ML20147E069

Annual Results & Data Rept,1987

ML20147E069
Person / Time
Site:	Point Beach
Issue date:	12/31/1987
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
CON-NRC-88-18 VPNPD-88-122, NUDOCS 8803040258
Download: ML20147E069 (103)

Text

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WISCONSIN ELECTRIC ANNUAL RESULTS AND

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POWER COMPANY 1987 A

POINT BEACH NUCLE AR PL ANT UNIT NOS.1 AND 2 l

l 1P7 1D U.S. Nuclear Regulatory Commission 8803040258 871231 Docket Nos. 50-2M and 50-301 (DR ADOCK 05000266 DCD Facility Operating License Nos.

DPR-24 and DPR-27

s PREFACE This Annual Results & Data Report for 1987 is submitted in accordance with Point Beach Nuclear Plant, Unit Nos. 1 & 2, Technical Specification 15.6.9.1.B and filed under Docket Nos. 50-266 & 50-301 for Facility Operating License Nos. DPR-24 & DPR-27, respectively.

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B TABLE OF CONTENTS E

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1.0 INTRODUCTION

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h 2.0 HIGHLIGHTS 2.1 Unit 1 1 1: 2.2 Unit 2 1 i 3.0 FACILITY CHANGES, TESTS & EXPERIMENTS 4

i 3.1 Amendments to Facility Operating Licenses 2 s 3.2 Facility or Procedure Changes Requiring NRC Approval 3 b 3.3 3.4 Tests or Experiments Requiring NRC Approval Design Changes 8

11 3.5 Temporary Hodifications 65 3.6 Core Reload 72 4.0 NUMBER OF PERSONNEL & HAN-REM BY WORK GROUP & JOB FUNCTION 76 .

5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION E 5.1 Unit 1 77

[ 5.2 Unit 2 78 6.0 REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES 6.1 overpressure Protection During Normal Pressure &

, Temperature Operation 99 R- 6.2 overpressure Protection During Low Pressure &

k Temperature operation 99 7.0 REACTOR COOLANT ACTIVITY ANALYSIS 99 E{E 1

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1.0 INTRODUCTION

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The Point Beach Nuclear Plant, Units 1 and 2, utilize identical 'h4 .f. _

(4[ if; pressurized water reactors rated at 1518 MWt each. Each turbine-generator is capable of producing 497 MWe net (524 MWe gross) of # . , a ',

electrical power. The plant is located ten miles north of Two Rivers, -

Wisconsin, on the west shore of Lake Michigan. 11*

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2.0 HIGHLIGHTS A_ ._;'5 ;,

$ 2.1 Unit 1 f(2h~r -

Highlights for the period January 1, 1987, through December 31, j 1987, included a 57-day refueling / maintenance outage, a brief

- outage to repair a steam generator level instrument root isola-

[ tion valve, and a forced outaged cause0 by a failed open pres-

. surizer spray valve which resulted in a reactor trip on pressurizer low pressure.

Unit 1 operated at an average capacity factor of 83.9 percent f and a net electrical / thermal efficiency of 32.6 percent. The F unit and reactor availability were 84.0 percent and 84.4 percent

! respectively. Unit 1 generated its 53 billionth kilowatt hour on March 14, 1987; its 54 billionth kilowatt hour on i

August 5, 1987; and its 55 billionth kilowatt hour on

[-

October 25, 1987.

2.2 Unit 2 s

b Highlights for the period January 1, 1987, through _

[ December 31, 1987, included a 46-day refueling / maintenance E outage, a brief outage to repair a reactor coolant system flow P transm.itter isolation valve, a brief outage to repair a moisture separator reheater and both feedwater regulator valves, a brief

' outage to adjust secondary chemistry necessitated by tube leaks T in the blowdown evaporator, a trip caused by lightning strikes, -

E and a brief outage to repair a 19,000 V potential transformer.

Unit 2 operated at an average capacity factor of 85.4 percent

- and a net electrical / thermal efficiency of 32.6 percent. The

unit and reactor availability were 86.0 percent and 86.6 percent respectively. Unit 2 generated its 52 billionth kilowatt hour E on January 1, 1987; its 53 billionth kilowatt hour on March 26, 1987; its 54 billionth kilowatt hour on June 6, 1987; and its 55 billionth kilowatt hour on September 9, 1987.

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3.0 FACILITY CHANGES, TESTS, AND EXPERIMENTS 3.1 Amendments to Facility Operating Licenses During the year 1987, there were five license amendments issued by the U.S. Nuclear Regulatory Commission to Facility Operating License DPR-24 for Point Beach Nuclear Plant Unit 1 and five license amendments issued to Facility Operating License DPR-27 for Point Beach Nuclear Plant Unit 2. These license amendments are listed by date of issuance and are summarized as follows:

3.1.1 02-02-87, Amendment 105 to DPR-24,  !

Amendment 108 to DPR-27 l These amendments revise the limiting conditions for operation (LCOs) for the component cooling water system to reflect the addition of a fourth heat exchanger to th. system.

3.1.2 03-03-87, Amendment 106 to DPR-24, Amendment 109 to DPR-27 These amendments revise the surveillance requirements for the main steam stop valves, main steam safety valves and pressurizer safety valves. The periodicity for testing main steam safety valves and pressurizer safety valves is changed from once each refueling to once every five years. Test conditions for main steam stop valves are changed from no-flow to low-flow conditions.

3.1.3 03-17-87, Amendment 107 to DPR-24, Amendment 110 to DPR-27 These amendments change the expiration date for the Unit 1 Facility Operating License No. DPR-24 from July 19, 2007. to October 5, 2010 and change the expiration date for the Unit 2 Facility Operating License No. DPR-27 from July 25, 2008, to March 8, 2013.

3.1.4 05-27-87, Amendment 108 to DPR-24, Amendment 111 to DPR-27 l

l These amendments modify the Technical Specifications to remove certain limitations on the repair of leaking fuel rods so long as the repairs proposed are justified by a cycle-specific reload analysis.

3.1.5 12-03-87, Amendment 109 te DPR-24, Amendment 112 to DPR-27 These amendments incorporate a change to Technical Specification 15.4.11, "Control Room Emergency Filtration," and revise the test parameters for the laboratory analysis of in-place charcoal adsorbant.

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s The amendments also incorporate administrative changes ,

to non-radiological Technical Specifications 16.1 and 16.5.

3.2 Procedure Changes There were no procedure changes made during 1987 beyond those authorized with license amendments as noted above, which required l Nuclear Regulatory Commission approval. The following procedure changes made at Point Beach Nuclear Plant during 1987, required a

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10 CFR 50.59 review:

3.2.1 ICP 10.39: Rod Bottom Bistable Relay Jumper Installation.

Summary of Safety Evaluation: Installation of rod bottom bistable relay jumpers is a change to the ,

facility as described in Section 7.3.3 of the FSAR and addressed in TS 15.3.10.D.3.a.

Section 7.3.3 of the FSAR describes the RPI rod bottom position detection system as a backup to the nuclear instrumentation system (NIS) for rod drop protection.

Additionally, the turbine runback and the rod withdrawal stop actuated by the rod drop detection circuits are considered "control" system actions versus "protection" ..

system actions. The purpose of these actions is to prevent exceeding design fuel limits, but they are not relied upon or required to assure safety. Also, TS 15.3.10.D.3.a allows for failures of one or more of , ;,

the RPIs.

Therefore, based upon the description of the rod drop protection system in the FSAR and the provisions made for failures of the RPI system in the Technical Specifications, it can be concluded that jumpering out one or more rod bottom bistable relays will not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety will not be increased. The change does 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 is not reduced.

3.2.2 IT-40B (Unit 1): Stroke Testing 1-850A&B While Containment is Pressurized at 30 psig (following Type "A" verification).

Summary of Safety Evaluation: An evaluation is required in accordance with the provisions of 10 CFR 50.59. The conduct of this test is not described in the FSAR.

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! The proposed test is being performed to verify the E adequacy of the Sump "B" valves and operating system. [hT,

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> It should be noted that there are no known or c3 'l f I anticipated problems regarding the adequacy or aC" i operability of these valves. The test basically s. .

. consists of opening the subject valves, one at a time, ;g. - -

[ while the containment is pressurized to approximately ;7 ' "

30 peig with air, following the verification phase of s the scheduled containment integrated leak rate test 4"o

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P (CILRT). IT-40B will administratively control the c e '{ y 7 proposed test and ensure that the necessary test gh, d prerequisites are accommodated. .2 '

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I. The subject piping system will not be subject to overpressurization since the containment pressure is only 30 psig.

l The RHR system will not be affected. This statement is based upon administratively ensuring that the associated

- 851 valve is shut (single valve isolation and that primary system pressure (also RHR suction pressure) is maintained between 50 and 100 psig. This will ensure that any possible leakage will be out from RHR suction toward containment; thus ensuring that the associated RHR pump suction is not subjected to a potential slug

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- of compressed air.

! It should be noted that slugging the operating pump with

[ air in this condition would cause a loss of pump suction and RHR flow (i.e., loss of core cooling). This would be limited to the pump associated with the tested 850 valve as the cross-connection path contains a check valve restricting flow to the second pump. Obviously, this is not desirable, however, AOP-9C addresses recovery.

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[ The potential for a loss of RHR could be minimized by

[ operating the pump not associated with the tested valve.

System alignment and draining will be administratively controlled to ensure correctness and to minimize the empty piping volume to minimize the resulting pneumatic i volume. This will also serve to minimize any potential piping slugging and keep the seal surfaces wet (to seal L as they were designed).

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$ The safety injection system feature is not needed during this condition since the reactor is in a cold shutdown h condition. Both trains of the RHR system are required

- by the Technical Specifications during this condition.

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The probability of occurrence or the consequences of an .

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accident or malfunction of equipment important to safety

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is not increased. The change does not create the lt . g possibility for an accident or malfunction which has not a sj a been previously evaluated. The margin of safety as , , ,*

defined in the Technical Specifications is not reduced. .-

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3.2.3 ORT #3 (Unit 2): Safety Injection Actuation with Loss 1 ? .s of Engineered Safeguards AC. Temporary hi ,i ^ . ,

.:U' Technical Evaluation: The intent of the revised ORT #3 Cf(?

