change added " atmospheric ['

steam dumps" to the list in Step 1 to prevent uncontrolled depressurization. This change i added Step 5 to " check secondary radiation" to :

determine if a steam generator tube rupture .

exists coincident with a secondary pressure !

l. boundary failure. Step 5_ replaces Step 13 of !

Revision 3, which the Manager's Supervisory l Staff previously required to be included in this ,

procedure. Therefore, Step 13 of Revision 3 was l deleted because the purpose of the step was ;

accomplished with WOG Rev 1A Step 5. !

i Step 21 of Revision 3 was placed inside of ;

Step 22 of Revision 4 because in order to establish letdown, one must first satisfy the i requirement cf pressurizer level. The word

" system" was removed in the foldout page because !

it is not used in PBNP plant teminology. The l RCS cold leg temperature was changed in the foldout page to ensure radiability of the i instrument on the ASIP. All pages were changed l to a new page structure and all cautions and ;

i notes were changed to the structure specified in .

the EOP Writers Guide. :

19. ECA-3.1: SGTR With Loss of Reactor Coolant-

'Subcooled Recovery Desired This revision moved Step 5 (from Revision 0) to before Step 4. This was done in response to .

operator comments such that operators would only have to move behind control board col once during the procedure. A note was added prior to Step 5 concerning DSS permission being required before opening containment isolation valves. :

This is consistent with standard step sequencing in other procedures. In Step 26 "RCS subcooling" ,

was changed to " break flow" to more accurately ,

describe the intent of the depressurization.

This was in response to operator comments ;

18

p

, i received during' training. The third substep in ;

Step 31 was removed because it is no longer possible to route the sampling drains to the facade sump. In Step 39, the last two substeps were deleted because we now operate with the subject transmitters in service during power ,

operation. A substep was added to step 38 to ;

energize the low range pressurizer pressure !

instrument because of the substeps deleted in i Step 39. Also, the "37" in the RNO was changed l to "38". This was an error in Revision O. In Step 31 the word " locally" was added because of [

the location of equipment. :

20. ECA-3.1: SGTR With Loss of Reactor Coolant- i L Subcooled Recovery Desired j This revision corrected the designation for i MOV-8364, B to A0V-8361, B in Step 10 to reflect l the proper nomenclature of the valves.

21. ECA-3.1: SGTR With Loss of Reactor Coolant- ,

Subcooled Recovery Desired This revision removed a space between " cool" and )

"down" in the caution step prior to Step 15. .i "Cooldown" is one word in other E0Ps. In 1 Step 15 the word " service" was replaced by [

" operation" to make the statement consistent I with other procedures. In the Step 20b

" response not obtained" column, the word ,

" pressurizer" was added for consistency !

throughout the procedure. In Step 40b, the l procedure name was changed to reflect revision of OP-138 to address secondary systems shutdown .

l only. Steam generator referr.nce pressure was l '

changed from 1025 psig to 1020 psig because 1025 psig is not within the capability of the meter. This setpoint change was accomplished in ,

I the conservative direction. In the Step 29 response not obtained column the ab" was changed

to an "a" to correct an error.

22. ECA-3.1: SGTR With Loss of Reactor Coolant- '

i Subcooled Recovery Desired This revision removes the reference to ECA-3'.3 Step 21 because it is no longer applicable. l Step 18 adds a reference to the RHR pumps to determine if any injection flow is being accomplished. In the red path summary for the t foldout page, the RCS cold leg temperature was e changed to ensure readability of the instrument.

In Step 12 a statement was added to obtain '

boron samples of the ruptured steam generator (s) 19

h !.

i b to avoid a possible localised boron transient upon starting a reactor coolant pump during a !

backfill situation. In the symptom and entry l

conditions, the entry condition from Step 13 of !

ECA-2.1 no longer exists because the step was i deleted with the Revision la changes. All pages- ;

were revised to reflect a new page structure and ;

all cautions and notes were revised to the structure specified in the E0P writer's guide.

{

23. ECA-3.2: SGTR With Loss of Reactor Coolant- i Saturated Recovery Desired -l t

This revision in Step 20 deletes the substep to route sampling drains to the facade sump. It is no longer possible to perform this substep. In l Step 29 the last two substeps were removed i because the subject transmitters will remain in !

service during power operation. A substep was !

added to Step 28 to energize the low range pressurizer pressure instrument because of -

the Step 29 changes. In Step 20, the word

" locally" was added to the high level step to identify the location of the equipment. !

i

24. ECA-3. o SGTR With Loss of Reactor Coolant-Saturaled Mecovery Desired

In this revision Steps 10 and 11 were revised to l change pressurizer level from (14%[34%)) to ,

c (24%[44%)) in order to maintain consistency with '

l' the ERGS due to errata data changes in Revision 1 l of the ERGS. *

25. ECA-3.2: SGTR With Loss of Reactor Coolant-

. Saturated Recovery Desired [

h This revision changed the word " thirteen" to

" seventeen" for the number of steps which should ;

l be performed in ECA-3.1 before continuing with .

this procedure of Note 2 prior to' Step. The change was made to achieve consistency with the ERGS.

26. ECA-3.2: SGTR With Loss of Reactor Coolant- ,

Saturated Recovery Desired

'L . In this revision, the word " service" was replaced with " operation" in Steps 5 and 31 ~

to be consistent with other EOPs. The word

" pressurizer" was added to the Step 9b response ,

not obtained column to be consistent with other -

EOPs. The wording in the Step 10a response not obtained column was revised to be more consistent with other E0Ps. The word " point" was moved :

?

20 ,

p j

l down one column in the Step 20 anticipated /

espected response column to maintain the ,

two-colus.n format. In the Step 26 response not !

obtained column, steam generator reference was j changed from 1025 psig to 1020 psig to remain !

within'the meter's capability. This setpoint l

change was accomplished in the conservative ;

direction. In Step 30s, the reference to

procedure OP-13B was changed to reflect a j revision to that procedure which changed the .j title. j

27. ECA-3.2: SGTR With Loss of Reactor Coolant- )

Saturated Recovery Desired This revision adds RHR pumps to Step 7 which j will determine if any injection flow is being j accomplished. RHR pumps were not included in the previous revision. RCS hot leg temperature l

specified on the foldout page was changed to

!- ensure readability of the instrument on the ASIP. The character "s" was removed from

" heaters" in high level Step 8 to correct a j typographical error. All pages were revised to ]

reflect the new page structure and all cautions -

and notes were revised to reflect the structure ,

specified in the EOP Writers Guide. I

28. ECA-3.3: SGTR Without Pressurizer Pressure (

Control This revision changed Step 17 to omit the substeps associated with routing sample drains to the facade sump because this evolution is no 4 longer possible. The phrase " restore pressurizer '

level to >14%.[34%]" were added to the note before Step 21 to properly describe.the entry- :

conditions to E0P-3 in order to avoid having to reinitiate safety injection. This change was made in response to WOG ERG maintenance program t t

suggestion DW 85-049. The last two substeps of Step 26 were deleted because the subject transmitters are energized during power operation and a containment entry would not be necessary to energize them. In Step 17 the word " locally" was inserted to identify the .

location of the equipment. ;

29. ECA-3.3: SGTR Without Pressurizer Pressure Control This revision changed the position of the word " points" in Step 17 to maintain the '

two-column format. Step 35 changed the reference to OP-13B because the procedure was 21

__ <__.-m , ___ _ _ _ _ _ _ _ _ _ _ . _ _ ___ _ _ _ _ - _ _ _ _ , _ . . . - _ _ . , - , , _ - - . - - - . , , - -

~

i revised to perform secondary systems shutdown l only. Steam generator reference pressure was ;

changed from 1025 psig to 1020 psig so pressure ;

to accommodate the capability of the meter.

This setpoint change was accomplished in the conservative direction.

30. ECA-3.3: SGTR Without Pressuriser Pressure control This revision changes the reference of E0P-3 in j the note prior to Step 21 because an additional ;

step was added to E0P-3 (Step 32 versus Step 31). .i A change to substep a was made to add, " reactor )

vessel narrow range level" and a transition to !

Step 10 was made in the RNO. The reactor vessel j level system was added to make the transition J back to Step 10 of this procedure. The transi-l tion back to Step 10, and eventually to ECA-3.1, i eliminates an extra transition out of this l procedure. RCS cold temperature in the integrity red path summary foldout page was changed to ensure readability of the instrument on the ASIP. All pages were revised to reflect the new page structure and all cautions and notes were ,

revised to reflect the structure specified in l the E0P writor's guide, j l

c. Emergency Operating Procedures (EOPs) j

1. E0P-0: Reactor Trip or Safety Injection This revision changes the word " verify" in Step +

25 to " check" based upon EOP definitions of these '

words. The transition in Step 32 Substep d RNO from 32a to Step 34 was a recommendation of the WOG ERG maintenance program, iten DW 85-03.

Step 37 and Step 38 were changed in order of presentation based upon operator comments. In Appendix B, Step 3.2.4, the words "and R11/R12 pump" were removed because this pump is no longer considered in containment isolation. In Appendix A, Page 3, Section B, valves CV-2042 and ;

CV-2045 were deleted because the "T" signal and the high radiation auto-close signal were removed.

Thus, the valves are no longer considered to be containment isolation valves.

2. E0P-0: Reactor Trip or Safety Injection This revision replaces circles around Steps 1 through 22. The circles were inadvertently-omitted from Revision 1. The change also replaces the box around Step 44, which was also inadvertently omitted from Revision 1.

22

3. E0P-0: Reactor Trip or Safety Injection This revision added a shorter version of l Appendix B to Steps 12, 16, 27, and 28; with i elimination of Appendix B. Appendix C was {

relabeled as Appendix B. Appendix A was rearranged to enhance control room manipulation.

The Appendix B reference wa added to the title page to inform the operator of its existence.

In Step I the bypass breakers were added. At the time of the initial issue of this top, no bypass breaker indication was not provided.

Step 11 was reworded to change " appropriate" to "necessary" to reflect an error in the initial issue of the document. Step 37 was added to address the problem of restoring lube oil flow to a turbine generator upon loss of D02 or D01.

Step 18 was modified to reduce the number of ,

running component cooling pumps because of the 1 relatively small heat load at this time in the j procedure. !

4. E0P-0: Reactor Trip or Safety Injection This revision changes the note prior to Step 30 to a caution because it denotes a potential hazard to equipment (Tavg instrument). Adding 2 "Tave may not be accurate" more clearly explains why we may not want to use this instrument. In ,

Step 24 there is a format change to which the

RNO is being present.- The manner in which the l equipment is operated is not changed. This .

change was accomplished in accordance with the !

EOP Writers Guide. Step 16 changes Hgg{p ghg ;

CCW pumps" to " verify the CCW pumps tripped"

( because the CCW pumps should already be tripped when there is an SI coincident with safeguards ,

undervoltage. This change does not change the intent of the procedure. The word "only" was added to high level Step 13 to satisfy the original intent of the step to only have one L

pump running after this step.

A change to Step 9 RNO manually trips the main feed pumps, condensate pumps, heater drain tank u pumps, and manually shuts the feed reg valves if the feed reg valves are not shut by this step.

l The intent of this change, per Night Order of February 19, 1988,- is to limit additional feed to the steam generator from the condensate and the heater drain tank pups during an SI caused -

by a steam line break. This RNO would accomplish this, but will have adverse effects on the secondary plant. Without these components 23

1 L j I

I t running, the cooling for various heat exchangers i and the condenser is not accomplished. This j will cause relief valves to lift and rupture i discs to rupture. '

5. E0P-0.0: Rediagnosis This revision changes all pages to the new page ]

structure and changes all cautions and notes to !

J the structure specified in the 30P writer's guide, i

6. E0P-0.1: Reactor Trip Response v This revision changes the setpoint for the j suction switchover to the refueling water t storage tank from the volume control tank from i 2% to 4% in Step 12.b per setpoint change sheet :

86-029. ;

7. E0P-0.1: Reactor Trip Response ,

This revision added Step 8 to restore lube oil flow to a turbine generator in the response not-

obtained column because of.the problem caused by loss of D01 and D02. In t'en Step 3 response ;

not obtained column, totalizer and boric acid storage tank level indication were added because [

the maximum rate could be above indication in the control room.' Thus, local reading of boric !

acid storage tank level would provide a backup l

reading. Step 8e was revised to reflect the correct labeling of the seal oil package.

8. E0P-0.1: Reactor Trip Response This revision adds substeps 4 and 5 to Step 4a RNO to complete the RNO and to prevent verifying charging after already accomplishing it in Step 4a Substep 3. The substep numering in Step 2 RNO was changed ta letters to maintain '

proper format per the E0P writer's guide. This revision changes all pages to-the new page structure and changes all cautions and notes to the structure specified in the E0P writer's guide.

9. E0P-0.2: Natural Circulation Cooldown '

This revision adds the word " limit" to make :

the notes prior to Steps 15 and 30 more under- '

standable. In Steps 21 and 39, the substeps "to place RCP seal water differential pressure transmitters in service" and " place pressurizer low range pressure transmitter in service" were removed because these transmitters will now ,

24 e -

.-- - + . - - - - -_ _ _ _ _ _ _ . _ _ _ _ _ _____-____ -

remain in service during power operation. Thus, a containment entry will not be required to accomplish this evolution during the 20Ps.

Substeps were added to Step 23a and 42a to energize the low range pressurizer pressure instrument. These steps were added becaube of ,

the remova). of the Step 21 and 39 substeps.

10. EOP-0.2: Natural Circulation Cooldown In this revision, Figure I was revised to add !

the words " acceptable region" to indicate the ;

region where natural circulation with both control rod shroud fans operating is acceptable.

11. E0P-0.2: Natural Circulation Cooldown _;

In this revision, the word " entire" was removed '

from Steps 26 and 46 and replaced by, "as indicated by core exit thermocouples, hot leg ;

temperatures, cold leg temperatures, and pres- !

surizer temperatures." Since we do not have ,

temperature indication in the upper reactor ;

vessel head region, we listed the temperature :

indication we do have. Figure I was replaced with a new figure having 50 psig lines. In :

Steps 24 and 43, the refernece to OP-138 was :

changed to reflect a revision to that procedure !

which resultes in it now addressing secondary l

systems shutdown only. The entry condition from ,

ECA-0.1 was changed because a step was added to -

l ECA-0.1. The word " allowed" in Step 46 was !

l changed to be " permitted" to make the step a standard step.

1? E0P-0.2: Netural Circulation'Cooldown ,

t l This revision makes 2.77% shutdown at 350'F in Step 4 obsolete because it is not given in the l ROD book. A value of 1% shutdown at 70'F boron value is used because it is the most conservative ,

value. The high level in Steps 17 and 32 was-- .

redefined, which checks for steam voids in the '

reactor vessel not the pressurizer level.

Pressurizer level is a check for steam voids so it was placed in the substeps. Also, this l revision changes all pages to the new page :

L structure and changes all cautions and notes to the structure specified in the E0P writer's guide.

1' 25

= _ _ _ - _ . - - . - _ _ , - _ . . -

1 f, -

g l

l

13. 50P-0.3: Natural Circulation Cooldown With Steam Void in Vessel i

This revision removes the last two substeps q from Step 9 as the reactor coolant pump seal- r water differential pressure and the pressurizer l low range pressure transmitters will remain in ;

service during power operation. A containment j entry will not longer be required to energize l these transmitters. A substep was added to Step 8 to energize the low range pressure instrument because the substeps were removed from Step 9. j Figure 1 was renamed to " Pressure and Temperature Limitations for Natural Circulation Cooldown I with Steam Void in Vessel" because the previous ,

title was confusing. An "LTOP knee" was !

included in the Figure 1 curve because the LTOP :

system should be in service prior to going below i 354'F . ,

14. E0P-0.3: Natural Circulation Cooldown With Steam l Void in Vessel This revision corrects an error which occurred -

during reproduction of Figure 1, Revision 1. ;

The line should be read as " Pressure and Tempera- i ture Limitations for." i

15. E0P-0.3: Natural Circulation Cooldown With Steam [

Void in Vessel In this revision, the word " entire" was removed i from Step 16 and replaced by, "as indicated by l core exit thermocouples, hot leg temperatures, ;

cold leg tenqperature, and pressurizer temperatures."

Since we do not have temperature indication in the upper reactor vessel head region, we list the temperature indication we do have. In Step l 13 a change was made because OP-138 was changed .

! to address secondary systems shutdown. A note l prior to Step 15 was changed u a caution within >

l a note to correct a misprint :m the previous revir:an. Step 16b was made a high level step because it is our practice to make transitions '

in the anticipated / expected response column high level steps.

16. E0P-0.3: Natural Circulation Cooldown With Steam .

Void in Vessel This revision changes the RCS cold leg tempera-ture in the integrity red path summary of the fold-out page to ensure readability of the i

26

I f

?

instrument on the auxiliary safety instrumenta-tion panel. In addition, all pages were changed to the new page structure and all cautions and notes were changed to the structure specified in the E0P Writers Guide. l

17. E0P-1: Loss of Reactor or Secondary Coolant !

This revision moves Step 9 of E0P-1 Revision 0 before Step 8 of EOP-1 Revision 0. This revised t order will require the operator to go behind i main control board only once. l

18. 50P-1: Loss of Reactor or Secondary Coolant ,

i This revision adds the identifiers "MOV" and !

"A0V" to valves in Step 14 to properly identify '

the valves. .

19. E0P-1: Loss of Reactor or Secondary Coolant f

In this revision, the word " injection".was removed and "$1" was inserted to properly designate the core deluge used in Steps 26 and '

27. In the symptom and entry conditions, the ,

word " pressurizer" was removed before the word "PORVs" for consistency with other E0Ps. The ,

step numbers from which this procedure is entered from E0P-0 because of steps added to l that procedure. Similar changes were performed -

with the entry conditions from ECA-0.2 and l CSP-C.I. A note was added prior to Step 14 which references a requirement for sodium '

hydroxide prior to the step that asks if the ;

i operator wants it. In Step 25a a procedural guide was added to direct the operator to go to CSP-I.3. t

20. E0P-1: Loss of Reactor or Secondary Coolant. l This revision changes the cover page because of steps added to ECA-2.1, E0P-1.1 and CSP-H.1. '

In Step 16a, a RCS pressure check was added in the anticipated / expected response (AER) in the case the. steam generator pressure are stable and there is a faulted steam generator which is

.depressurized at the time the SI termination criteria were checked. The contingency in the l

RNO was removed because of the coverage in'the.

i' AER. Step 24 was added to isolate accumulators to prevent nitrogen into the RCS. Step 26 was revised to add a note to properly satisfy the L requirem:nt of HED #424 to inform the TSC of the 27

--v ,- ~ -

1 i

)

possibility of using part of CSP-I.3 in this l

step. The page structure and all cautions and :

notes were changed to conform to the E0P writer's guide.

21. E0P-1.1: SI Termination !

This revision moves Step 4 of Revision 0 to ;

Step 3. This change was made as a result of -

operator comments received during training to minimize the number of times the operator must go behind main control board C01. The change !

will require this to be done only one time. In !

Step 12 the word "of" was changed to "or" to '

correct a typographical error. In Step 17, the capital "B" was changed to a lowetense "b" to be '

consistent with the E0P Writers Guide. 'In Step 26, the adverse containment condition of :

pressurizer level was changed from "24%" to "34%"

L to correct a mistake contained in Revision O.

L The transition from CSP-H.1 on the cover page E was changed as a result of step number changes 4 in that procedure. .

22. E0P-1.1: SI Termination This revision changed the auto bus transfer to manually perform the transfer in the response ngt obtained column for Step 6. This change was made to restore lube oil flow to the turbine generator upon loss of D01 or D02. In Step Sb, ;

the word " locally" was addod to delineate '

operator action for the euxiliary operator. ,

Step 6 was reworded to be more consistent with ,

language used in E0P-0.1 Step 8. In Steps 5 and 11, substep b was split into actions performed from the control room and actior.e performed by the auxiliary operator. This was done to assist the operator in deciding what work has to be completed.

23. E0P-1.1: SI Termination This revision added steps to reference CSP-H.1 in the symptoms and entry conditions.

Step 13 was added to check containment spray pumps if they are running and are no longer needed since there is a possibility of entering this procedura and not checking spray pumps running. Step 14 was added to ensure all control rods are inserted for adequate shutdown margin for the same reason. The step for checking pressurizer level was deleted because ;

it is part of operator knowledge when establishing ,

letdown. Step 15 adds a pressurizer level 28 6

ww , + - - - - - . ,.

l I

)

i check prior to establishing letdown, which is a l required action for the operator. All pages )

were changed to the new page structure end all )

cautions and notes were changed to the structure 1 specified in the 50P Writers Guide. j

\

2.4 . E0P-1.2: Small Break LOCA Cooldown and j Depressurization i This revision changes high level Step IS from, j

"Depressurize RCS to minimize RCS subcooling" to "Repressurize RCS to minimize breakflow." 1 The step was changed as a result of operater 1 suggestions for this step to be more consistent '

with similar steps in other procedures. . Step 15-j substep c changed the setpoint for RCS subcooling from "<35'[80*F]" to "<45'[90*F)" to avoid the need to restart the safety injection pumps in Step 17 and to conform to the ERG guideline che to the ERG maintenance program. In Step'24 a i substep was added to energize the low range pressurizer pressure instrument. This action is beneficial because the last two substeps were i removed from Step 25 which energized these transmitters. The transmitters will now remain l.

in service during power operation.

25. E0P-1.2: Small Block LOCA Cooldown and Depressurization i This revision underlined the word "not" in Step 9 in accordance with guidance provided in achinistrative procedure P8NP 4.19, and added the word " pressurizer" before " level" to prevent operator confusion. In Step 27, the reference l l to OP-13B was corrected to reflect a change to '

l OP-13B, which now addresses secondary systems shutdown only.

26. E0P-1.2: Small Break LOCA Cooldown and

Depressurization This revision provides a mearn of maintaining '

l pressurizer level by either the SI pumps or the.

l RHR pumps in Step 7. Step 13 was revised to add the possibility of using the RHR pump 6 in-the loop to refill the pressurizer. Pressurizer '

l level wac added to Step 18 because of tha- . ,

I possibility of getting out of the loop and not ,

checking level. Step 18 was also revised to add contingencies for <400'F and for A 00*F to allow isolation of the accumulators when the 29 k a,--,- * , - -,-. - -.c-, .- . . . -

- , - v y ,e.m w.. - . -

i: y j conditiore; are met. The page structure of the a procedure and cautions and notes were also l modified to meet the structure specified in the !

EOP writer's guide.

27. E0P-1.3: Transfer to Contairment Susp Recirculation ,

In this revision, Step 3 was revised to reflect the new normal operating mode of the component cooling system (2 heat eachangers for each unit ;

in service at all times). This change requires ,

the identifier in Step 2 to be changed and the :

second caution before Step 3 to be omitted. The +

second caution prior to Step 4 was also changed j because of the new heat exchanger. The Step 4 ;

and Step 5 tables were revised to change the word " tag" to " valve" to more accurately describe '

the number in the tables. In Step 11 the name for the 876 valves was changed to reflect the -

nameplate information and the word " locally" was !

added to the 897 valves because the operator

must now operate these valves locally. On :

Figure 1 the positions of the shields were changed as a result of MR 84-17. The word i

"down" was also changed to "up" with respect to stairway traffic to correct an editorial error.

28. E0P-1.3: Transfer to Containment Sump Recirculation .

This revision places Figure 1 before the .

fold-out page to correct an inadvertent error made in Revision 1.

i

29. E0P-1.3: Transfer to Containment Sump Recirculation This revision changes Step 12 Substep a (Page 9) from " verify containment sump >11 inches" to " verify containment sump B >11 p inches" to clarify which sump to read the level :

y indication. !

30. E0P-1.3: Transfer to Containment Sump Recirculation

[

l? In this revision, in order to allow more

<Ame for realignment of each train of RHR and SI L

to be accomplished prior to reaching the 6% RWST level, the RWST actuation setpoint for realignment to containment sump recirculation was changed -

L from 10% to 28% in Step 13 and the second ,

caution prior to Step 13. ,

l 30

i Because the original description for the substeps ,

in Step 14 was not explicit enough on what exactly i was to be done, the shutting of the SI-897A,B ,

valves had to be rearranged for a more explicit description. j Step 15 changed the " verify" to " check" to more emenplify the E0P Writer's Guide definition in ,

substep a. In substep c, an RNO was added to

" locally shut valve" because the SI pusp discharge j valves do have the capability to be locally shut. .

Substep d changed the transition to E0P-1.4 because ;

the original transition would have caused a total i loss of injection flow which is what this procedure ,

is not trying to avoid. Substep e was rearranged to the format described in the E0P Writer's Guide.

Step 17 a) RNO added the words "I,ocally shut l valve" because the SI pump discharge valves do i have the capability to be shut locally. Substep c was changed to confom to the standard step usage l

done in other EOPs.

This change makes a new page structure and l changes the cautions and notes to conform to the '

- E0P Writer's Guide format. This change also !

changes the vord "each" to "any" in Step 3d RNO l to make the proper logic in the IF/THEN statement. '

The first caution prior to Step 4 was moved.

from Step 10, E0P-1.3 Revision 4, to be closer to the front of the procedure where it will be more useful.

Step'6 was added to check and record waste holdup tank levels so at the end of the procedure when i l

checked again the operator will have a bar,is for what is happening to the RHR pump seals. ;

In order to increase the time available to crossconnect the RHR discharge to the SI pump suction prior to going on containment sump ;

recirculation, a 38% level was chosen for Step 7. :

l Step 8 was added because of the new procedure, l E0P-1.4, Transfer to Containment Sump l

Recirculation, one Train Inoperable. This step

!' will be the transition point to that procedure.

Step 9, 24, 25, 26 are all steps that were added because of the split of Step 21, Revision 4 of E0P-1.3. The step was split to make more-logical sense and flow through the procedure.

31

! i i

.In the first caution prior to Step 13, the word r " centrifugal" was added so the operator has warning to shut off all pumps (spray, RHR, SI) taking a suction on the RWST. Charging pumps t

are not included because they do not have any !

NPSH requirements. Step 28 changed the description in the transition because the i transition was changed in this revision due to the split up of Step 21, E0P-1.3 Revision 4. '

Due to the addition of a new procedure in the !

E0P-1 series, the foldout page is no longer part of the E0P-1.3 procedure. It is now part of the 50P-1.4 procedure.

31. E0P-1.4: Transfer to Containment Sump Recirculation This new procedure provides the necessary instructions for transferring the single operable safety injection train (low head and ,

high head) and containment spray system to the i recirculation mode. It is generally entered ,

when a fault in an SI or RHR system component is ;

identified and a train cannot be operated to supply cooling to the reactor vessel, after the safety injection and containment spray systems !

are realigned for containment sump recirculation, the operator returns to the guideline and step ir effect for further action. >

l The performance of this procedure will result ini (1) An establishment of a charging flow to both normal and auxiliary charging to maximize ,

the injection water into the core no matter i

where the break is located. The problem with I. the use of charging pumps is that they are L not environmentally qualified equipment. This is an attempt to provide some flow to the core from an alternate source. (2) RHR pump flow being checked to be less than the value that would be required to provide a suction source to the SI pump and deliver low head SI Sov without i causing RHR pump runout. (3) Crossconnecting l

' the operable SI pump to the RHR system, with the SI and RHR pumps operating, while preventing the RHR pump from runout. (4) Alignment of the operable train of SI for containment sump recirculation, with the SI and RHR pumps '

operating, and isolation of the RWST from the SI and RNR pump. All activities done before level in the. RWST decreases below the NPSH requirements

' of the pumps and after the containment sump level satisfies NPSH requirements. (5) Vigilance on RWST level and containment sump level for 32

~

actuation setpoints to establish the containment sump recirculation and alignment of the containment spray system for containment sump' recirculation. (6) Realignment of the suction j source of the spray pumps to the discharge of 'j the RHR pumps to maximize the amount of RWST j water to the core and in turn, inject all required NaOH into the system. (7) Determination of RHR pump seal status as to the condition for-failure and evaluate the seal as to the benefits j to isolating the leak against the degraded .l containment recirculation of flow that would be J initiated. l In this procedure when going on containment sump !

recirculation, the RHR suction check valve is '

relied upon to prevent contaminated water from -

the sump from backleaking into the RWST. This situation should exist for only a few minutes l

and is felt to be an acceptable risk.

This procedure is also concerned with RHR pump '

runout when switching lineups while-the pumps are running (SI and RHR pumps). Since there is no indication of RHR pump flow, the runout of-the pump is monitored by the discharge pressure i of the pump. {

I A change to the Technical Specifications is not involved. This procedure may violate aspects of our Technical Specification. This is recognized l by the NRC per LER 89-004-00 and does not required Technical Specifications to be changed. ;

This is within the bounds and intent of 10 CFR 50.54(x)&(y) due to predicted conditions :

a for use of the procedures.

32. E0P-2: Faulted Stear Generator Isolation This revision changed the symptom and entry conditions from E0P-0 because two steps were added to E0P-0 before this transition occurs.

O 33. EOP-2: Faulted Steam Generator Isolation

This revision changes all pages to the new page structure. and changes all cautions and notes to the structure specified in the E0P yriter's guide. In Step 4, " isolate blowdown" was '

changed to averify steam generator blowdown-isolation valves shut." The step is accomplished in this manner because the valves should already :

be shut by the containment isolation signal 33 i

p

_ y i

fb '

prior to entry into this procedure. There were '

improper substeps used per the EOP writer's l' guide in Step 3. The substeps were changed to dashed items. t

34. E0P-33.. Steam Generator Tube Rupture j i

This revision adds a caution before Step 4 to ensure the operators understand not to feed a v faulted steam generator unless it is needed for ;

cooldown. This was a suggestion from the WOG ERG maintenance program, DW 85-061, and was also :

observed as useful during simulator training l sessions. Step 11 of Revision 0 was placed ;

before Step 10 as a result of operator comments.

received in training sessions. The operators will not only have to go behind main control board C01 one time. In Step 34, " locally" was -

added to the high level step to reflect the ;

location of components. The third substep of ;

Step 3e was deleted because it is no longer .

possible to route sanqpling drains to the facade ,

sump.

l

35. E0P-3: Steam Generator. Tube Rupture This revision changed the steam generator reference pressure because controller increments did not agree with the pressure stated in Step 3b. The setpoint change was accomplished !

L in the conservative direction. Step 16a was :

rewritten to use ruptured steam generator pressure to determine-target core exit-temperature.

This clarifies to the operator which variable in

(. the table is dependent. Step 16e was similarly '

l- revised to identify that core exit thermocouple y

temperature should be less than target temperature.

A word in Step 34 was moved down one line to .

maintain the two-column format. In the caution prior to Step 15, steam generator vas made steam generator (s). The original justification for -

removing the potential for plural steam generators in a previous revision was determined to not be edequate. 'An entry condition from E0P-0 was changed to reflect steps added to E0P-0. The -

word " locally" was added to Step 24 to describe to the operator who should perfom the required e action. Steps 25 and 26 were similarly revised. '

36. E0P-3: Steam Generator Tube Rupture This change adds the expected response to Step 24, the SI line flush. When in a steam generator design basis tube rupture and an SI signal is received, the boric acid storage tanks 34 l

41 89 gy --w-- 3,y- -gy- -

re #ewwwiw,y-.-w- ,we w, - - e. --.ym+----mr-- 7 -- = > - -_m m. --

would empty and the switchover to the refueling i water storage tank would be accomplished. At j this point, RWST water would be pumped through l the SI lines. It is calculated that a 1% i decrease in RWST level on the plant process I computer system (or 2% on the control board 1 indication) is a sufficient volume of liquid to i flush the SI lines to meet the 30-minute time i limit to shut off the SI pumps. If this volume [

of liquid has not been injected through the system, then a 3-minute flush is required to prevent boric acid precipitation. The note !

prior to Step 25 is. incorporated into the !

response not obtained (RNO) of Step 24 because !

we will want to complete the 3-minute flush !

prior to stopping the pumps. Step.25 has been i revised for the possiblity of doing the RNO in '

Step 24 where the SI test valves are operated. !

j 37. E0P-3: Steam Generator Tube Rupture -

This revision removed the words "an intact" from the symptom and entry conditions as they ,

are carryover words from the high level Step 5 l in ECA-3.3 and they are not needed in this .,

procedure. The order to substeps in Step 3 was !

changed to ensure the isolation of the ruptured .

steam generator prior to possible exit into !

ECA-3.1.

L i Step 4 was changed because it used to call for i feeding the ruptured steam generator until level '

is in the narrow range, but assumed the ruptured steam generator is not faulted. Now the feed ,

flow is isolated to minimize any cooldown from '

the faulted / ruptured steam generator. There :

is a change to the caution prior to Step 15 which will alert the operator that the ruptured steam generator should be isolated from the ,

intact steam generator. Isolation of.the ruptured steam generator from the intact steam .

generator should be completed before decreasing ;

the intact steam generator pressure. Complete >

isolation must be performed of the ruptured i steam generator, but is not required prior to '

depressurization of the intact steam generator. '

l In Step 18 RNO a change was made to add, l " differential pressure between the ruptured steam generator and the intact steam generator

used for cooldown cannot be maintained >250 psig; ,

250 psi is required at no-load temperature so l primary-to-secondary leakage can be stopped ,

while maintaining RCS subcooling. A standard note was added to Step 28 for the possibility of '

35

T

i i

getting to E0P-3 without checking if the !

containment spray pumps should be stopped. A !

editorial error was corrected in Step 16 which i

changes "450" to "445". Step 2 of-the symptom l and entry conditions was revised because a step '

was.added to CSP-H.3. All pages were changed to a new page structure and all notes and cautions were changed to the structure specified in the ;

E0P writer's guide. l I

38. E0P-3.1: Post-Steam Generator Tube Rupture -l

, Cooldown Using Feedwater !

i This revision deletes the last two substeps of f Step 12 because the RCP seal water differential ,

pressure and low range pressurizer pressure ;

transmitters will remain in service during power i operation. A substep was added to Step 11 to '

energize the low range pressure instrument to ;

address the omissions from Step 11. ;

t

39. E0P-3.1: Post-Steam Generator Tube Rupture ,

Cooldown Using Feedwater +

i In this revision the purpose statement was revised to clarify the intent of the procedure.

In Step 13 the reference to the title of ob 138 :

was revised to reflect changes made to that

procedure. In Step 7 steam generator reference i pressure was revised because.the previous value 'j was not consistent with the capabilities of the meter. This setpoint change was accomplished in i the conservative direction.

P

40. E0P-3.1: Post-Steam Generator Tube Rupture .I Cooldown Using Feedwater This revision changes all W p3 to the new page [

structure and changes all ';.ations and notes to the structure specified in the E0P writer's '

guide. Due to adding a step in E0P-3, the symptom and entry conditions were changed to incorporate the E0P-3 change. In Step 8b RNO, a '

change was made to change "to control level to "as necessary" for better clarification of the step, i

41. E0P-3.2: Post-Steam Generator Tube Rupture Cooldown Using Blowdown This revision omits the last two substeps of Step 10 because the RCP seal water differential pressure and low range pressurizer pressure transmitters will remain in service during power ,

I 36 v-.w. ,, - - - - . ,.

t I

f operation. A substep was added to Step 9 to ,

energize the low range pressurizer pressure instrument as a result of the Step 10 substep i deletions. l

42. Eor-3.2: Post-Steam Generator Tube Rupture [

RToldown Using Blowdown !

This revision changed the MPC evaluation to a I Technical Support Center responsibility in the l caution prior to Step 1. In Step 18 the reference i

.to the title for OP-138 was revised to reflect- '

changes to OP-138. Steam generator reference l pressure was changed because the previous value ,

was not within the readability of the meter, i Step 12 changed the pressure of the ruptured l steam generator from 1025 psig to 1020 psig so l the pressure can be read on the gauge. This setpoint change was made in the conservative direction.

43. E0P-3.2: Post-Steam Generator Tube Rupture I e Cooldown Using Blowdown i This revision changes all pages to the new page i structure and changes all cautions and nctes to .

the structure specified in the EOP writer's '

guide. Because a step was added to EOP-3, the i symptom and entry conditions added a step in the. i E0P-3 reference. The Step 14b RNO changed "to

. control level" to "as necessary" for better

[ clarification of the step.

l 44. E0P-3.3: Post-Steam Generator Tube Rupture Cooldown Using Steam Dump ;

This revision omits the last two substeps . :

of Step 10 because the RCP seal water differential i pressure and low range pressurizer pressure ,

transmitters will remain in service during power -

operation. A substep was added to Step 9 to energize the low range pressurizer pressure .

l instrument as a result of the Step 10 substep deletions.

45. E0P-3 r Sost-Steam Generator Tube Rupture Cooldcun Using Steam Dump i l

This revision changes the title of Section 1 of the fold-out page for the E0P-3 series of procedures from "SI Termination" to "SI .

Reinitiation." This was an inadvertent error contained in Revision 1.

i 37

=a

~

46. 50P-3.3: Post-Steam Generator Tube Rupture Cooldown Using Steam Dump This revision corrects a typographical error to Step Sa RNO which changes 2000 gun feed flow of auxiliary feed flow to 200 gun.

47. E0P-3.3: Post-Steam Generator Tube Rupture Cooldown Using Steam Dunqp Revision 4 changed the reference to the title of OP-13B in Step 12 because the procedure had been revised to address secondary systems shutdown only. Steam generator reference pressure was conservatively changed in Step 12 because the previous value was not within the readability capabilities of the meter.

48. E0P-3.3: Post-Steam Generator Tube Rupture

.., Cooldown Using Steam Dump L.

l' L

This revision changes the RCS cold leg temperature in the integrity red path summary to ensure readability of the instrument on the auxiliary safety instrumentation panel of the foldout page. All pages in the procedure were

changed to the new page structure, along with all cautions and notes, which were revised to the structure specified in the EOP Writers Guide. E0P-3 added a step in its procedure so the symptom and entry conditions have added a step to the E0P-3 reference. Step 14b RNO was changed "to control level" to "as necessary" for better clarification of the step.

d. Status Trees (STs)

1. ST-2: Core Cooling I

This revision placed brackets around adverse containment conditions. The parentheses used in the original issue of the document were not in accordance with established format. An editorial error was corrected which changed "t" to "T".

2. ST-3: Heat Sink This revision changed steam generator reference pressure in the conservative direction to enable readability of the instrument. The steam generator narrow range level setpoint was changed to correct a typographical error.

Adverse containment condition information was added to be consistent with other STs.

38

, _ _ _ . . . . . _ _ _ _ _ ~

i

3. ST-4: Integrity ]

This revision changed the temperature of both !

reactor coolant system (RCS) hot legs from 283'F i' to 285'F for instrument readability on the auxiliary safety instrumentation panel (ASIP).

Figure 1 was also revised for this reason. The ;

setpoint change was accomplished in the i conservative direction. l

4. ST-5: Containment i This revision added "B" to specify which I containment sump indication is to be used on the ;

auxiliary safety instrumentation panel (ASIP). l

3.2.2 AOP-14A

Main Power Transformer Backfeed {

AOP-14A provides guidance for backfeeding power to either i unit's non-vital buses (A01, A02) through the main step-up transformer (X01). Implementation of this procedure '

assumes the plant is in at least a stable hot shutdown condition after a loss of all Ac power to the station ,

auxiliaries accident.

The procedure alerts the operators to monitor the unit auxiliary transformer (X02) to ensure that it is not t overloaded. Under the direction of the technical support ,

center, duty and call superintendent, and/or the DSS, and p pursuant to 10 CFR 50.54(x) and (y), either unit's vital >

buses may be energized as needed to provide auxiliary power.