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procedure and LW-61 and 62 configuration change is to { .Tt ,

perform ORT #3 without securing or significantly hj C,4.{ @{;

perturbating the blowdown evaporator. This approach is being taken to prevent the potential for blowdown p 1-evaporator bottoms leaking into the steam side of the  %.k (N reboiler in view of the tubing condition. j;j.,;]  ;

c Radwaste steam usage will be minimized. The gas [v s,,. 3 strippers will not be in service, and the blowdown gfg N evaporator reboiler level vill be raised to minimize the boiling rate. The blowdown evaporator internal Mj w ?t ^

pressure will be maintained at approximately 5 psig. 3 y p .l p ,

Just prior to running ORT #3, the steam supply will be  ;%. ( ,/

aligned to Unit 2 (no steam) and ORT #3 will proceed. Q Upon cycling of the steam supply valves from an SI s '. ".~ ' ',

signal, the steam supply will be transferred back to . . t ;.J i, Unit 1. ,:g.(: f Thus, steam will be lost for only a minimal amount of .

time (approximately two minutes). This should be of i short enough duration to prevent bleeding the reboiler steam pressure to less than the blowdown evaporator pressure (50 psig to 5 psig).

The service water isolation valves to the blowdown evaporator vent condenser will be temporarily reconfigured per the procedure to allow for manual operation. The intent is to prevent losing service water to the blowdown evaporator to ensure that no upward internal blowdown evaporator pressure transient occurs. If pressure was to increase, the vent condenser relief would still keep the blowdown evaporator at a pressure lower than the steam pressure of 50 psig.

The temporary change is non-QA and will be adequately administratively controlled by the procedure. System restoration and post restoration testing are addressed in the procedure.

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( for a release by maintaining the reboiler steam side  :[#

pressure above the blowdown evaporator system pressure. N( 3,'

1o If any condition is experienced (stuck steam valve, etc.) which presents the potential for losing this j,9 reverse pressura differential, operations will be 1

. a prepared to immediately shut the blowdown evaporator n ,

down and isolate radwaste condensate. Thus, the proposed procedure change (s) are adequate. p( .l..

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.] Summary of Safety Evaluation: The proposed change E modified the ventoff port on the instrument air solenoid 4[4.W .

valves controlling service water valves TV-LW-61 and 62. 4 3g4 The modified ventoff port line will have an isolation valve, a pressure gauge to monitor air on the valve 4dC operator (when the vent is opened), a quick disconnect t to allow repressurization with instrument air, and the

$ necessary tubing and fittings. This arrangement will h allow operations to take manual control of TV-LW-61 and k 62 to accommodate the performance of ORT #3 during U2R13. The intent is to take manual control during the performance of ORT #3 and maintain the subject service .

water valves open to maintain the blowdown evaporator in yI operation. This will preclude any potential for a blowdown evaporator fluid release through the reboiler a,5~g V g

L to the radwaste steam / condensate system. i K .'

e L TV-LW-61 and 62 provide service water isolation to the j ,[-

[ blowdown evaporator distillate cooler, blowdown 1, <. .

evaporator overhead condenser, and the Unit 1 and 2 jf ] e blowdown vent condensers in the event nonessential '%

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= service water isolation is required. .Qg The proposed change is adequate in view of the '

following

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. .Q h a. It is temporary in nature. k.~

4 Its use is controlled by plant procedure (s).

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f c. While it is installed, it will be assigned a b[,

j dedicated operator and will be monitored by an t/. ' ).'

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d. The configuration will allow for manual closure of -

the affected service water valve.

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e. The configuration will not alter the failure mode of the service water valve itself (fail closed).

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Tre configuration will meet the design requirements of ,

'i t pV toe instrument air system. Component pressure ratings 'y p rF will be compatible. The configuration will not be ip ,.

seismic, however, it will not impact the selsnic rating  ?-; ,{ .. .

of the service water system. The change does not py,4 present an additional PAB flooding concern.  : . F. r "

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The change does not pose an unreviewed safety question. 4.%fg(a,"

The probability of occurrence of the consequences of an ['_j~s' accident or malfunction of equipment important to safety j .. ,s.' .,g -

is not increased. The change does not create the fgg, r fj;g possibility for an accident or malfunction which has not p f ~y been previously evaluated. The margan of safety as defined in Technical Specifications is not reduced. .t ^)4;(.j f  !

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Subsequent to aporoval of the above procedure, an evaluation was conducted of this procedure in accordance with the requirements of 10 CFR 50.59.

Summary of Safety Evaluation: WMTP 11.19.2 proposes to install vacuum hoses to the valve bonnets of valves 2MS-235 and 2MS-237, which are located on the steam generator safety salve header This evolution constitutes a potential change to the facility of its operation as described in the FSAR. 2MS-235 and 2MS-237 supply steam to the radwaste system and the Unit 2 steam-driven auxiliar y feedwater pump during

.iormal operation. The procedure will be performed with the reactor in cold shutdown and bor ated to refueling shutdown concentration. Reactor coolant system pressure will be between about 250 psig (the minimum required to supply flow to the No. 1 reactor coolant pump seal) and 415 psig (the low temperature o'erptessurization protection system setpoint). RHR will be in operation, and both reactor coolant pumps will be running. Liquid drawn through the vacuum hose will be routed to the facade sump Gases will be dischar ged thr ough a compr essor exhaust to atmosphere.

The procedure provides for periodic surveys and samples to monitor potential r eleases via this pathway.

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In the unlikely event that the vacuum hose should fail,

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- a situation similar to a steam line break would occur.

I This accident is analyzed in FSAR Section 14.2.5.

The conditions postulated in the analysis are much more r

severe than the actual conditions of the procedure h (primary system temperature less than 200 F), as the FSAR analysis assumes operation at normal operating E

temperature when the break occurs. Therefore, the PSAR analysis bounds the conditions of the procedure.

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e Recent steam generator samples have indicated that l

activity levels on the secondary side are less than  ;

R MDA. Also, extensive steam generator eddy current leak testing of tubes has established the integrity of the

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primary-to-secondary system boundary of the steam N generator tubes. ,

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Other leakage from either the primary auxiliary building or facade is expected to be minimal and thus, insignificant radioactive leakage from either location

[f to the discharge of the compressor is expected.

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The change does not pose an unreviewed safety l

a question. The probability of occurrence or the F--

consequences of an accident or malfunction of equipment-important to safety is not increased. The change does F not create the possibility for an accident or

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malfunction which has not been previously evaluated.

The margin of safety as defined in the Technical f Specifications is not reduced.

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3.3 Tests or Experiments g The following tests or experiments performed at Point Beach g Nuclear Plant during 1987 required a 10 CFR 50.59 review:

e 3.3.1 SMP #844: Perform Torsional Response Test on ITG01

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(2TG01).

The torsional testing on the turbine generator, as

& controlled by SMP #844, required bringing turbine speed 7 up to 1900 rpm. In order to do this, IOPS was reset to 1950 rpm, the mechanical overspeed protection was l[ manually overridden by an operator when turbine speed is 7

21850 rpm but $1950 rpm and the auxiliary governor overspeed protection circuit was in the test position.

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In addition, the generator protective relaying for low field current, neutral overcurrent, and phase l" differential current was defeated to prevent a turbine trip. The generator output disconnect was locked and tagged open.

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[' Changing the IOPS setpoint to 1950 rpm and manually g overriding (operator holding overspeed trip lever in ,

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test position) the mechanical overspeed protection .j),.. t.N 7 ,- - 4 P between the 1850 and 1950 rpm, along with the conditions Q g %. L .

of the EMP (disconnect switch open and small generator h9-

. load) did not significantly increase the probability i -

,A for the turbine generator to exceed maximum speed of / 1.:; Q '

132 percent as specified by Sections 10.2 and 14.1.12 of - . , ,. ..,.

J the FSAR.  : ~ /.

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The only mention of the generator protective relaying ip _j.l is given in Figure 8.2-2 of the FSAR. This protection p. 4 %

I is for generator safety and is not related to reactor ;jW / ' lie

,$ operability or safety.

f The test did not pose an unreviewed safety question.

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.3 j The probability of occurrence or the consequences of an j{? :V.Q.. ,k.

M accident or malfunction of equipment important to safety Il.] f was not increased. The test did not create the  ; ' .A. (- .

possibility for an accident or malfunction which has M '[

.X not been previously evaluated. The margin of safety as defined in the Technical Specifications was not reduced.

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'S 3.3.2 WMTP 9.16 (Major), Steam Generator Moisture Carryover  % e

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Determination, Revision 7, dated January 19, 1987. *

(Permanent) #;>~

Summary of Safety Evaluation: The procedure describes J 9-the conduct of a test or experiment not described in the j ] .:j-

. FSAR. A liquid spill of 300 millicuries outside the yf g:-

-f primary auxiliary building does not constitute an .h "h'Q; accident because a monitored release from the plant of C

' this size is within allowable discharge limits of [ 7'{i J,k

, 10 CFR 20. The chances of a spill are low and .- .

J precautions are taken to minimize the spread. Na-24 [q[g2 7

-Q also decays quickly (T 1/2 = 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />). :3 $

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J" Na-24 injected into the steam generators would not mask '? f!-

the occurrence of a primary-to-secondary leak. High I,; 1 ', . e U radiation levels (2E-3 pCi/cc) due to Na-24 injection .U[jj ..,

f occur only in the steam generator. The condenser air {rg. .

ejector monitors are essentially unchanged by the j, ./ y addition of Na-24. If a primary-to-secondary leak ,, - ' y, i i developed, it would be noticed immediately on the c. "r condenser air ejector and primary auxiliary building i i .' . '

vent stack monitors (as mentioned in the FSAR). .I *"'

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Release of the 300 millicuries of Na-24 (which is > ..

, equivalent to 0.103 curies of Co-60) from the steam  % .; '.

generator to the environment due to a steam line rupture  ;

would not constitute an increased consequence of an J, analyzed accident. The steam rupture accident was ~.

analyzed with a tube rupture just prior to the steam ' f. ., , .'

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rupture. -

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Primary system activity was assumed to be equal to j i equilibrium 1% fuel defect during analysis. Seventy ye

[ thousand pounds of reactor coolant are assumed to enter '

F the steam generator. Eleven thousand curies of Xe-133 x

! equivalent and 80 curies of I-131 equivalent are -

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released to the secondary side during the analysis. _

! The 70,000 pounds of reactor coolant add 1140 curies of

[ Co-60 equivalent. The addition of 0.103 curies (Co-60 I, equivalent) of Na-24 does not increase the consequences "?

h of the accident. _

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) In summary, the test does not pose an unreviewed safety

]) question. The probability of occurrence or the kg consequences of an accident or malfunction of equipment _E important to safety is not increased. The test does not EU

( create the possibility for an accident or malfunction J

[ which has not been previously evaluated. The margin of -l J safety as defined in the Technical Specifications is not  ;;

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WMTP 12.12: Electric Auxiliary Feed Pump P38B Flow 2 3.3.3 Test, w

Summary of Safety Evaluation: The procedure proposes a "$

F test or experiment which is not described in the FSAR.

r The purpose of this test is to obtain data to validate 3 I

calculations P87-01 and P87-03. The test will require 2 i operation of the pump (s) beyond their normal flow rate

[ (200 gpm) and operation of the motor (s) above their su normal power output (210 hp). The pump and motor will ;E be closely monitored and shutdown if any unusual vibrations, temperatures, noises, etc., are encountered.

{l 14 E Hotor loading will be kept below the motor capability 4 curve (nameplate hp x service factor at room ambient).

[ This test will be done on a shutdown unit. Reactivity 2 4 addition due to temperature effects resulting from ,,

b auxiliary feed addition will not significantly affect ,

the amount of shutdown. --

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The probability of occurrence or the consequences of an

$ accident or malfunction of equipment important to safety  ;;

in not increased. The change does 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 is not reduced.

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,, ,i There were no plant modifications made during 1987, beyond those .Q.' K.O authorized with license amendments as noted above, which required ' ' . ? ,-

Nuclear Regulatory Commission approval. ( A; t .y

. .: - e 7 The following modifications made at Point Beach Nuclear Plant h '. . .; -

_ during 1987 required a 10 CFR 50.59 review: l ., j -

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3.4.1 M-140, Manipulator Crane Flying Bridge. The portable %d; structure was temporarily attached to the manipulator p$,9,,

f crane of either reactor manipulator during the phase of '@/q F refueling when the internals lifting rig is in service  ! (ji% '-

and access over the rig is difficult for certain operations. Past practice has required unhooking the

$ lifting rig from the polar crane to accomplish comparatively minor operations; this being heavy and time-consuming work. $

a Swnmary of Safety Evaluation: The staff conducted a

[_ safety evaluation of this modification from both a

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nuclear and industrial safety point of view and approved

- the use of the bridge subject to the following

_ conditions: (a) the flying bridge is to be considered P a refueling tool and must meet the same requirements I for installation and removal as any other tool, i.e.,

these operations will not be conducted above an open reactor vessel; (b) portions of the bridge being F aluminum, the bridge will not be stored in the L containment when the reactor is in operation; (c) from an industrial safety aspect, users of the flying bridge will wear a safety lanyard whenever they are on the

_m y bridge for whatever purpose.

3.4.2 M-210 (Unit 1), Securing of Containment Fan Volume 7 Dampers. This modification was submitted as a result of Downey company ventilation system balance checks

[ during simulated accident conditions. The report noted i that branch main volume dampers could be eliminated by securing them in the wide open position. The staff pointed out that system air flow is balanced at y individual branch ducts, and therefore, these volume dampers are not required.

Summary of Safety Evaluation: The safety evaluation 7

accepted the contractor's system balance report as sufficient evidence that the dampers in question served no useful purpose and could be secured open or removed.

a 3.4.3 M-661 (Unit 1), Reactor Coolant System Vent System. The modification required installation of a system which will provide a means to vent the reactor vessel or E pressurizer to the pressurizer relief tank or containment from the control room. The system is required by NUREG-0737.

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E h Summary of Safety Evaluation: System design and

[ installation met or exceeded original plant criteria t and therefore, the integrity of the reactor coolant

. system were maintained. Orifice couplings provide a

class break and flow restriction between the reactor E coolant system and downstream piping. Flow paths from

[ the pressurizer and reactor vessel head to the B pressurizer relief tank and containment were provided

[ with in-series solenoid-operated isolation valves which h fail closed upon loss of power.

3.4.4 M-714 (Common). Emergency Diesels. The modification

, provided prelubrication to the diesel engine lube oil circulating pump systems for emergency fast start ,

requirements in accordance with the vendor, EMD, f_ . ,]g recommendations. It was intended to prevent turbocharger pajf0 bearing damage upon restart of the unit between 9%

e 15 minutes and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after a load run. -N*

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$ Summary of Safety Evaluation: If the equipment [KE, installed by this modification were to fail, the  ;[g.d capability for the diesel to fast start will not be A.

affected. This modification provides additional alarms that will indicate failure of the oil circulation i ((

,_ q system that presently do not exist. The modification i F.r 0, y also reduces the risk of turbocharger bearing wear [l.l; '

caused by a fast start of the diesel between -

15 minutes-3 hours after an extended run. No Technical Specification changes were required, j 5 ;-

L ~3S h 3.4.5 E-206-01 (Unit 1), White & Yellow Instrument Bus (dh$h

[ Upgrade. Addendum 1 to the modification removes

eight annunciators from Cabinet F of CO2 and transfer f them to the Unit 2 ASIP Panel 2C20.

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3 Summary of Safety Evaluation: Eight annunciators s

associated with the present 125 V DC emergency power system were removed. These annunciator removals were

( combined with the startup of the new battery systems which incorporate both the old and new 125 V DC g

y emergency power systems. The required alarm indications

/ associated with the eight annunciators were incorporated

into new annunciators located on ASIP 2C20A and 2C20B.

[ Since an SMP tas used to make the changeover, necessary y safety precautions were incorporated to ensure that loss of any annunciator indication is only momentary.

Relocation of the annunciators does not affect any L Technical Specifications and does not involve any change in system function or operation not previously covered by the parent modifications.

F 1

v 12 F

-mens-

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, m . .

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1

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~: y, 3.4.6 E-206-02 (Unit 1), Instrument Bus Upgrade. Addendum "l , g% .

No. 2 to the modification renumbered the new battery b %.

charger distribution panels and revised the tripping dy

^^

logic of the supply breakers (on SI or undervoltage) to  :'

tripping the individual battery charger contactors on ,

WS . -

SI from either unit concurrent with a loss of normal 8 ' .'

power supply to the A05-A06 bus.  % . ;t, 7 g : ,.

Surnary of Safety Evaluation: Transfer of the tripping  ; .Jf t.

function from the battery charger supply breakers to / V the individual charger contactors was considered an j equivalent alternative. The change of logic from trip '

M{

on SI or undervoltage to trip on SI with a concurrent yp loss of offsite power ensures that continuity of power i e

is maintained if offsite power is available, yet also ;gr{V 7; .-

prevents an overload of the emergency diesel generators ig- ..

if offsite power is lost while a safety injection p!),i];l sequence occurs. This improves reliability of ~ Vj-maintaining the DC power train by improving the . . c. [ '; '.

availability of the battery charger system. . . Q: ..

3.4.7 E-207-01 (Unit 2), White & Yellow Instrument Bus k.Q["?l~

1 Upgrade. (Approved in MSSH 80-104). Addendum 1 to the

., .c.$

modification removed eight annunciators from Cabinet F of CO2 and transfered them to the Unit 2 ASIP

$[.N,; dI[-l' Panel 2C20. k . 2. ' '

Summary of Safety Evaluation: Eight annunciators +*

associated with the present 125 V DC emergency power '..,,

system were removed. These annunciator removals were '

.. '3 d(Yk combined with the startup of the new battery systems

~"

which incorporate both the old and new 125 V DC j emergency power systems. The required alarm indications associated with the eight annunciators were incorporated into new annunciators located on ASIP 2C20A and 2C20B.

Since an SMP was used to make the changeover, necessary safety precautions were incorporated to ensure that loss of any annunciator indication is only momentary.

Relocation of the annunciators does not affect any Technical Specifications and does not involve any change 7 in system function or operation not previously covered by the parent modifications.

3.4.8 E-207-02 (Unit 2), Instrument Bus Upgrade. Addendum No. 2 to the modification renumbered the new battery =

charger distribution panels and revised the tripping logic of the supply breakers (on SI or undervoltage) to tripping the individual battery charger contactors on SI from either unit concurrent with a loss of normal power supply to the A05-A06 bus.

K 13 4

I mz -

.p.

Summary of Safety Evaluation: Transfer of the' tripping .y 7s'ji-,

function from the battery charger supply breakers to _,

the individual charger contactors was considered an #. +['g equivalent alternative. The change of logic from trip f ., .

on SI or undervoltage to trip on SI with a concurrent yl ? .' '

loss of offsite power ensures that continuity of power T is maintained if offsite power is available, yet also prevents an overload of the emergency diesel generators ,.

{1 if offsite power is lost while a safety injection j[

sequence occurs. This improves reliability of ;f$o maintaining the DC power train by improving the M availability of ti.e battery charger system. #@Q y s; 3.4.9 E-231 (Unit 2), Electrical Penetrations. This 1 modification replaced electrical penetration 58 and ~ hi installed new penetrations in spare positions 20 and 22. $[ d_

Summary of Safety Evaluation: The modification does T-?

not have a deleterious effect upon plant safety. The J5 s-ib penetration design is equal to or better than original M: '

plant design.

3.4.10 E-244 (Unit 2), Containment Penetration. The modification installed a new containment penetration in g ..

spare position No. 1 to provide leads for approximately f(.f .

40 new fire detectors which will be installed in  ; " 'd '

containment. '+ k .