Disablir.g generator protective relaying is necessary to achieve backfeed of the main stepup transformer. This !

does not pose a loss of protection for the generator since [

! it is physically disconnected from the incoming bus. ;

[ Disabling this relaying does not degrade any reactor '

protection systems. :

p 3.2.3 PT.R-6: Periodic test procedure PT.R-6, which accomplishes >

a service test of station battery D06, to demonstrate that

l. battery D06 is not beyond the point in life where it is '

operable and capable of perfonning its function as stated '

in the FSAR and Technical Specifications.

Summary of Safety Evaluation: An evaluation is required I because the proposed evolution could affect the function L

or method of a system, structure or component described in ;

the FSAR. TS 15.3.7.A.1 requires four (4) operable

L battery chargers with the chargers carrying the DC loads -

L on each DC main distribution bus to ensure the batteries will always be at full charge in anticipation of a loss of AC power incident prior to making either reactor critical.

39 a=___________-_______-_ __ - . - - . - -- -- - .

1 During normal power operation, the battery (D05/D06) may be inoperable for a period not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided the other batteries and four battery chargers remain operable with one charger carrying the DC loads of each main distribution bus. -

1 Imposing the test criteria of projecting >l hour battery !

capacity remaining at the current service test discharge, ;

or securing the battery service test, ensures that the I battery (D06) remains operable throughout the test and capable of providing its function as defined in the FSAR and Technical specifications.

1 i

The charger will remain operable throughout the test with (

the AC breaker shut and the DC breaker open with voltage ;

adjustment turned down. Throughout the test the DC .

breaker shall remain capable of being closed to supply the !

DC loads for D06 should' conditions require this.

It is therefore the conclusion of this evaluation that the intent of the above referenced TS is met during the period that the battery service test is in progress. This inter- ;

pretation is consistent with Standardized Technical Specifications, which require that the batteries and their associated chargers remain operable, and if one charger becomes inoperable, the battery must remain adequately charged.

3.2.4 Routine Maintenance Procedures (RMPs) i

a. RMP 7B: Opening "B" Steam Generator Primary Side, .

Unit 1(2) .

This RMP allows the installation of RAD doors on only one manway per steam generator during reduced inventory (3/4 pipe).

Summary of Safety Evalue, tion: Although not required, a safety evaluation is considered to be appropriate ;

for this case because of the potential impact upon the '

loss of RHR during reduced inventory (mid-loop) operations. This accident is not addressed in the' FSAR but is a current industry concern.

The steam generator radiation attenuating doors (RAD)-

are used during refueling outages while the steam generator channelhead manways are removed to reduce l local radiation dose rates on the channelhead platforms.

These doors are used in conjunction with the steam generator ventilation HEPA units, which reduce 4 moisture in the tubes during eddy current testing and purge the channelheads of radioactive gases.

Installation of RAD doors on both manways prior to both nozzle dans being installed might violate the required size of the vent path.

40 L4 l_.. --

I k

Mid-loop operation requires a hot leg vent path with a cold leg opening to preclude a loss of RCS inventory.

NRC Information Notice 88-036 addresses concerns regarding a hot leg vent path while open cold leg vent paths exist. Due to the possibility of an increase of ,

reactor vessel pressure causing water-to be driven out

, of the cold leg opening, a hot leg vent path is required. The installation of both RAD doors can have 6 an effect on this concern at mid-loop due to the restriction they impose on the manways.

To ensure that this'does not occur, RMP #7A, Opening [

"A" Steam Generator Primary Side, Unit 1(2), Revision 0, l and #78, Opening "B" Steam Generator Primary Side, .

Unit 1(2), Revision 0, allow only one door to be installed on each steam generator. Regardless of which manway the RAD door is installed on, an i acceptable vent path exists due to the communication '

of the hot and cold legs via the tubes. ,

i !

This RMP does not adversely affect the consideration i

of RHR loss / loss of RCs inventory during mid-loop l operation. l t

b. RMP 23A: 480 V Breaker Maintenance, Unit 1 -j RMP 23A procedurally controls the annual maintenance of the 803 and B04 normal feeder breakers to ensure the breakers operate within the Westinghouse-specified limits.

Summary of Safety Evaluation: An evaluation is required because performance of this activity will ;

alter a' system, structure or component described in the FSAR.

4 l However, to place the plant in a condition to perform :

l- this maintenance requires electrical breaker swapping '

and out-of-normal electrical plant lineups. The maintenance is to be performed when the unit is in

cold shutdown and potentially while decay heat removal, ;

by RHR, is of utmost concern.

l The procedure utilises the opposite unit B03/B04 bus tie breaker as a substitute for and while performing maintenance on a normal B03 (or B04) feeder breaker.

While the opposite unit's tie breaker is in the normal feeder cubicle, the normal feeder breaker will now

have a reduced overload short delay time characteristic.

l This is acceptable because all load breakers' l characteristic curves,'on B03 or B04, are below the L B03/B04 tie breaker characteristic curve. Therefore, ,

breaker coordination is maintained for the bus whose l normal feeder breaker is under maintenance.

l 41 f-

0 G

The requirements as described in the RMP significantly minimize tbt time for this-potential to esist, and since this is a maintenance concern, its intent as ,

described in the FSAR to provide emergency power is I achieved with minimal impact on the operating unit. !

Standardized Technical specifications typically I authorize up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of having the buses tied j together, but this RMP will_ minimize this time to typically less than 5 minutes. A Technical Specifi- l cation change is being drafted to address this situation j where the 803/B04 bus tie breakers can be shut for :

maintenance purposes. All of these situations are )

significantly more restrictive than what is presently l authorized by TS 15.3.1.A.3.a.(1)(c. and d.)* note concerning RHR operability. Therefore, the maintenance l should be authorized pursuant to the intent of TS and !

the FSAR maintenance description of the tie breakers,

c. RMP 23B: 480 V Breaker Maintenance, Unit 2 RMP 23B procedurally controls the annual maintenance ,

of the 803 and B04 normal feeder breakers to ensure the breakers operate within the Westinghouse specified .

limits. This maintenance is already controlled by following the Westinghouse technical manuals and l requirements. However, to place the plant in a ;

condition to perform this maintenance requires electrical breaker swapping and out-of-normal "

electrical plant lineups.

summary of Safety Evaluation: An evaluation is required because a system, structure or component described in the FSAR will be altered. Furthemore, the proposed evolution constitutes a potential change to the facility or its operation as described in the .

FSAR. l The requirements at described in the RMP significantly minimize the time for this potential to exist, and t

since this is a maintenance concern, its intent as l described in the FSAR to provide emergency power is achieved with minimal impact on the operating unit.

Standardized Technical specifications typically authorize up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of having the buses tied together, but this RMP will minimize this time to '

typically less than 5 minutes. A Technical Specifi-cation change is being drafted to address this ,

situation where the B03/B04 bus tie breakers can be shut for maintenance purposes. All of these situations are significantly more restrictive than wh.t is presently authorized by TS 15.3.1.A.3.a.(1)(c. and d.)*

note concerning RHR operability. Therefore, the main-tenance should be authorized pursuant to the intent of TS and the FSAR maintenance description of the tie breakers. r 42

_ .. _ _ _ _ . ._ ___ ~ - . _ . , _ _ _ . _ _ . . - .

The maintenance is to be performed when the unit is.in cold shutdown and potentially while decay heat removal, j by RNR, is of utmost concern.

The procedure utilises the opposite unit B03/B04 bus

-tie breaker as a substitute for and while performing. !

maintenance on a normal 803 (or 804) feeder breaker. i While the oppc,dite unit's tie breaker is in the normal l feeder cubicle, the normal feeder breaker will not have a reduced overload short delay time characteristic. !

This is acceptable because all load breakers' ,

characteristic curves, on 503 or 804, are below the 4 B03/B04 tie breaker characteristic curve. Therefore, ;

breaker coordination is maintained for the bus whose i normal feeder breaker is under maintenance.

Appendix R concerns are addressed (on the Unit I side) '

by requiring the availability cf self-powered radios f which effectively eliminate the requirement for the l 1804 to MCC B43 power supply path to the installed plant radios. Thus, the Unit 1803/B04 bus tie breaker would not be required to be available for Appendix R consideration and no further compensatory actions are needed for Unit 1. Since the Unit 2 803/B04 bus tie breaker remains in place, there is no Appendix R concern for Unit 2. Therefore, the Appendix R :

concerns are met by the RKP.

d. RMP 30: Opening of Pressurizer Manway 1(2)T-001, (

This tesqporary change will specify and describe the temporary cover to be installed over the pressurizer -

manway.

Summary cf Safety Evaluation: An evaluation was considered to be required because installation of the cover over the pressuricer manway could have an impact >

l on the consequences of a loss of RHR accident when in

l. mid-loop operation. This accident is not described in the FSAR but is the subject of NRC Generic Letter 88-17.

Installing a temporary cover over the pressurizer manway has the potential of impacting the assumed vent ,

path for RCS protection during loss of RHR during mid-loop. Westinghouse did an evaluation to determine acceptable hot leg vent paths to prevent RCS )

pressurization and ejection of fluid from a cold leg opening. The analysis states that a vent capable of maintaining RCS pressure <2 psig following boiling in the core is required to minimize or prevent s spillage out of a cold leg opening.

The resistance for the pressurizer vent path was calculated to be 26 ft 4 with the surge line being the limiting path. The cover over the manway will result 43

t: i

}g. 'I i l I

in a pressure buildup in the pressurizer until the i cover is lifted off. This pressure buildup will- I reduce the available pressure drop through the surge !

line. The static pressure required to move the cover 'j' is. calculated to be 0.039 psi.

As the cover tips, the required force (pressure) will continue to decrease until it is blown off. j This pressure differential would only reduce the j

' available pressure drop for the surge line by about 2% i and thus, has an insignificant impact.

[

The pressurizer aanway will be considered an j acceptable vent after 108 hours0.00125 days 0.03 hours 1.785714e-4 weeks 4.1094e-5 months assuming an initial ;

RCS temperature of 140'F. l The boiloff rate at 108 hours0.00125 days 0.03 hours 1.785714e-4 weeks 4.1094e-5 months after shutdown is about I 5.5 lbm/sec; 5.5 lbm/sec of 212*F steam blowing through the manway would have an impact momentum force ,

of 18 lbf.

There la sufficient force to blow the cover completely off. A cover weight of up to 18 lbf (a cover weight ,

of up to 18 lbs/Cos45' E25 lbs could be used) is !

considered acceptable. This would be a 3% reduction-i' in the AP available for the surge line. ;

Installation of this cover over the pressurizer manway l

l opening as a safety precaution and to prevent foreign !

" objects from entering the pressurizer is acceptable j

and has no impact on the consequences of a mid-loop l l

loss of RHR accident provided the time since reactor t shutdown has been >108 hours,

e. RMP 30: Opening of Pressurizer Manway 1(2)T1.

The evolution will perform a hot retorque of the manway for pressurizer 1(2)T1. l Summary of Safety Evaluation: An evaluation is-required because the proposed evolution could affect ;

the function of a component described in the FSAR.

t The evaluation considered the following three aspects:

i l

1. The fact that the bolt stress will be increased L from 1600 ft-lb to 1800 ft-lb.

l

2. The fact that the torquing will be done at an elevated temperature and pressure versus ambient

(~70'F) and atmospheric pressure.

3. The fact that one bolt at a time will be removed for relubrication prior to torquing to

1800 ft-lb.

44 ;

- - l ?.* L __ L - . _ - - . . _ . - , , . - . . . - . . -, , , , - -

O: ,

l

, 1

-1 x

The 1800 ft-lb is considered acceptable because this 2 was the value recoereended by We:tinghouse for a period :

of time. They have reduced the recommended value to j g 1600 ft-lb because of galling concerns, not because of overstress concerns. It is expected that the actual.

bolt stress obtained will be. lower than would be espected at a cold condition because the lubricant-will be evaporating due to the hot condition as the bolt is installed and torqued. This.will increase the friction coefficient, which will result in a-lower.

applied tensile stress.

The higher temperature between 400 and 450'F, besides.

impacting the lubricant, affects the. thermal growth of-- !

L, the bolt and cover. The bolt will be cooler than the l cover when reinstalled and torquad. As the. bolt heats !

up and expands, the bolt stress will be reduced. It ;

a is espected that the bolt will not cool off completely '

y while it is removed and is being lubricated, and~will >

heat up quickly when screwed in, and the impact will

~

1 be slight. Also, a final torquing pass will help t compe wate eud the bolts will be hot by that time.

The stress will be on the conservative side and not be l an overload concern.

L ,

L Some deflection does occur as pressure is increased, !

L'-

resulting in flange roll about the gasket contact l surface. This flange roll will result in slightly-l- reducing the bolt tension. This-fact is considered in-r the specification of the bolt torque. The fact that L the pressurizer will be pressurized to about 450 psig.-

l> vill recult in some flange roll that will act to L increase the bolt stress when the pressure is relieved.

The pressure deflection at this load will be considerably less than at design pressure pd is not l - considered to be a significant impact. Also, the -

l friction' coefficient' increase and thermal expansion >

factors mentioned above will tend to offset this condition. Further, the potential increase in~ stress l' would only occur at the cold, depressurized condition.

L calculations show that the bolt stress in the remaining- :

bolts is not a concern with one bolt removed at 450 psig, and the possibility of a manway failure

l. during the performance of the job is not a concern.

h The only potential is the possible increase in the leak ,

[? rate, which does not pose a safety-related problem. -

~,

The possibility exists that the actual bolt clamping load may be reduced becruse of some of the facts mentioned above. This is not a safety-related concern because the structural strength would not be impacted; L only the leaktightness, which would be detected by the

fact that the leak was not stopped.

l 45 k'

- t .

e ,

e l The cover should be removed, the bolts cleaned and

,f inspected, and the gasket replaced at the' neat a: refueling outage.

E f. RMP 111: ISI-866A(B) Valve Operator Maintenance, l ValveISI-86NBisnormallyopenduringnormaloperation. N This valve provides isolation _of the_"B" high head SI -

pump discharge. .This valve may be shut for either-an .:

SI line break or for controlling core injection with loss of containment recirculation.

Removal of the valve operator for maintenance will require the unit to be.in a cold or refueling shutdown. ;

a. The valve is to be clasped in the open position per r RMP #111 such that requirements for mid-loop (reduced inventory) operation are met (one train of safety injection should be available as a backup for RHR when in mid-loop operation).

l :.

Summary of Safety Evaluation: An evaluation is -

required because performance of this activity will alter a system structure or component as described in l the FSAR.

, calculations show that the stem clamp has been demonstrated to be adequately designed. The stem clamp will be-installed to keep the valve open.

Removal ~of the valve operator will reduce the mass of the valve, therefore reducing seismic stresses. The seismic loading on the 1-1/2" unsupported valve stem is not considered excessive by engineering judginent.

t

g. RMP 118: SI-852A(B), Valve Operator Maintenance, l Valves ISI-S52A(B) are the interface vtives between L the RHR system and the reactor vessel (RV) low head-safety injection (SI) core deluge headers. This interconnecti:.n is used during a large break LOCA, and for long-term post-LOCA core cooling by recirculating the contents of the _ containment sump through the ~ RHR

}Gs end back into the RV.

L These valves are normally shut and open upon SI actuation. During perfonnance of routine maintenance or rebuilding of the Limitorque motor operator per +

RMP 118, the motor operator must be removed, and '

operationally,~the valve should remain shut to prevent bypassing RHR cooling flow through the RCS hot legs instead of back to the RCS cold leg and across the core.

46 a .

I

..'.I- , , ,

~

a.

l ;

1 Summary of' Safety Evaluation: The stem clamp has been l;' designed to provide adequate force to keep the valve shut at maximum RHR system pressure (setpoint'of relief valve 151-861A), and is designed to not damage-the valve stem / operator shaft.

Valves ISI-852A(B).will only-be clamped shut during cold shutdown or refueling outages when automatic-actuation of SI (SI-852A&B opening) is not required.

C1' aping either valve shut will not edversely affect the operability of the RHR system, and only one. valve will be clamped shut at any one time.

1 The seismic qualifications of the system will-not be adversely affected because-removal of the motor operator will reduce stresses on the valve and associated piping,

h. RMP 126: Electric & Water Supply for Hydrolancing-Unit 1(2)

RMP 126 will allow water from the condensate storage

- tanks (CSTs) to be supplied to hydrolancing pumps.

The RMP is evaluated with respect to interaction with the safety-related function of the CSTs; Since the CSTs serve as the source of auxiliary feedwater, TS 15.3.4.A.3 requires that a minimum reserve of

.10,000 gallons per operating unit be available in_the ;

L CST (s). _l s !

L The proposed RMP also involves connecting power to hydrolancing pumps from motor control center 1(2)B31.

These MCCs are not described in the FSAR and provide ;l no safety function. I

/ Summary of Safety Evaluation: An evaluation is

I b

required because use of the procedure involves a potential change to the facility or its operation.

i Evaluation of the propcsed RMP demonstrates the worst L case situation involves rupture of the hsoe running from the CST. This situation is essentially the same as evaluated in a prior SER (safety evaluation of- the i i temporary modification to provide DI water from the CSTs to DI. water header during water treatment plant modification). Calculations described in this review determined that 22 minutes are available from the CST low level alarm to a level corresponding to 20,000 gallons of feedwater reserve if failure-

, , occurred at the 2" temporary connection. This is sufficient time to allow for operator intervention to prevent violation of the applicable Technical

. Specification.- No other mechanism related to the RMP has been identified as a potential safety issue under the provisions of 10 CFR 50.59.

i 4

, 47 ,

!ll u _

~ . . . _ _ - . .

E f

3.2.5 SLP 2: Items Lifted by Unit 2 containment Polar Crane The safe load paths for ' safety-related handling devices ]

procedures' manual was originally drafted:in response to ,

mIREG-0612; The intent of SLP 2 is to provide procedural l guidance for limiting the travel of heavy loads over the !

reactor vessel.

An NSSS reactor vessel head drop analysis shows that it is possible to buckle a reactor vessel (RV) support if a 30,000 pound load was dropped on.the RV. Travel.of heavy ,

loads over the RV cannot be avoided.. The NRC has found ,

-the probability of a crane failure to be low enough that a

single failure-proof cr'ane, as described in WVREG-0544, is-not required. ,

The changes to SLP 2 Revision 3 are within the scope of the PBNP NUREG-0612 response to provide procedural guidance for moving heavy loads in containment. The ,

changes more clearly define the paths of heavy loads that

- must be carried over the RV.

A new path was added for the. reactor coolant pump (RCP) motor flywheel.to be carried over/near the RV to a laydown area that is not in the way of operationsLrefueling -

activities. This is'the only change which increases the ;

number of lifts carried over the-RV.

Summary of Safety Evaluation: An evaluation is necessary-although there is no specific reference.to the containment

, < safe load paths in the FSAR. However,.since PBNP is l L committed to NUREG-0612 and the changes made have the.

" potential to increase the probability of a heavy load drop, a safety evaluation is considered to be prudent. 3 Movement of the ."A" RCP flywheel to the north laydown area is a new path which adds movement over/near the reactor-vessel. It is noted that this is the same path as that +

used for movement of the RCP motor-(with the flywheel)'.

It could be said that a flywheel drop would be bounded by-a drop of the motor / flywheel assembly; however, there were ;

no specific analyses performed to assess the consequences -

of a dropped heavy load in containment. The weight of-the - flywheel is 14,000 pounds, which would be bounded by-

'E the NSSS's 30,000 pound number, however. The. drop of any heavy load in containment is considered to have unacceptable results. ,

The added path for the "A" RCP flywheel does increase the o potential of a heavy load drop that could damage the RV just by the fact ~ that we have increased the' number of loads that are lifted in the vicinity of the RV. However, the potential for dropping a heavy load has not increased 48

. ' ' ' _ - _ ___I__._.___ ~ --

d j

i because the same precautions and rigging requirements _.

apply to this load. Also, the frequency of handling the ._

flywheel is not increased by this change.

o ;

Because the flywheel weight is less than the 30,000 pound I '

weight that the NSSS said could buckle the RV supports,-

y all the precautions for liftinga heavy load will be met and this path is an siternate path (not the preferred i path), the change does not constitute an unacceptable-increase in the potential of a dropped heavy load or the consequences associated with it.

3.2.6 Special Maintenance Procedures (SMPs) {

'I a.

SMP 961: 13.8 kV Field Modifications, Modification.

Package 87-002*G, Unit 1 The 4160 V power supply configuration during certain phases of the U1R16 refueling and maintenance outage ,

will be as described in the special' maintenance procedure. ~This configuration has been evaluated (with and without a fast bus transfer on Unit 2) for-its effect on the FSAR, the Technical Specifications e and the NRC commitments as found in the safety _ +

evaluation reports.

! (

Section 8.2.2:of the FSAR states-that auxiliary power required during unit shutdown is supplied from the unit's XO3 and XO4 transformers. This will not be the case for Unit I as its auxiliary power will'be supplied ,

from 2XO4. However, Section 8.2.2 also. states that =t buses 1A03 and 1A04 can be supplied through crossties to 2A03 and 2A04, respectively, if 1XO4'is removed from service. This will be the case for Unit 1, and the FSAR will not be significantly_affected.

The configuration is controlled by TS 15.3.7.on- the auxiliary electrical systems required-for plant ~

operation. TS 15.3.7.A.1 includes tht requirement that the XO3 and XO4-transformers associated'with the ,

reactor to be taken critical are in service. Since 2X03 and 2X04 will be in service while Unit-2 is. -

operating, and Unit I will be at cold shutdown, the TS is met. t A loss of all AC power to the auxiliaries for either

.t unit, as presented in the FSAR Chapter 14 safety analyses, would not be affected by this SMP. . An initial assumption in the analysis of this event-is a-loss of all power to the station auxiliaries. A loss '

of-power to Unit I will have no effect because IXO3 and 1XO4 will be out of service. A loss of power to Unit 2 will create the situation analyzed in chapter 14 of the FSAR. 'As this is an analyzed condition, the SMP will not adversely affect the safety analysis.

49

+

Plant safeguards electrical loads for both units and

--nonsafeguards 480 v loads fer Unit 1 will be' carried by the 2X04 transfomer while the fast bus transfer capability for Unit 2 will be maintained.- In the

event of a' fast bus transfer, 1X04 will also carry the '

Unit 2 nonsafeguards loads.' This configuration' differs slightly from the normal outage lineup and therefore must be evaluated. Unit 2 will be at power.

and Unit I will be in cold shutdown during the perfomance of this SMP.

Maintaining'the 4160 V fast bus transfer capability.

for Unit 2 while all Unit 1 buses and all Unit 2 '

safeguards buses are sqpplied by the 2X04. trar.sformer requires bypassing an interlock for breakers 2A52-44 ,

(2A01 to 2A03 tie breaker) and 1A52-48.(2A02 to 2A04:

tie breaker). This is an abnormal configuration and-as such also_ requires evaluation. As the fast bus l- transfer capability itself is not required to maintain <

l

' safety-related functions, its operation does-not require analysis. However, the potential for overload-of the 2XO4 transfomer and -its related hardware must -

be analyzed to ensure that the preferred safeguards power source is not adversely affected.

A conservative load analysis assuming 4 safeguards- ,

3 actuation and a fast bus transfer on Unit 2 while at.

full power and Unit 1 at cold shutdown was perfomed to support the safety evaluation performed for RMP_#48.

(see SER 87-016),,- The 4160-V. configuration in RMP #48 ;

is identical to that=in this SMP.

l

' The load analysis assumed the following: Normal full L

power 4160 V_ loads for Unit 2,-Unit 2 4160 V safeguards loads..(i.e., two safety injection pumps), and full-ratedt .

loads on all Unit I and Unit 2 480 V buses (safeguards

- and nonsafeguards). This' analysis excluded the Unit 4160 V safeguards loads due to the cold shutdown condition.

Results of the analysis indicate a total' load on 2XO4-L and its related hardware of 35.85 MVA. Since 2X04 is rated for-37.7 MVA with forced ventilation.and having: .

t forced ventilation is a condition of the SMP, the performance of 2XO4 will not be compromised. Using the load analysis, the loads on each of the bus supply breakers (2A52-45 and-2A52-47) can be shown to be well within their 3000 ampere continuous ratings. ~At one

. point in the procedure, 1X04 and 2XO4'will both.be in service and supplied by 2X03. Unit 1 reactor coolant pumps may be needed while this configuration is in place. The maximum rating of the 2X03 transformer (41.8 MVA at 65'C rise) is more than adequate to supply this load. To avoid running the 2X03 -

transformer near its maximum rating, the procedure 50

- .-. ~-- . . - - - - - - - - . - - . - - .

requires.that the gas turbine generator (G05) must be <>

in operation and supplying power to bus H01 to start a ,

reactor coolant pump. Also, 11 a Unit 2-trip and fast .

bus transfer occurs, operations is required to secure ;

the Unit'l reactor coolant pusps. The worst case loading on 2XO3 falls well within the capabilities of the transformer; therefore, its operation will not be I. :

compromised. .

The analysis shows that the worst case loading of a-Unit 2 safeguards actuation, a Unit 2 fast busitransfer, and Unit 1 at cold shutdown will not be detrimental to l

~

the preferred safeguards power _ source. A spurious Unit l' safety injection signal'will not affect the.

load analysis because it already assumes maximum loading on the Unit 1 480 V. safeguards buses, and-because the two Unit I safety injection pumps will not ~

operate automatically. Even with a: fault in the 2XO4 transformer, the backup safeguards power source (i.e., ,

emergency diesel generators) will be available to _;

-supply.the power requirement of one complete set of 't safeguards equipment for Unit 2 and to provide

sufficient power to allow Unit 1-to remain in cold:

shutdown. This worst case scenario is bounded by the i analyzed loss of all AC power to the_ auxiliaries event' - ~l presented in Section 14.1.11 of the FSAR. The reliability of the electrical system will not be ,

compromised because the system will be. operated within' its rated capacity.

l While in the configuration discussed above, we will -

establish an additional unusual bus configuration.

This is done to add test switches for= lockout-relays-1-86/A03 and 1-86/A04.- Adding the switches involves cutting holes:in the switchgear cubicle doors below the lockout relays. Because'of concern over the possibility of vibration caused by the cutting,.

resulting in tripping of protective relays mounted on ,

the same cubicle doors, a number'of precautions are 3 taken to minimize the consequences of the relay trips.

The following protective relays are mounted on the 1A03 (IA04) cubicle door:

.1-86/A03-(1-86/A04) 1A03 (IA04) bus lockout relay 1-87/A036 (1-87/A04) 1&O3 (1A04) A9 bus differential relay 1-87/A03 (1-87/A04) 1A03 (1A04) BS bus differential relay 1-87/A03C (1-87/A04) 1A03 (1&O4) Cp bus differential-relay o .

The bus differential relays trip on a bus fault. They input to the bus lockout relay which trips all the breakers on the bus.

a e

_i\

'~

To minimize the consequences of a bus lockout, we.

,_- = eliminate all the necessary and safety-related loads W from 1A03 and 1&O4 buses during cutting of the holes. ]

This 12- accomplished by supplying the 1&O5,1803 and 1

'1B01 (IA06,1B04 and 1802) buses from the G01 (G02) . I emergency diesel generator. Prior to running either '1 diesel generator as an island to supply these. buses, 'I y' we assure that the other diesel is available by. l

" performing a full operability test,-TS-02 (TS-01). .

~

I Also, after running a diesel as an island, we again ');

perform a full operability test. Prior to running either diesel generator as an island, we designate J the opposite train safety injection pump for mid-loop J operation. Also, prior to running a diesel generator 1 as an island, we determine that no other required ,;

equipment is or will be made inoperable as defined by. t Technical Specifications.

Summary of Safety Evaluation: An evaluation is . *

o. required because work performed via the SMP could h alter the function of a system structure or component .

L as described in the FSAR and/or could result in a i potential change to the facility or its operation'as:

' described in the FSAR.

Since existing plant hardware will be used to carry .

'less than rated loads, and the configuration described i' in this SMP will not affect the Technical Specifications or the inputs to the FSAR analyses, this SMP does not involve a significant increase in the probability or_ ,

consequences of an accident previously evaluated. No significant hardware changes are associated with this o SMP; therefore, it does not create the possibility of -

H a new or different kind of accident from any_ accident previously evaluated. Finally,'this SMP does not involve a significant reduction in a margin of safety ;

because the' loads on the electrical system will not exceed the rated capacity.for which it-is designed. l This SMP will therefore not involve a significant L hazards consideration. i

b. SMP 978: Station Battery DOS Changeout, Design Package C of MR 88-074 is to change out station battery DOS. The. interim battery that was installed l

by Design Package B of the same MR will be used to L_ provide DC bus D01 backup power during the changeout '

i process. Swing battery charger D09 will be disconnected from D01 and its tie breaker used for connecting the interim battery to the bus.

Summary of Safety Evaluation: An evaluation is required because the MR, as implemented by the special maintenance procedure, will alter a system, structure or component as described in the FSAR.

52

TS 15.3.7.A.1 requires all four batteries.to be operable to make one or both reactors critical.

TS 15.3.7.B.1.f allows one or both reactors to generate power with either DOS or D06_ inoperable for a period not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided their associated chargers are operable;and supplying power to the.

loads. Since DOS changeout activities will require more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the interim battery installed in.

- the auxiliary feedwater pump room will be used as the bus backup power supply in order to not exceed these

, Technical Specification requirements. The interin' battery in the auxiliary feedwater pump room is qualified as Class 1E and Seismic Class 1.

Disconnecting DO5 from bus D01 could cause perturba-tions on the bus which could cause a spurious reactor trip. To avoid such a spurious reactor trip,.the interim battery will be paralleled with the existing DOS battery before DOS is disconnected from the bus.:

Swing battery charger D09 is provided to allow-disconnection of either D07 or D08 battery chargers from their associated buses for maintenance purposes.

It is not anticipated _that D09 will be_needed to allow D07 maintenance while-DOS is being changed out, so swing charger D09 will be-disconnected from its D01 tie breaker and the tie breaker (D01-4) will-be_used for connecting the interim battery to D01 before disconnecting battery DOS. If D07 is lost, the battery will supply bus pswer for -a sufficient amount of time to reestablish power from the charger.

Breaker D01-4 is rated at 600 amps, one-half of the 1200 amp rating of the main feeder breaker D01-1 for battery DOS. By taking the Unit 1 turbine bearing emergency lube oil pump (1P37D), the main generator air side seal oil pump (1P598), and the main feedwater pumps' lube oil pumps (1P73C&D) out of-service, the maximum load that would be seen on bus D01 is reduced to allow the use of breaker D01-4 as the main supply breaker without it tripping due to the maximum load (reference Calculation N-89-013). These pumps are essential for safe operation of Unit 1 balance of plant equipment so changeout should be done while Unit 1 is shut dan.

Breaker D01-1 will presently alam in the control room when it is opened. During changeout, D01-1 will be open so its alarm input will be defeated. (Note that the alarm is shared between DOS and D105). D01-4 will have to be verified closed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to ensure TS 15.3.7.B.1 is not violated.

In order to minimize the chances for a spurious reactor trip on Unit 2, the precautions that are taken for-shifting the red instrument supply bus feeder, as 53 A 1 -. Q j ,

delineated in 0I-37, will also be taken for the l battery paralleling process. (Note that the red. '

instrument bus is fed by bus D01.) In addition to these precautions, the Unit 2 A train. reactor trip l bypass breaker will be racked in and closed and the. -J normal Unit'2 "A" train reactor trip breaker opened.

~

This bypass breaker-is fed by B train. -Technical ;

Specifications require operability verificatir,n of the !

undervoltage trip function on the bypass breaker -

before use (reference TS Table 15.4.1-2). This ;

verification will be made before using the A train- i reactor trip bypass breakers. ]

The configuration of having the A train reactor. trip I breaker open and the bypass breaker closed will occur for a short duration, which.is allowed under Section 7.2.1 of the FSAR for maintenance and testing activities.

In the event that bus D01 is inadvertently lost during battery-to-bus connections and/or disconnections, adequate protection for Unit 2 will be provided via

'i

, "B" train.- It is likely that Unit 2 would trip in '

l-this event due to a steam generator low level'and/or ~

steam flow / feed flow mismatch since the automatic control circuitry for steam generator level is powered.

by A train (D01). Such an event has been previously analyzed and poses no unreviewed safety question, as evidenced by the existance of train redundancy and y allowable use of the bypass breakers. (It was

~

- subsequently decided to leave-the' reactor trip breaker-alignment per the normal at power arrangement.)

The new battery to be_ installed as the new DOS is made '

l: of Exide Style 2GV-23 cells, which are rated with a '

capacity equal to the existing C&D battery. The new Exide battery and racks will be Seismic Class 1 and nuclear' grade Class'1E.. The existing DOS battery room will be used and is' adequately sized, ventilated,-

temperature controlled, and protected from fire. The existing cables between the bus and battery are also

adequately sized so they will be used for this connection.

c. SMP 1000: CVE Relay Replacement; Unit 1 (MR 87-165) -

SMP 1000 controls'the replacement of 1A01 to 1A03 and 1A02 to 1A04 fast bus transfer synchronizing relays l (Westinghouse Type CVE replaced with GE Type.SLJ relays). The new relays provide an adjustable drop-out voltage setting which will allow the relay to remain picked up (and fast bus! transfer capability.

available) during an undervoltage condition. The relays will' drop out on a severe undervoltage or no-voltage condition (such as in the case of a fault on one of the supervised buses) or if synchronism between the supervised buses is lost.

~

i .

L 54 K

+ _ . _ . - . . . .

aj - -3 Sussnary of Safety Evaluation: An evaluation is: 1 required because the change constitutes a change to a system, structure or component descr hed in the FSAR. j y

SMP 1000, if worked with Unit I at power, will require j temporarily' disabling the fast bus-transfer capability ?

of 1A01 and 1&O2 and bus undervoltage protection for-

!' 1A03 and 1A04, one bus at a time. The loss'of the. . '

l fast bus transfer capability for either 1A01 or 1A02--

is acceptable because no safety-related loads are ,

supplied from these buses. .The duration of the loss 1

of this capability is expected to be less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

per bus. 'The loss of the undervoltage protection for '

either 1A03 or IA04 is acceptable because undervoltage' ;

protection is available on the 1A05 and 1A06 buses ' .

supplied from 1A03 and 1A04, respectively. -Bus 1A04 also supplies a lighting transformer for the south-gatehouse which is not a voltage sensitive or safety-related load. The duration of the loss of undervoltage protection on 1A03 and 1&o4 is expected to be less b than 5 minutes per bus, and therefore, exposure to a-i: possible undervoltage condition is minimal.

r d. SMP 1002: Relay 2-273/B03, 480 V Loss of Voltage Relay-Inspection and Repair:

SMP 1002 will remove 480 V undervoltage relay . i 2-273/B03 from service ~for inspection and repair.

Summar[of' Safety?. valuation: An evaluation is required because a system, structure or component I: described-in the FSAR will be altered.

Removal of 480 V undervoltage relay 2-273/B03 from service while Unit 2 is at power requires that relay L 2-273X/B03 be energized to meet the minimum degree of I; redundancy specified in Technical Specification Table 15.3.5-3 Section 4.c for loss of voltage (480 V).

This places Unit 2 in-a more conservative condition where an actuation of I of 2 additional 480 V.

undervoltage relays will result.in an undervoltage condition on 2B03. A single failure of one of the -

remaining undervoltage relays could result in an isolation of 2803,-but this single failure is compensated for by redundancy provided by equipment

~

4 supplied by bus 2B04 and diesel generator'G02.

L.

P The controlling special maintenance procedure requires .

that equipment on 2B04 and diesel gene.rator G02 be )

operable and procedurally controls the removal, -1 inspection and repair, and return to service of relay 2-273/B03. This also ensures that relay 2-273X/B03 remains energized to a et the minimum degree of redundancy.

55 l

e' 1~

1

m . _ _ . _ _ _ . - .___ _ _ _ _ . _ . _ _ _ . __ _ ,

y i

e '. SMP 1007: Installation & Testing of Mit 60-062, Unit 2

- Modification Request 89-062 will reduce the opening time of the PORVs with nitrogen actuation.. The modification involves reincating the nitrogen regulators for the PORVs, l and replacing the solenoid valves t'hich actuate ;the PORVs.-

The design of the MR was previously evaluated by a prior SER.

Sununary of Safety Evaluation: An evaluation ist l required because the proposed evolution could affect j the function or method of a system, structure or ;

l- component described in the FSAR. I

I This addendum evaluates SMP 1007, which coordinates .1 and directs the installation. Because requirements for PORV operability under different plant conditions 'j can be-confusing, relevant Technical Specification requirements are summarized for reference in Table 1 l

of thc SMP. -The SMP is organized into five separate work sections, to accommodate the work which must be i done at various times and conditions.

l Sections 3.0 and 4.0 cover moving the nitrogea L regulators for the two PORVs. This vork will be done L at power before shutdown of the unit. Only one PORV :

will be-worked at a time. The work will not affect i~

PORV operability for the following reasons: 1) The nitrogen system to the PORVs:is normally isolated at power; 2) The PORV can still be opened with' instrument air (IA); 3) a check valve between the IA and nitrogen portions.of each system will prevent IA from leaking r

,out the open tubing;-and'4) PORV operability as j defined in the Technical Specification 15.D.l.A' Basis '

will not be affected.- ,

Section 5.0 of the SMP will reduce-the preload'of 'het ,

closing spring on one of the PORVs in order to-further L

L reduce the opening time on nitrogen. This vill be. an ,

l- interim measure to make the PORV operable for 2 W v d will be controlled by a temporary' modificah or. ie work will be performed with the reactor shut down a +

RCS temperature above 354'F, a condition where there is no PORV operability requirement defined by~the r Technical Specifications. During spring adjustment, '

l-'

4 the other PORV will be available to control RCS l pressure if needed. The PORV will be tested satisfactorily and declared operable for LTOP before LTOP operation is required. Acceptance of the stroke 3 times will be determined by. the Responsible Engineer.

With the spring preload reduced in this interim configuration, the PORV will be operable-for LTOP but not for normal power operation since the temporary modification.will have the effect of reducing the 56 m

f l--, , - _, _ - _ . . . ._

u; '

f; : 6 l

pressure at which the PORV starts to open. Since RCS pressure is-under the seat of the PORVs, reducing the

spring preload will reduce the RCS pressure against

^

which the PORV can close., The pressure at which this-will start to. occur is conservati"ely estimated to be above 1400 psiig' This~does not c w ee an operational-problem becaure the work will be performed with the block valve red tagged shut, and cautions in the SMP m warn the operator not to clear the tag on the block valve unless RCS pressure is below 1000 psig.

=

Technical Specification 15.2.2 and FSAR Table 4.1-3 both list the PGB'? setpoint as 2335 psig. This refers-to i.he pressure setpoint required on 2/2 pressure channels to open each PORV. Although the PORV in interim condition may actually lift earlier, the.

setpoint referred to above will not be. changed..

.-.!(

A pressure rise in the RCS after the block valve is opened may cause the PORV in the interim condition to lift prematurely. This does not pose a safety concern since the block valve could be reclosed if necessary.

RCS overpressure protection would still be provided by <

the same PORV receiving an open signal at 2335 psig, the other PORV, and the pressurizer safety valves.

This is during the situation when there is a pressure excursion and the block valve'is not shut.

The SMP keeps the block valve shut until RCS pressure is below 1000 psig. In actuality, the plant will be in n eparation for going on LTOP protection when the block valve is opened. Thus, the actual setpoint for lift will be 425 psig and RCS pressure lifting-of the valve.is not a concern. The above two paragraphs address the interim condition that could exist following a reduction in RCS pressure below 1000 psig and the time that LTOP is actuall.y arined.

Section 6.0 of the SMP will replace the solenoid valve.

to the PORV which did not have the spring preload

+ reduced in Section 5.0. This work will not be started until the other PORV is declared operable for LTOP.

When LTOP becomes required, this work may be in progress during the resulting 7-day LCO. Testing will be successfully performed before declaring the PORV-back in service. Acceptance of stroke times will be determined by the responsible engineer. When this section' is cociplete, this PORV will be fully operable for both normal and LTOP conditions.

Section 7.0 of the SMP will restore the spring tension of the PORV placed in interim condition and replace

the SOV. If this work is performed when LTOP is required, a 7-day LCO will be in effect. The 1

57

.-_L__-___-__-_- _ _ _ _ _ _ . _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ . _ _ _ _ _ _

_ _____ _ __ __ _________________l

w procedural' temporary modification will be closed'out, !

and the PORV will'be stroke tested. Acceptance of stroke times will be determined by the Responsible, l Engineer. When this, section is complete, this PORV l will also be fully operable' for both normal and LTOP '

conditions. *

f. SMP 1009: Relay 2-273X/B03. Reset Wire Replacement, i Unit 2 ,

SMP 1009 will remove 480 V undervoltage auxiliary '

relay 2-273X/B03 from service to replace the reset ,

wires between W relay 2-273/B03 and the auxiliary relay, 2-273X/B03. These wires have shownlto exhibit higher than expected resistance in the reset circuit.

Summary of Safety Evaluation: An evaluation is required because a system, structure or. component described in the-FSAR will be altered.

Removal of 480 V undervoltage auxiliary relay-p 2-273X/B03 from service while Unit 2 is at power o requires that its output contacts used in the ..

. respective 2/3 W matrices are made up-(by installing:

temporary jumpers across the contacts) to meet t'v:

l minimum degree of redundancy specified in Technic.41 Specification Table 15.3.5-3 Section 4.c for loss of 3 voltage (480 V). This places Unit 2 in'a more '

conservative condition where an actuation of 1 of 2 additional 480 V:undervoltage relays will result ,

L in an undervoltage condition on 2803. A single failure of one of'the remaining undervaltage relays L

L could result'in an isolation'of 2B03,'but this single-failure is compensated for by redundancy provided by L equipment supplied by bus 2804 and diesel. generator L G02.

U .

l

.The controlling special' maintenance procedure requires that equipment on.2B04.and diesel generator.G02 be operable and procedurally controls the replacement of the reset wires, removal of W auxiliary relay 2-273X/B03 from service, and jumpering of the relay p , output contacts used in the W matrices. to meet the '

minimum degree af redundancy.

g. SMP 1020: 13.8 kV Field Modifications, MR 87-002*H,

[', Unit 2 The 4160 V power supply conff y ration during certain-phases of the.U2R15 refueling and maintenance outage will be as described in the special maintenance procedure. This configuration has been evaluated (with and without a fast bus transfer on Unit 1) 58 7;

h ,,

t s

for its effect on the FSAR, the Technical ;

I Specifications, and the NRC commitments as found in.

the safety evaluation reports. j Summary of Safety Evaluation: An evaluation is- ,

required because the proposed evolution involves a >

potential change to the' facility o- its operation as ,

described in the FSAR. The activines perfomed via SMP 1020 constitute a p tential, change to the facility' .;

of its operation as described in the FSAR.

The provisions of SER 88-076, prepared and approved.

.for SMP 927 (hinor), 2XO4 Outage for 13.8 KV Modifi-cations, MR 87-002, Revision 0, October 5, 1988, apply '

in their entirety to this special-maintenance procedure' .

with the following exceptions .

l Information associated with the' system-configuration.necessary to enable reactor '

coolant pump operation should be deleted. Thus, the paragraph dealing with' IXO4 is revised, to ,

h read as follows:

"Results of the analysis indicate a total load on IX04 and its related hardware of 35.7 MVA.

Since 1X04 is rated for 37.7 MVA with forced- ;

ventilation and having forced ventilation is a condition of the SMP, the perfomance of 1XO4 u will not be compromised. Using the load-analysis, the loads on each of the bus supply.

breakers (1A52-36 and.1152-56) can be shown to be well within their 3000 ampere continuous ratings."

In addition to the above the entire following paragraph, which discusses failure of G05 while a Unit 2 RCP is running should be deleted. ,

Also, to address the concerns of the possibility that

- a Unit 2 safety = injection pump could be running during.

p' a design basis event, the load analysis needs to

( consider this load addition (maximum addition of 0.725 MVA). This could result in a 1XO4 load of 36.42 MVA, which is still below the rating of- t 37.7 MVA. Therefore, even with.a Unit 2 safety injection pump running, the IXO4 transformer will.

e operate within its rating.

h. SMP'1028: Temporary Installation and Testing of

!- Hydrogen Recombiner The Rockwell hydrogen recombiner will be install on- -;

'El. 46' of the primary auxiliary building near the l 4'z4' equipment hatches just east of the new fuel storage area.- .J 59 w

51_:

YN'

l b.

c Summary of Safety Evaluation: An evaluation'is required because the proposed evolution constitutes a, ',j g potential change to the facility or:its operation as' i described in the FSAR. The recombiner will be I

J mechanically connected to the modified post-accident.

containment ventilation system piping. The connec - I tions for the piping will be made using flexible j stainless steel hoses. The hoses are designed for;a ,

minimum 100 psig and 150'F. . The recombiner is- l designed to operate at 30 psia at temperatures up 5' to 1450'F with an inlet and outlet temperature of 150'F and has a test pressure of 75 psig at ambient. :

The electrical connections are to a special connection ,

box which is powered from 1831.

{

" After the hydrogen recombiner has been placed in service, it will extend the containment boundary and' provide a potential accident containment atmosphere i~:

recirculation path. The extension of the contairment boundary is designed so it exceeds the' temperature and >

pressure . requirements of the contaiment during a ,

cold / refueling shutdown so the probability of occurrence or the consequences of an accident of .

malfunction of equipment important to safety is not increased. -Test cennections are provided in the L existing PACVS so leak testing can be performed prior to connecting the equipment into the containment boundary.

The location of the hydrogen recombiner will not affect the fire protection system at'PBNP. The placement of the hydrogen recombiner, the power and L, control cables, and the removal of PAB El. 46' floor equipment access-hatch were discussed with responsible personnel.

The hydrogen recombiner has'previously been seismically $

qualified but because it will be attached to the non-seistic portion of the PACVS, its qualified status l- will not be taken credit for. It is considered

! non-seismic and therefore, the two manual containment

isolation valves would be closed to satisfy containment integrity. -l During this entire test it is anticipated that Unit 'l >

L- will be on line at 100% power. Conduct of this test will not affect the operating unit.

i. SHP 1036: D105 Battery Performance Test The new special maintenance procedures provide guidance for performing a constant current capacity test of D105 and D106 per the requirements of IEEE Standard 450-1987, "IEEE Recommended Practice for Maintenance, Testing and Replacement of Large Lead 60 p

]

y )

Storage Batteries for Generating Stations and. l Substations." This test is also required per -!

TS 15.4.6.B.4.

Summary of Safety tvaluation: An evaluation is !

required because the proposed test (s) will alter a j system, structure or component described in the FSAR.

TS 15.3.7.A.1 requires all four batteries to be l operable to make one of both reactors critical. ;

TS 15.3.7.3.1.g allows one'or both reactors to -i generate power with D105 or D106' inoperable for.a period not exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the other three batteries and four battery chargers remain '

operable with one charger carrying the DC loads of each DC main distribution bus. D105 (D106).'will be. ,

removed fron service for the performance of this' test -

but will be returned to service in less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Disconnecting D105 (D106) from bus D03 (D04) could cause perturbations on the bus,~To avoid such j perturbations, the temporary battery will be.

paralleled with the existing D105 (D106) battery before D105 (D106) is disconnected ftis the bus. ,

i p Swing battery charger D109 is provided to allow i disconnection of either D107 or D108 battery chargers'- :

from their associated buses for maintenance purposes. .,

It is not anticipated that.D109 will be needed to

- allow D107 (D108). maintenance while D105 (D106)'is removed from service, so swing charger D109 will~be disconnected from its D03-(D04) bus tie breaker and-tie breaker D03-2 (D04-2)~will'be used for connecting D03 (D04). -The temporary battery will be tied to the

! .- D03 (D04) bus for filtering purposes for D107?(D108).

l L Although.D107 (D108) output voltage ripple may be low enough to supply D03 (D04) without. voltage problems,: S the availability of the: spare battery made an .i attractive conservative option in maintaining a stable .

! D03 (D04) voltage. Also, breaker D03-2 (D04-2) is l' rated the same es-the existing breaker at D03 (D04) for battery D105 (D106), which will be disconnected.

No credit will be~taken for the temporary battery.

(72-hour LCO in effect) as being-available to supply bus loads-in the event of a loss of D107 (D108).

Although the temporary battery installation meets all-QA and seismic' requirements, the cable connecting the" ,

temporary battery to D03 (D04) will not be installed J in accordance with the seismic criteria and credit for .

this installation will not be taken.

The routing of cable will follow existing conduit and cable tray runs except in the areas by security MUX #4 -

to the battery rooms where the cable will be run without a raceway and from the temporary battery to 61 c ,

P i~ i ___E_s.___1__t___

4' ,

gni .

[', tray'2AQO3, where the cables will be run above'the j l ceiling of the auxiliary feedwater pump tunnel.- The ;

L cables have been routed to avoid interaction with-safeguards cables. The routing does not provide any i

I ;4 h combustible pathways between safeguards trains.- Thez L cable-trays selected for the routing do not contain ,

any safety-related' cables. All fire zone penetrations:

will be handled per PBNP 3.4.11, " Penetrating Barriers,"

l}

and all penetrations will be tesporarily' sealed.

Cables will be run through the doors of the white (yellow) inverter room to D03 (D04) and appropriate - +

fire watch measures will be initiated to account for-the open door. -

The D109 battery charger circuit breaker D03-2 (D04-2) '

provides the connection point of the temporary battery to the D03 (D04) bus and provides a seismic boundary ;

between the bus and the battery. The interconnecting >

cable at this point will be run along the-floor and a would have minimal movement duriug a seismic event.

1.

The cable that will be installed will be.QA and will meet ~IEEE standards for current capacity.and fire

ratings. Fault currents available along the temporary- '

cable will not be any greater than previously analyzed I

for the setup during D05/D06 replacement. The actual current may be less, depending on the fault location,- ,

due to increased cable lengths (more resistance). The cable will be connected to the battery for a period of less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for each test and the probability .

- of a fault during this period is assumed to be minimal. ,

i

. . SMP 1037: D106 Battery Performance' Test The new special maintenance procedures provide-guidance for performing a constant current capacity test of D105 and D106 per the requirements of 'IEEE Stantrd 450-1987, "IEEE Recommended Practice'for

~

Maintenance, Testing md Replacement of Large Lead ,

Storage Batteries for Generating Stations and Substations." This test'is also required per TS 15.4.6.B.4. .

fl

! Summary of Safety Evaluation: An evaluation is required because the proposed test (s) will alter a system, structure or component described in the-FSAR.

TS 15.3.7.A.1 requires'all four batteries to be operable to make one of both reactors critical.

TS 15.3.7.B.I.9 allows one or both reactors to generate power with D105 or,D106 inoperable for a period not exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the other 1 three batteries and four battery chargers remain

< operable with one charger carrying the DC loads of 62

h w - - - - - _ * - - _ - - _ - - - - - - _ _" _ _ - _ _ _--- --a-___

N( -^x ;

each DC main' distribution bus. .DIOS (D106) will be

, removed from service for the performance of this test but will be returned to service in less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . :

&1sconnecting D105.(D106) from bus D03 (D04) could i cause perturbations on the bus. To avoid such perturbations, the temporary battery will be paralleled with the existing D105 (D106) battery ]

before.D105 (D106) is disconnected from the bus. ]

1 1 Swing battery charger 0109 is provided to allow 1 disconnection of either D107 or D108 battery chargers I from~their associated buses for maintenance purposes. 1 It is not anticipated that D109 will be needed to J

~

allow D107 (D108) maintenanceLwhile D105 (D106) is removed from service, so swing charger D109' will be '

disconnected from its D03 (D04). bus tie breaker and-

. tie breaker D03-2 (D04-2) will be'used for connecting D03 (D04). The temporary battery will be tied to the '

D03 (D04). bus for filtering purposes for D107 (D108).

Although D107 (D108) output. voltage ripple may be' low enough to supply D03 (D04) without voltage problems, the availability of.the spare battery made an attractive conservative option in maintaining a _ stable .

D03 (D04) voltage.', Also, breaker D03-2 (D04-2):is 't L ' rated the same as the existing breaker at D03 (D04)'

for battery D105 (D106), which will be disconnected.

No credit will be taken for the temporary battery c (72-hour LCO in effect) as being available to supply bus. loads in the event of a loss of D107 (D108).

Although the temporary battery installation meets all?

QA and seismic requirements, the cable connecting the temporary battery to D031(D04) will not'be' installed' s in accordance with the seismic criteria and credit for.

this installation will not be:taken.

The' routing of cable will follow existing conduit and cable tray runs except in the areas by security MUX #4-to the battery rooms where thu cable will be run without a raceway and from the temporary battery to- "

tray 2AQO3, where the cables will be run above the=

ceiling of the auxiliary feedwater pump tunnel. The cables have been routed to avoid-interaction with ,

safeguards cables. The routing does not provide any. ,

combustible pathways between safeguards trains. The-cable trays selected for the routing do not contain

any safety-related cables. All fire zone penetrations .

will be handled per PBNP 3.4.11, " Penetrating, Barriers,"

and all penetrations will'be temporarily sealed.

Cables will be run through the doors of the white L, (yellow) inverter room to D03 (D04) and appropriate L fire watch measures will be initiated to account for the open door.

jfg 63 ya L.

k. , ' ?}__ - . :- .

1 i

The D109 battery charger circuit breakur D03-1 (D04-2)

~

provides the connection point of the temporary battery J to the D03 (D04) bus and provider a seismic boundary- i between the bus and the battery. The interconnecting > J cable at this point will be run along the floor and I' would have minimal movement during a seismic event.

The cable that will be' installed will be QA and will-1 meet IEEE standards for current capacity and fire l ratings. Fault currents available along the temporary '

cable will not be any greater than previausly analyzed- ,

for the. setup during D05/D06. replacement. .The-actual. )

current may be less,. depending on the fault location, due to increased cable lengths (more resistance). The

,; - cable will be connected to the battery for a period or less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for each test and the probability of a fault during this period-is assumed to be minimal.- .

k. SMP 1038: Post-Accident Containment Atmosphere Sampling System, MR 89-132 (Unit 2)

MR 89-132 (Unit 2) will replace the containment.

atmosphere post-accident sampling system sample pump drain trap glass bowl with a new trap designed to operate-at a higher pressure-(>20 psig).- The purpose. .

of this special maintenance' procedure is to perform functional / acceptance testing for the installation of .e a new drain trap and the' installation of a; check valve in the bypass line of P707A.

Summary of Safety Evaluation: An evaluation is required because the proposed change could affect the :

I function or method of a system, structure or component 1 described in the FSAR. The SMP tests the containment atmosphere sampling system at design pressures. The-i system will be set up in the sample septum mode.. This-L mode isolates detectors RE-211/212 from the sample stream. The system will be operated at-15 psig and at 20 psig (for 15 minutes only) to verify operation at ;

the high pressures. .

Special caution is used to prevent pressurization of RE-211/212. The detectors (RE-211 and.212) have a sample pressure range of 10" of Hg vacuum to 5 psig.

The detectors can be damaged at sample pressures

p. beyond this range.

l Technical Specifications have a LCO of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months when one of the leak detection systems is out of service.

The testing should take the system out of service for about 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . In the unlikely event that RE-211 or RE-212 is damaged,'the damaged detectcr can be

replaced in about 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . At the end of the testing HPIP 7.51.6 is performed to verify that the detectors have not been damaged. Unit 2 testing will be

+ 64 o

s

, e s i

- - - .- -. - .-. .~. - _ .

3 %

2' performed during conditions that do not require the i

RE-211/212 system (cold shutdown with no refueling

operations). Unit 1 testing will only consist of a >

'u

-low pressure tube fitting leakage check controlled by- :

maintenance work request, based upon an acceptable functional test of Unit.2 and identical system design. l 3.2.7- Setpoint Document (STPTs) l a .' STPT 1.1 (Unit 1): The proposed change involves ,

raising the source range high flux trip setpoint to R 5x105 cps to avoid a situation of relying upon operator response time to prevent a reactor trip. '

Summary of Safety Evaluation: An evaluation is i required because the change constitutes a potential' change to the facility or its operation as described-in the FSAR. The core design for UIC17 is comprised of a low-low leakage loading pattern and the implemen-tation of part-length Hafnium rods to shield vessel welds; As a result of this core design change, the

! intermediate range nuclear instruments and source ,

p range nuclear instruments will be exposed to fewer '

neutrons due to the decreased radial leakage. The intermediate range may be affected more than the source range instruments. Therefore, during a startup the potential exists'that the count rate on the. source ,

range instruments will be-close to the source range high flux trip value (1x10 cps) 5 prior to the P6 block permissive from the intennediate range instruments

being activated (1.5x10 ~10 amps).

The Technical Specification for the source-range high flux trip states that the setpoint shall be within the span of the instrumentation. The Basis for this specification states that'the source range high flux reactor trip prevents a startup accident from sub- ,

critical conditions;from proceeding into the power range and that any setpoint within the range of,the.

. instrument would prevent an excursion from proceeding to the point at which significant thermal power is generated.

l Section 7.4 of the FSAR provides a description of the.

source range high flux trip and P6 permissive functionality end states that the trip setpoint is-between the source range cutoff power level and the maximum source range power level.

Section 14.1 of the FSAR reiterates the information of Section 7.4. A review of the FSAR analysis for uncontrolled RCCA withdrawal.from a subcritical ;

condition indicates that the transient analysis does I not take credit for the source range high flux reactor !

l: trip. The conclusion from this analysis is that the .l B e I

I 65

/

s . l L #~

u.'

i l

i e

DNBR remains above the limit value and that nominal l full' power values for the peak coolant temperature, thermal power and heat flux are not exceeded.

Analysis for uncontrolled RCCA withdrawal from a = l subcritical' condition ~ indicates that the new transient analysis does not take credit for the source range high flux trip; minimum DNBR remains above the limit- :

value; no fuel or clad damage is' predicted; and there is no change to the Technical Specification or Basis for'the source range high flux trip. q Since the range of the source range instrument-is 0-108 cps and the proposed setpoint of.5x105 eps is ,

within the range, the change does-not pose an -

unreviewed safety question.

b. STPT 1.1 Unit 2: Reactor Trips.- 'I o The core design for U2C16 is comprised of a low-low _;

L leakage loading pattern and the implementation of part-length Hafnium' rods to shield vessel _ welds. -As a result of this core design change, the intermediate range nuclear instruments and source range nuclear-instruments will be exposed to fewer neutrons due to the decreased radial leakage. The intermediate range.

may be affected more than the source-range, instruments.

During a startup the potential exists that the count rate on the source range instruments will'be close to-the source range high flux trip value (1x10 5cps)-

prior to the-P6 block permissive from the, intermediate range instruments being activated (1.5x10 10 amps). . ,

Therefore, it is proposed to raise:the source range '

high flux _ trip setpoint to 5x105cps-to avoid a situation of relying upon operator response time:to-prevent a reactor trip. .

l-Summary of Safety Evaluation: An evaluation is required because the proposed change constitutes a potential change to the facility or its operation as described in the FSAR.

b The Technical. Specification for the source range high flux trip states that the setpoint shall be within the span of the instrumentation. The basis for this specification states that the source range high flux 1 reactor trip prevents a startup accident from sub-

" crJtical conditions from proceeding into the power range'and that any setpoint within the range of the instrument would prevent an excursion from proceeding to the point at which significant thermal power.is generated.

, 66

< L.

7 25 , ~ . . , _ . .

, - . . - . ~ - - . . . - - - - . . . . . --

I

> j Section 7.4 of the FSAR provides a description of_the l source range high flux trip and P6 pemissive _ . l functionality and states that the trip setpoint is ;

between the source range cutoff power level and the maximum source range power level.

Section 14.1 of the FSAR reiterates the information of Section 7.4. A review of the FSAR' analysis for ;

uncontrolled RCCA withdrawal frov a subcritical condition indicates thatlthe transient analysis does not take credit for the source range high flux reactor ~

trip. The conclusion from this analysis is'that the i DNBR remains above the limit value and that nominal p full power values for the peak coolant temperature, !

thermal power and heat flux are not exceeded.

{

Analysis for uncontrolled RCCA withdrawal from a 4 suberitical condition indicates that the new transient. '

analysis does not take credit for the source range I l: high flux trip; minimum DNBR remains above the? limit ,

' value; no fuel or clad damage is predicted; and there is no change to the TS or Basis for the source range-high flux trip.

Since the range of the source range' instrument is

0-108 cps and the proposed setpoint of 5x10s cps-is within the range, the change does not pose an unreviewed safety question.

c. STPT I I.2 and 1.3 (Unik 1): .The proposed change will

(~ . lower the red. channel ATo value from 56.l'F to-54.9*F.

l , Based upon channel RCS flow measurements performed by Technical Services in11owing the refueling outage, the.

UIC17 red channel AT is ~55.2*T. This represents a reduction of ~1.2'F from the previous cycle to the present time. With an indicated 53.2*F AT and a ATo of 56.1*F, the red AT setpoints.are not less than the

. Technical Specification requirements. ,

Summary of Safety Evaluation: An evaluation is required because the proposed change could affect the function or method of a system, structure or component described in the FSAR. The present Unit I red channel AT setpoints are based on a ATo that is greater.than '

the actual loop AT, as indicated by the red temperature channel. This results in the overtemperature AT trip *

. being nonconservative in comparison to Technical Specification requirements. The change to the red ATo

l- for STPT 1.2 and 1.3 will result in the overpower.AT-L and overtemperature AT setpoints for the red channel being lower than required by Technical Specifications.

This will ensure that the margin of safety as cafined iL in the Technical Specifications is met.

67 y'

~

'1 n

W ,

l l

.The setpoint change will affect only the Unit I red instruents and will not result in an unreviewed safety question.

E ,

d. STPT 1.2 & STFT 1.3 (Units 1 & 2).- For the over-temperature formula, the TS constant for T' will be L. changed from 574.2'F to-573.9'F for Unit 2 and the TS s

constant K2 will-be changed from 0.0150'to 0.0200 for. r Unit 2. In addition, the f(AI) penalty term will be j E removed from the overpower AT setpoint formula ~for 1

'both units.

Summary of Safety Evaluation: An evaluation is F required because the proposed changes could affect the I i function of a system, structure or component described-in the FSAR. The changes to STPT 1.2 and l'3 are necessary to implement recent changes to Technical Specifications (in effect for Unit 1; to.be effective on November 1, 1989, for Unit 2). The TS revision-changes the AT setpoint formulas.

4 For the overtemperature AT setpoint (ATspl) the changes in the formula were: K2 was increased from 0.0150 to 0.0200 and T8 was reduced from 574.2*F to 573.9'F. For the overpower AT setpoint (ATsp2), the changes in the formula were: Removal of the f(AI) function.and reduction in t" from 574.2'F to 573.9'F.

The change in T' for both ATspl-and ATsp2 results in >

the setpoint values-being reduced at a lower Tavg.

-Therefore, protective action would' occur earlier '

during a transient. The changes to the ATsp1 and-s ATsp2 instruments required to implement:the T' change ,

consist of-adjusting the instruments' bias and gains.

These adjustments are within the capabilities of ^ the instruments and will not require a hardware modifi-cation. Therefore, the T' change will not affect the.

failure mode or the probability of failure of the' ATsp1 and ATsp2 instruments.

The change in K2 for ATsp1 was evaluated under SER 89-056 for MRs 89-045 and 89-082; and the removal of f(AI) from ATsp2 was evaluated under SER 89-110.for MRs 89-029 and 89-083. Additionally, all changes were evaluated in the safety evaluation reports for the -

associated TS change requests.

1 (' } .

68 l

I , n.g am _ ,

i , -g N

i

'E 3.3 Tests or Emperiments Following is a list of tests or experiments performed at Point- ,

j Beach Nuclear Plant during 1989 that required a 10 CFR 50.59 l review. In each case, the safety evaluation determined that the l 0 test or experiment did not pose an unreviewed safety question'. .!

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety was not increased. The change did not create the' possibility for an accident or malfunction which had not been previously evaluated.

The margin of safety as defined in the. Technical Specifications j was not reduced, q i

3.3.1 WMTP 11.54: Functional Test of Fuel Oil Transfer System 1

An evaluation is required because a system, structure or ;

component as described in the FSAR may be altered. The test will consist of a fuel oil transfer pump functional test using a temporary flow meter and a level control system functional test. The. operability of the emergency diesel generators will not be affected. -Fuel oil from the y sump tanks and the day tanks will be available automaticC ly if needed. Fuel oil-from the' emergency fuel oil tank will'

-be available, but may require operator action to restart the pump, or open the day tank supply valve MOV-3930 or l MOV-3931. Since a 3-hour supply will be available from l" ' the sump tank and day tank, this will not present a l-problem.

L The temporary flow meter will be connected between.the.

L, emergency fuel oil tank pumpout connection and the main -i L storage tank fill connection. If this temporary line breaks, fuel oil will be pumped out of the emergency fuel oil tank onto the ground. This can be easily stopped by securing the fuel oil transfer pumps. Since an operator will be watching this area, he can immediately direct the

~

control room to secure the pump.

p 3.3.2 . WMTP 11.55: D106 & D106 Battery room Ventilation System ,

Test An evaluation is required because a system, structure'or o component as described in the FSAR could be altered. The ,

purpose of the test is to determine-the amount of negative pressure which can be developed in the battery room air handling unit enclosures by varying the inlet damper position.

To accomplish this, a_ vacuum gauge will be installed ~on the drain pipe or an air handling unit operating in high speed, and the damper in the supply duct from the turbine hall to the unit will be cycled from full open to full closed. The amount of vacuum developed at each of these positions will be recorded.

g 1_I*2__'_i_____._.____ _ _____ _ -___

_ _ _ . _ _ - - =w

i i,,

. While thisLis taking place, the remaining air handling-unit will be-in the auto mode. If low flow is detected at

+ the discharge of.the unit being tested, the unit in auto will automatically start. This is per the existing ,

control system. At.no time will the cooling capability of--

the units be compromised.- -

The test alignment is the same as normal operating conditions. The drain line from the unit being tested will be temporarily plugged during the. test. This does not pose an operational problem. .

3.3.3 - WMTP 11.56: Investigation of Rad Levels on El. 66' of the Unit 1 Containment During Fuel Movement Through the Fuel-Transfer Penetration An evaluation is required because' the proposed evolution

- will alter a system, structure or component described in the FSAR. WMTP 11.56 was initiated to test radiation '

levels on El. 66' of Unit 1 containment during-fuel.

movement while lead shielding is installed on the' fuel transfer tube.(Penetration C-3), and as an option, on the- t intermediate level (floor immediately kbove penetration).. ,

There is a comunittent to install lead shielding to minimize radiation levels' on El. 66', Lead blankets are !

presently installed during the U1R16 outage on El. 66'-

over the fuel transfer penetration along the liner and >

edge'of the floor. 1 The data collected per WHTP 11.56 will be used to-determine the feasibility of installing permanent lead shielding at the penetration'in lieu of the present practice. The proposed installation of lead blankets on the fuel transfer tube was evaluated and found to be. .

acceptable with loadings up to.1500 pounds. The lead 'l

. shielding will be installed during an outage and will-be removed prior to startup.

WMTP 11.56 has several cautions, notes and Health Physics ,

L signoffs to minimize personnel exposure and to prevent l.

personnel from entering high radiation areas during fuel movement.

h !

N i t E 70

]

3.4. Design Changes There were no plant modifications made during 1989, beyond those '

authorized with license amenenents as noted above, which required !

Nuclear Regulatory Commiss, ion approval.

-Following is a list of modifications made at Point Beach Nuclear- t Plant during 1984 that required a 10 CFR 50.59. review. In each !

+ case, the safety evaluation determined that the design change did not pose an unreviewed safety question. The probability of ,

occurrence orlthe consequences of an accident or talfunction of equipment important to safety was not increased. The change did ,

not create the possibility for an accident or malfunction which :

had not been previously evaluated. The margin of safety av defined' in the Technical Specifications was not reduced.

3.4.1 M-785 and M-785*A, Computer Room and Reactor Engineering Office Modifications.

The modification request is for changes required to the~ +

computer room and engineering offices. to accommodate the cable routing for the new Foxboro Spec 200 racks, new ;

power supplies, the new computer system and other instru-mentation being installed in these areas.

Summary of Safety Evaluation: This modification pertains T

only to installation of-seismic supports for process racks and computer system components; other. portions of ,

the modification are not nuclear safety related.

The seismic racks will be anchored to the control building, a seismic structure. Evaluation and/or analysis will be performed as required to ensure the seismic qualification of the control building is not-degraded.

3.4.2 M-785-02, Fire Protection.

The addendum provides a Halon automatic fire suppression system for the computer / instrument rack room.

A Summary of Safety Evaluation: The modification is a potential change to the facility or its operation as

( described in the FSAR. The FSAR does not address fire p protection requirements for the computer / instrument rack room. The new automatic fire suppression should be- .

? interlocked with the new control room HVAC system. The L;' final design of the fire suppressions system has not been determined.- The new system should be interlocked with the smoke removal system and with the return air damper .

from the computer room. This is to prevent exhausting I the Halon before the fire is suppressed and from returning smoke and Halon to the air handling unit and discharging it to the control room. The return air 71

~ ,

damper'should be capable of being reset from control- .i panel C67. The Halon system should also discharge into the raised floor area to suppress any fire in'the- +

' instrumentation and data cables. ;

' 3.4. 3 '83-021, and 83-021*B&C, Post-Accident Containment

. Ventilation System. ;

The MR designs and installs the electrical and piping interfaces necessary to install a hydrogen recombiner. ,

Summary of Safety Evaluation: The proposed MR constitutes a change to the facility as described in the ,

FSAR. The modification would extend the containment. ,

boundary and provide a potential post-accident _ l containment atmosphere recirculation leak path after the hydrogen recombiner has been placed into service. The= 1 extension of the containment boundary would be designed- +

and fabricated to the same standards as the existing-l PACVS extension so.the probability of occurrence or the~ ;

. consequences of an accident or malfunction ~of equipment ;

important to safety is not increased. Test connections are provided which enable laak testing to be performed prior to connecting the equipment into the containment i boundary.

7 l 3.4.4 - 83-021*D, Temporary Installation and Testing of 1 Hydrogen Recombiner.

3 The hydrogen recombiner will be installed on El. 46' of ;

L the crimary auxiliary building near'the 4'x4' equipment hatches just east of the new fuel storage area,

h. ,

k Sumary of Safety Evaluation: . An evaluation is required because- the proposed evolution constitutes a potential change ~to the facility or its operation as described in

. the FSAR. The recombiner will be mechanically connected-- t to the modified post-accident containment ventilation .

system piping. The connections for the piping will be made using flexible stainless steel hoses. 'The hoses are designed for a minimum 100 psig and 150*F. The recombiner is designed to operate at.30 psia at temperatures up ,

to 1450*F with an inlet and outlet temperature of 150*F '

and has a test pressure of 75 psig at ambient. The

~

electrical connecticas are to a special connection box which is powered from 1831.

L L The installation of the piping system was evaluated in a previous safety evaluation report.

After the hydrogen recombiner has been placed in service, it will extend the containment boundary and provide a potential accident containment atmosphere recirculation path. The extension of the containment boundary is designed so it exceeds the temperature and pressure 72

3 l !

.J t

- requirements of the containment.during a cold / refueling ;

shutdown so the' probability of occurrence or the !

+ consequences of an accident of malfunction of equipment important to safety is not increased. Test _ connections. d are provided in the existing PACVS so leak testing can be '

performed prior to connecting the equipment into the jt containment boundary.

The location of the hydrogen recombiner will not affect [

the fire protection system at PBNP. The placement of the ';'

hydrogen recombiner, the power and control cables,.and the removal of PAB El. 46' floor equipment access hatch were 4 Ldiscussed with responsible NSEAS personnel. Procedure .

PBNP 3.4.8, " Transient Combustible Control," was reviewed.

to verify that the hydrogen recombiner will not be located ,

in a safe shutdown area. 7 t

The hydrogen recombiner has previously been seismically '

qualified but because it will be attached to the non-seismic portion of the PACVS, its qualified status will not be taken credit for. It is considered non-seismic and 7 therefore, the two manual containment isolation valves would be closed to satisfy containment integrity. '

During this entire test it is anticipated that Unit I will be on line at 100% power. Conduct of this test will not affect the operating unit.

3.4.5 84-099, Fire Protection. 7 The modification will change the fire detection and Halon a system alarm from a series configuration to a parallel-configuration to provide fire detection capability when' e.he Halon system is not available or'is out of service.

3 Summary of Safety Evaluation: The modification will increase the margin of safety.of the fire detection system.

3.4.6 84-136 and 84-136-01, Fuel Handling System Addition-of a 3 pent Fuel Pit Holst Load Monitor.

E The addendum involves the incorporation of several minor changes to the spent fuel pit. bridge. The major item of this addendum is to festoon power to the auxiliary hoist on the bridge. This hoist is not part of the original <

equipment but was added later. This two-ton coffer hoist is used to move and handle miscellaneous equipment but is not used for the moving, positioning or lifting of new or spent fuel. The auxiliary hoist is not used when fuel is being handled with the main hoist.

Summary of Safety Evaluation: The proposed HR would result in a potential change to the facility or its operation as described in the FSAR. The hoist weighs 73

- - ~ . -

i

.g ,

I .

,, 333 pounds and based on engineering judgment, does not - ,:

affect' the structural capabilities of the spentLfuel pit !

6 . bridge. The entire bridge is. rated for 4000 pounds so the live loads, including the auxiliary hoist, will not~

_ exceed the limits of the bridge and track. The trolley r >

l configuration of the auxiliary hoist-is compatible with the track system that the bridge.uses.

c The spent fuel pit bridge and hoist lwere seismically- ;

4 designed to prevent falling into the spent fuel pit. $

Although the auxiliary hoist was not specifically ;

"~

seismically qualified, the trolley arrangement is'the same

~

. as the existing seismically-qualified Harnischfeger hoist.

Thus, it can be concluded that the auxiliary hoist will not fall during a seismic event. : Note the ' auxiliary ^ ,

hoist is not normally stored in a position over the spent '

fuel. The power.that is being festooned comes from the existing bridge power supply. The installation of the ,

festooning will be controlled by contractor procedures . -4 during the addition of the load indication system on the spent fuel pit hoist.

The spent fuel pit bridge is addressed in' Section' 9.5.2 s of the FSAR, but the addition of this' hoist and the festoon of the power to it does not change any conclusion contained therein. ,

3.4.7 84-254*A&C, Plant Process Computers.

The modification will replace the existing computer system with a new system (including multiplexer hardware, host computers, printers, CRTs, and support peripherals) in order to' meet NUREG-0696 requirements.

Summary of Safety Evaluation: This computer complex will perform the function of the safety assessment system L (SAS) as required in NUREG-0737 and'the plant process l computer system (PPCS).- Redundant-displays will be available in the TSC, the SBCC, and the control room. A seismically qualified front-end data acquisition system ~ <

will also be included to provide a tie-in to Class 1E systems.

t The additional electrical load imposed on the white and yellow instrument buses due to this modification was previously analyzed and factored into the bus upgrade design for Modification Requests E-206/E-207. Automatic-switchover between corresponding Unit 1.and Unit 2 instrument bus power sources associated with each component will be available as part of_the data i ..

acquisition system. The equipment in the host computer complex will have power distribution such that loss of an instrument bus will result in the lss of only half of the

!; redundant hardware with functionality maintained for both L units.

L 74 L

L >

[ , 11 0

-)

, a s

D.

The SAS/PPCS complex will obtain radiation data from the RMS control terminals-(CT) via a pair of redundant data- )

3 links. The links will use RMS CT ports independent of i those used by the existing RMS smart terminals. 'The RMS l data vill be available for use by SAS and within normal i PPCS monitoring.j !

Several seismic requirements have been incorporated into -

the computer system designt j

a. The additional floor loading due to the computer and related equipment-added per Modification Request 84-254 is being cosidered and'the resulting configuration will-not degrade the seismic acceptability of the structure,

b. The data acquisition components, four mulfiplexers-and the. fiber optic modem, have been seismically qualified as they represent the interface betwen the .j computer and Class IE systems. J

c. The multiplexers located in the computer room will be l seismically supported so.there is no potential for t

missile impacts on the Spec 200 equipment.

d. A seismically-qualified system with redundant readout capability will be available to display certain data-directly from the data acquisition system. -The -

readout devices will be mounted in the auxiliary safety instrument panels (ASIP) in the control room.

p The additional heat load imposed upon the control L building heating, ventilating, and air conditioning l system will be assessed separately and HVAC system changes will'be incorporated under Modification Request 85-301.

The new PPCS (including data acquisition) will equal or ;

1 exceed the existing computer's accuracy and the margin of

? safety will not be reduced. The.new system will increase monitoring of normal and critical safety functions and provide displays of normal parameters, trends, accident-conditions,-and present an analysis of accident proba-bilities. The' critical determinations of core subcooling and reactor vessel water -level will be calculated by SAS/PPCS in parallel with the spec 200 instrumentation.

The design of SAS/PPCS is consistent with present operating philosophy of providing supplementary information and does not change any accident analysis 4 presently considered in the FSAR.

75

~'

d

, 3.4.8 - '85-199,7 Instructions for Removal of Obsolete Atcor I Equipment, i 1

x Summary 'of Safety Evaluation - This modification ~will~not jeopardize the safe operation of this plant, nor will it l involve.a potential radioactive release. The equipment l to be removed has been cleaned enough to allow easy

)

proe m ing as low-level waste and avoid any possibility 1 of a release. . These changes will not require any change to plant Technical Specifications. ;

4 3.4.9 85-213, 85-213*B&C-(Unit 1) and 85-214, 85-214*B&C (Unit 2), Anticipated Transients Without Scram Mitigating System Actuation circuit.

AMSAC is designed to trip the main turbine and start the motor-driven and steam turbine-driven auxiliary feedwater.

pumps when a loss of mein feedwater flow is detected.

I Summary of Safety Evaluation: An evaluation is required because the proposed MRs will alter a system, structure or l component or the function of a system, structure or component described in the FSAR. AMSAC is required by 10 CFR 50.62. AMSAC is not safety-related.

Loss of main feedwater flow is detected by sensing that both main feedwater pump motor circuit breakers are open or'that both main feedwater control valves are closed.

AMSAc is designed to perform its function when the unit is above 40% power. The output signals are delayed internally by 30 seconds to allow the reactor trip circuits to actuate prior to AMSAC.

3.4.10- 85-243, Instrument Air System.

The overall effects of the modification are to split the Units 1 and 2 instrument air and to increase the' system capacity.

l l Summary of Safety Evaluation: The mechanical work l constitutes changes to the facility as described in the l FSAR. When the modifications are complete, the system !

I will be more reliable due to the flexibility available-for isolating and cross-tying the headers. The installation of the piping and connections to tie-ins L installed during the fall 1986 outage may be done without special procedures. Others tie-ins to the existing l- system will be controlled and monitored by special maintenance procedures. Tie-ins to auxiliary feed pumps 2P28 and P38B will be done during LCOs. The tie-ins to L the existing system and the modifications to the existing l piping will not affect the safe operation of the plant.

L The electrical design package will result in a potential change to the facility or its operation as described in the FSAR. Each replacement compressor will increase the 76

a p .

, a load on its power supply by approximatelyl25 amps.- The adequacy of the compressor power supplies will be

. verified prior to cosqpressor installation.- The- '

compressors are locked out following an undervoltage trip so they will not be an initial load upon the emergency diesel generators. The compressors can be restarted manually if the operator determines that.the diesel generator can supply the additional load.- .,

3.4.11 86-004 (Unit 1) & 86-005 (Unit 2), Reactor Coolant System.