Summary of Safety Evaluation: The modification does i .a not have a deleterious effect upon plant safety. The &@ 2 +-

penetration design is equal to or better than original plant design.

3.4.11 E-251 (Unit 2), Containment Electrical Installation.

This modification consists of the installation of electrical cable and required raceways necessary for completion of modifications required per TMI Lessons Learned. These include reactor vessel head venting; isolation valves; fire protection penetrations; primary system pressure, temperature and level instrument upgrades; as well as high range radiation monitoring.

Summary of Safety Evaluation: This modification consolidated the installation of conduit an.1 cable for numerous other modifications and has no effect upon existing plant systems. The design and installation of the cabling systems meets or exceeds applicable criteria causing an upgrading of some existing systems.

14 <'

. ,( -

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3.4.12 E-252 (Common), Auxiliary Building Electrical .

! Installation. This modification consisted of the - 4:a m installation of electrical cable and raceways necessary jkf- '.%

for completion of modifications required per THI Lessons Learned. These included reactor vessel head "*q d)..

venting; isolation valves; fire protection , , c ,. f penetrations; primary system pressure, temperature and -

^--

level instrument upgrades; as well as high range  ?- ..

radiation monitoring; service water, and feedwater s, ,. ,

• i system pressure upgrades. t -x s,+

i.

s..y Summary of Safety Evaluation: This modification ~^ 4 igj 1

consolidated the installation of conduit and cable for dh4 numerous other modifications and has no effect upon gA/

existing plant systems. The design and installation of J.Ed7e the cabling systems meets or exceeds applicable

! criteria causing an upgrading in some existing ((." ,$

2 systems.

3.4.13 E-255-01, Fire Detection System. Addendum 1 to the  %..f.',

modification changed control room annunciation to Q~'f".i; '

operate such that the temporary alarm bell is disabled and "trouble" on panel D400 annunciates the control room alarm. e. y y g y.

F ,.; 3 :

Summary c,; Safety Evaluation: The modification improves '.L <

operator awareness of fire detection system problems. . .

No other fire detection system controls or functions are v affected by this change. Work was performed per special .. , ,-

maintenance procedure to ensure fire detection system fn

operability requirements are met. $t[k 3.4.14 IC-214 (Common), Radiation Monitoring System. The request installed Eberline RMS-II monitors on the auxiliary building vent, drumming area vent, and gas stripper building Radeco pallets with readout to the control room (THI-related).

Summary of Safety Evaluation: The modification was required to provide monitoring of potentially high range radiation levels in post-accident conditions.

3.4.15 IC-215 (Unit 1) & IC-216 (Unit 2), Radiation Monitoring System. The requests installed Eberline RMS-II monitors on the purge stack suction and discharge lines to the R11/R12 radiation monitoring system with readout to the control room (TMI-related).

Summary of Safety Evaluation: The modifications were required to provide monitoring of potentially high radiation levels in post-accident conditions.

I 15

-i ~

m e

5 3 9 E 3.4.16 IC-217 (Common), Radiation Monitoring System. The request installed an Eberline RMS-II monitor on the g

_g 3

r combined air ejector discharge line with readout in the g control roca (THI-related). _5

) Summary of Safety Evaluation: The modification was N

=

t required to provide monitoring of potentially high q range radiation levels irg post-accident conditions.

3.4.17 IC-220 (Unit 1) and IC-221 (Unit 2), containment Pressure. The modifications installed wide range j containment pressure transmitters with continuous p readout available to the control room per Short-Term 1 Lessons Learned requirements. I .,

Summary of Safety Evaluation: Non-nuclear safety 4 i related. .

C l' y 3.4.18 IC-222 (Common), Temporary Technical Support Center B Instrumentation. The request installs a total of 59

? parameters in the temporary TSC located in the plant conference room. Three paratreters were displayed; 4 22 parameters were recorded; and 34 were data logged.

The modification was required to meet Phase "A" NRC i y requirements contained in "Lessons Learned."

Summary of Safety Evaluation: Not nuclear safety

• related.

3.4.19 IC-222-01, Temporary TSC Instrumentation. Several of i the existing instrument loops which supply the TSC have i been replaced with new Spec 200 channels. This "

> addendum provides specific details for the design and ."

E implementation of the revised TSC inputs. The design used presently unused Spec 200 computer outputs to j

g g

provide the TSC inputs. This alleviated the concern of having the new control board inputs connected in series I with the TSC inputs. This addendum interfaces with

} Addendum 1 of Modifications IC-256 and IC-257 which -

[I covers a number of interim configurations for Spec 200 g loops including the TSC inputs.

b

, Summary of Safety Evaluation: Electrical isolation was 71 provided for all Spec 200 outputs. This ensures that E b

L the feeds to the TSC cannot adversely affect the =

Spec 200 loops. Instrument power supplies to the TSC 3 are not affected by this modification. g 2.4.20 IC-244 (Unit 1), Reactor Vessel Level Indication. The _

modification installs a reactor vessel level indication -"

J system as required by NUREG-0737. The system used two narrow range and two wide range differential pressure 3

transmitters and connected a spare instrument -

, penetration on the vessel head for the top connection and an incore detector thimble guide tube for the lower j

"_ connection. y a

16 7 [4 mmm-im---- - -- i

~ , .

, .,l' $/, . -

[, .I.',. . ,4<.

Sunima r y oi Satety Evaluatlon: Approprlate flow Qg j' restriction was provided at top and bottom tap ;gll.p g+

connections to limit extension of the Class I piping i, .;

boundary and to minimize the consequences of a rupture. Design and installation requirements

.t 2 :

j.t'[ s met or exceedea those applicable to original plant - , t design. The system provided indication only. no .t - 7 ,. i c ntrol functions. The modification is nuclear safety . ' , ' . ' . .'

related because it affects the reactor coolant system ?HA,, f ' ;

," pressur e boundar y . /;,[ Q (

1. , 8 : , . W il-t i 3 4 21 IC-256 (Unit 1) & IC-257 Rnit_2]. Post-Accident g .9p.W: . -

e Monitoring Racks. The moditications Install five $ gj, .. f .

Foxboro nuclear seismic Spec 200 analog equipment racks y  :.

[

per unit in the area above the control room to ( }g j r, ,4 t_

accommodate new pr ocess instrumentatton installed per

}p,?E[

~

r i-- NUREG-0578 r equir ement s One rack was powered from p the red bus and two each was powered by the white and ( y. ,

- ye1 low buses  ; kg ', . ',']

v. *M summary of Safets Evaluation Pr ocess acks ate part g' - "

f of add 1tlonal instrumentatlon iequired by NRC i regulations intended to improve nuclear plant safety 1, Design and inst alla t ion r equirement s meet or exceed 2 M.!,

those applicable to orAginal plant design. E, ,r f .

.py ., '%

Additional electrical loads on instrument buses are ,

f e, addressed in modification requests E-206 and E-207 ;w . .u , s Individual instrument loops are addressed in separate ' . t . - '4 s modification requests [ $5 3 4 22 IC-256-01 (Unit 1) & IC-29-01 (Unit /). Fost-Accident -

Monitoring Racks A numier of existing ins t r ument 4 is P -y, .' ,

loops are being replaced with new Spec 200 loops. Due a/& 'rM!iU to the delay of the computer system, a number of interim configurations were nec essar y to maintain existing functions These are old F-2 w iomputer inputs and the TSC t empor ar y displays (. -222) In addition. It is necessary to maintain inputs on both a

- t empor ar s and permanent basis t' o r Reac tor Engineer)ng's E conti01 rcom patch parie1 r

S- Summary of : a t elv Evaluat ton Llect r ic al 1so lat lori is provided ior al1 the Spec 20U outputs to prevent adverse attet ts on the Spec JUU i n s t r ume n t loops The

[ slight Inc r ease in loids on the instrument buses Is negligible K

w

w s....

c

[ 3.4.23 IC-261-02, Radiation Monitoring System. Addendum 2 to ".Q,9,$y j the modification changes all sample pump control .; j 7. .,

l circuitry from a 120 V AC to a 5 V DC circuit. This  !- j c removes the power to the K1 relays in the DAM, changes yjJ out the relay in the AC power box on the pallet, and iM^

rewires the relay output contacts at the motor to the  ?- '

' pallet. The change is required to eliminate spurious tj.*.~

l alarms caused by noise in the AC control circuitry. An i-SMP was required to perform the work.  ;. ' - [

l n_s, y-l Summary of Safety Evaluation: Not required.

%s 2

IC-261-13, Addition of Manual Isolation Valves for 3.4.24 CV-3200EE and 3200FF. This modification addendum k- b:

provides manual gate valves to isolate check valves Q, ki 3200EE and 3200FF. These check valves were found to have unacceptable leak rates during a system leak check.

ff%

p gj g

5. ;, ej. . -

Summary of Safety Evaluation: Addition of the manual iQf'4.g f $[

j valves provides a positive means to isolate the sample system for post-accident conditions. The new yg' l'" '

valves are administratively controlled and have no other effect on the operation or performance of the L y, RE-211/212 radiation monitoring system. [ 9, T p. '.

l 3.4.25 IC-288 (Unit 1) & IC-289 (Unit 2), Pressurizer Safety F , .V '

4 Valve Direct Position Indication. These modifications .' 7 ~;

require the installation of position switches on each l

pressurizer safety valve to provide direct indication .,'. .

of valve position. g, g, Summary of Safety Evaluation: The system only provides indication; no control functions, and is

@h therefore, not nuclear safety ralated. The system's importance to safety is recognized and therefore seismic, environmental qualification and single failure criteria vill apply to the extent practicable.

i 3.4.26 IC-288-01 (Unit 1) & IC-289-01 (Unit 2), Pressurizer Safety Valve Direct Position Indication. LISA Installation in ASIP. Installation and operation of the Foxboro Spec 200 equipment is acceptable based upon the following:

a. Both the supports and Foxboro Spec 200 equipment were included in the ASIP seismic analysis performed,

b. The weight used in the analysis for a full Spec 200 nest, without the power supply, was 35 pounds.

c. Both the static and dynamic analyses were examined.

In neither evaluation were the maximum stress levels associated with this equipment.

18

... ==

~

d. The stress maximums were all less than 4,000 psi i for structural components. The highest was 18,000 psi for the subpanel studs (back wall of N.!

gl,. ((d f tt "',.

panels). The 2"x2"x\" angles had maximums below ,22 .

2,500 psi.

Py-'. 7

e. The maximum panel structural steel stresses are 13'6

^

j .,

of the material yield strength for A36 steel, and therefore, satisfy the specification requirements. 4 .! <

d.i a $ '

Therefore, even a doubling of the Spec 200 weight %1,8 to account for the power supply addition would only M& .;) r!

result in stresses at one-fourth of A36 yield #.- ,

strength if the equipment were located at the point n!{,,d Q of maximum stress, which it is not. Jjfgp' IC-292-01 (Unit 1), Main Control Boards. Addendum 1 to h;t6J

' b,2'E 3.4.27 the modification disassembled a portion of the south Z g{;

control room door and a portion of the attached wall ' fer ..,

and installed a temporary structure. This work is required in order to install ASIP panels in the control Q.f.' *p 's ;f room. e Summary of Safety Evaluation: The fire, security, j s. ..M ventilation and radiological shielding aspects of this modification have been addressed satisfactorily. [ved .f *.y,l(

Removal of the door frame and adjacent metal panels *' L-7 did not degrade the seismic characteristics of the wall.

'(,,i a' '

Yg O $

The temporary wooden structure has been designed to withstand the 0.258 psi pressure differential which  ;

would occur in the event of a main steam line break as described in FSAR Appendix "E," Section 3.2.2. This modification does not change Technical Specifications nor does it involve an unreviewed safety question.

3.4.28 IC-301-01 (Unit 1) & IC-302-01 (Unit 2), Safety Injection. Addendum 1 to the modification installed freeze protection and instrument enclosure heating for ,

refueling water storage tank level transmitters 1(2)LT-972 and 973.

Summary of Safety Evaluation: The additional freeze protection on the affected refueling water storage tank instrumentatica lines increased the reliability and availability of these lines. This also provided redundant power sources from IB32 (2B42) for each line.

- The addition of this load to the diesel generators will not result in overloading due to the small magnitude of the load (less than 0.1*. of rated capacity). This does not result in an unreviewed safety question.

7 19 i

= .

r E

9 3.4.29 IC-312 (Unit 1), Incore Thermocouple Upgrade. The y modifications replace existing reactor vessel 1 incore thermocouple connectors with qualified

connectors. It split thermocouples into separate redundant white and yellow channels and be routed to

_ new penetrations using qualified cable, thus bypassing k non-qualified junction boxes. The cable from the penetrations were routed to the new computer input

racks. Completion of these modifications satisfies j NUREG-0737 Item II.F.2 which requires that incore

% thermocouples be used for redundant safety grade e subcooling meters as an indication of adequate core cooling.

Summary of Safety Evaluation: The upgraded

, thermocouple system does not involve changes to the

- existing thermocouples or penetrations in the reactor head. Therefore, the modifications do not affect the g reactor coolant system pressure boundary. The systen g provides indication only and therefore, is not g nuclear safety related, although its importance in monitoring post-accident core conditions is recognized.

3.4.30 IC-312-01 & IC-313-01, Incore Thermocouple Upgrade.

• Addendum I was presented which provides changed

[ thermocouple core locations, connector stalk locations, quadrant boundaries, and the corrected distribution r into the redundant white and yellow channels.

P D Summary of Safety Evaluation: Not required.

3.4.31 IC-312-02 (Unit 1), Incore Thermocouple Upgrade.

= Addendwn 2 was presented to relocate the reference

junction boxes from containment to the computer room to J provide an environmentally qualified incore thermocouple h system on Unit 1 by the end of the steam generator P replacement outage. Control room indication on ICl21 and input to the plant computer is still maintained.

[

F h Summary of Safety Evaluation: The change will provide

[ an environmentally qualified incore thermocouple system.

E Existing control room indication on IC121 and computer

{r inputs are maintained. The junction box mounting support was evaluated and found to be structurally p sound.

3.4.32 IC-313-02 (Unit 2), Incore Thermocouple Upgrade.

6 Addendum 2 to the modification relocates two reference junction boxes from the Unit 2 containment to the

- existing computer room, including recont. action of cables. When the new computer system is installed and

. operating, the reference junction boxes will be entirely removed.

20

, *s,

^  ;

Summary of Safety Evalurtion: The change will provide kg. ..l ' ' J

! an environmentally qualified incore thermocouple system. Existing control room indication on 2Cl21 and $y'i.W  :. -

computer inputs are maintained. The junction box .%.' ^ '

mounting support was evaluated and found to be .

j structurally sound. /

3.4.33 IC-336 (Unit 1), Main Control Boards. The modification <-

changed the main control boards such that new redundant

.M l channels and new instrumentation could be installed to 5. , . f meet regulatory requirements. The instrumentation ,. r, 1 f includes new redundant channels and displays. 7 (J.3 - Q.~f Summary of Safety Evaluation: This modification Q . f: h '

consolidated the numerous control board changes N. "

required by additional instrumentation being installed .h. j'k per NUREG-0737 recommendations. All work on the main .i . ,

control boards was appropriately controlled and . ,' j ja scheduled to minimize potential impact upon safe plant .< A'-

operation. N . .f i1 '

3.4.34 82-080 (Unit 1), Atmospheric Steam Dump. The modification installed stem bushings in the valve M( ,.

"/[.

bonnet of the steam dump valves and reduced closing spring precompression by removing or modifying the spring adjusting plate. The modification is necessary

• because the atmospheric steam dumps tend to stick in the closed position, particularly if steam pressure is high. The horirontal mounting of the positioner / valve stem compounds the problem.

Summary of Safety Evaluation: The proposed changes reduced the force required to open an atmospheric dump valve, but will not prevent reclosure since there is still adequate spring force available for valve shutoff. Reducing frictional forces in the actuator should improve actuator operation in both the opening and closing directions.

3.4.35 82-085 (Unit 1), Containment Personnel Airlock. The modificatione provide a vacuum pump, isolation valve and asr.ociated hardware necessary to provide a system for leak testing the door seals within three days of opening the containment airlocks at E. 66'.

Summary of Safety Evaluation: The modification resulted in a slight change to the airlock door secondary seal's pressure boundary. This minor boundary change, past the first "0" ring seal in each door, consists of replacing the pipe plug boundary with a short run of stainless steel tubing, associated fittings and a manual isolation valve.

21

h Administrative controls will require that the proposed D isolation valve be normally closed when the opposite  :

f airlock door is opened.

The modification has no effect upon the structural integrity or other design features of the containment g airlocks. Material and installation will meet or exceed original system specifications and code requirements.

3.4.36 82-088 (Unit 1), Transformers. The modification

} replaced the existing 1X01 and IX03 overexcitation relays which have a fixed setpoint with variable i response relays. An additional overexcitation relay

was t.lso installed in the main generator voltage

$ regulator PT circuit.

i

) Summary of Safety Evaluation: Not nuclear safety I related.

i 82-091 (Common), Containment Personnel Airlock. The t 3.4.37 j modification provided a portable system including a i vacuum pump and gauge, in order to leak test the lower containment airlocks (El. 26') in both units, 4

i Summary of Safety Evaluation: The modification resulted y in a slight change to the airlock door secondary seal's pressure boundary. This minor boundary change, past the first "0" ring seal in each door, consists of replacing

< the pipe plug boundary with a short run of stainless

{ steel tubing, cssociated fittings and a manual isolation valve. Administrative controls will require that the 4

7 proposed isolation valve be normally closed when the opposite airlock door is opened, y

[

P The modification has no effect upon the structural

[

integrity or other design features of the containment airlocks. Material and installation will meet or exceed h original system specifications and code requirements.

I f_ 3.4.38 83-122 (Unit 1), Reactor Trip Breakers. The

[ modifications install a shunt trip relay (STR) in each

- of the reactor trip breakers in parallel with the undervoltage trip relay. Actuation of the STRs are I in addition to and as a backup for actuation of the L undervoltage trip attachment. The modification is necessary to meet the requirements of Generic Letter 83-28.

- Addendum to the above modification adds six test points to provide for time response testing of the reactor trip breakers and to facilitate testing of the control board manual pushbuttons. In addition, pushbuttons 1(2)PB/RTR, 1(2)PBA/RT and 1(2)PBB/RT were replaced with devices allowing for physical and electrical

. separation between the "A" and "B" trains on the same

_ pushbutton.

22

Summary of Safety Evaluation: A plant-specific unreviewed safety question was determined to exist ,

concerning these modifications. The responses contain proposed Technical Specification changes which were <

reviewed by the NRC. Installation of the modifications was controlled via an SMP.

l 3.4.39 83-141 (Unit 1) & 85-189 (Unit 2), Auxiliary Coolant

[ System. The modifications add a 2" bypass (globe valve) f around 1(2)-742.

Summary of Safety Evaluation: The modifications did not affect the safe operation of the plant because all work was done in accordance with the requirements of B31.1-1967 and NRC IE Bulletin 79-14 (for seismic -

requirements). The full flow to the refueling water storage tanks will not be affected.

The bypass valves are locked shut to prevent inadvertent ,

crossflow between the containment sumps and refueling water storage tanks. The bypass globe valves have as good or better shutoff / sealing characteristics as the 742 valves. The flow is from under the seat and will =

only see RWST pressure. The valves will be shut during an accident: therefore, no double packing and leakoffs -

will be provided. No Technical Specification changes are required.

I,

• 3.4.40 83-145 (Common), Emergency Diesel Generators. The modification installs double-groove water outlet elbows at each diesel cylinder head per the j manufacturer's recommendation. The modification applies to G01 and G02. -

[ Sunnary of Safety Evaluation: The change replaces a j standard OEM part with an improved OEM part of similar construction. The change is recommended by the manufacturer to improve engine reliability. No Technical Specification change is required. Failure of the improved component would not in any way be different or cause any different effects than a failure of the presently used component. The margin of safety as defined in the Technical Specifications was not reduced.