The MRs will replace the existing rod position indication ~ l (RPI) and control rod drive mechanism-(CRDM) pigtail -!*

connectors and field cabling with stainless steel pigtail connectors; stainless steel sheathed intermediate cables (called HLR cables); two stainless steel patch panels,-

and new. field cabling with pre-installed stainless steel connectors.

Summary of Safety Evaluation: An evaluation is required because the MRs constitute a potential change to the ,

facility or its operation as described in the FSAR. The i

-new installation will decrease the containment aluminum inventory by removing the 66 pairs of CRDM and RPI "

connectors. The new-installation will also result in an insignificant increase in the free volume of containment due to the-removal of excess-cable, which is presently coiled up in cable trays. The new patch panels are designed to remain intact during a safe-shutdown earthquake, and will not impact any safety-related-

l- equipment. The combined weight of the new HLR cabling and I

the patch panels is approximately 85 poundo per Combustion Engineering calculation. Any additional weight.will not affect the RV head nor the head lift rig since the change in weight is not-significant (by engineering judgment).

I-The replacement connector system reduces the probability of an RPI or CRDM failure due to a superior connector mating system and.use of inorganic materials for the HLR l- cabling. ~ A failure of the new equipment carries the same consequences as a failure of the existing connectors and ,

cabling. Strict administrative controls will assure-proper cable / core location relationships are maintained.

3.4.12 86-051, Fire Protection System.

The modification will install a wet pipe automatic sprinkler system in both levels of the fuel oil pumphouse supplied by a valved connection to the three inch lead-in supplying the foam' deluge systems.

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR.

Manually-operated bypass valves and piping were installed around the fuel oil transfer pumps in order to ensure oil ,

77

. z- ~

I 5 a transfer capability to the diesels if the fuel oil transfer pump power supply was unavailable. This~was done to ,

satisfy Appendix R. Under Appendix R, mechanical compo- 1 nents are normally determined to withstand the' effects of fire by convective cooling effect of the contained fluid

-and the low combustible: loading within the area.

,. 1 L The conditions stated above are normally true in the fuel l ' oil pumphouse. However, if a fuel oil leak occurred with a fire, the combustible loading would be very high. Although the pumphouse has automatic fire detection, manual fire extinguishment could be delayed because .of the' remote location. Therefore, the proposed system is recommended to ;

ensure rapid. fire extinguishment and adequate-cooling of. '

the mechanical pipe and valves. l The proposed system will not affect plant operation and: '

should be installed. Because of the remote location and i small size of the fuel oil tank in the pumphouse, and oil i leek or suppression system actuation.coule cause oil to spread into the yard area. Therefore, a sump pump which; discharges inside of the fuel oil storage tank diked area-should also be installed. This will be included within the scope of this modification request.

3.4.13- 86-056 and 86-056*B, Open Circuit Self-CJntained Breathing Apparatus.

This modification will change the existing air ports in C01 i and CO2 from directly-supplied by service air to supplied by dedicated air tanks located in the north service building.

'The requirements are to provide 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of breathing air to 6 control room personnel. This is based on fire protection o considerations established by WE. Industrial Health & Safety.

Summary of Safety Evaluation: An evaluation is' required because the change will alter a prior documented technical commitment to the NRC. ;

An earlier SER evaluated the change of breathing apparatus at Point Beach. In addition, it evaluated-the manifold ;

that would supply dedicated air to the control' room. As a l continuation to that evaluation, the following details can be evaluated.

The control room habitability study and the associated NRC safety evaluation report (dated August 10, 1982) states in regard to fixed emergency breathing air requirements:

l l.

l "The control room is also equipped with provisions for supplied air. There are three outlets on each side of the control room and six full face masks available.

Although the service air compressors are stripped on safety injection signal, they can be normally restored if loading permits."

/

l 78

?-_-_L_--_____________ - - _ - _ , _ . -. . - --

1 t

i I

?

This will remain unchanged except.that MR 86-056 will remove the direct tie-in to service air.~and add-the independent emergency breathing air to. control. The j service air tie-in will be available (via manual- ,

isolation) as.a potential. backup to the emerger.cy. .

breathing air. The* manifold will be used with the new air packs which have been previously discussed in the .

~!

i) '

original evaluation. Two lines.will be routed to thel control' room from the north service buildir.g. > These lines will be separated in the turbine building to assure that- ;

they cannot be-affected by the same fire. :Srparation will' '!

not be applied in the cable spreading room'or auxiliary '!

feedwater pump room'as fire protection is available in -

these areas. l q The electrical' power for the compressor and the associated control panel will come from non-vital sources. An

' annunciator will be provided in the control room to-l: indicate when the emergency breathing air has been ll depleted by 75%.

The manifold poses no significant additional' safety risk to the plant because it has the same characteristics as l the existing service-air line._ These characteristics include.the same pressure rating (a relief valve will protect the piping from overpressurization), and a better material in the st.minless steel versus the carbon steel.- ^

I The line will.be supported seismically in the auxiliary

~

L feedwater pump room, cable spreading room and the control; boards. The specification for seismically supporting L conduit.will.be applied.to this line in these areas as the '

size, weight and material of this.line fits into the

' assumptions that the specification was based upon. The new line will be built to the: requirements of B31.1. The high pressure portion will be located in the north service i1 building away from equipment important to plant' safety.

3.4.14 86-105 (Unit 2), Instrument Buses.

I

[ The modifications will install three banana jacks in each of the Westinghouse instrument bus control panels 1(2)Y01/101, 1(2)YO2/102 in order that the normal and ,

l- alternate instrument bus power sources-are in synchroni-

zation before a manual transfer is made. The modifications do not apply to the remaining plant instrument buses L because these buses incorporate an automatic synchroni-L zation feature.

Summary of Safety Evaluation: The modifications constitute a change to the facility as described in the FSAR. The banana jacks installed in each panel will indicate " normal supply hot," " alternate supply hot," and

" common neutral." A test device will be fabricated which contains a synchrocheck relay and an indicating light.

l 79 l

l l^

7 . ._-_ . . _

. . . . _ _ . _ . , - . .m.,- .m_ . m w

's

}

When the test device is connected, the light will l 4 indicate if the two sources are in synchronization'and.

O the transfer can be made.

The test jacks will be fused to provide isolation between

,P the two incoming sources. In addition, fuses in the test i device will also ensure that isolation occurs when this device is in'use and a fault is present.

3.4.15 87-002-and 87-002*E,G&H, 13.8.kV Switchgear. ;

This modification will replace the existing H01 switchgear [

-with three new switchgear sections.. The physical separa- ,

tion and isolation of the new equipment will greatly .

increase the reliability of the 13.8 kV system and will-also provide separate feeder breakers to:- (1) provide power from the 13.8 kV system to the new north service building; (2) provide alternate shutdown powers (3) pro-vide capability for supplying a future common low voltage n station auxiliary transformer; and (4) provide three spare j breakers for. future load additions. ,

Summary of Safety Evaluation: The proposed modification is a change to the facility as described in the.FSAR.

The only safety analysis that-is applicable to this modification is Section 13.1.11', " Loss of all AC Power to Auxiliaries." The worst event that could occur as e . [

result of this modification would be a loss of all power to both low voltage station auxiliary transformers.. This ,

~

event is covered within the scope of the modification.

,. The new alternate shutdown power feed from the new 13.8 kV l' switchgear is being provided to assure capability to shut.

down the reactors-in the event of certain fires. The 1 Appendix R scenario that results in this requirement explicitly states that we do not.need to postulate any- -

accidents coincident with the fire. Thus, this portion- ;

of the modification does not affect any safety analysis. "

3.4.16 87-002*F, 13.8 kV Electrical System.

Design Package F creates an interim configuration of the 13.8 kV system by reconnecting the supply to transformer i: 2X04 from 2XO4 through new bus section H03, including two l new circuit breakers. The automatic transfer of 2X04 to l' be supplied from IXO3 through existing 13.8 kV breakers H52-02 and H52-03 will be maintained. Minor relocation and changes to protective relaying to support the above will also.be made.

Summary of-Safety Evaluation: An evaluation is required because this MR design package constitutes a change to the facility as described in the TSAR.

80

.u_ _ _ - . _ _ _ _ _ _ _ _ _ _ . - .

e

... +

.The only safety analysis that is applicable is FSAR Section 1.4.11, " Loss of All AC Power to the L Auxiliaries." The worst event'that could occur due to .

failure of.any of.the components added by this design ,

package or which are part of the existing 13.8 kV system ,

would be loss of' power-to both low voltage station :

auxiliary transformers. This event is covered within the scope of safety analysis 14.1.11. '

I^ No changes to the Technical Specifications are required for this design package. However, the installation and .

timing will be governed by the limitations of TS i Section 15.3.7. :

3.4.17 003 and 87-003*A, HVAC. >

This modification installs drain traps and a trap primer on the condensate drains from the control room ventilation 1 system (W13B) and from the cable spreading room tentila-L tion system (W13A).

Work Package "A" Summary of Safety Evaluation: . Work Package "A" is a potential change-to the facility or its j operation as described in the FSAR. Drain piping will be. l the same material and design as the existing equipment-drains. . The trap primer piping will be an extension of the existing potable water piping. Trap primer. piping will be equal to or better than the original Bechtel 1 specification for potable water (piping class JD).

l. The HVAC equipment drains and the potable water piping is- q ~

l- not considered seismic. The new drains and trap primer-r' piping will not affect any siesmic equipment or piping located in the HVAC equipment room.

The drain traps and trap primer piping are extensions of existing systems located in the HVAC equipment room.

There is no change in the flooding effects from the i original installation. Any flooding will overflow to the floor. With the exception of the duct chase, which is protected with a curb, there are no penetrations through.

g the floor into the control room. Any flooding of the .

equipment room floor will run under the north door into the turbine building. .;

The building static pressure gauges are to provide indication of relative building static differential pressures. Differential pressure indication is to verify L proper pressurization of the TSC and the control room during emergency operations. Fire barrier penetrations required for the pressure sensing tubing will be controlled via PBNP 3.4.11. The control room / turbine building penetration will be protected by the existing window fire shutter.

81

l 3.4.18- 87-006 and 87-006*B&C, Main' Steam System.

This modification will replace the existing turbine-supervisory system with'an IRD Mechanalysis system. l Summary of Safety Evaluation: An evaluation is required because the modification will affect equipment described in the FSAR (i.e., the main control board).

s The changes to the main control boards will be the removal'

- of the original NSSS equipment (with the exception of the. ,

valve position / speed drawer) on the rear of C03 and the

addition of a new indication panel on the rear'of C03.

The indication panel will be constructed from 3/16" steel. ;

plate (same as the control board) and will cover the' holes ;

left in the board by the removal of the original' NSSS equipment. This'will reduce the load supported by the rear panel of the control board by 170 pounds. ,

Additionally, the new panel will be mounted ca the >

L external side of the control board and will be larger than 1: any opening into the control board. This will prevent the new panel from falling into the control board. Based on these two factors it is judged that the modifications- to the control board will not affect the seismic characteristics of the control board. ]

3.4.19 87-034*C, 480 V Electrical System. !

i The MR provides for replacement of pneumatic trip devices-with solid-state trip devices on our 480 V NSSS Type DB circuit breakers. -

All power for the new overload protection is provided by the current transformer type sensors, thus making the new protection independent of any outside power source. Since the new protection is solid state versus the existing which is electro-mechanical, there is significantly less l

l probability of system malfunction or erratic operation.

l.

L All units on safety-related buses (B03/B04) are qualified for seismic installation per IEEE 344. Those units

'.I- installed on nonsafety-related buses (B01/B02, although not meeting the requirements of.the IEEE standard, do meet the seismic criteria for PBNP as defined by a SQOG-l report. All installations will be per NSSS instructions and addenda for retrofitted.DB breakers. This will also assure that problems experienced with the direct trip actuator as described by NRC_IE Information Notice 88-054 are corrected. Also, at least for safety-related breakers, full current testing will be performed after the modifi-cation to ensure all settings are correct and the new everload protection functions as expected.

l l

l 82

3.4.20- 87-134 and'87-134*B,1 Fire Protection System. -

The modification replaces the existing pneumatic heat-

.: activated devices-(HADs) on the fire protection deluge ,

systems. The new' '

detectors will be electric, heat-only detectors.

Fummary of Safety Evaluation: . The deluge valves are not i i described in the FSAR. However, the supervisory air to the deluge valves is shown on Detail A of Figure 9.6-1.

The supervisory air to the deluge valves will be removed-as part of this_ modification. The new detection, ,

supervisory and release system are all electric. It should be noted that the electric actuation system is presently used (solenoid) for manual remote deluge valve operation.

The third paragraph of FPER Section 2.4.2 indicates that !

L power for the detection system is from the normal y lighting system. The new detection system for the deluge l system will be powered from 1YO5.

The change has no significant impacts upon instrument air.-

3.4.21 87-151, Addition of a Hanual Isolation Valve in the Drain Line Between the Blowdown Evaporator Building and the Sump-Tank.

This modification proposes to. add a manual isolation valve in the drain line from the blowdown evaporator building to the sump tank. This will be done to help prevent primary auxiliary building flooding in the event.of a Unit 2 facade flood.

Summary of Safety Evaluation: A diaphrap valve was chosen for this application because presently, diaphrap valves are typically used throughout this system. Body and diaphra p material types were selected based on types presently used for this sytem. The temperature and pressure ratings of the valve are acceptable for its l.- intended use.

1

! The valve will be located in the overhead of the Unit 2 L residual heat removal pipeway. A support will be added near the valve to ensure that the addftional weight of the valve will not cause the drain line to fail during a seismic event.

u The valve will be normally open, and will be administra-L tively controlled. It will not affect the functionality l- of the system.

I?

L 83

.O.___.___--J'__K --l--_2- -

.3.4.22 190, Containment Jib Crane.

This modification; installs a jib crane in the containment.

Summary of Safety Evaluation: An evaluation is required- f because a system;. structure or component described in the FSAR was altered. The jib crane was-designed to be used

~

during periods of unit outages only. During normal plant ~

operation, the crane is secured by a seismic restraint to prevent it from swinging around during a seismic event.

If the crane would become unsecured ~ during a seismic. >

event and swing around, it would not be consequence since ,

the crane would not impact or. damage any safety equipment. The crane was designed to not fall-during a. ,

seismic event. ,

3.4.23 87-218 and 87-218*A&B(Unit 1),. Main Control Boards.

< The MR installs limit switches on the Unit 1 pressurizer spray valves and indicating lamps on IC04 for control room position indication. In addition, a solenoid valve is installed in the spray valves' air line and a selector switch is mounted on IC04 to provide _the. control operator-a means to cause a Unit 1 spray valve to fail closed.

Summary of Safety Evaluation: The modification constitutes.

a potential change to the facility or its operation as described in the FSAR. Section 7.7.3 of the FSAR says control stations on the main control boards (MCBs) are ;

packaged in a modular concept and are grouped according to function to minimize operator error. It also' states the l- vertical section of the control boards incorporates t L instrumentation, trend recorders and annunciator panels and that the console ~section contains control devices and indicating lights. These concepts are employed in this !

design except the indicating lamps are located on the~

lower vertical section of IC04 just above the spray valve controllers. This is consistent with the intent:of minimizing operator error and packaging control station in 9

~

a modular concept.

3.4.24 87-219 and 87-219*A (Unit 2), Main Control Boards.

The modification installs-limit switches on the Unit 2--

pressurizer spray valves and indicating lamps on 2004 for control room position indication. In addition, a solenoid L valve is installed in the spray valves' air line and a E selector switch mounted on 2004 to provide the control operator a means to cause a Unit 2 spray valve to fail closed.

Summary of Safety Evaluation: The modification constitutes a potential change to the facility or its operation as described in the FSAR. .%ction 4.2.2 of the FSAR says the spray valves limit pressure during load 84

I v r transients'and that they can bel manually operated from. a the control room. This will remain unchanged after- ;

installation of this' modification. -

Section 7.7.3 of the FSAR says control stations on the.

main control boards are packaged in a modular concept and; i are grouped according to function to minimize' operator

. error. It also states the. vertical section of the i control boards incorporates instrumentation, trend ,

recorders and annunciator panels and that'the console l section contains control devices and indicating lights. >

These concepts are employed in this design except the

-indicating lamps,are-located'on the~ lower vertical' i; , - -

section-of 2004 just above the: spray valve controllers- -

[' This is consistent with the' intent of minimizing operator- i error and packaging control stations in a modular?

concept.

3.4.25 87-232 (Unit 2), Main Control Boards.

MR 87-232 replaces the SI spray ready / spray active and containment isolation status light panels in main control board C01. There are four panels per unit,' located near-' !

the top of the vertical section of MCB C01 in the control room. The existing panels _will be replaced with panels that: (1) Have larger windows, allowing larger, more readable test; (2) utilize the CHAMPS database equipmentn numbering system; (3) have the windows functionally-grouped for. pattern recognition; and (4) operate on the -

on/off/ push to test sequence rather than the dim / bright ;

l: concept currently in use.

Summary of Safety Evaluation: An evaluation is required because the proposed change could affect =the function or method of a system, structure or component described in '

the FSAR. The new panels will enhance the operators' ability to determine SI system and containment isolation status by presenting the same information as the existing panels in a clearer format. The most significant improvement will be that it will be much easier to tell' which windows are on and which are off. .

q 3.4.26 87-233 (Unit 1),87-234 and 87-234*A&B (Unit 2), Safety Injection System.

These modifications will remove the full stroke seal-in feature of the high head SI pump discharge motor-operated.

valves, 866A&B, to allow use of these valves in a demand (throttling) capability. They will also replace the existing valve position two-rotor limit switch with a l

j four-rotor limit switch to provide better valve position l:

indication and still maintain equivalent valve opera-l bility. The new limit switches will be nuclear grade and R QA for this application.

l l

1 b 85 y

1, . _ _ .- _ _____ _ ____ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _

'l

>w' ;

?

susunary of' safety Evaluation:- The modifications constitute a potential change _to the facility or its operation as described in the FSAR. These valves are normally open and maintained open under_ static ;

conditions. They do not change position and are not s acted upon by i,n SI signal, and are passive devices during the SI injection phase of an accident. They are 1 procedurally shut and potentially reopened during the- ,

recirculation-phase of a loss of coolant accident (LOCA). Since these valves are in separate independent !

trains, a single failure does not violate design criteria- .

of the SI system for either the' injection or high head recirculation phase of an accident.

3.4.27 87-240 (Unit 1) & 87-241 (Unit 2), 480 V Electrical System.

These modifications replace the CV-7 480.V loss of voltage l relays on 1(2)B03 and 1(2)B04 with ITE relays. Instal- ;

lation will be performed during a refueling outage on each

unit.

Summary of Safety Evaluation: An evaluation is required because an SSC, as described in the FSAR.will be altered.

During installation of the new relays, the load shedding-feature for the affected Me V bus (1[2]B03 or.1[2]B04) will be disabled so that y a loss of offsite power, loads on the bus would not be stripped. The diesel generator, after starting, would close on the bus with many loads .,

still connected. This'is an acceptable situation because during an outage the safeguards buses are relatively "

lightly loaded. The major loads that will not be running or potentially needed are SI pumps, charging pumps, ,

auxiliary feed pumps, spray pumps,.and pressurizer heaters. In addition, instrument air compressors ~ trip independently'on low voltage.and require manual reset and restart. The major loads that will probably be connected ,

include service water pumps, RHR pumps and component ,

cooling pumps.

The new relays are of the same type as those presently being used for the 4160 V loss of voltage relays.

i Installation shall be done per an SMP. .The SMP shall also

. control loading of the bus which is being affected via the-L MR to ensure it is lightly loaded to preclude damage to

( the emergency diesel generator.

3.4.28 88-009 (Unit 2), Main Control Boards.

1.

f' MRs 88-008/009 are to rearrange the nuclear instrumenta-l tion (NIS) meters on 1(2)C04, respectively, in order to L

resolve HED #350. The rearrangement process will involve L

removing the meters from the control board and remounting 1- them in the existing NIS cutouts using the existing I

86 )

1 l

/.s==*.6 __

1:

}

[ ,

j

} ~

components. Some main control board (McB) internal wires I

)

between the meters and risers will be replaced due to inadequate lengths.

p Summary of safety Evaluation: An evaluation is required because the proposed change constitutes a change _to the 1

facility as described in the FSAR. Further, the proposed. '

change constitutes a potential change to the operation of the facility as described in the FSAR.

Each meter is electrically isolated from its associated input or protection _ channel via an isolation amplifier. ,

E This configuration allows meter rearrangement without l

[4 taking any associated protection channels out of service. :J FSAR Section 7.4.1. states that;"startup rate indication for the source and intermediate range channels ~is provided ]

on the main control boards." :Section 7.4.3 also specifically states that there are source: range count rate, intermediate range current, power range % power, power range delta flux,i and post-accident NIS meters on the main control boards. While these FSAR statements will-not hold true during the modification installation i_ process, their intent will be met by having all these-E meters mounted on the control boards before and after the modification work is performed.

The work will be done while the associated reactor is-shut ,

down for refueling so no power range or intermediate range- t indication will be needed. ' All associated protection channels will remain operable. The audible count rate will be available both in the control room and containment. <

t Visual and audible annunciation of any abnormal increase in. core activity will remain available in the control. room :

as well as audible annunciation in containment.

TS 15.3.8.3 states that,-" Core suberitical' neutron flux shall be continuously monitored by-at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the . .,

containment available whenever core geometry is_being changed. When core geometry is not being changed, ~ at least one neutron flux monitor shall be in service."

The meters will, therefore, only be rearranged when fuel ,

movement is not in progress. !

FSAR Section 7.4.2, under " Protection Philosophy," states,

" Separation of redundant protective channels is maintained from the neutron sensor with its associated cables to the signal conditioning equipment in the control room with its associated output wiring, indicating or recording devices, and protective devices." This statement is taken to mean that all components and wiring of a channel are physically separated from all components and wiring of a redundant channel. The isolated output signal cables from the r

87

, 1 f

y W . signal conditioning equipment to the NIS meters' on the MCB L. do not meet;this criteria nor do'they need to since the p

. outputs'are isolated. The new wiring required for these; b . modifications also will not meet this criteria; the existing wiring paths will be used. Therefore, the above statement -

on separation-of redundant channels should be changed to

clarify the actual wiring configuration such that wiring l separation is not required for non-protective portions of <

the circuitry. ,

FSAR Section 7.4.2, under " Equipment Design Basis," states ,

that for the wide range detection channel, Hall electrical >

equipment is seismically supported." The seismic support :'

-configuration will not be degraded by this modification.

b The meters will be remounted using the same hardware. : Making l the post-accident meterst mounting plate- flush with the rest of the control board will increase the control board '

strength by filling an existing hole.-

FSAR Sections.14.1.1'and 14.2.1 analyze an uncontrolled RCCA withdrawal from a nuberitical condition and fuel handling accidents, respectively. The NIS monitors reactivity and provides protective inputs under abnormal conditions. Since these protective features will not be affected by these modifications due to isolation via . .

isolation amplifiers, neither of the analyses is affected.

3.4.29 88-010-01 and 88-010-02 (Unit 1),. Containment. Ventilation-System.

The MR and Addendum I will replace'the volume control and backdraft dampers in the Unit 1 containment ventilation

(, system. Addendum 2 will remove _the pitot tube grids from j the system. (These MRs are= identical to MR 88-011 and addenda which were performed on Unit 2 during the U2R14 outage in 1988.)

Summary of Safety Evaluation: An evaluation is required ~

'because the MR and addendums'will alter a system as-de-scribed in the FSAR and/or' involve a potential change to the facility or its operation as described in the FSAR.

Replacement of the volume control and backdraft dampers will

l. not result in a change to the the original-system design as L described in the FSAR nor adversely affect the operation of the plant since the installation will be performed under administrative controls (e.g., SMP #923) and during a refueling outage.

H Addendum 2 will remove pitot tube grids from the containment ventilation syster.. These grids were added to the system as a plant betterment, however, the grids are currently in h

p 1 o l l- 88 ,

1 L

L

m ,

a' state of disrepair and do not provide accurate data. -;

Following removal of the grids, the ductwork will be returned: ;

,to its original construction using equivalent gauge sheet.

metal. .

3.4.30 88-012 and 88'-01'2*D, Circulating Water System.

l There are several design packages associated with the proposed systems.'. These would include one each for.

mechanical, electrical and other tie-ins. '

i r The differences between the originally proposed MR and the MR design as it exists at the present are as follows:

a. The hypochlorite tank in the pumphouse will be- -

located in the southwest corner. A dike must be- t installed around the tank, with dike height being about 6-1/2' high. ;

b. The sodium bisulfite tank is tentatist 'y planned to be located in the pit north of the Unn 7 condensate -

cooler,

c. Dilution water for the hypochlorite system will be q taken from the top of the screen wash header. The '

decision was made to not utilize the service water- .!

header. The bisulfite tank, however, will obtain dilution water from the service water line:to the water treatment plant.

d. There will be some PVC piping within the.pumphouse from the southwest corner of the pumphouse to north of the service water header. This decision was arrived at based upon cost considerations. An evaluation will need to be performed because of fire

^~

zone considerations. .

3.4.31 88-012*A, circulating Water System.

Design Package "A" covers installation of the mechanical-and civil portions of the chlorination and dechlorination ,r systems.

L Summary of Safety Evaluation: The MR constitutes a l' change to the facility as described in the FSAR. The i

following areas were addressed in the safety evaluation:

a. Fire Protection System. The planned chlorination of the service water system can potentially expose the .

1. fire water system jockey pump and motor-driven fire pump, as well as all system piping and valves, to a chlorinated water source. The current plans are to chlorinate the service water at a concentration of 0.3 ppm. This will not impair the ability of the fire protection system to perform its design 89 L

o l l

i function. The city water used as a source of fire. ,l water'at a system.fossile~fue1~ power plant.' typically .

I has a~ chlorine concentration of 0.3-0.5 ppm, and no: ;l detrimental effects have every been observed. In ,

addition, since the system is stagnant'most'of the l time, the chlorine will eventually convert to" j chlorides, and the lake water that is used-in the fire l water system at PBNP typically has a chloride . J concentration of 9 ppm. An additional chloride l concentration of 0.6 ppe from chlorine conversion would be insignificant.

b. FSAR. The FSAR will need to be revised to reflect the addition of permanent chlorination and dechlori-nation systems at PBNP. The FFDSAR contained one sentence about the original. chlorination. system.in Section 10.2. 'The NRC has received a copy of the '!'

latest WPDES permit for PBNP which allows chlorination and temporary dechlorination, and will !

receive a copy of the formal DNR approval for'the l permanent dechlorination system. '

c. DNR Approval. Chapter NR 108 of the Wisconsin Admini-strative Code (WAC) requires'that the construction of a permanent dechlorination system at PBNP receive-

prior DNR' approval, because PBNP would'then be classified as an industrial wastewater facility. ,

Formal DNR approval to proceed with the proposed. ::

project was received under DNR Approval No.88-205. !

d. Pipe Rupture. A potential exists for a rupture of

h. piping inside the service water pump room. The i L technical evaluation of this situation is somewhat similar to that for inadvertent: fire suppression system actuation. The rupture could affect no more

-than three service water pumps. A'significant number '

l. of pumps would remain available to meet normal j operating or safe shutdown needs. 7

e. Flooding. Flooding' damage to vital equipment-is not-credible for the chlorination and dechlorination'--

l systems. Sodium hypochlorite released due to a_ >

b storage tank rupture will be contained by.a specially-designed ~ dike. Sodium bisulfite released due to a storage tank rupture would be contained by the Unit 2 D condenser pit. Fluid from a pipe rupture in the service water pump room would drain to one of the circulating water pump pits through the steel grating l in the west corners of the room.

Technical Specifications. There are no technical l'

f.

Specifications affected by the subject modification.

)

1 I

90

.A

+

j

~

m '

g. Sodium Hypochlorite Spill. In the event'of a' rupture .[

of the= sodium hypochlorite storage tank, all of the 1 contents'(about 5,000 gallons) would be. contained, <

within a specially-designed dike around the, tank. ,

Two other subjects of. concern are personnel protection >

i .

and spill response

.i R 1. Personnel Protection: In the event of a tank rupture, unprotected personnel should be evacuated ;

from the immediate area. For spill response' ,

I personnel,: protective clothing should be worn, including s NIOSH-approved self-contained breathing apparatus (SCBA), as necessary. One or' both of the pumphouse roll-up doors'should be i opened to improve ventilation. A permanentx >

emergency eyewash / shower station will be: installed ;

within 25' of the-storage tank and pumps.

l-lD 2. Spill Response: The steps in AOP-12A should be~

l: followed in response to a sodium hypochlorite L spill. The chemical should be pumped from the dike into ' drums or a tanker truck for proper disposal. The area should then be flushed with water to remove the trace residue.

h.

Sodium Bisulfite Spill. In the event of;a rupture of the sodium bisulfite storage tank, all of the contents (about'4,250 gallons) will1 be contained within the Unit 2 condenser pit.

.Three subjects are of concern in this situation; a personnel protection, spill response and control room habitability.

1. Personnel Protectione In the event of a tank-rupture, unprotected personnel should'be evacuated from the-immediate area. For spill response personnel, protective clothing 'should be-worn, including either a NIOSH-approved SCBA or a full-face respirator with a cartridge capable of. l filtering sulfur dioxide gas. A permanent ;

emergency eyewash / shower station will be' I installed on El. 8' of the Unit 2 turbine hall l within.25' of the sodium bisulfite tank and pumps.

2. Spill Response: 40 CFR 116 and 302 list. sodium bisulfite as a hazardous substance. Parts 117 and 302 identify the reportable quanitity of this chemical as being 5,000 pounds, which equates to about 1,200 gallons for the 38% codium bisulfite that will be used for the dechlorination system. l l

1 91

4

t

. 1

3. Control Room Habitability: The'only event A involving the chlorination and & chlorination- l systems that has a potential for affecting- >

control room habitability is a rupture of the sodium bisulfite storage tank. i s An evaluation shows that a rupture of this nature L would not create a control room habitability problem.

1. Appendix R Evaluation. The-sodium hypochlorite !

storage tank, which is made of high density cross

linked polyethylene, will be located in the j circulating water pumphouse (fire zone 553). ;1 1

The sodium bisulfite storage tank, which is made of high density cross-linked polyethylene, will be l-located in the north end of the condenser pit in the Unit 2 turbine hall (fire zone 584) within a-sprinklered area.

The presence of the tanks in these fire zones will not significantly increase the combustible loading in the fire zones, nor present a significant hazard to equipment located within the fire' zones. '

A portion of the chlorination system CPVC piping will be routed inside the service water pump room (fire-zone 552). However, all CPVC piping within this room will be routed inside steel piping. The CPVC piping i

is therefore not considered to be an intervening ,

n combustible. The room is also protected by photo- ,

electric smoke detection and wet pipe sprinklers. 4

L i The CPVC piping would not provide a combustible path-l- way through.the service water pump room floor, nor-would it be classified as an intervening combustible -

because it would not be: exposed. The penetration .

E around the steel pipe sleeve will be sealed,:thereby. '

maintaining the fire area boundary.

Neither sodium hypochlorite nor sodium bisulfite is ,

flammable or represents an explosion hazard. Neither chemical will adversely react with conventional-L extinguishing media. Personnel-fighting fires in the vicinity of these chemicals should wear SCBA in accordance with normal fire fighting procedures, j . Chemical Delivery Accidents. Without appropriate measure being taken, the potential exists for I chemical solutions to be introduced into the wrong i

l l -

l 92

7, p

K-l

[" ,4 storage tanks during chemical deliveries.- Some'of I e- the possible mixtures that could result are as-follows: t

1. 12.5% Sodium Hypochlorite and 38% Sodium .

] '

Bisulfite: The heat of reaction resulting from-the mixture of these two chemicals would probably : j melt the host polyethylene storage tank and could j lead to an explosion from water vaporization. l J

2. 50% Sodium Hydroxide and 12.5% Sodium Hypochlorite: The mixture of these two chemicals )

would not present a safety hazard. Sodium-hydroxide is often used in the manufacture of ,

sodium hypochlorite. !

7 3. 50% Sodium Hydroxide and 38% Sodium Bisulfite: '

. The mixture of these two chemicals would not j

!- present a safety hazard. Sodium hydroxide is g often used in the manufacture of sodium j 1

bisulfite.

1

4. 98% Sulfuric Acid and 12.5% Sodium Hypochlorite: 'l The mixture of these two chemicals will result in l

!. the generation of molecular chlorine gas.- )

l

5. 98% Sulfuric Acid and 38% Sodium Bisulfite: The heat of reaction from the mixture of these two chemicals would be great enough to melt the :)

!: sodium bisulfite storage tank. I Mixtures of this nature will' be precluded by:

4 (1) Properly labeling each of the fill connections for operator identification; (2) inspecting the shipping list'of each chemical delivery prior to; unloading;'and (3) testing a sample of the' chemical' for-identification prior-to unloading'.' t

k. Miscellaneous. The chlorination ~ system will use-

1. dilution water from the. screen wash system rather than

1. from the seismic portion.of the service water-system. ;

L In addition,.the dilution water for the dechlorination }

l system will be taken from the service water line to I

the new water treatment plant, downstream of motor-operated safeguards isolation valve MOV-2817 for the line. Therefore, the proposed systems in no way L. affect' the seismic portions of the plant service [

water system, and will not unnecessarily draw from

=

the service water system in emergencies during periods of high service water demand.

93 a < 4 v4

s _ _ __ .

4 <

j.

,M wm ;

3.4.32( 88-038t (Unit 2),: Reactor Coolant System.

This modification will install thermocouples on the shells ,i" of the steam generatots and will provide local readout i

capability. ' , ,

)

Summary of Safety Evaluation: An evaluation is required because the MiI involves in a potentiki change to the 4 facility or its operation at described in the'FSAR. ,

Installation of these MRs has very little impact upon any. I safety-related equipment. The~ attachment of the-thermo- ;

couples is via high temperature epoxy cement and the steam.

generator shell is carbon. steel. . This' arrangement will have no deleterious effectr on the shell metal or its.

ability to retain secondary system pressure. .In addition,- e the conduit for these MRs is being installed seismically' :

and is mostly located inside the steam generator shield wall. Due'to its location and small physical size, there are no postulated failure modes which may'cause any impact 'j l- on. safety-related equipment. '

l

. 3.4.33 88-041 (Unit 1), Reactor Vessel. -l lJ Reactor cavity neutron dosimetry will be installed in the l annular air gap between the reactor vessel (RV) and the ;

primary concrete shield in each PSNP unit via MRs 88-041/. ,

042. The dosimetry-installation consists of aluminum dosimeter capsules-connected to and supported by ;tainless steel bead chain. j Summary of Safety Evaluation: . The MR constitutes a -

potential change to the facility or.its. operation as .[

described in the FSAR. A safety evaluation for this 2 modification has been' performed by the NSSS. U i

J The dosimetry is supported by a stainless steel frame which will be' installed around one of the RV outlet nozzle support shoes. The frame itself will rest on the ring girder. The support frame is installed with a nominal 1/8" clearance around the support shoe and is restrained by the two frame braces which have 1/2" bolts ,

tightened against the underside of the ring girder. The total weight of the support frame, bead chains, .

dosimetry, and miscellaneous hardware is approximately 33 pounds. The bottom termination of the bead chains j consists of chain clamps connected to stainless steel eye '

nuts threaded onto studs in the sum wall.

The specific aspects of the mechanical evaluation are as follows:

a. Loose Parts: The NSSS generic safety evaluation determined that there were no adverse effects on the ls RHR system pumps or valves due to the ingestion of l

stainless steel dosimetry bead chain.

i 94 L

l

!I -

w.S. . ' '

.)

I=i As described on Page 6.2-9 of the'FSAR, the sump "B" 'j screens prevent the entry of objects-larger than 1/8". This effectively eliminates any chance of.the-e, bead chains entering.the system since the; nominal .I diameter.of,the beads is'3/16".- 1 I

jf Post-accident use of'this'line could possibly result .

E in leaking valves subsequent to use due to debris- 'l (note leakage only in view of the. valve type and )

Sump "A" grating). Use of this line in this' fashion a is not anticipated as the need_is questionable and j access (unless a mild LOCA,-low radioactivity in- ;

coolant) would be prohibitive. If use was pursued, I it certainly would be after the containment pressure subsided; thus little driving force would be j present. In any event, the RHR system would provide i

, for necessary sampling for Sump "B" recirculation.

i b. Missile Hazards: The design of the support frame prevents the dosimetry system from constituting a ;

missile hazard by two means. First', the physical

. size of the frame assembly is too large,to permit' passage through the narrow annular' air. gap betweeni the RV insulation and biological shield liner plate.

Second, the support frame is physically restrained by ,

virtue of the fact that it is assembled-around one of the RV nozzle support shoes.

In the event of chain failure, the dosimetry capsules and gradient chains are prevented from becoming missiles'above:the elevation of the nozzles by the geometry of the air gap and the presence of the RV insulation support channel.

r The dosimetry capsules and gradient chain do not' constitute a missile hazard to the incore detector thimble conduit tubes for two reasons. First, the

~

dosimetry is not installed above the incore tubes and ,

second, the mass of a capsule is approximately 'l one-third of the most conservative acceptable impact mass.

The dosimetry capsules and gradient chain do not i constitute a missile hazard to the containment-liner plate. The liner plate is protected by 18" of . ,

concrete in the region below the RV, and by a minimum I of 12" of concrete on the floor of the incore l p tunnel. The exposed liner plate in the incore tunnel i L is.not at risk because of the geometry of.the tunnel l and due to the small mass of the dosimetry capsules. I p c. Effect of Added Mass: The installed hardware does not affect the capacity or operation of the RV support system or the RV insulation during normal operation or during seismic events.

95 l'

y+ ,

t i

d. NIS Excore Detectors: The reactor cavity neutron' ,

dosimetry system is totally passive and does not ;

interact either neutronically or mechanically with-the active excore neutron detectors. This is due to 'l the fact that the reactor cavity neutron dosimetry j system is designed to produce free field measurements

'with little or no local perturbation of the neutron flux and that the excore detectors are physically .

1solated by the steel positioning tubes and the j reactor cavity liner plate.

e. Materials Used: With.the exception of>the~ aluminum .

chain stops and dosimetry capsules (approximately l 3.5 pounds), the reactor cavity neutron dosimetry j system is constructed of stainless steel. 'j The installation was also reviewed with respect to the mass of aluminum in containment. The. amount of aluminum i in containment-is restricted since in'a post-LOCA 'l environment'the aluminum corrodes, forming hydrogen gas 1 which is a flammability concern. As indicated in FSAR Table 5.6.2-2, the present mass of aluminum in the l containment is 566 pounds (of which 100 pounds is j identified as contingency). The total mass of aluminum ' -

that would be added to the containment by the addition of the dosimetry is about 3.5 pounds. Not only is this a small fraction of the contingency value that has been taken into account, it is also a very'small increase in the total mass of aluminum. Additionally,'per FSAR- . 9 Appendix B,' Table 9,- alupinum corrosion contributes only a about 10% of the total hydrogen production. Thus, the. L assumptions regarding the times of hydrogen purge to control hydrogen buildup remain .anchanged and the-calculated doses due to containment purge that;are 4 reported in Appendix D of the FSAR.are-also unchanged. J Table 5.6.2-2 will be updated to include the addition of 3.5 pounds of aluminum to the containment inventory.