3.4.41 83-147 (Unit 1), Fuel. The modification repaired Unit 1 '

fuel assembly N02 by removing the failed fuel rod in Position 6 on the south face.

Summary of Safety Evaluation: The mechanical repair of assembly NO2 was successfully completed in January 1984. The NRC safety evaluation for amendment Nos. 27 and 32 dated November 2, 1977, covered repair for both the inert rod and the waterhole options, both of which are covered in Technical Specification 15.5.?.A.1, "Where safety limits are not violated."

23 .

v~ -

An evaluation was performed of the use of Assembly N02 in the' Unit 2 Cycle 11 core with results being reported in a revision to the U2C11 reload safety evaluation.

The change in bottom nozzle' stresses due to the different location of the machined = slot and the ' effects of small changes in. localized crossflow were assessed and found to be acceptable. Similarly,:the small changes in the nuclear design due to relocation.of the few fuel assemblies and BPRAs have no ' adverse impact on

~

, the parameters used in the U2C11 accident analysis.

Thermal and hydraulic considerations were evaluated and since Assembly NO2 has a higher burnup than:the.

surrounding assemblies and since the peak Fxy locations have' changed, Assembly N02 was not in a high power, limiting core location any time during the cycle.

3.4.42 83-158 (Unit 2), Fire Protection. The modification installed alternate shutdown instrumentation including necessary power supplies and transfer switches for.

reactor coolant loop "B" pressure, hot leg and cold leg temperatures; pressurizer level, and steam generator level and pressure. The instrumentation is displayed on 2N04 and 2C205. Sound powered phone capability was'also installed between the local control stations. The.

modifications are required to meet 10 CFR 50 Appendix R alternate shutdown capability commitments.

Summary of Safety Evaluation: Switching channels 2L426 or 2P420C to local without proper administrative controls could result in a Technical Specification violation for failing to meet a minimum degree of~

redundancy.

However, if these switches are selected to local'for testing while at power, their bistables will first be placed in the trip condition *per procedure. .To. assist-in preventing the switches from being selected to local while at power, it was recommended that a cover be placed over the switches which, if removed, would annunciate a main control-board and/or ASIP alarm via un attached microswitch. This would occur at the time of switch access and prior to actual switch operation.

The panel cover has annunciation on the main control board when the cover is opened.

Routine testing will be scheduled during shutdowns.

Emergency operation of the switch would only be necessary when fire has disabled the normal control room instrumentation or made it inaccessible. In either case, the reactor will be in het shutdown as a minimum.

Installation of these modifications was controlled per special maintenance procedure.

I 24 v . -

t y

- 3.4.43 83-159, Emergency Ligliting. The modification provides permanently-installed DC emergency lighting at the

_ designated-alternate shutdown local control stations

.(1&2C205, IN11 and 2N04) and in access and egress routes thereto.

Summary'of Safety Evaluation: Not nuclear safety 2' '

related.

3.4.44 84-024 (Unit 2)', Fire Protection. The modification wraps conduit 2P426 between 2LT-426 and the El. 46.0" floor. It fire seals all conduit connections to the junction box in this conduit area.

Summary of Safety Evaluation: This modification will

-improve / maintain reliability in a fire event and not affect systems operability. ,

1 3.4.45 84-045 (Common), Emergency Diesel G02. The modification installed a test isolation valve for pressure switch PS-3057B and 3058B and an isolation valve between the pressure bleedoff valve and diesel regulator. Identical components already exist on.G01.

Summary of Safety Evaluation: Installation of these valves on G02 will make the starting air systems for both emergency diesel generators identical. Fittings and valves will meet or exceed system pressure /

temperature ratings and system integrity will not be degraded.

3.4.46 84-061 (Unit 1), Steam Generator Channelhead Drain Header. This modification installs a low pressure drain collection system in containment that is routed to sump "A." This drain header facilitates (but is not limited to) the following:

a. Low point drains for isolation valve leak tests;

b. System draining for maintenance activities;

c. Instrumentation fill and vent; and

d. Steam generator primary head draining for inspection.

Summary of Safety Evaluation: Not nuclear safety related.

25

r:

y -. ,

A.

1r o

{5 3.4.47 '84-097 (common), Boric Acid Heat Tracing. This

$ 9. r modification added alarm. indication and a reset button to the boric acid-heat tracing recorder. 'This_was incorporated by. replacing the six leads and recorders-with~a new multibank-recorder. LThe new recorder contains both an alarm light andLa printout of which points are in alarm. The alarm can be' acknowledged.at this new recorder, and a disconnect-function will allow:

the alarming point to be removed from the alarm circuit.

Summary of Safety Evaluation: Addition of an alarm

< indicating light and disconnect function enables the operator to identify alarming points faster. . The disconnect function also allows a point on the recorder to be disconnected from-the control room alarm when maintenance or testing is being performed.

The practice of verifying.the tank and flow path temperature at least once per eight hours will' continue' to be followed.

3.4.48 84-154 (Common), 125 V DC Distribution. The modification replaced the existing spare.70 amp 2-pole 125 V DC breakers with 30 amp 2-pole breckers at positions 23 and 24 for panel D11 and positiens 23 and 24 for panel D13.

Summary of Safety Evaluation: The 30 amp breakers provide proper protection for the No. 10 AWG "A" and "B" train 125 V DC supplies to new ASIP panels 1(2)C20.

This meets or exceeds safety requirements of the National Electric Code Article 140-3 and-Table 310-16.

The increased load does not appreciably decrease the load carrying' capability of the batteries during an emergency condition and still allows the batteries to meet the design load evaluation as specified in the FSAR.

Also, the increase in emergency-diesel loading following a LOCA, provided offsite power is not available, is less than 0.05% during the recirculation phase when the diesel is selected to assume all DC loads via the battery chargers. This also d.es not degrade the emergency diesel loading capability. Therefore, the increased load is satisfactory as concerned to the safety and loading analysis.

3.4.49 84-164 (Unit 1) and 84-165 (Unit 2) Nuclear Instrumentation System. The modification permanently installed a third source range detector per unit.

26

Summary of Safety Evaluation: Installation of one new

. qualified excore neutron flux, detector per unit with displays over the range of 10 6% to 100% reactor. power meets the NRC criteria for the neutron flux variable

. described in Regulatory Guide 1.97. Addition of this detector with supporting inputs and outputs meets all requirements of Category 1 items as specified in the-regulatory guide, such as environmental.and seismic qualification, with the exception of no redundancy.

With this installation, the requirements of 10 CFR 50, Appendix R for source range neutron display on a safe shutdown panel was realized. Also realized was a third permanent source range detector-to be located in a spare excore well, thus eliminating the necessity of a third detector for use during refueling and the exposure

resulting in its installation and removal.

Installation of the detector and all supporting instrumentation and wiring was done'with fully qualified equipment and in accordance with a special maintenance procedure (SMP).

The modification does not constitute an unreviewed safety question, but~will require a change to the FSAR.

Present Technical _ Specifications concerning source range H nuclear instrumentation will not have to be changed.

3.4.50 84-169 (Unit 1), Condensate /Feedwater. This modification replaces the Unit 1 Nos. lA/2A & IB/2B (IHX-17A&B) feedwater heaters. The new stainless steel duplex tube bundles should alleviate copper in the feed train caused by the corrosion of the existing Admiralty tubes. The existing heater shells have been reused.

Summary of Safety Evaluation: Not nuclear safety related.

3.4.51 84-170 (Unit 1), Condensate /Feedwater. This modification replaced the Unit 1 Nos. 3A & 3B (1HX-19A&l98) feedwater heaters. The new stainless steel tube bundles should alleviate copper in the feed train caused by the corrosion of the existing Admiralty tubes. The existing heater shells were modified and reused.

Summary of Safety Evaluation: Not nuclear safety related.

3.4.52 84-237 (Unit 1), Main Steam. This modification installed moisture preseparators in the exhaust leg piping of the high pressure turbine. The preseparators remove sufficient moisture to prevent erosion / corrosion of the crossunder piping and also improve efficiency because of reduced pressure losses in the crossunder piping.

27 m .., , , - - + -

y - -

+ .

s Summary'of-Safet? Evaluation: Not nuclear safety related.

3.4.53_ 84-284 (Unit 1) and 84-281 (Unit 2), Reactor Coolant System. The modifications remove all isolation valves and vent and-drain valves from the RTD bypass manifold piping in "A".and "B" reactor coolant loops, as well as changing existing: orifices to a welded instead of flanged configuration.

Summary of Safety Evaluation: Calculations show that-

.the increased flow will not exceed vendor limitations.

All work was performed in accordance with code requirements. Seismic loadings will be reduced due to the removal of various components. Welded-in orifices have the same flow characteristics as the existing orifices.

The change removed all of the manual valves from the reactor coolant bypass loops. This includes.the isolation, vent, and drain valves in the hot, cold, and-common return lines. Also, the flanged-in flow orifices were removed. The 3" common return metering orifice and the 2" cold leg restricting orifice were replaced with "welded-in" flow orifices of identical fluid hydraulic characteristics.

It should be noted that the hot leg bypass line only contained the flanges for a flow orifice, but no flow orifice. Straight sections of pipe replace the items removed (isolation valves and hot leg orifice flanges).

This change was made to increase system reliability and reduce plant personnel exposure by removing the potential for:

a. RCS leakage due to numerous valves and flanged fittings.

b. Unscheduled outages due to valve plug separation resulting in loss of bypass loop flow.

c. Significant personnel exposures encountered when working directly on the bypass system-or other equipment in the general area (reactor cco. ant pumps and steam generator sludge lancing) l The only drawback of this proposed change is the loss of the ability to isolate the RTDs during a hot shutdown condition. This has been deemed an acceptable operational constraint in view of the performance history of the RTDs in the past, and the fact that installed spares are available.

I 28  ;

tr .:

t t

We have not ever had to isolate and make repairs /

replacements in a hot shutdown condition. The annual refueling schedule presents an adequate maintenance window for accommodating RCS RTD work. If repair was required at another time, the evolution would not present a safety concern, but would impact plant availability.