L -3.4.34 88-048 (Unit 1) and 88-049 (Unit 2), 4160 V/480 V ;

Electrical Systems. '

MRs 88-048/049 will install test jacks and knife blade ~

test switches and require revisions to RMPs #71-76 to' =I r assure complete monthly testing of the 4160 V and 480 V L loss of voltage and degraded voltage relay channels, l

including the automatic undervoltage (UV) initiation of the turbine-driven auxiliary feed pump. These modifica- ,

tions and appropriate RMP revisions will assure the lL complete monthly testing of the following: '

a. Degraded voltage and loss of voltage relay contact operation,

[-

L

b. Slave relay coil integrity, t 96

-- n .

l h j

c. Slave relay contacts which are not individually !

verified annually as part of a refueling outage -j test, including the degraded voltage time delay relay j p contacts, j

> i

d. Integrity of' the W matrix output relay coils, !

including the time delay relay for the turbine-driven !

auxiliary feedwater pump initiation circuit. ;

i~

Summary of Safety Evaluation: An evaluation is required Fecause these MRs represent a potential change to the :

facility or its operation as described in the FSAR. A large resistor in series with each ground test jack !

provides assurance that equipment will not be damaged or j inadvertently actuated during the test due to a faulty :

test meter or an improperly placed jumper wire. A sheet !

metal cover over each test jack panel will prevent an external jumper wire from defeating relay channel ,

operation. The 14 AWG SIS wire, the knife blade test

switches, the terminal blocks, and the test jacks have all

+ been specified as QA material. The test jacks are a nylon t body-type, which provides resistance to vibration, i electrical insulation and overall durability. Test switch- ,

assemblies are seismically qualified per the IQ database. - j Existing train separation will be maintained by this ;

installation. Note that the test jack terminals will be insulated with electrical tape or heat shrink tubing. ;

A05, A06, B03 and B04 are seismic class 1 components per the QA Policy Manual. The addition of the test jacks, t test switches and external cover will not degrade the "

seismic capacities of these cabinets based upon the following: '

a. The relatively small mass of the test jacks, test ;

t switches and sheet metal cover will not cause a shift l in a cabinet's center of mass.

b. Installation of the test equipment on the cabinets I will require that a 2-1/2ax6" (approximate) l- rectangular hole be cut for each test switch assembly and a 1/4" hole be drilled for each test jack. By ,

evaluating the size and thickness of the cabinet ,

panels, ong with the proposed number of jacks and switches to be added to each cabinet, engineering >

judgment concludes that the proposed installations y will not degrade the ability of a cabinet to support L itself under seismic loading. All equipment will be mounted on doors which are not structural components.

1

c. The test switch assemblies to be installed are seismically qualified per the EQ database.

P 97

1 :

i

! i l' ;

A single failure of the test jacks or test switches will )'

not prevent the safety function of the W relay channels.

Failure of a test switch would be indicated by W alarms failing to clear upon recovery from the W condition.

Since the test switches are QA and are seismically qualified, the probability of their failure is considered 1 to be no greater than that of an existing relay wire !

becoming detached from a teminal. Such a failure would !

not prevent a safety grade function. ,

3.4.35 88-052, Fire Protection System.

The MR proposes to change the' fire pump test header to allow testing without use of additional equipment. The i existing configuration is a header on the west pumphouse wall. To perform the pump test, six 50' rubber hoses are ,

connected to the header. The rubber hose pressure drop is too high to allow the required 150% flow rate. The change proposes replacing the existing header with an integral .

header / playpipe assembly. !

Sunnary of Safety Evaluation: The change constitutes a l' potential change to the facility or its operation as described in the FSAR. The new assembly will not add ~

additional piping runs to the system. Its addition does l not affect the normal operation of the fire pumps.

Additional fixtures will be of Bechtel Piping Class KB and will meet existing pressure and temperature i specifications. A vehicle bumper will be included to -+

prevent accidental damage by collision.

3.4.36 88-067 (Unit 1) and 88-068 (Unit 2), Reactor Safeguards i System.

An evaluation was done to determine the acceptability of the ORT #3 test panels and to propose any changes needed ;

to make the panels acceptable.

Summary of Safety Evaluation: The purpose of this evaluation is to evaluate the safety impacts of the installation of the ORT #3 test panels. The panels were installed in the early 1970s to accommodate testing ,

described in TS 15.4.6.A.2. The purpose of the testing is

to ensure that the diesel generators will start and assume '

load as described in FSAR Section 8.2.

The panels monitor the diesel loading by monitoring the position of vital load breakers, along with other relays.

Cables connect the test points to either auxiliary contacts ;

or cell switches on the breaker to monitor its position.

A recorder is connected to the panels during ORT #3 tests ,

to reccrd the information. '

98

l 1

l l

I The test panels do not introduce a hazard to the various l

l. breakers and relays they are connected to because the ;

cables are only energized for a short time during ORT #3 l testing, which is done while the respective unit is in 1 cold shutdown. While the unit is at power, the cables are l deenergized. . l The testing circuitry does not interconnect with any accident-detection or mitigation circuits and will, therefore, have l no effect upon the potential for or consequences of an 1 sceident. The cables do not substantially increase the fire !

loading in the cable spreading room or the vital switchgear roce and therefore, do not present a fire hazard. ,,

t

. 3.4.37 88-074*B (Common), 125 V DC Electrical System.

This modification temporarily install a 125 V Class 1E '

station battery in the auxiliary feedwater pump room, east !

of the tunnel. Removal of the battery will be controlled l by a different design package. ;

Summary of Safety Evaluation: An evaluation is required ,

because the proposed change alters a system, structure or ;

component described in the FSAR. The basic function of l this battery will be to provide backup emergency power to buses D01 and D02 while the existing emergency power '

sources, DOS and D06, are replaced. This installation is l required since changeout activities for one battery, which >

l will be controlled by other design packages, will take longer than the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed for having cne of DOS or i

D06 out of service while one or both reactors are

! generating power. '

)

TS 15.3.7.B.I.f delineates the 24-hour LCO by. stating, L "One of the batteries DOS or D06 may be inoperable for a .

period not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided the other three batteries and four battery chargers remain operable with ;

one charger carrying the DC loads of each DC main distribution bus." While DOS and D06 are specifically called out in this LCO, the intent of having a Class IE battery connected to the main distribution buses will be met during battery changeout. i Sections 8.1.1, 8.2.2 and 8.2.3 of the FSAR describe the DC distribution system and Figure 8.2-10 shows the system I diagrammatically. All descriptions given will still apply while the_ battery of this design package is connected to the bus (es) except Figure 8.2-10 which shows both DOS and '

D06 and 60-cell strings. This is not consistent with all other FSAR data nor is it consistent with the existing '

installations which show both DOS and D06 are made of 59 cells connected in series. The battery to be installed will also be made of 59 cells. Therefore, Figure 8.2-10 should be changed to indicate both DOS and DOE are 59-cell batteries, not 60-cell batteries.

99

_ e

I The design bases criteria for a station battery, as l documented in the FSAR, is that it must be able to withstand, without undue risk to public safety, the forces :

that may be imposed by natural phenomena such as earthquake, :

tornado, flooding, high wind, or heavy ice. The station i battery to be installed will be able to withstand the forces of a tornado, flooding, high wind, or heavy ice !

since it will be installed in the class 1 portion of i the control building.

Some forces due to an earthquake will be transmitted to the battery. The battery system will be qualified to withstand the maximum expected seismic forces with a 100% ^

margin. Existing equipment in the area of battery installation may not, however, meet this criterion ;

creating the potential for this equipment to fall on the '

battery system, rendering it inoperable.

l All equipment that has a potential for falling on the battery was evaluated to determine the seismic mounting adequacy. The evaluation shows only two fluorescent lights, two service air rubber hoses, and a cover over the W46 pulleys are inadequately supported to withstand a '

seismic event. The lights and hoses will be removed for ,

the period the battery is actually connected to a bus and the pulley cover mounting will be improved in order to ,

remove these seismic hazards. Removing the lights will .

cause no personnel hazard since lighting wil.L be available !

from other fixtures in the room which have no potential '

for falling on the battery. Emergency lighting for the _,

area, which is seismically mounted, will remain intact. ;

The design bases criteria for either D05 or D06, as delineated in Bechtel's Information Exchange on Electrical Design for PBNP, is that the batteries must provide power to all essential DC loads for one hour upon loss of r station AC power and supply DC power required to actuate ,

and operate the engineered safeguards system at any time including the one-hour period following the complete loss of station power. Bechtel determined the battery size necessary to meet these criteria by considering the worst case (i.e., loss of station AC coincident with an SI actuation) and included a 100 A contingency throughout the required one-hour duration. The battery to be installed ,

will have the capability to exceed these maxi:num power ,

delivery requirements.

Section 14.1.11 summarizes the evaluation of a loss of all AC power to the station auxiliaries accident. Step 1 of the turbine trip sequence states that vital instruments are supplied by emergency power sources. The emergency power sources for the red and blue instrument buses are the emergency backup batteries connected to buses D01 and D02. This modification will not change the probability of this accident occurring since power availability at the 100

. - . - - . . _. . - , - - - . .. -. 1

I station auxiliaries is independent of the emergency power source. The consequences of this accident remain unchanged since the battery system is designed to supply !

the necessary loads under these circumstances. '

No other documented design basis criterion exists.

t A steam line break near the battery could possibly render i the battery system inoperable. The only steam lines that will be near the battery are the 3" lines feeding the steam-driven auxiliary feedwater pumps, downstream of the i normally closed, motor-operated stop check valves. The ,

highly unlikely event of a break in these lines would not ;

constitute a steam line break (rupture) accident as described in Section 14.2.5 of the FSAR since these lines :

are normally depressurized and are not considered in the ;

auxiliary steam supply to auxiliary feedwater pump turbine -

piping analysis in Appendix E, Section 5.2 of the FSAR. !

A hydrogen gas fire and/or explosion due to an excessive ',

hydrogen concentration (i.e., greater than 2%) from the battery could hypothetically cause a steam line rupture, calculation shows that it would take no less than +

133 days with no ventilation to reach a 2% concentration i of hydrogen. The redundancy and air circulating ,

capabilities of the existing HVAC system for the area assure such a concentration buildup will not occur. In the highly unlikely event that all air circulating capabilities were lost, the hydrogen generation from the ;

battery could be stopped before creating any hazard by 4 removing its float voltage since the battery only needs to be able to supply emergency power for no more than several weeks. Therefore, the probability of a steam lise rupture is not increased by the installation of this design package.

Since the service water headers in the auxiliary feedwater -

pump room are not sleeved, and these headers are sleeved

in the existing DOS and D06 battery rooms, the battery will be more_ susceptible to a malfunction caused by a service water header leak. Such a service water header '

malfunction can be mitigated by isolating that portion of the pipe. Neither bus D01 nor D02 operability is required to provide such isolation. Therefore, no safety issue is raised by placing an emergency backup power supply for either D01 or D02 in a room with unsleeved service water headers since either or both units can'oe operated safely without such an emergency backup power supply available assuming battery loss does not cause loss of the bus. If tne bus is lost, the operating unit would trip due to the loss of the associated instrument bus.

101

i i

ir Installing the battery in the auxiliary feedwater pump room will not cause the existing Halon fire suppression !

system to be overloaded. Halon is the only recommended :

type of fire suppression system for batteries and is used ;

in the existing, DOS and D06 battery rooms.

All Appendix R feeder cables powered by the bus to which the battery will be connected are routed through the same ;

fire zone that the battery will be installed in, one fire i could cause all Appendix R loads powered by this bus to !

lose their power source by burning these feeder cables. ;

If this were to occur, there would be no Appendix R need '

for the power source, and therefore, no Appendix R issues are raised.

The installation location of the battery will impair access to the local control stations mounted on the i auxiliary feedwater pump room east wall. Sufficient room j will be left such that no special equipment (such as !

ladders) will be required to access the panels. j 3.4.38 88-074*C, Station Battery DOS Changeout This modification changes out station battery DOS. I i

Summary of Safety Evaluation: An evaluation is required -j because the MR, as implemented by the special maintenance !

procedure, will alter a system, structure or component as ;

described in the FSAR.

The interim battery that was installed by Design Package B of the same MR will be used to provide DC bus D01 backup ;

power during the changeout process. Swing battery charger i D09 will be disconnected from D01 and its tie breaker used for connecting the interim battery to the bus.

TS 15.3.7.A.1 requires all four batteries to be operable to make one or both reactors critical. TS 15.3.7.B.1.f allows one or both reactors to generate power with either i DOS or D06 inoperable for a period not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided their associated chargers are operable and l L supplying power to the loads. Since DOS changeout ,

l activities will require more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the interim battery installed in the auxiliary feedwater pump room v vill be used as the bus backup power supply in order to not exceed these Technical Specification requirements.

The interim battery in the auxiliary feedwater pump room is qualified as Class 1E and Seismic Class 1.

i Disconnecting DOS from bus D01 could cause perturbations

! on the bus which could cause a spurious reactor trip. To l avoid such a spurious reactor trip, the interim battery I will be paralleled with the existing DOS battery before D05 is disconnected from the bus. Swing battery charger D09 is provided to allow disconnection of either D07 or 102

I i

DOS battery chargers from their associated buses for i maintenance purposes. It is not anticipated that D09 will I be needed to allow D07 maintenance while DOS is being changed out, so swing charger D09 will be disconnected from its D01 tie breaker and the tie breaker (D01-4) will .

be used for conriecting the interim battery to D01 before disconnecting battery DOS. If D07 is lost, the battery :

vill supply bus power for a sufficient amount of time ;

to raestablish power from the charger.

Breaker D01-4 is rated at 600 amps, one-half of the '

1200 asp rating of the main feeder breaker D01-1 for battery DOS. By taking the Unit I turbine bearing emergency lube oil pump (1P37D), the main generator air side seal oil pump (IP59B), and the main feedwater pumps' lube oil pumps (1P73C&D) out of service, the maximum load ,

that would be seen on bus D01 is reduced to allow the use of breaker D01-4 as the main supply breaker without it tripping due to the maximum load. These pumps are l essential for safe operation of Unit 1 balance of plant

(

equipment so changeout should be done while Unit 1 is shut down.

l Breaker D01-1 will presently alarm in the control room I

when it is opened. During changeout, D01-1 will be open so its alarm input will be defeated. D01-4 will have to l be verified closed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to ensure TS 15.3.7.E.1 is not violated.

The new battery to be installed as the new DOS is made of Exide Style 2GV-23 cells, which are rated with a capt. city equal to the existing C&D battery. The new ExiG. 1,attery '

and racks will be Seismic Class 1 and nuclear grade Class ;

1E. The existing DOS battery room will be used and is adequately sized, ventilated, temperature controlled, and -

protected from fire. The existing cables between the bus

and battery are also adequately sized so they_ will be used for this connection.

The description of DOS, as given in Sections 8.1.1, 8.2.2 and 8.2.3 of the FSAR will spply to the new battery installation without exception. Figure 8.2-10 of the FSAR shows both DOS and D06 as 60-cell batteries. This does ;

not presently apply to the existing batteries nor the new DOS battery which are (or will be) 59-cell batteries.

Fifty-nine cell battery banks is consistent with the written description of DOS and D06 given in Section 8.2.2 of the FSAR. DOS and D06 were initially 60-cell batteries so it is more than likely that Figure 8.2-10 updating was missed when the batteries were changed to 59-cell strings.

Therefore, Figure 8.2-10 should be changed to reflect the proper number of cells which make up these battery systems.

i 103

I i

3.4.39 MR 88-074*E, Station Batteries, t

Design Package E of MR 88-074 is to change out station i battery D06. The interim battery that was installed by j Design Package 4 of the same MR will be used to provide DC bus D02 backup power during the changeout process. Swing :

battery charger D09 will be disconnected from D02 and its ;

tie breaker used for connecting the interim battery to the !

bus, j Summary of Safety Evaluation: An evaluation is required because the proposed modification will alter a system, ,

structure or component described in the FSAR. .

TS 15.3.7.A.1 requires all four batteries to be operable ;

to make one or both reactors critical. TS 15.3.7.B.1.f -

allows one or both reactors to generate power with either. i DOS or D06 inoperable for a period not exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided their associated chargers are operable and i supplying power to the loads. Since D06 changeout ;

activities will require more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the interim 1 battery installed in the auxiliary feedwater pump room will be used as the bus backup power supply in order to >

not exceed these Technical Specification requirements.

The interim battery in the auxiliary feedwater pump room ;

is qualified as Class 1E and Seismic Class 1.

Disconnecting D06 from bus D02 could cause perturbations on the bus which could cause a spurious reactor protection system actuation. To avoid such a spurious RPS actuation, the interim battery will be paralleled with the existing [

D06 battery before D06 is disconnected from the bus, i Swing battery charger D09 is provided to allow disconnection of either D07 or D08 battery chargers from their associated buses for maintenance purposes. It is not anticipated that D09 will be needed to allow D08 maintenance while D06 is being changed out, so swing charger D09 will be disconnected from its D02 tie breaker and the tie breaker (D02-4) will be used for connecting the interim battery to D02 before disconnecting battery D06. If D08 is lost, the battery will supply bus power for a sufficient amount of time to reestablish power from the charger.

i Breaker D02-4 is rated at 600 amps, one-half of the 1200 l amp rating of the main feeder breaker D02-1 for battery D06. By taking the Unit 2 turbine bearing emergency lube oil pungp (2P37D), the main generator air side seal oil pump (2P59B), and the main feedwater pumps' lube oil pumps ;

(2P73C&D) out of service, the maximum load that would be ,

seen on bus D02 is reduced to allow the use of breaker D02-4 as the main supply breaker without it tripping due to the maximum load. These pumps are essential for safe i

operation of Unit 2 balance of plant equipment so changeout should be done while Unit 2 is shut down.

I 104 i t ;

j l

areaker D02-1 will presently alaam in the control room I when it is opened. During changeout, D02-1 will be open i so its alarm input will be defeated. D02-4 will have to i be verified closed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to ensure l TS 15.3.7.B.1 is,not violated.

In order to minimize the chances for a spurious RPS I actuation on Unit 1, the precautions that are taken for ;

shifting the blue instrument supply bus feeder, as delineated in 01-37, will also be taken for the battery j paralleling process. '

The new battery to be installed as the new D06 is made of )

Exide Style 2GN-23 cells, which are rated with a capacity !

equal to the existing C&D battery. The new Emide battery i and racks will be seismic Class 1 and nuclear grade Class j 1E. The existing D06 battery room will be used and is 1 adequately sized, ventilated, temperature controlled, and I protected from fire. The existing cables between the bus and battery are also adequately sized so they will be used ,

for this connection. l The description of D06, as given in Sections 8.1.1, 8.2.2 and 8.2.3 of the FSAR will apply to the new battery :

installation without exception. Figure 8.2-10 of the FSAR J shows both DOS and D06 as 60-cell batteries. This does- '

l not presently apply to the existing batteries nor the new D06 battery which are (or will be) 59-cell batteries. ,

Fifty-nine cell battery banks is consistent with the written description of DOS and D06 given in Section 8.2.2 of the FSAR. DOS and D06 were initially 60-cell batteries i so it is more than likely that Figure 8.2-10 updating .

was missed when the batteries were changed to 59-cell _

strings. Therefore, Figure 8.2-10 should be changed to reflect the proper number of cells which make up these battery systems.

L 3.4.40 88-086, ISI-866A&B Motor and Gearing Replacement. ,

This modification request will modify the Unit 1 safety l injection pumps discharge valves, ISI-866A and ISI-866B to l improve the operating characteristics of the motor operators.

Summary of Safety Evaluation: As presently installed, it may not be possible to remotely open the valves against design differential pressure during a degraded voltage condition. This is due to the motor operator not being ,

able to generate sufficient stem thrust to open the valve prior to torquing out. These valves are normally open valves and are ackninistratively verified as being open.

The modification will consist of replacing the motor operator motor and internal gearing. The new motor will be a 7.5 ft-lb motor that was originally specified for 105 i

,- v , ~ .vv. -, , , - - - - . _ . _ . - -_ . _

i I !

the operator. The new gearing will be selected to I provide sufficient stem thrust to open the valve under !

design differential pressure with degraded voltage. j Following the ' modification, ISI-466A and ISI-466B will be

[. identical to the corresponding Unit 2 valves. l The seismic qualification of the valves will not be affected because the new motors will restore the motor operators to the originally specified configuration. i (PBNP previously replaced the motors with 10 ft-lb !

motors.) Additionally, the gearing change will not affect the mass of the valves. l The proposed modification will result in an increased valve stroke time. The present stroke time for the t valves is approximately 25 seconds. Following the i

modification, the new stroke time is acceptable because i these valves are not required to function on a safety injection signal. The valve operators will still be able :

to cycle from one position to the other within the 120 seconds '

required by Section 6.2.2 of the FSAR. Note that the 866 valves are not considered the injection line isolation valves. ;

These valves are injection line CIVs outside of containment.

These do not receive any auto signals (SI or CI).

3.4.41 88-088 (Unit 1) and 88-089 (Unit 2), HVAC. ,

Chese modifications replace existing containment cooling fan discharge plenum access door handle / latch sets and install an inspection window in each access door, i l Summary of Safety-Evaluation: The failure modes are:

I- (1) Door dogs break; (2) Glass in window breaks out. .

Both the glass and door dogs are being sized for 2 psi

differential pressure. For the door dogs, the differential pressure is assumed to be in a direction to open the door.

FSAR Chapter 6.3 requires that the accident fan cooler components be capable of withstanding differential pressures which may occur during the rapid pressure rise to 60 psig in 10 seconds. Assuming this is the basis for

, the specification of 2 psi differential pressure, then the l differential pressure will act to hold the access doors closed during the initial phase of a LOCA. The opening forces on the door will be from the discharge plenum pressures (7.6" wg normal and 3.1" wg normal / accident).

Since the latches are being sized based on 2 psi differen-tial pressure acting to open the door, the safety factor for the latches will be in excess of 7.

Failure of both door latches on an access door will cause the door to open fully, providing an alternate flow path +

for the air. This will cause the fan to " ride the fan 106

l i

curve" to approximately 40,000 cfm/ fan normal and 34,000 cfm/ fan normal accident. This will still provide l the required containment cooling. However, the cooling l' will not be distributed as designed. This will not affect the abili,ty.cf the accident fans to remove the heat in containment following a LOCA.

The probability of both door latches failing is small. .l Even if they should fail, the ability of the containment i accident cooling fan to remove heat will not be diminished. l The resulting latch configuration will be stronger than j existing. This change does not pose an unreviewed safety j question. j 1

The failure of the glass in the inspection window would j increase the air flow through the fan cooler by

<1500 cfm/ fan. The increased fan load will cause the brake horsepower requirements to increase slightly. This j is still within the service factor of the 50 hp motor for l normal operation. The thermal overload may trip i depending on its actual setpoint. The thermal overload 1 can be reset at_1(2)B31. The operator of the 150 hp. I motor for the accident fan will not be affected. The '

tempered glass will be able-to withstand a 2 psi differential pressure under accident conditions (high temperature).

It is unlikely that the 3/4" thick tempered glass in the inspection will break during normal operation and the !

increased fan load will cause the 50 hp motor to trip.

l-In the event that this should happen, the broken glass can be replaced with a steel plate. This will allow the I fan cooler to be returned to service within the 48-hour i Technical Specification limitations. The installation of an inspection window in the access door does not pose an unreviewed safety question. -

3.4.42 88-094*B (Unit 2), Secondary Chemistry.

{

! This modification makes the tie-ins and line routings for batch tank water and chemical injection for the new Unit 2

( hydrazine and morpholine chemical injection skid. The ,

I tie-ins will be made and root valves installed during the l

1989 Unit 2 Outage. The line routings to the new chemical injection skid will be made after the skid is installed which will occur at a later date. A separate design package and associated safety evaluation will be prepared :

for placing the new hydrazine and morpholine chemical injection skid into service. .

Summary of Safety Evaluation: An evaluation is required because the proposed change will alter a system, structure .

or component described in the FSAR. The batch tanks' water supply will be provided by gravity feed from the condensate storage tanks (CSTs). A 1" line will be 107

! I L .

l r

connected . to an existing 6" line which is valved into the .;

CSTs by normally locked open valves. The new components j to be installed will meet the design pressure and j temperature ratings of the line class which is being tied ;

into. The new 1" line will be supported in accordance with Power Pipirig Code USAS B31.1-1967. A liquid penetrant examination of the welds up to the new root l valve will be performed along with an initial service leak j check. The leak tightness of the remainder of the new 1" i line will be assured by a post installation leak check at !

design pressure. 1 l

The future point of chemical injection into the condensate !

s'ystem from the new chemical injection skid will remain !

unchanged from its existing position at the condensate '

pump discharge heeder. A tie-in will be made and a new :

rcot valve installed in the existing 3/8" injection tubing upstream of existing root valve CS-21. In addition, a new l in-line valve will be installed upstream of this new root ;

l valve to facilitate the removal of the existing hydrazine ,

system once the new chemical injection skid has been ,

functionally tested and accepted. The new components to ,

be installed will meet the design pressure and temperature j ratings of the line class which is being affected. The new 3/6" injection tubing will be supported at least every l 6'-6" per standard plant practice. A post- installation !

leak check at design pressure will be performed on all i of the newly installed components. :

l 3.4.43 88-109 (Unit 1), Safety Injection System.

This modification proposes to install an isolation valve ,

L and calibration tee in each of the sensing lines for the accumulator level indication system. This will allow for >

calibration of the level element / transmitter without isolating the associated accumulator pressure transmitter.

Summary of Safety Evaluation: The MR constitutes a potential change to the facility or its operation as i described in the FSAR. The existing sensing line is 3/8" '

t-tubing and is part of the accumulator pressure boundary. l Present configurations on the sensing line employ Swagelok fittings and Whitey valves. Pressure and i temperature requirements of the system are much less than the design ratings of the fittings and valves.

Material types will be compatible with existing types. .i' Valve positioning will be administrative 1y controlled.

The accumulator level indication system is Seismic class 1. Installation will be based upon TMI tubing installation criteria for seismic qualification.

The proposed change will not affect the functionality of the system. '

108 l

3.4.44- 88-119 (Unit 2), Primary Plant Instrumentation.

The proposed design for the MRs for both units is to break up the string of 6 indicators (4 pressurizer level, RV dual level and pressurizer relief tank (PRT) level) by moving the PRT level indicator (LT-442) over a few inches to the space that was previously occupied by an LTOP key switch and indicating light. The RV dual level indicator (LI-447/447A) will then be moved to the position LI-442 was in. The indicators will be mounted in the same manner as existing and the electrical connections will be similar to existing; thereby the design and function of the instruments will not be changed.

Summary of Safety Evaluation: An evaluation is required because the proposed modifications will alter a system, structure or component described in the FSAR. To minimite the potential for installation impacts on plant operation, the work will be done during the respective unit's refueling outage.

Since the PRT level and the RV level indication will be unavailable for a short time while the indicators are being moved, the work plan will be written to minimize the time the indication is unavailable, and will identify the specific plant conditions required.

The indication will be removed from service while the cavity is flooded and the reactor vessel head is removed.

The RV level indication is not required during this time.

The PRT level indication is not essential for the short period of time the indicator is being moved because discharge into the PRT can be observed by monitoring the PRT temperature and pressure indicators.

No additional weight is being added to the control boards i and no structural supports will be cut so the seismic rating of 1(2)C04 will not be affected.

l 3.4.45 88-123 (Unit 1), Replace the SI block switch on col with L two train-specific switches, This modification replaces the existing SI block switch with two train-specific switches, to meet the intent of IEEE-384 for train separation. This change requires mounting a new selector switch on C01 and moving two L contact blocks from the existing SI block switch to the new one. The wiring design will not be changed; however, as-found wiring will be upgraded to safeguards system design practices.

Summary of Safety Evaluation: An evaluation is required as the proposed change will affect the function of a system, structure, or component as described in the FSAR.

L 109

o The only section in the FSAR affected by this change is

, Section 6.2-43, the operational sequence testing of the SI system. The FSAR describes the SI block switch being moved to the unblock position to initiate the test. This ,

will have to be changed to consider two SI block l switches. -

)

1 The change will not affect any of the system's operating i characteristics or accident response characteristics as j described in the FSAR. i l

The SI block switch is not mentioned in the Technical ,

Specifications and this modification will not have any impact on them.

3.4.46 88-131 (Unit 1), Main Steam System. 2 The modification proposes to change out the fourth pass i drain lines between the MSRs and the Nos. 5 feedwater )

i heaters. The piping would be changed from carbon steel to #

l stainless steel. An elbow failure and subsequent testing

-during the summer of 1988 revealed significant wall- i thinning in various locations in these lines. 'The i' stainless steep pipe is more resistant to erosion /

corrosion than carbon steel. .

Summary of Safety Evaluation: An evaluation is required !

because a system, structure or conqponent, as described in the FSAR, will be altered. The new piping will be installed in accordance with the requirements of B31.1.

The only.significant change will be the change in material. The piping class designation will change from EB to ED, with the same pressure and temperature ratings. ,

The replacement pipe will be the same size and schedule as the existing pipe, and will be routed in the same field configurations. Since the size and weights of the piping ,

will not change, the existing supports and support locations can be reused without further analysis, t Stainless steel does have a greater coefficient cf thermal j expansion than carbon steel, but the routing / support configurations allow enough flexibility that additional L thennal growth will not create excessive piping stresses.

The piping replacement will not change the functionality of this system, nor affect any of its failure modes. .

3.4.47 88-136*A&B, Main control Boards. .

The MR will remove two trend recorders in 1(2)C03 per unit, add one CRT in 1(2)C03 per unit and one charceter display in 1(2)C04 per unit.

110

Summary of Safety Evaluation: An evaluation is required because the proposed MR constitutes a change to the facility as described in the FSAR. The trend recorders being removed from 1(2)C03 are controlled via the plant computer. These computer outputs will be placed on existing hard-wired trend recorders in the Unit 1(2)

ASIP. The hard-wired parameters being removed from the ASIP to accommodate the computer outputs may be placed on the trend recorders by the operators, if desired.

The CRT being installed in 1(2)C03 in place of the removed trend recorder will be controlled by both the SAS and PPCS keyboards; making it useful during nomal plant operation as well as emergency situations. The 1x20 character ,

display will be an addition to the one that already exists '

in 1(2)C04.

All devices will be mounted Seismic Class 2 and all control board modifications will maintain the seismic response that existed prior to this change. ,

3.4.48 88-139 (Unit 1), Turbine Crossover Steam Dump System. .

This modification proposes to install a connection on the crossover steam dump system so service air can be used to resent any valve that sticks open. A similar change was installed on Unit 2.

Summary of Safety Evaluation: The FSAR states that the t resenting action is accomplished by using the reseat steam system. This system, however, has failed to adequately perform, and therefore,. service air will be used to >-

supplement the reseat steam if needed. A failure of the ,

proposed change would cause the dump valve to open, which !

is in the safe direction. This inadvertent opening of the dump valve would most likely cause reactor power to increase and would rely on the reactor protection system to mitigate the consequences. The probability of this happening, however, is no greater than it is at present, The proposed change does not affect the safety function of the dump valve (s). The material and components used for this change will have pressure and temperature ratings ,

l- required for use in this application. Materials used will be compatible with the existing valves.

3.4.49 88-152 (Unit 2), Feedwater System.

The MRs will install circuits to trip the condensate pumps by energizing their respective lockout relays and trip the l heater drain tank pumps by paralleling their stop circuits upon a containment high pressure signal. The MRs are I

needed to address a concern about a steam line break inside containment with a stuck open feedwater regulating valve. Termination of feedwater flow in the event of a steam line break inside containment and a stuck open 111

1 h

feedwater regulating valve is required to resolve LER 266/88-008. The containment high pressure signal which will actuate at or below 6 psig was chosen as the initiating signal. Tripping off the motors for the condensate punqps and heater drain tank pumps was chosen as the method of interrupting feedwater flow.

Sununary of Safety Evaluation an evaluation is required because the proposed MRs will alter a system, structure or component described in the FSAR.

The initiating signal is mechanized by using additional contacts on existing safety injection relays and switches.

The circuit used for safety injection is not interrupted or altered. Either tripping the heater drain tank pump 3 motors off-line or tripping closed a valve in their common discharge line could have been used to interlock a water flow from the heater drain tank. Tripping the pumps was chosen to eliminate the time delay associated with the valve closing. This results in a lower peak pressure during postulated accident.

l The circuit is mechanized to respond to a double-ended steam line break inside containment which results in a very fast increase in containment pressure. By calcula-tion, the containment high pressure actuation point is reached within 2 seconds after the break. For a smaller steam line break inside containment or a loss of coolant accident, it is possible that safety injection would have been actuated by another initiation signal and reset before containment pressure reaches the high setpoint. In these cases, the termination of feedwater flow would be provided by the main feedwater pump discharge motor operated valves.

The condensate and heater drain tank pumps could also be operated following a safety injection-reset. Under these circumstances, motor-operated valves on the discharge of the main feedwater pumps would have closed to provide the l redundancy to close the main feedwater regulating valves.

Also, the control room operators would have had time to analyze the accident situation and take appropriate actions.

l- If this circuitry is activated, it expected that the

! secondary system will be stressed. When water flow i

through the tubes of the Nos. 4 feedwater heaters is l.

stopped, the water in the tubes is expected to flash into steam, pressurize the system and open the safety valves.

Although this is not a desirable situation, it is no worse than what happens during a loss of AC power incident.

112 i

')

3.4.50 88-160*A (Unit 2), Containment Structure. l Modification 88-160 will install into Unit 2 a permanent penetration for both mechanical and electrical connections that are needed to support steam generator refueling maintenance ,

and testing work. This design package will install a !

preliminary penetration into an existing spare penetra- I tion; El. 32', pipeway No. 3. The existing caps will be j removed, both inside and outside of containment. The caps ;

will be replaced with 150 lb ANSI B 16.5 flanges. This !

penetration will contain the electrical cables and hoses ;

, needed for steam generator outage work. I J

Susunary of Safety Evaluation: An evaluation is required I because the proposed MR constitutes a change to a system, structure or component described in the FSAR. The revised configuration will meet the design, installation and ,

testing requirements of the original containment I

penetrations. This includes compliance with ASA N6.2-1965

" Safety standard for Design, Fabrication and Maintenance ]

of Steel Containment Structures for Stationary Nuclear Power Reactors" and NUREG-0737 containment isolation requirements (Section II.E.4.2). The penetration will be ,

painted on the inside of containment.

During times when containment integrity is required the flanges will be blanked off both inside and outside of '>

containment. The two blank flanges will provide the redundant containment barriers. To verify leak-tightness the assembly will be volumetrically tested. When ,

containment integrity is not needed but fuel motion is in ;

progress the penetration will be sealed with foam. The foam will be designed to provide protection for a refueling accident by providing a sealed penetration for HVAC pressure. Mid-loop concerns can be resolved by disconnecting the hoses and cables on the outside of containment and bolting on the blank flange in the auxiliary building within the half-hour timeframe of OP-4F.

In addition, the modification will provide access from pipeway No. 3 into the facade. This access will be to l:

l provide a path for the cables and hoses to get from the "

l_ facade into the pipeway and then into the new containment penetration. This will be done by boring two holes into the exterior pipeway wall. These holes have been evaluated and will not affect the structural integrity nor the structural barrier provided by the pipeway wall.

The holes will be covered or sealed when they are not in use. The FSAR discusses containment penetrations in several places. The following reviews the revised configuration in comparison to those section which are applicable.

113 l

.~ -

! I l

I 1

Section 5.1.2.2, Penetration Criteria - This section I states that the penetrations conform to kSA N6.2-1965.

The revised penetration will comply with this standard. l Section 5.1.2.4, Detailed Design - This section deals with i the evaluation of loadings on the containment by the.

penetration. For the revised penetration, the impact on i the containment structure has been evaluated by the con- j tainment designer and have found to not adversely affect i the containment structure. l Section 5.1.2.4, Penetration Details - This section' states ,

that the penetrations are built in compliance with the i ASME B&PV code Section III Class B, 1968 Edition all !

addenda. In addition this section states that the welds on i penetrations were radiograph, liquid penetrant, and local leak tested. All of these requirements will be met for i this revised penetration. j Section 5.1.2.6, Penetrations - This section restates !

previous sections. !

Section 5.1.2.6, Piping and Ventilation Penetrations - ,

This section states any potential leak path in the piping l also has provisions for individual testing. The revised penetration will meet this requirement. ;

Section 5.2, Containment Isolation Systems - The revised configuration of penetration from a capped off spare to a ,

permanently installed blank flange on the inside and

outside of containment will now classify the penetration. !

as a SPECIAL per the requirements of Note D of Table 5.2-1 of the FSAR. The revised penetration will be configured i similar to penetration 56 (containment pressure test) '

which is described in Table 5.2-1 of the FSAR. Thus the new configuration of the former spare penetration is i the virtually the same as existing penetration .

l configurations. The new configuration meets the redundant associated apparatus of discussed in the General Design Criteria, l

The new weld on the inside of containment will not have a leakchase channel. However, the weld will be volumetrically leak checked when the penetration is tested.

3.t.51 88-165 (Unit 2), Reactor Coolant System.

l The MRs will replace PT-420, the original reactor coolant system wide range pressure transmitter. Presently, PT-420 is a 0-3000 psig pressure transmitter with an accuracy of 115 psi. PT-420-is used in the overpressurization miti-gation system (OMS). When the OMS is actuated, PT-420 provides a signal to bistable PC-420C to open PORV-430 when RCS pressure increases >415 psig. The TS requirement is for actuation $425 psig. Therefore, it is possible to 114

violate TS with PT-420 in calfbration. To eliminate i this possibility, PT-420 will be replaced with a narrow range 0-1000 psig 12.5 psi Rosemount pressure transmitter. l Therefore, it is possible to violate TS with PT-420 in !

calibration. :

Summary of Safety Evaluation: An evaluation is required because the proposed modifications will alter a system, structure or cosponent described in the FSAR. The Rosemount pressure transmitter is qualified and suitable ,

for this application. It meets or exceeds all design ratings and characteristics of the originally installed transmitter with the exception of maximum overpressure. '

However, the transmitter has a maximum allowable over-pressure of 7500 psig, which is sufficient for use in -

this application. Additionally, the transmitter has been i seismically qualified in accordance with the SQUG guide-lines.

The transmitter is completely compatible with the power l supply and instrument loop for PT-420. Therefore, the :

replacement of PT-420 will not increase the likelihood ,

of a spurious actuation or the failure of the OMS. ;

i Since PT-420 will no longer be a wide range pressure channel, it will be removed from the RCS wide range !

pressure recorder, PR-420. It will be replaced by PT-420C, which is one of the three EQ wide range pressure transmitters. Since PT-420 and PT-420C share the same sensing location, there will be no change in RCS pressure- !

information indicated on PR-420. !

! A digital display will be added to the control board to provide indication of PT-420. This will add 5 pounds to the control board. Based upon engineering judgment, this ;

will not affect the seismic characteristics of the control room.

The replacement of PT-420 will not adversely affect the operation of the overpressurization mitigation system. ,

3.4.52 88-174 (Unit 2), MR 88-174 will change 2SV-466B and i 2SV-476B from a " universal" 3-way valve to a "normally closed" 3-way valve. .This change is to ensure the l l reliable operation of these valves. ;

A universal 3-way valve can have pressure supplied at '

! either port #2 or port #3. Port #1 is connected to the

" load." A normally closed 3-way valve has port #1 connected to the " load," port #2 connected to the supply air, and port #3 vented to atmosphere. The deenergized l condition is for port #1 aligned to port #3.

l' 115 s -

l Summary of Safety Evaluation: The function of the "B" solenoid valves is not specifically described in the FSAR.

Section 7.3 does refer to an override signal closing the feedwater valve when the average coolant temperature is below a given' temperature or when the respective steam generator level rises to a given value or upon a safety injection signal. Section 10.2 refers to the same signals closing the feedwater control valves.

Changing the "B" solenoid valve from a universal 3-way valve to a normally closed 3-way valve will not change the override signals indicated in the FSAR.

The solenoid valves and the feedwater control valves will be tested per ICP 5.11. ICP 5.11 is not specifically described in the FSAR.