The change does not present a safety concern in view of the following:

a. The increase in bypass line flow is within component (RTD) limitations as specified by the vendor.

b. The increase in' flow will provide a slightly quicker RTD overall response time, which is conservative from the safety analysis viewpoint.

c. Bypass loop flows were verified using an ultrasonic-type external flow measuring device to ensure adequate flow exists,

d. The resulting RCS pressure boundary does not rely upon valve packing and gaskets, thus significantly minimizing the potential for leakage.

e. All pressure boundary components (except the RTDs) were welded in place (no studs).

f. All work was done per the code and is equal to or better than original design.

g. Seismic loadings were reduced due to the removal of significant mass. The piping and support configuration remained the same.

h. The seismic adequacy of the resulting configuration was verified by computer analysis prior to acceptance of the modification.

i. The small increase in RTD bypass flow is insignificant with respect to loop flows and core bypass flow.

3.4.54 85-059 (Unit 1), Feedwater/ Extraction Steam. This modification replaces the Unit 1 Nos. 5A & 5B (1HX-21A&218) feedwater heaters in their entirety. The new stainless steel tube bundles should alleviate copper in the feed train caus2d by the corrosion of the existing copper-nickel tubes. It was noted that the new No. 5 heaters are longer than the existing ones but they fit in the epace available.

Summary of Safety Evaluation: Not nuclear safety related.

29

.s c

3.4.55 85-107 (Unit 1), Auxiliary-Coolant CVCS Crossconnect, l The modification'crossconnected the' reactor cavity drain via the-spent ^ fuel pool. cooling. system and.the CVCSJin order to:: :(1) allow for continuous' purification

~

of the lower cavity while the?RHR-system is not-available; (2)' provide-more circulation-for water purification in the lower cavity; Land (3)fprovide-greater flexibility'for-filtration.-

' Summary of-Safety Evaluationi In addition to considerations. contained _in the-accident analyses:

(1) all work was performed in accordance with ANSI B31.1-1967; (2)_the. modification did not affect the ability to purify from RHR; (3) the containment boundary was not affected by- the modifications; and (4) the safety flow paths of CVCS and RHR were'not affected by the modifications.

3.4.56 85-209 (Unit-1), Reactor Protection System. The modification relocated.the reactor trip breaker test pushbuttons for the shunt trip and shunt block functions on the panels.in the rod drive room.

Summary of Safety Evaluation: No evaluation required.

Relocation of the shunt trip and shunt block pushbuttons to the side of the shunt trip panel did not change their intended function or design purpote. The exact location of these pushbuttons is not described in the FSAR because-it is not-important to their function. The design of the shunt trip panel as seismic ~is maintained.

3.4.57_ 85-211 (Unit 1) & 85-212 (Unit 2), Radiation Monitoring

, System. The modifications changed the existing wells for 1(2)RE-217 to eliminate the need for milling the detectors to fit into the wells and to provide lead shielding at.least 2" thick to enclose the detectors.

Summary of Saf,etv- Evaluation: No evaluation was required. It was determined that no modification to the wells was needed. Four detectors were purchased with premachined housings that fit into the existing wells.

The detectors have interchangeable internals and reusable housings. Therefore, no additional detectors will have to be machined and modification to the wells is not necessary. Permanent, but removable lead shielding was added around the exposed section of the detectors to reduce the effects of background radiation.

The additional shielding was designed so the piping system remains qualified per B31.1 (Class 1 seismic).

The shielding does not interfere with the detector's function and it will not fall on any other equipment.

, No changes to the FSAR are required.

f 30

i:

e 3.4.58 85-279-02 (Unit 1), Containment. The addendum to the modification installed a leak verification system composed of tubing connected to the post-accident-containment ventilation system.

Summary of Safety Evaluation: The modification constitutes changes to the facility as described in the FSAR. The leak verification system provides a means to deliberately induce a leak on the containment during the verification phase of a containment integrated leak rate test. The system was permanently installed with the exception of the connections to the test panel and the containment penetrations. All connections to the containment boundary were made with removable spoolpieces. These spoolpieces will not be installed during periods when containment integrity is required. I In this manner, the containment boundary compromised by-a non-QA, non-seismic system.

Administrative controls, such as CL-1B, "Containment Integrity Checklist," will be used to ensure the spoolpieces are removed prior to setting containment integrity. Installation of this modification will have no effect upon the containment boundary during periods when containment integrity is established if the proper administrative controls are followed. FSAR Figure 5.3.1 will be revised to show attachment points of the leak verification system.

This change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.59 85-280-01 (Unit 2), Containment Integrated Leak Testing System. The original modification provided for permanently installing new ILRT instrumentation and supporting equipment. This addendum installed module (s) in electrical penetration 2Q01 to provide sufficient shielded cabling for the new ILRT instrumentation system.

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR.

Installation of one or more modules in electrical penetration 2Q01 for use with the ILRT system instrumentation does not involve a change in the facility Technical Specifications or constitute an unreviewed safety question.

31

i j

- o

. The: penetration-is located in an area among penetrations forinonessential services above Pipeway 3. .This is consistent with the guidance provided in Section 5.1.2.5 of the-FSAR,-under the section entitled, "Electric-Penetrations." This section states'that there is

. typically two.to five feet. vertical clearance between penetrations that supply-essential services (safeguards and reactor protection) and nonessential services, with the nonessential services on top.

Separation of redundant trains of safeguards and reactor -

protection are not jeopardized by this modification.

Integrity of the containment structure is not affected by this modification. Design, fabrication, installation, and post-installation testing was conducted in accordance with appropriate codes, standards, PBNP Technical Specifications, and FSAR.

3.4.60 86-006-(Unit 1),' Steam Generator Blowdown l System. The modification removed the "T" signal from valves 2042 and 2045 by eliminating relay 2045X contacts from the valve control circuit. This eliminates all other automatic close signals to the valves except.from PS-5955.

Corapletion of the modifications also eliminates the need g to perform Appendix "J" leak rate testing of the valves.

Summary of Safety Evaluation: Although the steam generator blowdown line is part of the Appendix ="E"

- analysis, there is no safety equipment in the area and the "T" signal on CV-2042/2045 has no effect on protection from this failure.

In accordance with FSAR Chapter 5.2, "Containment Isolation System," this line penetration is a class 4 penetration which requires at least one manual isolation valve located outside of containment.

With removal of.the "T" signal from CV-2042/2045, the combination of the CV-5958/5959 (inside containment) and the blowdown system being a closed system exceed the requirements for a Class 4 penetration. Therefore, removal of the "T" signal from CV-2042/2045 does not involve a change in the Technical Specifications nor does it constitute an unreviewed safety question.

3.4.61 86-008 (Unit 1), Condensate /Feedwater Systems. The modification replaced the existing trim on the main feed regulating valves with Copes-Vulcan "hush" trim.

l 32

, *~ , + .-- -,,,e o e w ~ n r + - ~ -.-w, -<~,

r Summary of Safety Evalua~cion: The overall operational

-and failure characteristics of the main feed reg valves will'not.be affected by' replacement of the existing valve-trim with a "disc stack" anti-cavitation-type of-trim. The valves still are able-to either; fail open or fail closed. The proposal has no effect on the

-valve's pressure boundary.

Failed closed has been evaluated in FSAR

' Section 14.'1.10, "Loss of Normal Feedwater." Fail-open will cause steam generator level to rise until the operator recognizes the event and taken appropriate action. Feedwater isolation would be generated at' 70% steam generator level. .The valve operator remains the same and provides a fail closed valve' configuration.

The proposed trim is balanced. The FSAR closing time of 15 seconds is not affected since the valve operator and control remains the'same. Valve stroke will not increase. The ICP 5.11 opening time of.5 seconds will also not be affected.

The proposed change does not increase the probability or consequences of a-DBA, nor does it present an unreviewed safety question. No changes to the. Technical Specificatior- " required.

3.4.62 86-008-01 ~

Feedwater Gystem. .The addendum replaces ths. .;gliated instrument air supply at 80 psig to the normally closed port of 1(2)SV-466/476A with a direct instrument air supply of approximately 100 psig.

Summary of Safety Evaluation: The modification is a potential change to the facility or its operation as described in Figure.10.2-2A of the FSAR. The P&ID change is insignificant and 'does not warrant a 10 CFR 50.59 analysis as it clearly has no safety implications (regulated versus direct instrument air supply). The proposed addendum allows a main feed reg valve opening time of 8-9 seconds being achieved upon a reactor trip when Tavg is >554*F (slugging feature).

This change was needed in support of the modification for retrimming the main feed reg valves which resulted in a longer valve stroke (approximately 3\"). The longer valve stroke resulted in longer time requirements to achieve full open. With the change, the resulting valve operating times have been shown to be approximately 4-4.5 seconds to close upon SI or high level trips; 8-8.5 seconds to open upon a reactor trip with Tavg >554 F; and 20 seconds to close upon a reactor trip with Tavg <554'F.

l 33

_w - -

The' change is acceptable and does not present an unreviewed safety question in view of the following

-considerations:

a. The SI.and iiigh level' closing times ^ remain at

<5 seconds. Fifteen-second closing times have been shown to be_ acceptable. -This is the only requirement which may have any association with the safety analyses-steam.line break cooldown concerns.

The SI trip of the main feed pumps also minimizes cocidown (excessive cooldown).

b. The opening time of 8 seconds upon a reactor trip with Tavg >554 F is very close to 5 seconds.

Five seconds is referenced only on the logic diagram, and not in the.FSAR; thus it is not a.

E commitment. This feature is intended to ease system

, transients after a reactor trip. However, it is.

not a safety feature and no credit is taken for it in the analysis. In the past, 8-10 second opening times have been accepted with no adverse consequences noted. The logic diagram will be changed to reflect this opening time.