The " slugging" feature provided by 2SV-466A and 2SV-476A is described in Section 10.2 of the FSAR. The MFRV l slugging is for operational considerations only. There I are no safety-related functions associated with the

! slugging. The primary concern is to maintain the inventory of water in the steam generator using hot feedwater and to keep the steam generator water level in the narrow range level. The NSSS indicated that the 5 second opening time for the slugging of the MFRV is not critical. The opening time can be increased as required.

If the opening time is increased, we could consider lowering the closing temperature to about 550*F. This will mainta'in the amount of water entering the steam generator. Care must be taken when choosing the closing temperature because too low a closing temperature could cause depressurization of the primary side (over cooling) and may initiate SI. Since all of the steam dumps are not required to open after a reactor trip, we may consider optimizing the steam dumps.

3.4.53 88-176*C (Unit 2), Steam Generators.

The steam generator primary manway diaphraps are removed

during each steam generator primary side inspection.

These diaphra ps are normally in contact with reactor

coolant and are highly contaminated. . Typical dose rates l

for a diaphrap are 10 R/hr gamma contact 100 R/hr beta contact; 1 R/hr gamma at 18", and 7 R/hr beta at 18".

! Past practice has been to store the diaphra ps behind a concrete shield wall inside containment; shield them with lead blankets; rope off and conspicuously post the area; and install a flashing light as a warning device. This practice, while allowable by plant procedures and Technical Specifications, is not desirable and may possibly not be accepted by the NRC for routine future applications (reference NRC Information Notice 88-079).

116 1

)

Therefore, a change in our steam generator diaphrap (

storage practice is needed. _ MR 88-175*C and 88-176*C will ;

provide lockable, shielded storage containers for i diaphraps on each contailment El.10' platform.

Summary of Safety Evaluation: The proposed MRs will alter a system, structure or component described in the FSAR; l and further, will alter procedures described in the FSAR. l Design requirements of the shielded storage containers j are: '

a. Four containers will be provided: Unit 1 "A" loop,

Unit 1 "B" loop, Unit 2 "A" loop, and Unit 2 i "B" loop.

b. Each container is capable of storing two diaphra ps and the plastic bags used for contamination control.

c. The container's shielding decreases the dose rate from the two diaphraps to <1000 mr/hr at 18" (the calculated dose rate from the container is equivalent to the normal containment El. 10' platform general j area dose rate).

d. The container is a seismically-mounted, poured-lead shield, with the lead contained in stainless steel,

e. The stainless steel container will not require the use- '

of post-DBA qualified paint. These design requirements i assure that the container does not introduce any new '

l hazards to the plant.

f. The evaluation and calculations provided in the design fi indicate that there will be no adverse effects as a '

result of this installation.

The lockable, shielded storage containers for the steam generator primary manway diaphraps meet the requirements of NRC Information Notice 88-079.

3.4.54 89-006 (Unit 1) and 89-010 (Unit 2), Reactor Coolant l System.

The MRs will provide a reactor vessel (RV) low level alarm for use during mid-loop operations. These alarms are being added in response to NRC Generic Letter 88-17, dated October 17, 1988.

Summary of Safety Evaluation: An evaluation is required because the MRs will alter a system, structure or component (SSC) described in the FSAR and could affect the function or method of an SSC. The alarm will be supplied by I replacing the existing alarm bistable (LC-447) with a dual .

level bistable. The new bistable is made by the same i 117 l

i

~ ~

i i[,

manufacturer as the existing single level bistable (RiS).

The existing, " reactor vessel level hi" annunciator window will be changed to read, " reactor vessel level hi-lo," and either a high or low level signal will cause the annunciator to alano.

The low level alarm setpoint will be set by the operators once the mid-loop operating level is determined, in the same manner the existing high level setpoint is determined.

A switch is also being added inside 1(2)C04 near the bistable to bypass the low level alarm so the alara is cleared during normal plant operation while transmitter LT-447 is isolated. The switch will be administrative 1y controlled to prevent the alarm from being bypassed during normal operation. The consequences of failure involving the new switch will be reduced by the addition of a' redundant RV level instrumentation loop being added per NR 89-023/024. The second alarm bistable will have a separate switch to bypass its low level alarm.

The addition of the low level alarm will not affect the

availability of the existing high level alarm or increase the possibility of failure for any existing equipment.

3.4.55 89-018 (Unit 2), Electric Generator.

The MRs will remove the 32% generator power relays and f will replace that circuit with a new circuit which will l trip the generator lockout relays when the generator I

output breaker opens at any power level. The new circuit will be able to be defeated during startup using the turbine trip-to-lockout switch.

Summary of Safety Evaluation: The MRs constitute a potential change to the facility or its operation as described in the FSAR. The only significant change resulting from these MRs is that the lockout relays (86-TG01 and 86-X01) will be tripped when the generator output breaker opens at any generator power level, instead of tripping only at power levels above 32%.

The reason for the existing circuit is to prevent damage .

to motors from overspeed resulting from turbine overspeed j upon loss of electrical load. To lower the setpoint, as '

l the new circuit will (to 0% effectively), would farther i

reduce the amount of overspeed, and therefore, would be more conservative.

The probability of the new circuit malfunctioning will not be greater than that for the existing circuit. Loss of power to the auxiliary relay in the circuit will not l

prevent the circuit from tripping the lockout relay. As with the existing circuit, two redundant trains will be used to increase reliability.

118

~~ ~'

_ _ _ __ .__ _ . _ _ _ a

7 J

l There are no sections of the FSAR or Technical l Specifications which refer specifically to the 32%

generator power relays. No changes to these documents :

are required.

Adequate train separation will be maintained. Seismic adequate of main control board CO2 will be maintained.

The MRs will not impact the turbine-generator overspeed ,

analyses since the change is below the load level in which an overspeed event can be achieved.

This MR will result in a change in operating philosophy in that the turbine trips to lockout switch will be placed in defeat before the generator breaker is opened for turbine i overspeed testing. This will prevent turbine trip events from causing a lockout. This would result in a reverse '

power trip if a turbine trip occurred while the generator breaker was shut. This should not be a significant '

reduction in protection as the trips defeated have a 60 ;

sec. time delay and the reverse power is similar. There are no safety concerns in view of low power level and the !

non-safeguards status of the turbine trip feature.

P 3.4.56 89-023*A*B* Add 01, Reactor Coolant System.

The purpose of MR 89-023A, 89-023B and 89-023 Addendum 1 is to: (1) Install a new reactor coolant system (RCS) ,

reduced inventory level channel (LT-447A) that is ;

basically identical to existing level channel LT-447.

This new level channel is being added in response to NRC Generic Letter 88-17. (2) Replace the instrument manifold valves for LT-447 to make the valving configuration the same as LT-447A. !

l Summary of Safety Evaluation: LT-447A will use the same connection as LT-447 for the variable leg. The reference leg for LT-447A will be independent from the reference leg I for LT-447 in that a single valve is not capable of isolating both reference legs. The new reference leg will tap into the pressurizer steam space sample line between l manual valve SC-950 and containment isolation valve SC-951 l

just downstream of SC-950 in the upper pressurizer cubicle. Welded connections are being used up to the first normally closed (during power operation) valve. New tubing with mechanical fittings will be run to LT-447A.

l In order that a pressurizer steam space sample is not inadvertently taken when LT-447A is in use, which could cause a perturbation in the level indication, a new in-line valve is being installed between the reference leg tie-in point and valve SC-951. This in-line valve will be procedurally controlled so it is shut when LT-447A is in use and normally open when LT-447A is not in use. Since pressurizer steam space samples are not periodically 119 m.- _.

i l

l l

required or taken, there is not a concern in isolating j this sample line during the periods of time when LT-447A 1 is needed. In addition, this new in-line valve will J provide a test boundary for performing ORT #37 (seat )

leakage test on SC-951) such that valve SC-950 and the new '

reference leg root' valve will not have to be shut to perform the ORT. J Mechanical fittings will be used throughout the instal-lation after the first nonnally closed valve. This !

. portion of the line will not experience severe thermal i transients. Welded fittings will be used in line portions !

that may experience severe operational thermal transients. !

The acceptability of mechanical fittings is based on their.

present use in the variable leg of LT-447 and in the pressurizer steam space sample line where Swagelok ;

fittings currently exist outside of the pressurizer cubicle both upstream and at valve ISC-951. In addition,

the newly installed fittings will be hydrostatically tested / leak checked at 2500 psig before returning the affected lines to service.

Review of FSAR Section 9.4.2 revealed the incorrect statement that socket-welded joints are used throughout the sampling system except at sample vessels, sample panels and sample sinks. It is felt appropriate to modify this statement in the FSAR to include the appropriate use j of mechanical fittings. There is not a safety concern with using mechanical fittings in 3/8" diameter lines connected to the RCS that do not experience severe operational thermal cycling since: !

a.. Mechanical fittings have been proven to be highly reliable by their extensive use throughout the plant,

b. As stated in FSAR Section 14.3, the rupture of a 3/8" line in the RCS can be accommodated for by the-charging pumps; thereby allowing an orderly shutdown.

More specifically, the FSAR states that typically one charging pump can accommodate for a 3/8" diameter hole in the RCS. The resultant radioactive impurities contained in the discharged coolant would be contated to the containment.

The installation of the mechanical portion of LT-447A and the changeout of the instrument manifold for LT-447 with-Whitey valves will be acconplished per ICP 11.419. The ICP requires the refueling cavity to be flooded before isolating LT-447. Since LT-447 is only required during reduced inventory conditions (3' below the reactor vessel .

flange or lower), the isolation of LT-447 will not violate its availability requirements. The ICP requires a post-installation leak check of all newly installed fittings at 2500 ps!g. The ICP also requires that the 120

. , _ . , _.r.- .-s.- a u - - . a - . .- , .. _

l L

[ variable legs of LT-447 and LT-447A be filled and vented, !

and the reference legs be drained after installation l

acceptance so the level channels are prepared to return to service. l The new level tr'ansmitter, power supply and alarm bistable l are of the same model or type as existing level channel ;

LT-447 which has proven to be a reliable instrumentation l channel. The level indicator is a dual input indicator j which will display the level for both LT-447 and LT-447A, ;

and is used in other applications on the control board. '

The new instrument channel (LT-447A) will be powered from

~the red instrument bus, while the existing channel ,

(LT-447) is powered by the yellow instrument bus.

Separation of the channels, in accordance with IEEE-384, will be maintained to the extent possible.

1 The changes made do not add any new failure modes or I channel functions (besides the redundant level indication channel) that did not exist in level channel LT-447.

Therefore, the consequences of failure are not increased.

In fact, the additional level channel provides added .

assurance that RCS level indication will remain available i during reduced inventory conditions.

1 3.4.57 89-024, 89-024*A&B and 89-024-01&O2 (Unit 2), Reactor l Coolant System.

Unit 2 is being modified in the same manner as Unit 1. J The MRs will install new level transmitter LT-447A and e change out the existing LT-447 instrument manifold with l l Whitey valves. In addition, there has been an increase in !

L scope for these MRs to accomplish the following: (1) l Relocation of the reactor vessel flange leakoff line t manual valve (RC-522); (2) addition of a drain valve in l this liner and (3) elimination of the tie-together line l between the reactor vessel flange leakoff line and LT-447s l- variable leg drain valve. ,

l Sununary of Safety Evaluation: An evaluation is required because a system, structure or component as described in ,

j the FSAR will be altered. The evaluation documented in 1

SER 89-054 applies in its entirety to this report except ,

as revised or supplemented in the preceding and following paragraphs.

Manual valve RC-522, which is the final valve to the RCDT L

from the RV flange leakoff line will simply be moved upstream 15' to alleviate ALARA concerns. In addition, the 3/8" tubing line between the RV flange leakoff line and manual valves RC-522A, 523 and 525 which allowed the RV flange leakoff line to be drained through LT-447s variable leg drain valve (RC-525A) will be eliminated, and.

a 3/8" drain valve will be added to the RV flange leakoff line downstream of temperature element TE-418.. The 121 -

1:

i n

i l

addition of this drain valve will allow i.he drainage of !

the RV flange leakoff line to be visually verified without l affecting LT-447 or LT-447A, and it will also allow a i procedure to be simplified (RP-1B), thus red'xing the I chance of valve mispositioning. These installations will !

involve 3/8" mec'hanical fittings and will be QA and seismic. An MWR will be used to control work and a post-installation leak check performed as practical to j 2500 psig. Since the modified portion of the RV flange j leakoff line is downstream of temperature element TE-419, j this change cannot significantly impact indication of RV j flange leakage, j l

The alarm bistable being used in level channel 2LT-447A is )

the same model which was installed in level channels 1 ILT-447 and ILT-447A, and which will be installed in level l channel 2LT-447 per MR 89-010.

]

A small manometer-type loop seal exists in the reference leg for LT-447A at the reference leg tie-in point at the >

top of the pressurizer. The existence of this manometer-type loop seal is justified based upon the administrative ,

controls applied to the use of this level channel.

Initially the reference leg is blown down with nitrogen in )

OP-4D, " Draining the Reactor Coolant System." A signoff

step in OP-4D verifies a positive outflow of nitrogen at .

the transmitter's reference leg drain valve. After the :

initial blowdown, if the pressurizer is flooded up offscale high, it will again be required to drain down ,

using OP-4D. A precaution is being placed in OP-4D which L states that the reference leg may not be self-draining since a small inverted manometer and a very long run of small diameter tubing are present.

In addition, the level indication from LT-447A will be cross-checked against the indications from LT-447 and the ]

tygon hose / local level indicator (when available), and be recorded in the operations log and in specific steps of' !

OP-4D. If a deviation of 3 inches or more is noted, action will be taken. It should also be noted that even if the reference leg was sloped completely downward, any water entering the line would still affect the level -

indication; and, after initial blowdown of the line, the >

upward sloping tie-in may, in fact, preclub water from -

entering the line.

3.4.58 89-027, Radiation Monitoring System.

MR 89-027 will remove bubble memory boards from the control terminals (cts). The design intent of the bubble ,

memory is to provide a CT-based backup memory for all radiation monitoring system (RMS) channel parameter flies ,

in the event that the memory (RAM) in a DAM /SP~NG was lost.

122 l "

L Summary of Safety Evaluation: An evaluation is required I because the change will alter a system, structure.or i component described in the FSAR and further could affect ,

the function of a system described in the FSAR. The channel parameter files will automatically download from j the bubble memot;y in a CT to a DAM /SPING upon a DAM /SPING ;

power-up evolution. The automatic downloading feature is intended to eliminate having to manually reinsert channel

parameter file data via the CT keypad if the RAM memory in !

a DAM /SPING is lost or intentionally erased. This is a :

very infrequent occurrence and if it is necessary to do, ,

the parameter files will be loaded manually.

Removal of the bubble memories will enhance the opera- i bility of the RMS. j 3.4.59 89-041 (Unit 2), Instrument Buses.

The scope of modification 89-041 is to swap the instrument bus supplies to 2RCSI (2C128) and 2RCS2 (2C129), such that !

2RCSI will be supplied from 2YO4 (yellow instrument bus) and 2RCS2 will be supplied from 2YO3 (white instrument bus), as in the case for Unit 1. This will eliminate the extension cord currently running between the two panels in order to supply some of the instruments inside 2RC$2 with white instrument power.

In doing this, instrument loop P420 bill be supplied frh ,

the yellow instrument bus so both trains of Unit 2 LTOP j will not be dependent on the white instrument bus, as t identified in NCR #89-010. This change will also make the '

analysis for loss of instrument bus the same for both i Units, in regards to the RCS racks. ,

Summary of Safety Evaluation: An evaluation is required -

because the proposed change could affect the function or method of a system, structure or component described in

the FSAR. The work required includes rerouting the instrument power supply cables to the appropriate racks and rerouting the computer signal input cables to the j appropriate computer MUX.

1' Proper cable separation, in accordance with IEEE 384 and-existing practices is being maintained. The safety-related cables going to 2RCS2 are already labeled as if they are associated with the white instrument bus and are routed in the same cable trays as other safety-related cables associated with the white instrument bus. The' ,

existing condition, in which some of these cables are actually associated with the yellow instrument bud, is wrong and will be corrected by this modification.

l 123

?

)

1 Redundancy of similar instrument loops is also being maintained in that all redundant instrument loops will be j supplied from instrument buses other than that of the '

redundant loop.

Installation wil'1 tiake place while Unit 2 is shut down for refueling and the primary system is depressurized. Under ,

these conditions most of the RCS instrumentation will not be required and can be removed from service. Installation 1 will be controlled by an ICP so the effect on plant operations can be controlled. 1 Swapping the instrument power supply to the two racks will ,

result in transferring approximately 6 amps from 2YO3 to :

2YO4. This will then increase the loading of station :

battery D106 by 6 anps.- Since the existing battery load j is substantially less than the 795 amp-hour rating of the batteries (at a one-hour discharge rate), the additional load will not affect the reliability of the batteries. ;

l Changing the power supplies to the instrument loops will I not change the function or reliability of any instrument loop, r 3.4.60 89-045 (Unit 1): Replace the 100 ohm dropping resistors for the input to the ATsp1 modules with 200 ohm resistors.

The change is required to develop the instrument gains necessitated by the change to TS constant K2, which was included in TS Change Request #127. The resistor change J will allow the ATsp1 module to develop a gain of 0 to 2 l l for the Tavg input. This gain range corresponds to.the i original design range of this instrument.

.l The proposed setpoint changes will revise the ATo's to

, more accurately reflect measured channel ATs for Unit 1. .

These setpoints are contained in STPT 1.2 and 1.3 of the ,

Setpoint Document. The new ATo's were selected based on ,

the lowest value of loop AT for each channel over the past few years of operating data for Unit 1<. Therefore, the I

generated AT setpoints will be conservative in comparison ,

to the TS-required setpoints.

Summary of Safety Evaluation: An evaluation is required ,

because the proposed change could affect the function or i

l. method of a system, structure or component described in L

the FSAR. ATsp1 and ATsp2 provide core protection by ^

L preventing DNB and exceeding 108% of design power density, ,

L respectively. They do not serve to prevent an accident, but to minimize the consequences of an accident.

The change to channel-specific ATo's will more accurately ,

reflect actual loop conditions as sensed by the individual instrument loops. The existing ATo's cause the reactor '

trip, runback and rod withdrawal block setpoints to be 124

.. i

,w

e - __ __ ___ __ _ _ _ _

.- t ,p

, , . 3 3: ,

1 n

o overly conservative.; The new values have been selected a such that the AT setpoints generated by the protection ,

system will be conservative compared to the Technical-Specification setpoint requirements. The change to channel-specific ATo's and the. replacement of the

, ', resistors do not' pose any unreviewed safety questions.- 3 The' resistor change will allow the ATspl modules to-g develop the inatrument gains necessitated by the change to TS constant K2, The resistors used are of the same-type as used elsewhere in'the reactor protection system. The ratings of the resistor'and the resultant' loop load- )

increase.are within the capabilities of the affected g equipment, and will'not cause'a failure of the equipment.

3.4.61- 89-048 (Unit 3) & 89-049 (Unit 2), Containment Ventilation.

The MRs will install Whitey valves'to be used as Mst ;

vents on both sides of the check valves on the instrument '

Sir lines to the purge supply and exhaust valve boot -

seals.

-Summary of Safety Evaluation: An evaluation is required because the MRs could result in a potential' change to the-

~

facility or its operation as described in the FSAR. One of the installations on each valve will be inside the ;

pressure boundary of the boot seal. This' installation is mounted seismically to assure that it does not adversely.

affect the pressure boundary of the boot seal during a seismic event. l All materials meet or exceed the original design :

specification for the system. Therefore, the probability r of failure is not increased; and therefore, containment' intcgrity will not be challenged.

3.4.62 89-053 (Unit 2), Sofety Injection System.

MRs 89-052/053 welds a threaded pipe stub into the valve body of safety injection accumulator nitrogen. fill vent valve, HCV-957. The threaded connection will allow-the volumetrics tester to be directly connected to the valve for the performance of ORT #32.

Summary of Safety Evaluation: An evaluation is required

~

because the proposed modifications will alter 'a system, structure or component described in the FSAR or alter the function thereof. HCV-957 is a remote operated valve which serves as a containment isolation valve (CIV) and as a relief valve protecting the piping between inside containment CIVs 834A and 834B and outside containment CIV i 846 from overpressurization from the nitrogen. supply 1 connected downstream of CIV 846. There it a remote possibility that the isolating components of HCV-957 could be damaged when welding in the pipe stub due to excessive 125 1

- a . --

c I r

I

4 - heating. To decrease the possibility of damage, the welding will be done with the valve open to lessen the l amount of heat buildup. Since the valve will be open j

while welding in the pipe stub,-the installation will be .1 e done with the plant in the cold or refueling shutdown l L'

condition at which. time containment integrity is'not l required by the Technical Specifications. After i

' installation and prior to returning HCV-957 to service, l HCV-957 will be leak tested per ORT #32 and stroked to l ensure the operability of the valve was not affected by the modification.

. J The addition of the threaded pipe stub to HCV-957' will :

have a negligible affect on its relieving capacity. ,

since the pipe stub's inside diameter-is the same diameter- t as the valve's inlet piping. It is not considered possible that the' installation work could affect the setpoint of HCV-957. Therefore, the stroke testing.of the valve is considered adequate to ensure that the valve operator j is free to move and will open at its setpoint if required. {

To prevent inadvertent capping or plugging of HCV-957, ~ a step will be added to ORT 32 and CL-7A which verifies that the pipe stub has not been capped or plugged. .In addition,.

a placard stating that HCV-957 is a relief valve and is not to be capped or plugged except during ORT 32 will'be hung from the valve. Also, ORT 32 will be modified to s ensure that the nitrogen spoolpiece is' removed or_the nitrogen 12-packs are disconnected prior to capping the relief valve. Also note that testing is done when -'

containment isolation is not required (normally).

Based on these controls, the probability of the relief port being inadvertently blocked is considered acceptably low.

Since the welded-in pipe stub does not breach the pressure boundary, no pressure testing is. required per PBNP 3.2.5.

. During performance of ORT 32, any leakage from the L threaded pipe stub would add to the leakage measured through the valve, which is conservative.

The threaded pipe stub will be made of stainless steel to -

4 preclude rusting of the pipe threads, Welding procedure L WP-4 will be used to join the stainless steel pipe stub to .

L the carbon steel valve body. Since the pipe stub is open to atmosphere and is a free end, the slight difference in thermal expansion coefficients of the two steels is not a

. concern.

The placard and pipe stub will add approximately 0.7 pounds to the valve which is considered to be an l insignificant weight addition to the seismically-supported li valve and is within the Bechtel small bore piping seismic >

guidelines.

126

3.4.63 89-054 (Unit 1), R xcter Cooltnt System.

NRC IE Bulletin 88-011, Pressurizer Surge Line Thermal Stratification, required licensees to visually inspect the pressurizer surge line for gross discernable distress or structural damage including piping, pipe whip restraints, and anchor bolts during the first available cold shutdown.

This inspection was performed on Unit I with the system at 500*F. The thermal gap between the surge line collar and wrip restraint hardware at location R-4 appeared to be cloh d and at R-5 was 1/8". No discernable support, collar, lug, or pipe distress was observed.

D nstigation into the specific design of the pipe whip a '.t.aints was initiated to determine the gap requirements.

'..a ARC IE Bulletin *19-14 stress analysis was reviewed to q fetermine t* fredicted thermal displacement at the two j locations. ance the restraints were installed for only ;

pipe whip, Westinghouse did not model them in the weight, l thermal or seismic analyses because they would be inactive H L with' sufficient gap. Further, NSEAS performed a thermal l

analysis to predict displacement in excess of gap design. ,

L Should the actual piping displace as predicted by the 1 analyses, the analyses would no longer be considered valid since R-4 and R-5 may become inactive in thermal stratifi ' ,

L cation and seismic conditions. -l L

! This modification will ensure that sufficient gap will.be i

! available such that contact with the whip restraint-hardware will not occur. Removing the east half of each collar is acceptable since.only the west half is, required by design to protect the steam generator from pipe whip.

Therefore, removing the east side section of each collar will ensure the original design will be maintained and the piping stress analyses are valid. -i Summary of Safety Evaluation: An evaluation is required j because the proposed change could affect the function or !

method of a system, structure or component described in l the FSAR.

Measurements of the. actual surge line displacements at l locations R-4 and R-5 will be made during RCS heatup (cold-to hot). These measurements will confirm whether.or not !

actual collar / restraint contact has occurred since original l plant construction at either or both of these locations. 1 Appropriate stress and fatigue evaluations will be performed. ;

should measurements conclude-that significant contact has ;

occurred. These evaluations, if necessary, will be .

i incorporated-into the stress and fatigue analyses required to respond to NRC IE Bulletin 88-11-for PBNP.

Since no physical distress or damage was observed, any ;

stress contributions from possible contact have not 'l resulted in any discernable yielding of restraint, collar, 127 i

W " " + t - - - __ _____ _ -

_______m __.____.______________.a

lug, weld, or piping material. .It must be recognized, however, that contact.could have induced local stresses-above code allowable and undetectable yielding may have

, occurred.. Based upon the physical geometry of the piping and support system (Sch 140 piping,1-1/2'? thick collar ;

and 1" thick x 8" long lugs welded on both sides), we would not expect significant yielding on~a local basis ,

L such that a safety concern exists. ' Additionally, I contributions to fatigue life of the surge line piping L could be more significant than previously included in the >

original design.

6 3.4.64- 89-061 (Unit'1), Reactor Coolant System. .f MR 89-061 will relocate the regulator for the nitrogen to- f 6 ,

the Unit 1 pressurizer power-operated relief valves .

(PORVs 430 and 431C). This change is being made to '

improve the opening time of the PORVs for low temperature s overpressure protection (LTOP) considerations.

Summary of Safety Evaluation: An evaluation is required r because the modification involves a potential change to the facility or its operation as described in the FSAR.

The resultant configuration will be the same as was a provided in the original' design as found in MR IC-147.

However, with the regulator. closer to the valve, the ,

amount of head loss in the 3/8" tubing will be reduced. l The only new consideration as'a result'of this change ~is- [

the extension of the 3000 psi pressure boundary into the -

pressurizer cubicle. This concern is minimal given that. -

J the line.is installed to the requirements of B31.1-1967; is capable of the pressure; and that the equipment is '

seismically supported. The regulator will not affect any i critical equipment.

The system will be leak checked as part of the post-'

installation testing and the PORVs will be stroke tested to ensure adequate opening time.

3.4.65 89-062 (Unit 2), Nitrogen System. a MR 89-062 will rearrange the components of the nitrogen gas backup to PORV-430 and 431C. The change is being made to improve the opening time of the PORVs for low tempera-ture overpressure protection (LTOP) considerations to meet-the opening time requirements per the LTOP analysis by ,

Westinghouse.

Summary of Safety Evaluation: The changes to the existing system involve the relocation of the nitrogen gas regulator, presently installed in the vicinity of the nitrogen cylinders on El. 66' of containment to the pressurizer cubicle. The new location, adjacent to the PORVs, will greatly reduce the friction losses associated with the 128 av

3/8" tubing. In' addition' to relocating the regulator, the )

solenoid ~ valves admitting air / nitrogen to the PORVs will~ I be replaced with a model that has the same electrical (

characteristics, but with much less pressure drop. The-

- new solenoid is purchased QA per the plant EQ program and will be seismically mounted. The tubing will be supported: '

in accordance with the seismic tubing support guidelines'. 1 The revised configuration will not. change-the system function; it only changes the sequencing of components on '

the nitrogen supply tubing to reduce friction'and assure that the PORVs will open within~the time required per the LTOP analysis.

The 3000 psi nitrogen tubing will extend into the pres- 3 surizer cubicle, but is a minimal concern given that the -

tubing is capable of the pressure, seismically supported, installed to the requirements of B31.1,'and will be leak tested as part of the post-installation acceptance testing. 4 The replacement solenoid will affect the head losses for the air supply system.- The opening time required per the LTOP analysis is applicable to operation by air and changes per MR 89-062 will serve to reduce the opening-time-of the PORVs when operated by air as well as by nitrogen.

SMP 1007 will control the installation and address Technical Specification requirements for taking the PORVs ,

out of service while performing the modification. !

I J Per Amendment #45 to Facility Operating License DPR 24 (Unit 1) and Amendment #50 to Facility Operating License DPR 27 (Unit 2), the LTOP system was approved by the NRC for installation at PBNP. The original safety evaluation report received from the NRC included the following design I -- considerations:

> a. System shall meet Seismic Class 1 and IEEE-279 standards,

b. Backup gas cylinder shall have enough gas-to cycle 139 times.

,, c. Opening time for subject PORVs shall be less than or li equal to two seconds. The time criteria was used by the NSSS in performing the LTOP analysis.

The modification does not change or affect the existing system from the standpoint of considerations 1 and 2.

. The proposed modification was initiated to meet design

.. consideration 3.

l 129

r ,

i l

l I

The two second opening criteria was established by the I NSSS. _The time is based upon an evaluation of RCS volume, LTOP Relief Valve Pressure Setting and mass inputs associated with the activation of one SI pump. The analysis has been reviewed using site specific input I (PBNP) and the t'wo'second opening time was valid for PBNP. I L The lower limit or minimum opening time for PBNP has not L been established. Based upon discussions with the valve manufacturer opening times as low as 0.75 seconds will not cause damage to the valves. The problem involves water-hammer to the downstream piping. The faster the valve E opens, the greater the water-hammer.

For a similar nuclear plant, the NSSS supplied a minimum valve opening time for 1.65 seconds for at-pressure !

conditions for water-hammer considerations. It is our engineering judgment that is a starting point for deter-

mining a minimum time criteria for PBNP. The NSSS'has i been contacted to provide a minimum opening time based

? upon their original generic analysis. If this time is more conservative than the proposed value (i.e., the=NSSS time >1.65 seconds), the new value will be used to establish an acceptable range of operation while con-sidering potential water-hammer effects on downstream c piping. ,

The issue of pressure assisted opening has been considered.

The valve will be tested per_the modification-with essentially no pressure under the seat, although during

t. normal operation the valve will have 425 psig under the seat. The system pressure will have the effect of assisting the valve in opening quicker.

L.

There are several varying approaches in allowing credit '

E for pressure assist. The NSSS investigated this aspect of l the modification during the analysis-of another nuclear plant's system. The factor used was in the range of 8% on-the 3.2 second maximum time. Since at PBNP, the maximum

(. opening time is established as 2 seconds, the factor can be applied in the range of 15%. If the test results were 2.3 seconds with no pressure under the seat, then it can be assumed that the actual opening time, with pressure under the seat would be 2.0 seconds or less. The actual time difference will be either calculated conservatively or actually tested during the Unit 2 shutdown. Once the pressure assisted time difference is established, it will ,

l. be added to the acceptable range and the SMP 1007 test results will be accepted on that basis.

l- 3.4.66 89-065 (Unit 1), Safety Injection System.

Modification 89-065 will remove two nonconforming 150#

l' flanges (reference NCR N-89-090) on the SI line, 4-SI-1501R-3 and -2 (with valves SI-D-2 and SI-D-13, 4

130 L

.-. . ~ - __. - -. . -.

, i a'" '

respectively). The NSSS piping' specification Class 1501 --

-requires'the use of a 1500# flange. The_modificat.on will l

remove the flange and replace it with a stainless steel l 3000 #, 3/4" threaded coupling. This configuration is in' H accordance with Westinghouse piping specification 1501 and- )

B31.1-1967.

J j

Summary of Safety Evaluation: An evaluation is required ;

because the modification alters a system described in=the .i FSAR. The cap will weigh less than the existing 150#

flange and will not affect the IEB 79-14 analysis of the SI line or the seismic capability of the drain connection.

Section 6.2.2 of the F5AR states that all joints are welded except those flanged connections tabulated in 4 Table 5.2-11. This new connection will not in itself !

cause the leakage values of Table 6.2-11 to significantly change (secondary recirculation fluid boundary).

3.4.67 89-082 (Unit 2), Reactor Coolant System. I The proposed change is to replace the 100 ohm dropping

resistors for the input to the ATsp1 modules with 200 ohm resistors. The change is required to develop the instru-ment gains necessitated by the change to TS constant K2,.

which was included in TS Change Request #127. The resistor change will allow the ATsp1 module to develop a gain of 0 to 2 for the Tavg input. This gain range '

corresponds to the original design range of this instru-ment. The change to channel-specific ATo's and the replacement of the resistors do not pose any unreviewed safety question.

Summary of Safety Evaluation: An evaluation is required because the proposed change could affect the function or method of a system, structure or component described in the FSAR. The provisions of the safety evaluation per-L formed for an identical modification to Unit 1 per MR 89-045, apply to this Unit 2 modification, s 3.4.68 89-083 (Unit 2), Reactor Protection.

The modifications will remove the delta flux function

{f(AI)} from the overpower AT setpoint (OPDT) for all four protection channels of each unit. The end result of the modification (s) is that the OPDT will only vary as a function of Tavg, f(Tavg), and then only when Tavg is

>573.9'F.

x Summary of Safety Evaluation: An evaluation is required because the modifications will alter a system, structure l or. component described in the FSAR. Furthermore, procedures referenced in the FSAR will be changed. The elimination of the f(AI) input to OPDT is necessary in order to utilize the full delta flux operating envelope 131

L l j

1

.).

requested under Technical Specification Change Request 1

127. The acceptability of removing the f(AI) term from

the OPDT setpoint with respect to plant / reactor operation l f

and safety was addressed in the TS change submittal. This safety evaluation report will address only the physical changes to plant' hardware required to eliminate the' f(AI) function.

The modification (s) will consist of reviring the signal ,

~ cables for the OPDT summer, the AI current source, the. i f(AI) high current selector and the overtemperature AT J (OTDT) compensator. Following the modification (s), the'.

~

f(AI) signal will only be input into the OTDT compensator.

The f(AI) signal to the OPDT summer will be replaced with a constant 10 ma input from the AI current source. This ,

10 ma input corresponds to.f(AI)=0. This input to the. .

summer is necessary to maintain the proper relationship between the f(Tavg) input to the OPDT summer and susuner '

output following the removal of the f(AI) signal.

This modification will have a small.effect on plant hardware because there is an increase in the load on the AI current source and a corresponding decrease in the load on the f(AI) high current selector. This load change is

- associated with adding the OPDT summer to the AI current source loop. The input resistance of the OPDT. summer is 100 ohms. This will result in a total load on the AI current source of 200 ohms. The 6I current source is designed to provide an output of 10-50 ma'into an output load of up to 600 ohms. Likewise, the load on the f(AI) .

high current selector will be reduced to 100 ohms.

Therefore, following the modification (s), the AI current sources and the f(AI) high current selectors will still be

' operated within their capabilities. ,

The physical changes will not result in an increased

~

l probability of component failure and will not reduce the i j margin of safety as described in any licensing document.

L 3.4.69 89-089 (Unit 1)', Main Control Boards.

L The modification will relocate the Tavg steam dump channel alert annunciator from IC031Dl-7 to IC031E-2 4-2.

Summary of Safety Evaluation: An evaluation is required because the proposed change involves a potential change to the facility or its operation as described in the FSAR.

Control boards are mentioned in the FSAR. The modifica-tion is being implemented to move the annunciator closer

, to its associated controls (steam dump mode selector switch and condenser steam dump controller). There is no L change in function or operability of the annunciator.

l

. 132 L

w4

'f G t

= .

3.4.70 89-090 (Consoon), Main Control' Boards.

The modification will relocate the Tavg steam dump channel alert annunciator from Row l' Column 5 of annunciator panel-

2003 2D to Row 4 Column 6 of annunciator panel 2003 2E-2.

Summary of Safety Evaluation: An evaluation is required '

l because the proposed change involves a potential change to ,

the. facility or its operation as described in the FSAR. .

Control boards are mentioned in the FSAR. The modification ,

is being implemented to move.the annunciator closer to its associatedsteam condenser controls dump (steam dump) controller modeisselector

. There switch and no change.in-function or operability of the annunciator.

3.4.71 89-100 and 89-100*A, Reactor Makeup Water System. ,

MR 89-100 proposes to' abandon in place the."A" reactor makeup water tank, but keep "A" reactor makeup water pump

[i in service. This is being pursued at the recommendation L u" l of INPO, to ensure positive control to prevent the inadvertent use'of the water in the "A" tank. Three p'

design packages will be used to do this. modification:

one will change the low' level pump trip function on the-

"A" pump to allow the "A" tank to be drained. The second y will physically isolate the "A" tank using blank flanges, o caps, etc. The third will remove the level indications.

from the contro1~ boards and make other miscellaneous.

electrical. disconnections. .

p l- Summary of Safety Evaluation: An evaluation is required-l' because the MR Will alter a system, structure or component L described in the FSAR and will also alter procedures described in the FSAR. The."A" tank is presently isolated from the rest of the reactor makeup. water system because the water in it is out of specification (too much oxygen),

and it would not be cost-effective to fix the bladder-in the tank. The tank is not needed for continued- 1 operation because there is enough capacity for the system in the "B" tank alone, and it has backup capability via l

direct connection to the DI water header.

The low level trip function on the "A" pump will be changed to operate off of the "B" tank ~ level. This change-will utilize the existing relay on the level channel for l_ "B" tank, which also provides the trip fun tion of "B" pump. This means that the failure of this one instrument loop (161) would disable both pumps. This failure, I however, would be in the safe direction because it would I. only prevent dilution of the RCS, and does not impact any li boration paths to the RCS.

l:

l Permanent piping changes will be made per B31.1-1967, and all materials will be stainless steel.

l 133 o ,

- r .- ; :__ . . _ _ _ _ _ _ . _ . _ _ _ _ . _ - _ _ _ . _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _

f

-i

' 3.4.72 124 (Unit 2), Safety Injection System.

MRs 89-123/124 propose to change out the accumulator level ;

nitrogen equalization line from 3/8" tubing to 1/2" ,

tubing. This would be done to facilitate proper drainage .

of this line should water accumulate in it. Recent' ,

problems.with accumulator level and pressure indication have been attributed, at least in part, to the failure of this equalization line to drain properly. .t t

Summary of Safety Evaluation: An evaluation is required because the proposed modification constitutes a

-potential change to the facility or its operation as described in the FSAR. .The proposed modification (s)'will x simply replace the 3/8" tubing with 1/2" tubing and' increase the slope on horizontal runs. Swagelok fittings .

will again be used on this line. The tubing and fittings-have temperature.and pressure ratings far in excess of the t: design requirements of this system. Stainless steel '

i material will again be used for all tubing and fittings.

The larger diameter tubing (same wall thickness) will' provide more bending strength than the original tubing, '

with an insignificant change in weight. Valves in the '

l tubing run will still be within 6" of a support, per the' L TMI era seismic tubing guidelines. Therefore, the L proposed change will not affect the seismic qualification of the tubing run.

h The change will not affect the functionality of this; system, but in fact, should make it more reliable. No control functions are included with the instrumentation.

l Instrument response should improve in level instrumentation due to no water hangup in the tubing. >

l.

3.4.73 89-131 (Unit 1) and 89-132 (Unit 2), Radiation Monitoring-System.

Modification Requests 89-131/132 upgrades the pressure rating for the grab sample flow path for the containment

. atmosphere post-accident sampling System (PASS or PACSS). .

L The modifications replace the glass bowl drain trap with i an all metal drain trap.

l. Summary of Safety Evaluation: An evaluation is required l'

because the proposed modifications will alter a system, structure or component described in the FSAR. The design bases contained in Section 6.5, " Leakage Detection Provisions for the Primary and Auxiliary Coolant Loops,"

Subsection 6.5.1, " Leakage Detection Systems," are as follows:

Monitoring Reactor Coolant Lea _kace Criterion: Means shall be provided to detect signif4 - J uncontrolled leakage from the reactor coolant 2 asure boundary. (GDC 16) 134

i 5

Positive indications in the control room of leakage of )

coolant-from the reactor coolant system to the containment ;

are provided by equipment which permits continuous '

monitoring of containment air activity and humidity, and of runoff from the. air recirculation units. This equipment provides indication of normal' background which l is indicative of a basic level of leakage from primary .l systems and components. Any increase in the observed- J parameters is an indication of change.within the containment, and the equipment provided is capable of monitoring this change. The basic design criterion is the. ;

detection of deviations from normal containment ;

L; environmental conditions including air particulate J activity,. gaseous activity, humidity, condensate runoff 4

E and, in addition,.in the case of gross leakage,.the liquid ;

inventory in the process systems and containment ~ sump..

Monitoring Radioactivity Releases Criterion: Means shall be.provided for monitoring the containment atmosphere and the facility effluent discharge paths for radioactivity released from normal operations, from anticipated transients, and from accident conditions. An environmental monitoring program shall be maintained to confirm that radioactivity releases to the environs of the l plant have not been excessive. (GDC 17)

The following are monitored for radioactiv'ty i concentrations during normal operation, anticipated

transients, and accident conditions: the containment 7 atmosphere,...

During normal operation and anticipated reactor transients, the containment air radiation monitoring system (1/2RE-211/212) is employed to help detect leakage from the reactor coolant system.

The containment air radiation monitoring system is a subsystem of the radiation monitoring system described in Section 11.2.3 of the FSAR. The primary purpose-of the-equipment is to sample and monitor containment air for l radioactive particulates and noble gases. Additional L- capability for sampling, venting and quantifying releases. >

L and release rates is also provided by the system. The system is shown in Figure 6.5-1 of the FSAR.

i, The system is comprised of a particulate and noble gas i sampling pallet, valving controls, flow instrumentation, L tubing, and air pumps. The design of the system allows for continuous sampling of containment air, sampling of the containment purge exhaust stack, obtaining a grab sample of containment air, venting the containment air, and flushing lines with service air or facade air. The h

mode of operation for this system is remotely centrolled from the control room. The switches used to obtain the desired sampling modes are located on the auxiliary safety 135

t I 1

a instrumentation panel (ASIP) for each unit in the control room. The only manual _ operations required are those asso-ciated with obtaining a grab sample and connecting service air to the system. The equipment for_the above manual.

operations is located in the facade on El. 52' of each unit outside the RE-211/212 sample cubicle. Containment penetra- ,

tions. serve to provide containment air to the equipment and a return path for discharge to the containment.

The control function of the containment air monitors is to initiate containment ventilation isolation (CVI)._ The initiation of CVI is based upon a high alarm signal from-the noble gas monitor (RE-212) only. The reason for,using only the noble gas monitor vice using both monitors (particulate and noble gas) is that the particulate monitor (RE-211)-is a fixed-filter monitor which would- ;

require'an alarm output based on a trend.

LL NRC NUREG-0737, II.b.3, Criterion 1 requires that a H

H containment atmosphere sample be completed, including analysis, within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after'the decision is made to obtain a sample. Conservatively allowing for. analysis time, we need to obtain a sample within I hour of the ,

decision to sample.

- The containment atmosphere post-accident sampling system l-(PASS) is presently limited to a maximum of 5 psig operating pressure. The limiting components are the RE-211 and RE-212 monitors; the drain trap; and the filter. The monitors can be bypassed and a sample taken ;

L from the septum. The drain trap and the filter have glass *

[ bowls that have a 5 psig limit. The filter is on the y

monitor bypass line and will not be pressurized when the system is configured to sample through the septum. The ,

sample pump (P-707A) has-a 15 psig limit for-continuous t operation and a 20 psig limit-for intermittent operation.

At 1-1/2 hours following a double ended break in

+ containment, the containment pressure will be below 18 psig. This is within the operating range of the sample y pump, P-707A.

In view of the accident response activities, it would not be reasonable to expect a request for.a containment sample within 1-1/2 hours of the event. This philosophy has been discussed with the NRC and they are in general agreement.

Thus, a 20 psig intermittent pressure rating for PACSS is adequate.

To increase the pressure rating of the sample path the glass bowl drain trap will be replaced with a metal drain trap. A commercial' drain trap rated for 20 psig minimum is not available with the flow cutoff feature of the existing trap and appropriate materials. A new drain trap will be fabricated.

136

_ _ _ _ _ ~ _ . _ _ __ _ -. _

4 I

i The new drain trap will be constructed of 2" Schedule 40

)

stainless steel pipe and pipe caps. The new drain-trap- 3 shall be hydrostatically tested to assure it will not fail i at accident operating pressures. -l The drain trap is designed to hold approximately 12 fluid ounces. This is the amount of water that would be H expected if all of the water vapor in the air (at 20 psig, l 250 saturated mixture) condensed in a'15-minute time period. The 15-minute sampling time is based on'the duty.

cycle of the sample pump, P-707A, at'20 psig operating-4 pressure. 1 If the trap collected more than-12 fluid ounces of water,

, the ball will close the discharge opening' stopping the air- ,

flow to the sample pump. Low flow alarsas will indicate the flow has stopped. A drain valve is provided'to' allow draining water from the trap.

1 L The new drain trap is Seismic class 3. This is in 1 accordance with the original installation. The new trap ,

has been designed to meet the requirements of Seismic-class 1 in the event the system is upgraded in the future.

The new drain trap does not involve the radiation'-

detection / monitoring characteristics of RE-211/212. ~The ,

flow through the detectors should remain the same. The design basis for the flows for RE-211/212 has not been 3 determined, so it is not known what affect minor changes:

in air flow have on their sensitivity. Since there is an unmonitored bypass _line around the detectors, the ' flow through RE-211/212 is assumed to not be critical.

QA requirements for material is limited to the primary sampling _ flow path and the first isolation valve on branches off of the primary flow path.

All new material used for the new drain trap will be'QA with the exception of the support and hose clamps that secure the drain to the support. Parts of.the existing.

drain trap (ball float and discharge cone) will be reused.

The new drain trap will not effect the function of PACSS.

The primary change is only to the pressure boundary. The

- change does no'. change the design basis of the plant not ,

any of the Technical Specifications.

Addendum to SER #89-123.

This addendum evaluates the installation of a manual i isolation valve in the bypass line of P-707A.

(.

l- Summary of Safety Evaluation: A pressure check of the grab sample path for IP-707A determined that there is some '

backleakage through the bypass regulator. The backleakage 137 L

l r W , _

rg- .

, . could pressurize the RE-211/212' detectors, causing failure,

-when a post-accident containment atmosphere. grab sample-(via the sample septum)_is obtained at containment pressures above 5 psig. To reduce the probability of backleakage through the bypass line, a manual isolation-valve will be installed between the bypass regulator and. ,

the bypass filter.

The bypass line is intended to provide a " relief" path .

around RE-211/212 in the event that the filter in the-

particulate monitor, RE-211, becomes plugged. This-prevents damage to the detectors due to high vacuum. The -

detectors, RE-211 and RE-212, have a sample pressure range of 10" of Hg vacuum to 5 psig.

The manual valve will add some additional flow resistance-to the bypass line. Since the bypass regulator senses the pressure only on the main flow path, it will adjust,to:

allow the correct flow through the bypass. This will not affect the operation of the system. Flow-transmitters FIT-3288 and FIT-3400 will show a flow difference l' indicating bypass flow. FIT-3288 will also alarm atr 1.25 cfm indicating excess bypass flow. Alarm response instructions are being prepared by Health Physics.

The. manual valve has little weight, approximately 4 ounces. The low weight will not affect any seismic c concerns for the sample pump. The manual valve has a l working pressure of at least 1000 psig. This is greater that the maximum design pressure of the PACSS.

3.4.74 89-156 (Common), Emergency Diesel Generators.

L The emergency diesel generator (EDG) stack supports-L (HB-29-H6,-H7,~H13 and H14) were not qualified correctly-by a contractor for seismic loads. The calculated seismic lateral load on the U-bolt of each support was compared against the normal load capacity'of the bolt rather than -

the lateral load capacity. Replacement.of the' supports by ones which are designed for the lateral loads and meet - ;

code allowables is necessary. Installation of the supports will bring the as-built condition of the system

, ~into compliance with the analysis results.

Summary of Safety Evaluation: An evaluation-is required.

because a system, structure or component described in the l' FSAR will be altered and a potential change to the L facility or its operation is involved. For personal safety concerns, each EDG must be taken out of' service while the stack supports for that diesel are being replaced.

The modification is a direct replacement of four supports.

The new supports will be installed within 2' of the existing supports at each location. After installation of l

138 l

each new support is complete, the existing _U-bolt suppo't l< r

L will be disengaged. With this method of installation, the Ly : system will not be in'an unanalyzed condition at any time

? during the installation.

3.4.75 89-167 (Unit 2),. Component' Cooling System.-

MR 89-167 documents the removal of a welded' fitting on a L ,

component cooling (CC) system isolation valve test connec-1: tion (2CC-1413), and replacing it with a threaded fitting,- ,

to. accommodate operations refueling testing, Summary of Safety Evaluation: The FSAR states that.CC i

1. piping is installed with welded fittings except at L locations where flanged joints are necessary for L maintenance. This threaded fitting will be used because ;

it would be impractical to use welded or flanged . .

1 connections for this-application. Threaded fittings are.

used elsewhere in this system,-in similar applications.

The threaded fittings are rated well in excess of system design ratings-, and will not affect-the seismic classifi '

cation of the CC system (by judgment based on the- 3 insignificant change in mass).

L l' The change will result in sealing a TVTC with a threaded L pipe connection verses another mechanical type of design. '

L The leak tightness of the joint is equivalent. It.is not

i. tested for_ Appendix J (used to test containment isolation '

valves). Any leakage would be small in view of closed--

valve and cap and would'be detected in normal containment inspections. ,

If a leak developed at this fitting, adequate valving is I available to' isolate the leak, CC surge volume plus makeup water will provide time to isolate the leak before

cooling is lost to essential components. . Isolating.this c 1 section of CC would render the excess letdown heat l exchanger out of service. This heat exchanger,-however, ,

I does not serve any safety function, and its loss-would i therefore not affect overall plant safety.

3.4.76 89-171, Feedwater Control System.

L MR 89-171 proposes to replace the existing 1(2)LM-463C L 10 m potentiometer with a 5 m potentiometer in series L with a 2.4 m fixed resistor. The potentiometer is used to adjust the lag time for the "A" steam generator. level :

program dynamic compensator and is being replaced because I the 10 m potentiometers are no longer available.

L Summary of Safety Evaluation: An evaluation is required l because the proposed modification constitutes a potential l change to the facility or its operation as described in j the FSAR. The modification will limit the adjustment :

range of the lag unit's time constant from 0-18 seconds 139

< i

, 1 1

1 to 4.3-13.3 seconds. The time' constant required, as.

calibrated per ICP 5.22, is 10 seconds, which is well within the range of the modified circuit.-

1 The modified circuit will perform exactly the same as the existing circuiti 'The change will not affect the progranuned steam generator water-level; and the required lag tim econstant will still be maintained. The l reliability of the circuit will not be;affected because the potentiometer is made by the same manufacturer and is of a similar model to the existing. The resistor is a 1/2 watt carbon composite resistor, which is commonly used=

, in the Foxboro boxes. The power rating of the new components is substantially conservative for.this -

-application.

The addition of the fixed resistor will result in an- '

insignificant mass increase to the instruments. Also, there will be no additional moving parts added to-the '

instruments. The circuit board mounting of the fixed resistor will be identical to the existing wiring.

Environmental' qualification of new parts is not a concern since the' control room is a mild environment. The ,

seismic qualifications of the instrument will not be affected.

s 3.4.77- 89-179 and 89-179*A&B (Unit 2), Residual Heat Removal System.

MR 89-179 addresses the reconfiguration of the Unit 2 l residual heat removal relief valve 861C discharge. The L relief valve discharge ~was installed to connect to the El 21' floor drain ~ system. These_ drains then discharged.

into the sump A. 'The relief valve shall be modified to discharge to the El.~8'--floor of< containment-(sump B). -

The El. 8' (sump B) drains, via floor drains, into sump A.

l This effort will involve the shortening of the 3" inlet L- piping to the relief valve, the reorientation of.the L relief valve, the capping of the existing drain line-tie-in and the building of a relief valve discharge standpipe.

The modification is being done to decouple the analysis forces of the floor drain system (non-seismic) from the seismic residual heat removal suction piping as transmitted through the relief valve.

This valve (861C) is described in several places in.the FSAR. i The reconfiguration of the discharge of the 861C will not 1:

alter any of its'overpressurization protection capabilities as described in the FSAR.

l-l-

p 140 1

I'

d i 4

.The fluid that could be discharged will not affect j i

' adversely El 8' of containment. This area normally '

serves as a collection point for discharged fluid in its role as sump B. The fluid will drain via the El. 8' floor drains into sump A. The-current configuration' drains '

discharge to 'sumip A. The equipment in this area that is _j important to safety is post-LOCA qualified up to El. 10' .l above the floor of the containment. . The discharge of this: !

valve would not empty any more inventory than is R postulated in LOCA evaluations. Thus this change is bounded by the existing LOCA accident evaluations. 1 The discharge from this relief valve would not directly be (

considered a LOCA unless the residual heat removal system-was required. The relief valve can be isolated from the reactor coolant system. At which point cooldown could'be accomplished by the use of~the steam generators or by use

-of containment sump recirculation.

The changing of the relief discharge to the El. 88 floor of containment from the drain line with cause a personnel safety concern; however, design package B of the ;

modification will install a standpipe for the relief '

discharge. This standpipe will serve to mitigate the 1 concerns to personnel safety.

Installatica will be done to the requirements of B31.1, to the original NSSS design requirements and the requirements of modification M-46. The entire revised configuration including the 10" AC-601R-2 residual heat removal piping has been reviewed for design loadings including pressure, ,

dead weight, seismic and relief valve discharge. '

During the first phase of this installation and while the '

head of off the vessel, no' analysis has been done that evaluates the 2-phase blowdown loadings of the relief '

L valve. This. evaluation will be done prior to inserting '

l the studs for the RV head. The installation will be seismically qualified for decay heat removal service when the RV head is not bolted down. Further qualification will be performed prior to bolting the RV head.

3.4.78 89-188*A&B (Common), 125 V DC System.

. The MR design package replaces the four DC breakers (two' per unit) on D11 and D13 for the crossover steam dump system with breakers having thermal / magnetic trip elements.

The breakers presently installed (original construction) have only thermal trip elements.

Summary of Safety Evaluation: An evaluation is required l because the MR design package will alter a system, structure or component described in the FSAR. Thermal /

magnetic trip elements are fast acting and protect DC buses D11 and D13 from a common fault which may occur L 141

a y> .

i downstream of the breakers being replaced. This maintains 1 adequate train separation between D11 and D13 and ;

isolation of the safety-related DC distribution sy. stem d y from the crossover steam dump system.

The replacementabreakers, which supply the Unit I and I t- -Unit 2 crossover steam dump system; are of lower current i H

trip setpoint than the breakers being removed. The l setpoints are adequate to protect downstream wiring and .l L are calculated to be sufficient to allow the turbine crossover steam dump' system to operate without supply beaker tripping. This will be verified by a post- J installation test of one unit's steam dump circuitry.

The retest may occur during'the test of station battery D105. Actuation of steam dump is expected to have no effect upon the D105 test or on post-test recovery.

3.4.79 89-188*C (Common), Station Batteries.

l .This design package replaces breakers on D11 for DC control power to Unit 2 buses A01, A02, B01, and B02 with r breakers having thermal / magnetic trip elements. Previously installed breakers had only' thennal trip elements which do -

not have an instantaneous trip rating. The new breakers have thermal / magnetic trip elements which have an instan-taneous' trip. The instantaneous trip is necessary to prevent a single fault occurring on the nonsafety-related load side of the breakers from resulting in a loss of the safety-related DC supply.

Summary of Safety Evaluation: An evaluation is required because the proposed modification will alter a system, structure or component described in the'FSAR.

~

Currently installed DC control power (both normal and alternate from D13-and D11, respectively) for Unit 2 A01, A02, B01, and B02 breakers on D11 with thermal trip elements only are being replaced by breakers having thermal / magnetic trip elements. The breakers have equivalent ratings and the trip elements have an instantaneous trip feature on high fault currents. This-provides isolation between nonsafety-related Unit 2 buses A01, A02, B01, and B02,-and the safety-related DC-distribution system. A single fault on a nonsafety-

+< related will not cause a loss of function of the safety-related DC distribution system. A single fault in a

-nonsafety-related DC circuit or a safety-related system will also not result in failure of feed line isolation upon safety injection of containment pressure condsnute ,

isolation. 1 I

l 142 l

.. . l

~

f 9 ,

Supplying Unit 2 buses A01, 102, B01, and B02 from the alternate control power supply on'D11 is acceptable per i plant design. Currently, Unit 1 buses A01, A02, B01, and '

B02 are supplied from the " alternate" breakers on D13.

Battery loading vill be in accordance with our previous

~

method of operation, with one set of A01, A02, B01, and-B02 buses loaded per battery although SER 89-134 t determined the additional loading would be acceptable. 3 3.4.81 89-191*A (Unit 1), Residual Heat Removal System. i

~

Modification 89-191A addresses the reconfiguration of the Unit I residual heat removal- (AC-601R-2) relief valve 861C discharge. The relief valve discharge was ir. stalled to connect to-the El. 21' floor drain system. These- drains then discharged into the Sump A. This design package will ,

decouple the relief valve discharge from the containment floor drain system, remove a section of the old relief L valve discharge line to ensure adequate seismic clearance-l' and install a temporary plug on the floor drain ~ piping in i the line.

Summary of Safety Evaluation: An evaluation is required-because the proposed modification alters a system,

9. structure or component described in the FSAR. This change is being performed due to seismic qualification-requirements for the relief valve and inlet piping configuration. This change will allow qualifying the line.-

for decay heat removal use until the subject piping _

configuration can be completely corrected during U1R17.

l The modification is being done to decouple the' analysis forces of the floor drain system (non-seismic) from the seismic residual heat removal suction piping'as- t transmitted through the relief valve.

The FSAR describes this valve (861C) in several places.

The reconfiguration of the discharge of the 861C will not alter any of its overpressurization protection capabilities as described in the FSAR.

The fluid that could be discharged will not effect adversely '

the El. 8' of containment. This area normally serves as a-collection point for discharged fluid in its role as Sump B. The fluid will drain via the El. 8' floor drains into Sump A. The current configuration drains discharge to Sump A. The equipment in this area that is important to safety is post-LOCA qualified up to 10' above the floor of the containment. The discharge of this valve would not empty any more inventory than is postulated in LOCA evaluations. Thus, this change is bounded by the existing LOCA accident evaluations. The discharge line will not i- aim directly at any safety-related equipment.

L 143 ac+ n --

Vl K :

i The discharge from this relief valve would not directly be considered a LOCA unless the residual heat removal system !

was required. The relief valve can be isolated from the reactor coolant system,-at which point cooldown could be accomplished by ,the use of the steam generators or,by use -;

of containment sump recirculation.

]

The_ changing of the relief discharge to the El'. 8' floor-

,. of containment from the drain line will cause a personnel safety concern; however, design package B of the modification will address this concern. Until that. time, administrative controls need to be in place, f The installation will be done to meet the' original design requirements. The entire revised configuration, including '

- the 10" AC-601R-2 residual beat removal piping, has been : "

' reviewed by NSEAS for design loadings including pressure,-

deadweight, seismic and relief valve discharge.

e P-144

p ,

t 3.57 - Tens >orary Modifications Following is a list of . temporary modifications performed at Point Beach Nuclear Plant during 1989 that required a 10 CFR 50.59 review.- In each case, it was determined that the temporary modification did not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety was not increased.

The change did not create the possibility for'an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications was not- ,

reduced.

3.5.1 TM 89-09: Circulating Water System. This evaluation -

amends the original safety evaluation performed for - i temporary modification TM 88-025 for the purposes of installing a temporary circulating water chlorination /

dechlorination system. The conclusions of that SER remain essentially valid.

3.5.2 TM 89-012 (Unit 1), Primary & Auxiliary Systems._ The TM installs temperature and system parameter monitoring associated with NRC IE Bulletins- 88-08 and 88-11 on the pressurizer surge line, the auxiliary spray line and the ,

auxiliary charging line. The work includes installation-of thermocouples and cables in the loops; installation of remote computer multiplexing equipment in containment; connection of 22 inputs from the PPCS to computer. ,

equipment ih the north computer room;'and operation of this equipment remotely from Milwaukee. All of the reference equipment will remain installed until the U1R17 outage.

Summary of Safety Evaluation: An evaluation is requh ed because the TM involves a potential change to the facility *

,. or its operation'as described in the FSAR. The 22 inputs to the PPCS will be paralleled off the MUX terminal strips to provide signals to the IEB 88-08/88-11 computer. The K analog inputs from the MUX are isolated from the ,

I corresponding instrument loops. Therefore, installation

, of twisted shielded pairs (jumpers) will not affect-L operation of the reactor control and reactor protection l- system. However, during installation and use of the jumpers, if the MUX input voltage signals are shorted, the voltage signals to the PPCS would be significantly p reduced. This reduces the value of the channels /-

parameters indicated by the PPCS and further affects the subcooling margin calculation and RTO calculation. Care will be taken to inform Operations when input termination work will start and finish. The isolation devices provide 240 V isolation at 650 Kohm input impedance which will affect parameters no more thLn 0.5%.

145 l~

t TS Table 15.3.5-5 (No. 4) requires that the RCS subcooling g monitor or backup be functional. The PPCS and SAS are the relied-upon backup displays of subcooling margin. The RTO:

calculation is used by operations to precisely control

reactor power,.although it is not included in the Technical Specifications. l 1=

l Grounding or shorting of a MUX analog' input signal would 1 l cause the' indication in the PPCS to go to zero. The computer would alarm in this condition.. Shorting or grounding would only typically be likely to occur during-the installation process. Hookup of the jumpers will be

controlled via MWR work plan- to eliminate the potential-F for spurious signals.

An isolation amplifier with 640 Kohm input impedance (both on and off) be utilized to separate the PPCS and IEB ,

88-08/88-11 monitoring minicomputer. With an 8 megohm input impedance parallel.with the 640 Kohm inputs of the '

i minicomputer, the' MUX input signals to the PPCS will be- .

H '

unchanged. Hence, the PPCS indications, subcooling margin.

indication and RTO calculation will not be affected. In case problems occur with PPCS displays, control board ~

displays can be utilized in controlling plant operation.

The primary differences between the work associated with this TM and the work done under Unit 2 TMs 88-67, 88-68 :

and 88-76 are as follows: (1) Fewer locations will be monitored. There are 29 Unit I locations as compared to 44 Unit 2 locations. (2) Less computer hardware will be ,

installed. A MUX in containment-and isolation equipment 'i in the north computer room are-the only new computer.

hardware being installed. (3) Unit 1 plant parameter ,

signal-taps-will be made in parallel to inputs to the PPCS~

and Unit 2 inputs will be moved upstream of the' "

Westinghouse-fatigue monitoring system isolation amps. +

The concern for electrical isolation and on/off state impedance is addressed by' the use of analog-isolation amplifiers being installed as part of this TM. The-calculated maximum effect that the new isolation devices .

will have on PPCS indications is less than 0.02%. The isolation devices provide 1500 V.'RMS input / output. '

Apart from consideration for the differences between this TM and the ones for Unit 2, the conclusions reached in SER 88-126 remain valid for the scope of this TM.

3.5.3 TM 89-019 (Unit 1), Containment Ventilation System. The TM will remove the 3-way pneumatic switch from the purge supply and exhaust valve operator control circuit. The 3-way switch has internal air leakage which could result in a decrease in valve boot seal air pressure in the event of a loss of instrument air.

146

4

)

I Summary of Safet7 Evaluation: An evaluation is required because-the proposed evolution could involve a potential-change to the facility or its operation as described in- 1 the FSAR. . The function of the 3-way switch is to stop the jl supply air and vent the boot seal air with the valve is opened. Since the valve is locked shut during power-

-operation in accordance_with TS 15.3.6.C, removal of the ;

3-way switch will have no adverse'effect on the _ 1 performance of the purge supply and exhaust _ valves. The .

3-way switch will be replaced with a threaded pipe ;

coupling, q This change will be performed on the.four Unit 1 .I containment purge supply and exhaust valves; VNPSE-3212, 1 3213, 3244 and 3245.

3.5.4 TM 89-020 (Unit 1), DC Power System. Temporary modification 89-20 will supply IP37D with alternate battery power supply which is rated to deliver the L required load for the one-hour loading requirement as i described in FSAR Table 8.2-3.

Summary of Safety Evaluation: An evaluation is required because the temporary modiffcation will alter a system L described in the FSAR. The use of this temporary battery 4 E '

is an economical choice and has no effect on safe operation of the unit. The economical choice is to provide an alternate lube oil supply to prevent turbine '

l bearing damage under the conditions of a loss of AC power.

The evaluation performed by the temporary modification l~ shows the adequacy of the battery and installation, and L also addresses industrial safety considerations.

L _

'. 3.5.5 TM 89-021 (common), 125 V DC Power System. The temporary L'

modification will provide charging power to new station battery DOS prior to placing itlinto service. This will be accomplished via a temporary battery charger supplied from a spare breaker. ,

Summary of Safety Evaluation: An evaluation is required because the TM constitutes a potential change to the facility or its operation as' described in the FSAR. The J only effect of this TM on plant systems is the temporary load increase on MCC 1B39 of ~12. amps at 480 V on two l phases and 21 asas on the third phase. All power supply l equipment (cables, breakers, transformers, etc.) upstream-

-of the temporary battery charger have enough reserve capacity to supply this unbalanced load. Sufficient isolation between the qualified bus and the unqualified chargers is provided via the qualified breaker.

The battery charger would increase the load on the emergency diesel generator by <1% if offsite power is lost. Initial load on the battery is the greatest and as the battery becomes charged, this decreases. Thus, the 147

1 longer this load exists prior to an accident, the smaller i its effect upon the post-accident load on the EDG (G01). ~

Any'long-term accident concerns can be met by tripping this load without any'effect upon the plant. A dedicated

' person will be available to trip this load.

All cables will be routed within the same fire zone to satisfy Appendix R concerns, except cable to the DOS room, which is not a problem since DOS is not connected to loads. The cables will be routed through the battery room door, degreding the fire barrier. An hourly fire watch inspection will be performed in accordance with .

TS 15.3.15.C.l.b.3.

3.5.6 TM 89-023 (Unit'1), Circulating Water System. The TM will provide for the connection of hoses and associated equipment for the chlorination and dechlorination of the circulating water system.

Summary of Safety Evaluation: An evaluation is required '

because - this evolution constitutes a potential change to the facility or its operation as described in the FSAR.

This temporary modification will be accomplished for Unit 1 in the same manner TM 89-09 was performed for Unit 2.

An evaluation determined that the summary and conclusions documented in SER 88-039-01 for TM'89-09 remain valid for this evolution.

3.5.7- TM 89-027, ICS-00466hA Hinge Pin Bonnet Leak Repair.

This temporary modification will accomplish;a temporary j repair of valve ICS-466AA as described in.MWR #892461.

The proposed fix for this valve includes. seal welding the hinge. pin bonnet flange to the valve body; drilling a vent hole in the bonnet flange, and welding pipe caps over the i L studs / nuts.

Summary of Safety Evaluation: The evolution constitutes a potential change to the' facility or its operation as.

described in the FSAR. Additional stresses due to.

changing the pressure boundary are calculated. Material' strengths and inservice (and design) pressure are compared to code allowables in the calculation. The' calculation /

evaluation'does not reveal a condition in which the integrity of the valve is endangered.

The additional weight on the valve due to the pipe caps and vent plug is insignificant and is not considered to impact the seismic capability. The volume created between the seal weld and pin gasket will form a small vessel, but pressure excursions due to a heatup of the trapped volume 148

.. . ...z. . . . . . _ . _ _ . _ . - =

should be eliminated by leakage into the main valve. Loss of normal-feed and high energy pipe failure are both considered in the FSAR.

The extension in the cover pressure boundary as a result:

_ of the seal weld raises the bolt stress to above that which is' allowable for. pressure' loads per Section III and-Section VIII. The valve was not designed and built to these codes so they are being used as more of a reference than as absolute limits.

Sections III and VIII limits for the bolting is 24,000 psi.

Calculated stress at a design pressure of 1310 psig is

=

42,730 psi. The specified minimum yield is 105,000 psi.

If some credit is taken for the seal weld, it can be shown the bolting pressure-stress would be within allowable.- ,

Seal welds are not typically considered as providing any i structural strength, however, and a portion of the weld-

[ will have to be made with water present.

An adequate margin of safety is considered to be maintained for this temporary repair, based upon the following:

- The design pressure of 1310 psig is not expected to be seen.

- An adequate margin to yield (2.5) is still available.

- The seal weld will provide some' strength.

- Section III allows-going to two times the allowable for special service conditions.

- The bolting will have been-torqued to 60,000 psi in an attempt to stop the, leak before performing the-temporary

_ modification; which is well above the stress resulting from the pressure load.

3.5.8 TM 89-033, Service Water System. TM 89-033 $natalls a o patch on an existing 8x14" reducer downstream of valve ISW-144. NDE inspections have revealed flow-induced cavitation damage (pitting) on the interior of the reducer. Valve 1SW-144 is used to throttle service water e

from a nominal 80 psig to atmosphere pressure which causes some downstream cavitation.

Summary of Safety Evaluation: The damaged reducer is a carbon steel A-234 WPB standard weight fitting. The fitting is required to carry seismic, pressure and deadweight loads. Pressure stress is essentially zero at normal operating conditions. The fitting does not leak but may be weakened by the pitting damage. The patch will 149 b

b- _ _ _ _ _ _ _ _ . . . . . . .. . . .

-1 1

act as a reinforcement to ensure that the seismic'and-deadweight loads can be carried. The patch will be ;

installed with an-open nipple to provide a tell-tale .i leakoff point.

_l This change will not impact the operability of the. service water system. The patch will restore any strength that-s may have been lost in the. existing fitting. This change a is being done es a conservative measure; the ability of j the existing fitting to handle service loads is difficult 1 to assess., Therefore, the reinforcement is installed to- -!

remove any doubt. )

l- .

j The additional weight of the reducer patch is. considered I to add negligible weight to piping. Seismic stressesiin l this area are very low (400 psi) and thus, considered-added weight from the reducer'to be negligible.

l ll The reducer will be split and full penetration butt-welded.

'along the seams over the existing reducer. This will -t restore the hoop strength of the reinforcement reducer '

completely. The reinforcement reducer will be welded to' the outer diameter of the existing reducer and-flange neck L with a long leg fillet to meet the 3:1 taper typically L used-at weld joints of different thickness to provide a ?

l smooth stress transition.

L E

The reinforcement reducer will provide equivalent hoop ,;

strength capability and nearly equivalent axial load- M carrying capability. Because the pressure and seismic. J stresses are very low and there would be essentially no thermal stress in this .line, the reinforcement patch is '

l considered adequate for this repair.

L To gain access for welding, it.may.be necessary.to' remove one bolt at a time from the flange. 1The flange is an 8",

150 lb standard flange with a bolt circle of 11-3/4'.',

connected with eight B7 3/4" diameter bolts. .The bolting.

pressure load would be 8.722 x n/4 x 100 psi' = 5972 lbs.

The bolting allowable working load, based'on an allowable stress of 25 ksi, would be 7547 lbf per bolt, or 60,381 lbf total. The pressure load and seismic stresses are i negligible in comparison to the capability of,the bolting.

Thus removal of one bolt at a time is considered

~

acceptable during the repair. t 3.5.9 TM 89-034 Unit 2, 4160 V. Electrical' System. The TM will l' '

remove the sudden pressure trip output off of transformer 2X01B$ from the 2-86/X01 unit lockout circuit. This will be accomplished by opening two knife blade switches at the i 2X01B$ control cabinet that passes the trip signal to the '

lockout circuit.

150

Summary of Safety Evaluation: An evaluation =is required because the proposed activity.will alter a system, .

3 n structure, or component'(SCC) described in the'FSAR and. t could affect the function or method of the SCC. Removal-of the sudden pressure trip output will mean that an' .

electrical fault inside'2X01B$ would have to be detected H L by the transformer differential relay (2-87/X01B$) which ,

L is, in fact, the primary protection for electrical fault:

! in the transformer. !

?

Although the sudden pressure trip provides early detection of minor growing failures ~inside the. transformer, it is !

not a replacement for.the. differential relays, whose d

, purpose is to protect the transformer and' system:from;

lifetime degrading faults. Although some protection will; be lost, the primary protection is maintained to support- i the transformer operation. Thus, the non-nuclear safety /

function of the transformer is maintained.

l -L , 3.5.10 '{'MJ,89-037 (Unit 1), Condensate System. TM 89-037 will accomplish a temporary repair of valve ICS-476AA. The '

proposed fix for this valve includes seal welding the hinge pin bonnet flange to the valve bodyr drilling a vent -

. hole in the bonnet flange, and welding pipe caps over the studs / nuts. .

Summary of Safety Evaluation: The evolution constitutes a potential change to the facility or its operation as described 1,n the: FSAR. Additional stresses due to ,

changing the pressure baundary are calculated. Material :

strengths and inservice (and design) pressure are compared to Code.allowables in the calculation. The calculation /

evaluation does not reveal a condition in which the:

integrity of the valve is endangered, ,

The additional weight on the valve due to the pipe caps ,

and vent plug is insignificant and is.not considered to impact the~ seismic capability. The volume created between '

the seal weld and pin gasket will form a small vessel, but pressure excursions due.to a heatup of the trapped volume should be eliminated by leakage into the main valve. Loss of normal feed and high energy pipe failure are both considered in the FSAR.

The extension in the cover pressure boundary as a result of the seal weld raises the bolt stress to above that which is allowable for pressure loads per Section III and Section VIII. The valve was not designed and built to :

these codes so they are being used as more of a reference than as absolute limits.

Section III and VIII limits for the bolting is 24,000 psi. _

Calculated stress at a design pressure of 1310 psig is '

42,730 psi. The specified minimum yield is 105,000 psi.

151

i If some credit-is takin for the seal weld, it can be shown i the bolting pressure stress would be within al3owable.

. Seal welds are not typically considered as providing any :

structural strength, however, and a portion of the weld. i will have to be made with water present.

An adequate margin of safety is considered to be

' maintained for this temporary repair, based upon the-followings.

The design pressure of 1310 psig is not expected to be seen.

.]

An adequate margin to yield (2.5) is still available.

The seal weld will provide some strength.

Section III allows going to two times- the allowable y for special service conditions, The bolting will have been torqued to 45,000 psi, in I an attempt to stop'the leak before performing the-temporary modification; which is above the stress resulting from the design pressure load. J 3.5.11 TM 89-039 (Unit 2), Reactor Coolant System. The TM will. I l

remove the motor feeders for the 2W3A&B control rod drive i mechanisms and install a 60 amp temporary power' supply ~1 from each motor power supply. This.will provide an a uninterruptible power supply for steam generator maintenance work during the U2R16 outage.

Summary of Safety Evaluation: An evaluation is required because the proposed change constitutes an alteration to a system, structure or component'(SSC) described in the FSAR and could affect the function or method of an SSC described in the PSAR.. 6 L

Unit 2 will be in the cold shutdown condition. The W3 ;

fans are used to cool the control rod drive mechanisms i

during operation. They are not in use during cold shutdown or refueling conditions. The temporary modification will be removed and the system will be returned to its normal configuration and operationally l

tested prior to commencing reactor coolant system heatup.

The CRDM fans are an integral part of the containment l'

' ventilation and, as such, are required to satisfy the (

design basis prior to RCS heatup.

I 1 3.5.12 TM 89-043, Reactor Coolant System. The TM blocked open ,

l power-operated relief valves 2RC-430 and 2RC-431C during ,

performance of the U2R15 containment integrated leak rate test. ~

I-l ,

l 152

h i

Summary of Safety Evaluation: An evaluation is required s

because the proposed change could affect the function or ,

method of a system, structure or component described in the FSAR. During low temperature operation, the power- ]

c operated relief valves (PORVs) provide overpressure l protection to the reactor coolant system by opening at the l LTOP setpoint. Since the valves will be blocked open, the ]

PORVs still function to prevent overpressure of the i reactor coolant system. l t

If mid-loop operation is entered with the blocks in place, the PORVs must be considered to be a hot leg vent, and appropriate actions and precautions will be taken per OP-4F. A hot leg vent is not a problem during mid-loop !

operation; a cold leg vent requires compensatory action. l The blocks will be designed and installed to be held ,.

g recurely in place during a des $gn basis event. The added -

mass is negligible and will not degrade seismic integrity 3 of the PCRVs. The biochs vill not dan, age the PORVs. (

1 l' 3.5.13 TM 89-044 (common), Cranes. Installation cf a debris, j catcher on the primary auxiliary building (PAB) crane !

bridge is being done to catch small particles ol' debris j

( which tre generated by the reroefing work on the PAB. ,

summayy of Safety Eve 1 nation: An evaluation is required because the TM constitutes a change to a system, structure :'

I or component described in the FSAR. The debris causes a l

concern in the area of the spent fuel pool (SFP). A number of different methods have been proposed to protect ;

the pool, but a debris cover on the crane was chosen as :

the best alternative. j The cover itself will be made of a Unistrut frame covered o by a tarp. The Unistrut frame will consist of a Unistrut L spanning from one bridge girder to the other on 6'-0" centers attached to the Unistrut, which is laid next to the trolley rails. The attachment of the spanning pieces of Unistrut to the pieces next to the rails will be

sufficiently rigid so racking of the frame will not be possible. Since the frame is located between the trolley :

t rails, movement of the cover is limited and therefore cennot fall down. Also, the weight of the cover is less than 1750 pounds arwt is therefore not defined as a heavy ,

load.

The tarp to be installed will be a fire-retardant, reinforced tarp with grommets located around the perimeter at about 4'-0" center-to-center. The tarp will be sufficiently large enough to span from bridge girder to bridge girder and will be attached to the crane with nylon j rope to each bridge girder walkway at a minimum of 5'-0" center-to-center. Each rope will have sufficient strength to hold the entire tarp. ,

153 >

)

..+ - c .. _ _ . - - _ _ _ _ - , _ _ _ _ ~ _ . _ . . - _ _ _ _. _ _

l

_u. ...a ,xx-aa-._ -w ---e,_+-n- .1.-a .

=..--a,,a. -

I~

other small areas (such as the gap between the walkways t i and bridge girders) will be covered with fire retardant j plastic sheets and installed such that they cannot fall from the crane. l It is concluded 'that the debris cover installed on the FAB crane will serve the purpose for which it to designed. l 3.5.14 TM 89-045 (Unit 2), Reactor coolant System. The proposed !

temporary modification involve installing a strongback on ,

the 2PIB reactor coolant pump in order to raise the pump ,

shaft such that the pump seals can be made up.

Sunmary of safety Evaluation: An evaluation is required ;

because the prcposed e':elution constitutes an alteration ;

to a system, structure er component described in the FSAR.

The conclusions of another 3EP are valid for the use of l the strongbacks te only hold the pump shaft end inre13er in position. The stronjoacks only need te hold the weAght ,

of the pump shaft and impeller, ir tenci of tha design ,

weight of the pump she.ft iveeller and motor rotating assembly.

Failure of the s.trongbcck, resulting in e leak of reactor [

coolant sysAem water would not allow tho kCS tr, drain low -

cnough to prevent residual heat removal operation ner would it drop low enough to enter OPr4F mid-loop ;

operations (55% :eactor vessel level).

As noted in SER 86-073-01, sufficient time is considered available in the tvent that a leak develops due to '

strongback failure to replace a fuel assembly back into ;

the reactor vessel if one is in the manipulator at the -

time of the hypothecized failure. ;

t. The strongback shall not be installed or removed during l

mid-loop operation unless an acceptable RCS hot leg vent exists in accordance with OP-4F. Mid-loop conditions may not be entered with the strongback in piace unless an acceptable hot leg vent exists in accordance with OP-4F.

The strongback is not capable of holding the shaft pump

( and RCS pressurization on a loss of RHR at mid-loop could lift the shaft and possibly damage the seals; and create a cold leg break.

3.5.15 TM 89-048 (Unit 2), Residual Heat Removal System. The TM will install a stem clamp on 2RH-700. This valve is closed during plant operation. It provides isolation from the reactor coolant system to the residual heat removal pumps and is interlocked such that it cannot be opened until RCS pressure is <425 psig. During refueling the valve is normally open to provide suction to the RHR pumps from the reactor coolant system.

154

'7 % - -- . - - + m- , . . , .-m.., , . . - -

- ~ , . . . , , . . . . . . - _ , . , , - ,-e .

! f summary of safety Evaluation: An evaluation is required because the proposed 7M constitutes a change to the !

facility or its operation as described in the F&hR and '

will alter a system, structure or component described in !

the FSAR. Removal.of the valve operator from 2RH-700 for maintenance will require clamping the valve open. l Clamping the valve open will prevent the valve from being i able to be closed to isolate the RHR pump suction, j however, 2RH-701 is in series with 2RH-700 and can be ~

closed remotely to isolate RHR.

f Clamping 2RH-700 in the open position while the RCS is i 400 psi and 350'F does not present an unreviewed safety -

concern. The valve stem clamp weighs less then the l operator; therefore, reismic stresses will be reduced. By 1 engineering judpent, the seismic 1cading on the 2-1/8" '

diameter valve stem when uar,upported by the operator is not a problem. !

4 The clamp will be installed on the valve whan the core is

! dafueled but the cavity is flooded. There are several ;

conditions that could occur during the time the clap As '

installed, bewever. These possibit condtions aree

a. Core defueled, enity (It,cded, core barrel and upper internals ane in their storage stands. ;

b. Core barrel in the reactor venel, upper internals in r its stand, upper cavity drained to allow work on RH-700, RH-720, SI-842h, and SI-867B. t

c. Cavity flooded and refueling in progress. !

d. Core loaded, cavity drained, RCS not in "mid-loop" ~

operation.

e. Core loaded, RCS level at "mid-loop." I

f. Core loaded, RCS filled and vented, and heatup in progress. Conditions are below those which would '

require isolation of RHR.

The valve is being blocked open to allow flow which is .

required or desirable for all of the above conditions.

The valve is not required to change position during any of i the above operating conditions. The only function of the valve is to remain open to allow flow and act as a .

pressure boundary. Clamping the valve open has no affect on the pressure-retaining capability of the valve.

Installing the clamp provides a positive means of holding ;

the valve open.

155

1 l

l l

The system will function as designed with the clanqp installed escept that the valve cannot be remotely i operated, once in the open position, there is no need to remotely operate the valve during decay heat removal or for a postulated accident. A backup means exists for isolating the RCS should a leak or rypture occur in the RHR system.

The clanqp is designed with adequate margin to maintain the valve in the open position, including seismic loads. The j seismic qualification of the system is not affected by j installation of the clamp; nor should removal of the 1 operator have any impact on the seismic capability of the j system. The series of drawings do not show any suppcrts/ i restraints attached to the va.?.ve operator.

t kestoration of the motoro.ipersted valvt is required befo?r. 'd returning.the unit to cperation, this is altvady contro?. led '!

by procedure, which requ$re.s isclot$0n of RHR from the RCS ;

by closiug of thf.s valve. I I

3.3.16 TM 89-049 Cifnit 2), Fesidual Hect rep. oval System. The TM !

will instal 1 a stem clamp on 2Ril-720. This selve can only .!

be opened when RCS pressure is less than 425 pr.ig. yhe i valve is open when the RHR system is used to provide RCS ;

cooling. This valve is normally open during a refueling - '

outage. There is a check valve in series with this valve I' to pro *;1de a redundant method of isolation from the RCS.

i Summary of Safety Evaluation: An evaluation is required because the proposed TM constitutes a change to the facility or its operation as described in the PSAR and !

will alter a system, structure or component described in [

l the FSAR. Removal of the valve operator from 2RH-720 for '

maintenance will require clamping the valve in the open :

position. Clamping the valve open will prevent the valve from closing should the RHR system require isolation, but i there is a check valve downstream of 2RH-720 which would provide isolation. Should the valve stem clamp fail, i 2SI-852A would provide a redundant path.

Clamping the valve open while the RCS is <400 psi and l 350*F does not present an unreviewed safety concern. The '

valve stem clamp weighs less than the valve operator. -!

Therefore, seismic stresses will be reduced. By engineering judgment, the seismic loading on the 2-1/2" diameter valve stem, when unsupported by the operator, is not a problem.

5 The alternate injection path through 2SI-852A would put water through the core injection nozzles to the top of the core. This is not an analyzed decay heat removal flow path and would rely on natural convection cooling of the core. This flow path has been used successfully_for short durations during Event V check valve testing. The decay 156 1

heat load will be low since the timeframe the clamp will be on is 4 weeks after shutdown and the core would be a reload core (if fuel is installed before the motor :

operator is reinstalled). Thus, cooling by core injection i and hot leg discharge should be relatively effective, if j required. !

i The clamp will be installed on the valve when the core is f defueled but the cavity is flooded. There are several :

conditions that could occur during the time the clamp is !

installed. These possible conditions ares ;

w. Cora defueled, cavity flooded, c: ore barrel and upper !

internals in their Storage etsids. '

t

b. Core 1arrel in, the vescel uppva intern =3s in its i (t ed, upper ;avity dreit.ad to alJoy ucrk oss T& 700, i RH 710, 51-M2A, ar,d Sb86%. *

, c. Cavst') zlooded and refueling in progress. l

[

d. Core loaded, r.avity dre.ix 3, RCS not ina rid-loop" operation. ,

e. Core loaded, RCS level at "sdd-loop."

f. Core loaded, RCS filled and vented, and heatup in progress. Conditions are below those which would ;

require isolation of RHR.

I- The valve is being blocked open to allow flow which is I l required or desirable for all of the above conditions.

The valve is not required to change position during any of the above operating conditions. The only function of the :

valve is to remain open to allow flow and act as a pressure i boundary. Clamping the valve open has no effect on the :

i- pressure-retaining capability of the valve. Installing the clamp provides a positive means of holding the valve open.

The system will function as designed with the clamp installed except that the valve cannot be remotely ,

operated. Once in the open position there is no need to '

remotely operate the valve during decay heat removal or '

for a postulated accident. A backup means exists for isolating the RCS should a leak or rupture occur in the RHR system.

L The clamp is designed with adequate margin to maintain the valve in the open position, including seismic loads. The seismic qualification of the system is not affected by installation of the clamp; nor should removal of the operator have any impact on the seismic capability of the system. The series drawihgs do not show any supports / ,

restraints attached to the valve operator. l 157 l~

]

\

Restoration of the motor-operated valve is required before '

returning the unit to operation. This is already !

controlled by procedure, which requires isolation of RHR !

from the RCS by closure of this valve. i 3.5.17 TN 89-054 (Unit 2), Feedwater System. . FSAR Table 10.1-2

{

shows that pressure vessels in the steam and power ;

conversion system comply with Section VIII of the ASHE Boiler and Pressure Vessel (5kPV) Code. Part UG-134 of Section VIII states that "when a single pressure-relieving device is used, it shall be set to operate at a pressure j

not exceeding the maximum allowable werking pressure of the vessel." The nameplates on 2MX-19A arid 2JO(-19B, '

the Utit 2 #3 feedwater heaters, give the mar.imum 4

' allowable working pressw e of the tabe side as 413 prig. -;

" The existing tube side re31ef v dva setpoints r.re ;

nameplated at 415 psig. h.s, the pressure prntection of ,

D, the Wit 2 63 feedwater heatern does not comply with the .

wrding in the code'. Compliance with tie sr. ate rahini- l

'__ r*.rative code is also in question, since the B&P# code is i incorporated by reference into Chapters ILHR 41 and 42..

The preferred way to return to cotipliance would be installing new relief valves with a setpoint of 413 psig '

or less, or resetting the esisting rehef valves to 413 '

psig. However, since the heaters are Section VIII l pressure vessels, Part UG-129 requires that UV-st&:nped f relief devices be installed. New 413 psig UV-stamped relief valves are not available to be installed before the end of the unit 2 outage. To reset the existing ,

valves, the Wisconrin State Administrative Code ILHR 42.35 '

requires that the work be performed by an organization holding the appropriate National Board VR stamp. There are very few VR holders with certification for liquid service valves, and none in the area. Additionally, there is no reasonable safety benefit to having these valves '

reset 2 psi.

This temporary modification documents and justifies retaining the 415 psig relief valves on the tube side of '

the Unit 2 #3 feedwater heaters until the Unit 2 outage in 1990. During this time, appropriate relief valves will be obtained and installed for full compliance with the letter of the code. To ensure that safety is not compromised, '

the existing relief valves with 415 psig nameplat-s will be setpoint tested to ensure that they meet the setpoint tolerance criteria for 413 psig relief valves, which is 13% or 401 to 425 psig. >

Summary of Safety Evaluation: An evaluation is required because the proposed TM constitutes a change to the ,

facility as described in the FSAR. The Foreward to the B&PV Code states that "The objective of the rules is to i afford reasonably certain protection of life and

158

'H:_. _

i.

f-property." By ensuring the valves nameplated at 415 psig meet 413 psig setpoint criteria, this temporary modification will meet the intent of the code.

3.5.18 TM 89-055, Containment Ventilation. The tosporary modification will remove the 3-way cam-operated air switch from VNPSE-3213 and 3245.

Summary of safety Evaluation: An evaluation is required because the evolution involves a potential change to the facility or its operation as described in the FSAR. The 3-way valves, which are in the air supply line to the boot of VNPSE-3213 and 3245, will be removed and bypassed with e pipe coupling or Swagelok fittings. The lines are 1/4" ceppt.r tebing. An unsupported tubir.g at the bypassed location vin be pecured vith hose clampe, to ensurn no y ederse seismic effects. This is acceptable per observation.

The 3-way air switch is actuated by a lever arm on the purge supply and exhaust valve. Actuation results in infletion of the boot seal. Per this temporary modification, the 3 %ey air switch will be bypassed: thus me. king the air supply to t.he boot sen1 con'inuous via a seal regala?.or (Fisher 164A). Note that this 71 sher regulator a ho controls air to the boot, but the 3-way valve is the cam-operated portion, while the Fisher 164A is air-operated.

TS 15.3.6.C requires the purge supply and exhaust valves to be locked closed during conditions other than cold

!. shutdown or refueling shutdown. When the valves are shut, the 3-way valve mpplies air to the boot. Thus, a bypass around the 3-way valve will accomplish the same function.

Operation of the valve without the 3-way valve, e.g.,

cycling, could result in boot damage, but the intent of the temporary modification is for implementation while the purge supply and exhaust valves are closed only.

The installation of a bypass around the 3-way air switch and removal of the air switch will increase the assurance of the boot remaining inflated if instrument air is lost.

After the valves are closed and the temporary modification is installed, a tag series will be issued to require that the temporary modification be cleared and the 3-way valve

! reinstalled (during cold shutdown or refueling shutdown) l prior to cycling the valves open.

l l

l 159 l

t i

3.6 Core Reloads 3.6.1 UIC17 Reload Safety Evaluation The U1R16 refueling outage began on April 2,1989, with criticality for . Cycle 17 scheduled for May 9,1989. This !

review covers the cycle 17 reload design and safety '

analysis work performed by Westinghouse. It also covers ;

further evaluations performed by Wisconsin Electric.

The Unit 1 Cycle 17 reload contains 12 fresh Region 19A -

upgraded optimized Fuel Assemblies (OFA) at 3.6 w/o,16 i fresh Region 198 upgraded 0FA at 4.0 w/o, 28 once-burned !

Region 18 0FA. 32 twice-burned Region 37 0FA, 28 thrice- i burned Region 16 OTA, and 5 Region 8 and 128 standard design assemblies. The four Recic,n 8 assemblies were initially loaded into Unit 2. The Cyt;1e 17 reload is ti.e ,

first reload containing a full region of up, graded 0FA fuel i for Point Beach Nuclear Plant (PBNP) Unit 1. Upgraded DFA '

fuel is the subject of Technical Specification Chtnge !

Request No. 127.

Eum_marv of 1Saf ty,Evsluation: An evaluatica is required because the fuel rr:1oM dill alter a system, structure or :

coraponent described in the FSAR and because the reload '

involves a potential change to the facility and its :

operation.

Technical Specification changes required for the operation of Cycle 17 have been reviewed by the NRC. The evaluations and conclusions presented herein are contingent upon NRC {

! approval of Technical Specification Change Request 127.

No special environmental considerations are involved and, therefore, evaluation of environmental effects by the NRC

staff has shown that the issuance of environmental impact or assessment statements is not required. >

The Unit 1, Cycle 17 reload design and safety analysis is ,

p acceptable and falls within the scope of earlier analyses, and indicates that operation of the Cycle 17 core does not involve a significant increase in the probability or l consequences of accidents previously considered, does not involve a significant decrease in safety margin, and does not involve a significant hazard consideration. Therefore, provided that the startup physics testing does not result in any discrepancies with the analysis assumptions, the ,

operation of Cycle 17 in accordance with Technical Specification Change Request 127 is acceptable based on its reload design and safety analysis.

3.6.2 Unit 2 Cycle 16 Reload The U2R15 refueling outage began on September 23, 1989, with criticality for Cycle 16 scheduled for about November

, 17, 1989. This review covers the Cycle 16 reload design 160

1 and safety analysis work performed by Westinghouse. It j also covers further evaluations perfonned by Wisconsin i Electric. I j

The U2C16 reload contains 12 fresh Region 1&h upgraded l Optimized Fuel Assemblies (OFA) at 3.4 w/o,16 fresh j Region 188 upgraded 0FA at 3.8 w/o, 28 once-burned Region '

17 0FA, 32 twice-burned Region 16 0FA, 27 thrice-burned Region 15 0FA, and 6 Region 10,12, and 13 standard design ;

assemblies. The two Region 12 and 13 assemblies were i initially loaded into Unit 1. The Cycle 16 reload it the j first reload containing a full region of upgraded 0FA fuel for Unit 2. Uppradtd OFA fuel is the subject of TS Change ,

Reques6 #127, which was approved by the NEC in their SER '

+

dated Hay 8,1989.

,, S y g et taiety Evaluation: An evaluttio<a in required be:ause the reload will alter a system, structure or j component deocribed in the FSAh. >

TS chstges required for the operation of cycle 16 have l been reviewed and approved by the NRC. No special i environmental considerations are involved ard. therefore, ;

evaluation of environmental effects by the NRC staff has sho'in that the issuance of environmental impact or 5 assessment statements is not required.

The Unit 2, Cycle 16 reload design and safety analysis is acceptable and falls within the scope of earlier analyses :

and indicates that operation of the cycle 16 core does not r involve a significant increase in the probability or- ,

consequences of accidents previously considered, does not ;

create the possibility of a new accident, does not involve t a significant decrease in safety margin, and does not. ;

involve a significant hazard consideration. Therefore, provided that the startup physics testing does not result ;

in any discrepancies with the analysis assumptions, the operation of cycle 16 in accordance with Technical Specification Change Requests 127 and 129 is acceptable ,

based on its reload design and safety analysis.

161

,m -

I C.0 NUBBER OF PERSCISIEL AND PERSON-REM BY WORK GROUP AND JOB FUNCTION Number of Work Function and Total Parson-Rem

' Personnel Greater Total Reactor Than rem For' Operations & Routine Special Weste Job Group 100 aren Job Group Surveillance Maintenance Inspections Maintenance Processing Refueling Station Eteloyees

operations 65 36.520 26.780 ------

7.130 ------ ------

2.610 Maintenance 47 78.820 ------

53.910 3.170 5.490 ------

16.250 Chemistry and Health Physics' 37 41.400 38.890 ------ - - - - - ------

2.510 - - - - -

Instrtamentation and Control 14 4.630 ------

3.460 0.070 0.250 ------

0.850 Technical Services 4 0.720 0.200 ------

C.100 - - - - - -----

0.420 Ackninistration, Engineering and Regulatory Services 23 9.230 2.660 - - - - -

6.310 - - - - - ------

0.260 Utility Employees 28 36.110 3.730 18.260 2.760 1.860 - - - - -

9.500 Contract Workers l and others 313 266.960 0.680 - - - -

19.30C- 236.900 10.080 ------

GRAND TOTALS 531 474.390 72.940 75.630 38.840 244.500 12.590 29.890 162

h I

, l 5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION j The results of the findings from steam generator tube inspections are l as follows: 1 1

5.1 Unit 1 -

5.1.1 Inspection Plan !

During the Unit 1 Refueling 16 outage, eddy current i testing was performed from April 4,19P9 to April 11, [

1989. An approximate 20% sample was inspected icli !

length and an additional sample was tested through the !

U-bend Teg! n. The extent terted in each steam i generator is as follows:

n! .

Edo g rry,t Insp,ee ger Finn .;

Extont of Inspection Numbey of Tubes Inspr:cted "A" SG "B" SG s Full Length 592 010 U-Bend 50 ___ E l i

i TOTALS 642 661 U-Bend - Through the U-bend to the #6 cold leg ,

support plate.

5.1.2 Inspection Results l The results of these inspections showed no reportable indications (>20% wall loss). The_18 extra inspections ,

done in the "B" steam generator were as a result of a ;

phenomena referred to as manufacturing burnishing ,

marks. These appear randomly throughout the tube bundles and are a result of attempts made to dress or 4 buff scratches in tubes which occurred during the '

initial insertion of the tube into the bundle during manufacture. These indications are not considered deleterious and are, therefore, not reportable, l 5.1.3 Repaired or Plugged Tubes l

l There were no repairs made nor plugs installed in I

either steam generator during the U1R16 refueling g outage.

h l

j 163 1'

i

I

i. . l

\

.g- '52

. Unit 2 5.2.1 Inspection Plan i s

During the Unit 2 Refueling 15 outage, eddy current l j

testing was performed from October 3 to October 13, !

1989. The extent inspected in each steam generator is ;

as follows: 1

[ Eddy Ct'rrent Inspection Plan Entent of Inspection Number of Tubes Inspected *

"A" SG "B" SG L

Hot Leg (Cold Leg) Hot Leg (Cold Leg)

U Full Length 138 150 Sleeves . 138 (12) 141 .(48) ;

n 'to Top of Sleeves (138) (412)

#6 Tube support Plate Cold Leg

20 (20) 21 (21)

1 Tube St:pport Plate 1420 (435) 1512 (229?)-

TOTALS 1716 (605) 1824 (2780) !

These exams performed in Rows 1-4 in the same tubes to constitute a full-length exam. ,

6 4

t 164

~

LL 5.2.2 Inspection Results The following is a summary of the eddy current inspection results listing the largest indication range

, per single tube per Icg.

g Eddy Current Inspection Results i

Hot Leg (Cold Leg)

. a.' so asa so i ;

j <20% 2(3) 15(87) 20-29% 4(1) 4(162) 30-39% (1) 5(77) 40-49% 1(5) 70-79% 1 80-89% 2 1

f 90-99% 1 -,

UDI 1 DI 1(3) 3(13)

NAI 14 1 SAI 6 1 i

TOTALS 32(8) 31(%4) ,

UDI - Indleations whnee quantitative analysis has not been possible but in previous instances have necessitated repair. ,

i DI - Indications whose quantitative analysis has not been possible but in previous instances have not <

necessitated repair.

MAI - Multiple axial indications which require repair. ,

SAI - Single axial indications which require repair. '

h

(

165

o 1 E. i

! 1 t ;

I~

' i 5.2.3 Repaired or Plugged Tubes The following is a list of tubes which were mechanically plugged as a result of indications found during the i U2R15 eddy current inspection. l Plugged Tubes in the "A" Steam Generator Row - Column Indication % Location 5- 4 MAI 4.74" TOR E 15- 7 SAI 3.20" TOR HL i 18- 7 MAI 5.53" TOR E 20-10 UDI 8.0" ATE E >

37-28 SAI 6.1" TOR HL l 1-30 SAI 0.2" ATS HL 1 J 47-31 RST #5 TSP E :

37-33 *18 7.7" ATE E 1 42-33 -RST #5 TSP HL i 32-35 90 3. 6" AIT HL i 37-40 MAI 0.9" TOR HL 37-52 85 10.3" Ats E l 39-54 MAI 1 4" TOR HL !

33-59 SAI

~

0 ,8a Tor. E 33-61 MAI 3.5" TOR ML 5-72 MAI 20.0" TOR E 31-75 MAI 0.2" Tok HL 35-76 MAI 1.7" TOR HL 6-77 '

MAI 10.9" TOR HL 8-77 MAI 7.3" TOR HL 9-79 SAI 10" TOR HL ;

6-80 MAI 5.0" TOR HL 11-84 MAI 0.1" TOR HL 1-85 MAI 1" TOR HL 10-85 SAI TOR HL 12-86 MAI 2" TOR HL 13-89 80 7.5" ATE HL ATE - Above Tube End ATS - Above Tubesheet HL - Hot Leg MAI - Multiple Axial Indication SAI - Single Axial Indication RST - Restricted TOR - Top of Roll TSP - Tube Support Plate UDI - Undefined Indication I.

1.

166

Plugged Tubes in the "B" Steam Generator Row - Column Indication % Location 6- 1 48 #1 TSP HL 6- 7 SAI 3.91" TOR HL 17-18 40 #1 TSP CL 31-19 41 #1 TSP CL 1-50 80 4.1" ATE HL 33-78 MAI TOR HL ATE - Above Tube End ATS - Above Tubesheet 11 - Hot L'kg MAI - Multiple Axial Indication SAI - Single A43a1 indication RST - Restrictel t 70R - Tor af Roll TSP -- T h s eport flate

a. Pitvar.tive Siceviog Program During the U2R15 outage, eclected "B" stear generator cold leg tubes were sleeved using the Westinghouse Rosa system. This was performed due to cold leg wastage progression noted in previous examinations. A total of 298 tubes were successfully sleeved. The selection criteria used was tube location, defect location, and defect severity. The following is the row-column listing of the sleeved tubes:

Row-Column Row-Column Row-Column Row-Column 12-15 8-30 17-32 15-36 11-24 10-30 21-32 16-36 7-25 11-30 8-33 18-36 14-25 12-30 11-33 19-36 5-26 13-30 14-33 4-37 6-26 14-30 16-33 6-37 9-26 35-30 11-34 9-37 6-27 16-30 12-34 15-37 9-27 17 'o0 14-34 16-37 11-27 19-30 15-34 17-37 16-27 4-31 16-34 18-37 3-28 11-31 11-35 20-37 5-28 12-31 13-35 4-38 7-28 14-31 14-35 7-38 8-28 16-31 15-35 8-38 17-28 4-32 16-35 11-38 8-29 8-32 19-35 12-38 10-29 10-32 4-36 13-38 13-29 13-32 6-36 14-38 17-29 16-32 8-36 15-38 167

Row-Coltunn Row-Column Row-Coltaan Row-Column 16-38 3-48 9-55 17-38 8 61 4-48 10-55 10-61 19-38 10 48 -

12-55 20-38 12-61 25-48 13-55 15-61 i

22-38 26-48 14-55 24-38 16-61 2-49 15-55 17-61 6-39 4-49 16-55 9-39 19-61 5-49 17-55 21-61 10-39 12-49 18-55 14-39 4-62 14-49 19-55 15-39 5-62 16-49 21-55' 6-62 16-39 17-49 11-56 17-39 10-62 7-50 12-56 13-62 18-39 8-50 13-56 i- 19-39 16-62 10-50 14-56 17-62 le 20-39 "y- 11-50 16-56 18-62 22-39 12-50 17-56 6-40 19-62 14-50 19-56 21-62 11-40 15-50 21-56 14-40 4 16-50 6-57 10-63 17-40 17-50 11-57 19-40 13-53 24-50 14-57 14-63 22-40 4-51 15-57 23-40 17-63 5-51 16-57 6-41 21-63 8-51 21-57 13-41 4-64 9-51 24-57 14-41 5-64 10*51 25-57 15-41 10-64 12-51 9-58 17-41 11-64 14-51 10-58 18-41 12-64 15-51 11-58 20-41 13-64 16-51 12-58 23-41 14-64 17-51 14-58 21-64 9-42 6-52 15-58 20-42 3-65 11 52 17-58 14-65 22-42 12-52 8-59 6-43 17-65 14-52 11-59 20-43 18-65 17-52 12-59 20-65 22-43 18-52 13-59 23-43 3-66 20-52 14-59 10-66 3-44 3-53 16-59 5-44 12-66 4-53 21-59 17-66 L 12-44 5-53 22-59 4-67 17-44 12-53 5-60 23-44 6-67 14-53 10-60 12-67 3-45 17-53 11-60 24-45 14-67 20-53 14-60 17-67 29-45 5-54 15-60 t

21-46 20-67

' 12-54 16-60 21-67 23-46 19-54 17-60

! 4-68 2-47 4-55 4-61

l. 3-47 5-68 6-55 6-61 8-68 24-47 7-55 7-61 21-68 1'

L 168 L

- 1 l-r

i. [

Row-Column Row-Column Row-Column Row-Column .

4-69 11-70 8-71 10-72 ;

13-69 6 9-71 6-70 7-71 , 11-71 ;

b. Plug Repair Program In response to NRC Bulletin 89-01, 190 mechanical .

, plugs susceptible to primary water stress corrosion :

cracking were either repaired by the plug-in-plug :

(PIP) process or replaced. The following is a list of those plugs which were repaired or replaced.

"A" Steam Generator PIPS i

Row-Column Row-Column Row-Colurin Row-Column ,

1- 3 9-80 33-25 37-47 ,

1-83 10-87 33-30 37-60 '

1-84 12-88 33-62 37-64 1-86 13-84 33-64 39-49 1-89 13-87 34-62 40-44 -

5-79 16- 7 34-67 40-63 4-17 16-83 35-60 40-64 4-73 18-86 35-62 41-13 4-75 20-83 35-63 41-56 4-79 21-85 35-64 41-63 6-76 22-62 35-65 41-64 i

7-36 23- 8 35-73 42-37 L 7-79 23-80 36-36 42-38 l 8- 3 31-66 36-68 43-40 '

8-86 32-33 37-36 44-40 8-88 33-18 37-39 44-41 9-54

! "B" Steam Generator PIPS i

Row-Column Row-Column Row-Column Row-Column 1- 6 6- 4 *7-46 *9-65 1-25 6- 5 *7-58 10- 2 l 1-40 6- 6 *7-59 *10-26 l 1-46 6- 8 *7-62 *10-27

l. 1-67 *6-45 *7-69 *10-45 E 2- 2 *6-46 *8-26 *10-46 L *4-59 *6-47 *8-45 *10-47 l *4-60 *6-48 *8-46 11- 2 l 5-18 *6-59 *8-47 *11-43

5-45 6-75 8-92 *11-45

5-46 *7-27 *9-32 *11-46

5-47 *7-30 *9-46 *11-47

5-48 *7-43 *9-47 *11-48 5-72 *7-44 *9-53 *11-68 5-73 *7-45 *9-59 *12-29 169 i

l I

I Row-Column Row-Column Row-Column Row-Column

12-41 *18-32 *21-48 23-77 l

12-46 *18-40 *21-51 23-78

12-47 *18-48 , *21-52 23-79

13-27 *18-49 *22-49 23-80 '

i

13-54 *18-50 *22-52 23-81

14-43 *18-51 *22-53 23-82 ;

14-46 *18-69 .*22-64 **24-52

15-47 *18-70 *22-67 *24-53 )

15-53 19-10 23-10 *24-55

15-70 *19-44 *23-51 *24-59 ;

15-71 *19-50 *23-52 *25-54 !

16- 4 *19-68 *23-54 *25-55 >

16-47 *20-49 *23-57 26-79 ,

16-53 *20-51 *23-60 *29-46 17- 7 *20-55 *23-65 31-78 17-12 *21-41 23-75 33-77 '

18-29 ;

Sleeve Plug PIPS i

- Attempt was made to install a PIP. It was removed, the txisting ,

plug was drilled out and replaced with a new plug. ;

5.2.4 Tubes with Indications - Not Plugged or Repaired The following is a list of tubes which had indications .

but were not plugged or repaired as a result of eddy !

current inspections performed in 1989. .

"A" Steam Generator Indications UB - U-Bend 1H or C - Tube Support Plate No. Hot TSC - Tubesheet Cold Leg Leg or Cold Leg TSH - Tubesheet Hot Leg RST - Restricted to Standard Probe DI - Distorted Indication 1-39 - Percent Throughwall Indication NOTE: ALL INCH MARKS ARE ABOVE THE REFERENCE LOCATION; NO INCH MARK INDICATES AT THE REFERENCE LOCATION.

Tube Indication Reference Row-Column Description Location Inch Mark 1-15 RST 6H 1-73 RST UB 1-78 RST TSH 3- 9 RST TSH 3-45 RST UB 3-82 RST UB 3-83 RST UB 4-16 19 TSH 5-16 14 TSH 6-26 16 TSC 1.1" 6-33 14 TSC 0.5" ,

170

-e.

"A" Steam Generator Indications (Continued)

UB - U-Bend 1H or C - Tube Support Plate No. Hot TSC - Tubesheet Cold Leg . Leg or Cold Leg TSH - Tubesneet Hot Leg RST - Restricted to Standard Probe DI - Distcrted Indication 1-39 - Percent Throughwall Indication NOTE: ALL INCH MARKS ARE ABOVE THE REFERENCE LOCATION; NO INCH MARK INDICATES AT THE REFERENCE LOCATION.

Tube Indication Reference Row-Column Description Location Inch Mark i

7- 1 DI IC 8-91 RST TSH 0-92 RST TSH 9-16 RST UB 9-38 25 TSC 1.B" 9-89 RST TSH

. , . 9-90 RST TSF, 9-91 RST 'JSH 10-88 RST TSH 10-90 RET TSH 10-91 RST TSH 11-86 RET TSH 11-91 RST TSH 12-87 RST TSH 12-89 RST TSH 12-91 RST TSH 13-88 RST TSH 13-90 RST TSH 14-88 RST TSH 14-89 RST SC 17-38 DI TSC 0.7" 17-50 RST 1H 17-82 RST TSH 17-87 RST TSH 18-24 32 6C 0.6" 20-37 13 TSC 0.5" 21-78 21 TSH 6.0" 25-10 RST TSH l

30-15 DI IC 33-16 RST TSH 34-18 RST TSH 36-55 RST 2H i 36-59 25 TSH 3.7" 36-59 14 TSH 2.2" 36-72 R51' TSH >

38-49 RST TSH 39-41 RST TSH 39-45 RST 2H 40-26 28 TSH 5.4" 40-40 RST TSH L

171 ,

l "A" Steam Generator Indications (Continued)

UB - U-Bend 1H or C - Tube Support Plate No. Not TSC - Tubesheet Cold Leg Leg or Cold Leg ,

TSH - Tubesheet Hot Leg RST- - Restricted to Standard Probe '

DI - Distorted Indication 1-39 - Percent Throughwall Indication ,

NOTE: ALL INCH MARKS ARE ABOVE THE REFERENCE LOCATION: NO INCH MARK INDICATES AT THE REFERENCE LOCATION. !

Tube Indication Reference Row-Column Description :

Location inch Mark ,

40-45 RST TSH 41-38 I RST 2H 41-39 RST TSH 41-50 RST TSH 41-59 RST TSH 42-34 RST TSH 42-35 RST TSH 42-36 RST TSH 42-57 RST TSH +

42-50 RST TSH 42-59 RST TSH ;

42-64 DI 1H 43-34 RST TSH 43-36 RST TSH 43-39 RST TSH 43-41 RST TSH 43-44 21 1H 43-56 RST TSH '

43-57 RST TSH 43-58 RST TSH 43-59 RST TSH 44-36 RST TSH 44-37 RST TSH 44-38 RST TSH 44-52 RST TSH 44-57 RST TSH 44-58 RST TSH i

l I

l 1

l l 172 I;

I

p t ..

"B" Steam Generator Indications UB - U-Bend 1H or C - Tube Support Plate No. Hot TSC - Tubesheet Cold Leg Leg or cold Leg TSH - Tubesheet Hot Leg RST - Restricted to Standard Probe DI - Distorted Indication 1-39 - Percent Throughwall Ir.dication NOTE: ALL INCH MARKS ARE ABOVE THE REFERENCE LOCATION; NO INCH MARK INDICATES AT THE REFERENCE LOCATION.

Tube Indication Reference Row-Column- Description Location Inch Mark 1- 1 RST UB 1- 2 19 1C 1- 2 RST UB 1- 7 RST UB 1*19 RST UB 1 22 12 1C 1-23 RST UB 1-24 RST UB 1-30 RST UB 1-45 DI TSH 1.2" 1-47 RST UB 1-68 RST l'B 1-49 RST UB 1-52 14 TSC 12.3" 1-55 RST UB 1-61 RST UB 1-61 RST UB 2- 1 18 TSC 8.9" 2- 9 DI IC 2-15 RST 6H 2-33 RST UB 2-61 RST UB 2-62 RST UB 2-63 RST UB 2-64 RST 6C 2-65 RST UB 2-66 RST 6C 2-67 RST UB 2-68 RST- UB 2-69 RST UB 2-70 RST UB 2-71 RST UB 2-72 RST UB 2-73 RST UB 2-83 RST UB 2-89 RST UB 3- 1 34 1C 0.1" '

3-20 16 TSC 3.4" 3-38 11 TSC 0.9" 3-38 33 2H 3-83 RST UB l

173

l.

i I

"B" Steam Generator Indications (Continued) ,

UB - U-Bend 1H or C - Tube Support Plate No. Hot TSC - Tubesheet Cold Leg Leg or Cold Leg TSH - Tubesheet Hot Leg RST' - Restricted to Standard Probe DI - Distorted Indication 1-39 - Percent Throughwall Indication -

NOTE: ALL IWCH HARKS ARE ABOVE THE REFERENCE LOCATION; NO INCH HARK INDICATES AT THE REFERENCE LOCATION.

Tube Indication Reference '

Row-Column Description Location Inch Mark 4-58 RST UB 5- 2 21 1C 5- 3 13 TSH 2.8" 5-17 13 TSH 5-83 RST 6H 5-88 RST UB 5-92 RST UB 6-15 38 TSH 39.1" .

6-16 DI fSH I U 6'89 RST 6H 7- 1 23 TSH 7.6" i 7- 1 13 TSH 6.2" ,

6-23 12 TSC 3.2" 8-23 25 TSC 4.1" 8-23 28 TSC 5.6" 9-74 RST UB 10- 3 DI IC 10- 3 19 TSH 3.6" 10-76 26 TSH 0.1" "

10-90 RST TSC 11- 3 RST TSH 11- 4 RST TSH

[ 11-11 14 TSH -

l- 11-29 16 TSC 6.6" 11-29 19 TSC 1.6" 11-29 24 TSC 0.9" 11-29 28 TSC 2.5" 11-90 RST TSC 12- 2 36 1C 0.1" 12- 3 RST TSH 13- 4 39 1C 4 14-13 RST UB >

15-13 RST UB 15-72 16 TSC 5.1" 15-73 11 TSC 5.3" 17- 8 34 1C 17-13 RST UB l 18-12 19 TSH 42.0" 20- 6 27 1C 0.2" 20- 9 DI IC 1

174 l

i "B" Steam Generator Indications ;

(Continued)

UB - U-Bend 1H or C - Tube Support Plate No. Hot TSC - Tubesheet Cold Leg Leg or Cold Leg ,

TSH - Tubesheet Hot Leg RST- - Restricted to Standard Probe DI - Distorted Indication 1-39 - Percent Throughwall Indication NOTE: ALL INCH MARKS ARE ABOVE THE REFERENCE LOCATION: NO INCH MARK INDICATES AT THE REFERENCE LOCATION.

Tube Indication Reference '

Row-Column Description Location Inch Mark 21- 7 30 1C 22- 8 19 1C >

22-12 33 1C 22-14 18 TSH 49.5" ,

22-36 24 1C 23-29 32 1C

?.3-69 DI 10 24-21 12 TSH 45.1" 26-27

{

19 1H 5.10 26-71 14 TSC 5 5" .

26-75 29 TSH 49.6" 26-75 35 TSH 50.3" 37-69 DI IC 27-83 37 TSH 1.7" 28-23 DI TSH 0.4" 29-16 26 1C 29-29 33 1C 30-49 29 1C 31-66 39 1C 31-73 16 TSH 10.5" 32-70 DI IC 33-19 27 1C 33-46 34 1C 33-48 34 TSH 0.7" 33-58 36 1C 33-73 DI 1C 33-74 DI 1C 34-45 17 1C 36-17 17 TSC 19.2" l 35-17 18 TSH .11.9" i 35-73 15 TSH 1.8" l 36-20 22 IH 21.5" 36-48 27 TSH 36.4" !

36-48 25 TSH 42.3" l 36-48 25 TSH 35.4" 36-69 DI 1C 37-21 28 1C 37-48 15 1C 37-68 26 1C 38-21 17 TSH 11.7" I i

I I

175

"B" Steam Generator Indications (Continued)

UB - U-Bend 1H or C - Tube Support Plate No. Hot TSC - Tubeshee; Cold Leg Leg or Cold Leg TSH - Tubesheet Hot Leg RST ~ - Restricted to Standard Probe DI - Distorted Indication 1-39 - Percent Throughwall Indication 5

NOTE: ALL INCH MARKS ARE ABOVE THE REFERENCE LOCATION; NO INCH MARK ;

INDICATES AT THE REFERENCE LOCATION. L Tube Indication Reference i Row-Column Description Location Inch Mark :

38-49 25 1C 41-55 RST TSH 42-36 RST 6H 42-52 19 TSH 13.7" '

43-59 RST TSH 44-55 RST TSH 44-56 RST TSH 44-57 RST TSH 45-54 RST 2H i

5.3 Errata - Steam Generator Plugging and Sleevino (Unit 2)

The following errors have been discovered in the 1989 Annual Operations Report (AOR).

5.3.1 R12C63 in steam generator A is listed in the AOR as being H/L sleeved during U2R9 (June,1983). In fact, the 1982 AOR and Westinghouse's Field Service Report *

(FSR) and raw data from U2R9 reveal that this tube was plugged during the previous U2R8 (April,1982). ]

5.3.2 R31C54 is not included on the steam generator A H/L !

sleeving list. Westinghouse's FSR and raw data from i U2R9 clearly show that this tube was sleeved during !

that outage. 1 5.3.3 R10C17 is included on the steam generator B H/L sleeving list. Once again, review of Westinghouse's I raw data and FSR from U2R9 reveals that this tube was I

" honed" as an alternative sleeving candidate, but was never actually sleeved.

5.3.4 R18C17 and R27C28 are not on the steam generator B H/L sleeving list. As in 5.3.2 above, this is incorrect.

1 176

i: ,

6.0 RE. ACTOR COOLANT SYSTEN RELIEF VALVE CHkLLENGES 6.1 Overpressure Protection During Normal Pressure & Temperature !

Qperation t

There were no challenges'to'the Unit 1 or Unit 2 reactor coolant system power-operated relief valves or safety valves at normal operating pressure and temperature in 1989. l t

6.2 overpressure Protection During Low Pressure & Temperature Qperation There were no challenges to the Unit 1 or Unit 2 power-operated I relief valves during low pressure and temperature operation in i 1989.

7.0 REACTOR COOLANT ACTIVITY ANALYSIS, r There were no indications during operation of Unit I and Unit 2 in !

1989 where reactor coolant activity exceeded that allowed by Technical Spec 3tications.

s l

l l

l L

l t

i 177 l

__

ML20012A185

Text

1.0 INTRODUCTION

3.2.2 AOP-14A

Navigation menu

Search