2

c. The closing time is approximately 20 seconds upon a reactor trip with Tavg <554*F. This is in accordance with the logic diagrams. This time is not described in the FSAR; and thus, it is not a commitment. It is not a safety feature. We are, however, meeting the previously accepted times,

d. The slightly longer opening time may even ease post-trip system cooldown transients in relation to the approach to the SI setpoint from reduced pressure operations.

e. All times are conservative since they are obtained from a fully closed to fully open_ position. In reality, events resulting in initiating the specific feature would start from a mid-stroke valve position. Pressure effects are negligible due to the force of the spring in relation to the steam diameter pressure force (balanced valve disc).

f. The original opening time was based on a 2\" stroke valve. The new valve stroke within the first 2h" of travel is relatively quick (comparable to the old valve).

g. Instrument air system quality without a regulator is acceptable (95-105 psig).

h. Normal valve operation is with regulated instrument air, and only the slugging (opening) feature is affected.

34 E

4^

i. All components are capable of the extra 25 psig.

j. Seismic mounting of the solenoid-operated. valves is not affected.'

p 'k. The valve operating logic will not' change.

1. The valves remain failed closed devices.

The changes do not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.63 86-015-02, Fire Protection System. The addendum installed 3-hour rated fire dampers in ventilation ducts which penetrate the control building roof. The roof is presently used to store miscellaneous turbine maintenance support material, making it impractical to cover under the safe shutdown transient combustible control procedure.

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR.

Outside air ducting to control building HVAC is the only system described in the FSAR. NRC commitments as found in safety evaluation reports require that the control building boundary be maintained at a 3-hour fire rating.

The security boundary will also be maintained. All four ventilation ducts were removed to install the new fire dampers and access doors. Security and fire watches were maintained whenever the duct penetration was open.

Fire damper access doors will comply with security requirements.

This modification does not pose an unreviewed safety question. There is no increase in the probability or the consequences of an accident or malfunction of equipment, nor is the possibility of an accident created different than those previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.

3.4.64 86-026 (Unit 1), Reactor Protection. The modification formalized Temporary Hodification 85-038 for Unit 1.

The modification installed a jumper between the R9 resistor mounting stud and the dual source casing in the reactor prottction drawer to send unwanted "noise" to ground.

35

Summary of Safety Evaluation: -The modification helps to provide a more stable.and reliable input signal to the

. reactor protection circuits from the bypass manifold RTDs. There are no circuit changes required. .The grounding of the mounting post will not . affect circuit operation.in any'way. The' modification does not-constitute an unreviewed safety question and does not-require a change to;the Technical Specifications.

i- 3.4.65 86-055 (Common), Acoustic Monitoring. The modification removed the acoustic monitoring systems ~previously installed under MRs IC-206/207 to provide pressurizer safety valve position indication. The acoustic monitoring systems were not environmentally qualified and could not be taken credit for.' Modifications IC-288/289 provide for direct position indication for the, safety valves per installation of the lift indicating switch assemblies.

. Summary of Safety Evaluation: Not required.

3.4.66 86-058 (Unit 1), Reactor Coolant System. The modification changed the reactor vessel lower internals by plugging existing holes in the core barrel and machining new holes in the top former plate to reduce the differential pressure across the baffle plates. The modification resulted in lower velocity and thus, prevents damage to-fuel assemblies stored in areas of the core where baffle joints are present.

Summary of Safety Evaluation: It was determined that the provisions of these safety evaluation reports (SERs 86-014-01 and 86-034-02) were valid for the Unit 1 modification.

In addition to rereviewing the safety evaluation reports, a detailed technical review of the activity was conducted in order to ensure that some of the problems

-experienced during the Unit 2 modification would be resolved prior to initiation of the Unit 1' modification.

The following is a brief review of the various issues raised during the course of the discussions:

a. Core Bypass Flow Limits. Excessive core bypass flow is not applicable to the Unit 1 modification since the limits have been satisfied.

b. Chamfer Depth. New criteria concerning chamfer depth was initiated based on the Unit 2 modification experience. Maximum chamfer depth is now 3/16" per chamfer.

36

i

-3.4.67 86-079 (Units 1 & 2), Main Control' Boards. The modification disconnected the power supply for the

-containment sump "B" level indication utilized prior to installation of a new system installed to meet-THI '

requirements. -The leve1' indication contained-in MCB C01 for the:affected instrument' channels (942 and 943) was removed.

Summary of Safety Evaluation: The modifications are a change to the facility as described in the FSAR.

Containmentusump "B" level indication is provided by level indicators 1&2LI-960 and.961. With the exception of being depicted on the SI system drawing (FSAR Figure 6.2-1, Sheet 1), level indicators 942 and 943 are not described in the FSAR. TS 15.3.5, instrumentation system, Table 15.3.5-5, instrument operating for indication, states that there are two channels for containment water level sump "B" continuous indication.

!? These two channels are provided by LI-960 and 961 which' will not be affected *7 the removal of LI-942 and LI-943. Removal of LI-942 and 943 did not affect associated system operation.

The changes do not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.68 86-085 (Unit 1) and 86-086 (Unit 2), Condensate System.

The modifications provided a controlled flow bypass around the condensate overboard valves CS-38 and CS-39 in order to discharge controlled and metered amounts of condensate overboard.

Summary of Safety Evaluation: The modifications constituted a change to the facility as described in Chapters 10 and 11 of the FSAR. The modifications did not create a new flow path, but rather provided for better flow control of an existing flow path. Liquid releases created by the bypass are monitored by the sampling of discrete volumes. In addition, air ejector and blowdown radiation monitoring will provide radiation indication in the steam / condensate systems.

The bypass was constructed to the requirements of the code. Valving will be provided to estiire leak tightness to prevent loss of vacuum to the circulating water system.

37 b . .

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1

_ ,s Use Jo'fithe . overboard b'ypass wil'llbe controlled.in al-

~

- mannerLidentical:tolthe existing. overboard line. -Thisi

> censists of samplingLto;obtain a discharge permit in.

~

, ~

accordance with. radiological! effluent Technical ~

iSpecifications1(RETS). . ; ItLshould be-noted that there,is J - ~

i cur'rently noL direct-in-line: RMS monitoring via detector.

~

ofLthis-discharge. flow-path. SNo, direct in-line;RMS

- detector monitoring:of.:the-bypass will be.provided.

Li

~

The change.does-not: pose,an unrev'iewed safety question'. -

, The probability of occurrence {or the' consequences. of ani

-accident or'malfunctionL of equipment:important to safety uis not-increased. The change,does:not create.the possibility for aniaccidentior malfunction which has ~

i not' >

- been previously evaluated.' 'The margin'of safety as

. defined in the Technical 1 Specifications is'nott ~ ,

- reduced.

3.4.69

~

.86-091'(Unit 1)'& 86-092'(Unit'2), ' Residual Heat Removal System._.The modifications install' pressure indicator, tubing, valves,.and fittings on_thefdischarge

. side of RHR pumps 1(2)P10A&B inforder.to meet accuracy, and range requirements of ASME Section'XI ,

Article IWP-4100.

Summary of Safety Evaluation: .The modifications constitute changes to the facility as described in the y FSAR. The pressure indicatorf will be used~ for the -

inservice testing program only and therefore,iwill be -

l

' isolated from the system during normal operation.

i ,

- The pressure-gauge was selected to meet the range'and I i> accuracy requirements of ASME.Section I Article >

IWP-4100. IWP-4100 requires'the range of the gauge to

' be less than three times'the reference pressure, 'Since ,

~

the' reference pressure of the RHR pumps is 152 psi, the gauges will be isolated during normal _ decay heat removal-operating conditions to protect them from over-ranging /over pressurizing. Proof / burst (150/500%

full-scale) pressures.for the proposed gauges was above d

system design pressure. The modifications are an

. extension of the containment boundary by means.of D tubing, valves, and Swagelok fittings. The addition was designed per B31.1 and was supported to comply with the

- Seismic Class I considerations of the RHR system.

The proposed indicator was installed downstream of

' the root isolation valve and does not adversely affect the functionality or leak-tightness of the system. ,

Present system leakage testing will further ensure ,

system integrity.  ;

c P

i 5

38

- ,-e -

4-'r" rw1- e o T-V-*'v'4' T F-FT5 9-*'* -W M F=- V'+m*r *w'rC1 w- 4 *-MwT'-~e'w--* 'r'W'*u* V W '= *-*-t--N - '--'= "'rr*W** 'W T

i 6?

!The change does not pose an unreviewed safety' question.

% .The probability of occurrence-orithe consequences of an

' accident or malfunction of equipment important to safety '

is not increased. The change does not create the

-possibility;for an accident or malfunction whichlhas not'

, been previously evaluated. .The margin ofLsafety as defined in the Te' chnical Specifications is not reduced.

3.4.70- 86-1091(Unit 1), Reactor Coolant System. . The' modification will add new sensing points for-reactor-vessel level transmitter ILT-447. ILT-447 was provided with a reference leg. tie-in to the pressurizer vent path.in the. reactor coolant gas. vent system'(RCGVS) and a variable leg tie-in to a new reactor vessel level indicating system (RVLIS) common variable leg manifold. These new sensing points should improve the performance of ILT-447, particularly ,

while draining the reactor coolant system-to 3/4 pipe.

Summary of Safety Evaluation: The modification is a change to.the facility as describe'd in the FSAR. All materials used are compatible with both the RVLIS and RCGVS and meet or exceed the original design specifications for both systems. The connections to both-systems are at originally-installed vent or. drain o valves which are normally closed and serve as system QA boundaries. These valves will continue to be normally closed-following this modification and a second normally closed. isolation valve has been added in each sensing line to isolate 1(2)LT-447 from the RVLIS'and RCGVS

'during normal power operation. Therefore, the probability of a reactor coolant system leak from either the RVLIS or the RCGVS is not be increased by this modificat' ion.

Additionally, as stated in the design reviews for the RVLIS and the RCGVS, any leak is limited to within the capability of one charging pump due to the small tube diameter of the RVLIS sensing lines and the 7/32" orifice in the RCGVS pressurizer vent line. Also, during shutdown conditions when ILT-447 is cut in, any potential leak is limited by tubing and RCS pressure, and is not a safety concern as it does not significantly impact the draindown analysis. ILT-447 will be isolated from the RCS intermediate leg by red-locking shut and administratively controlling valve 1(2)RC-523. TP is will prevent inadvertently creating a core bypara flow path and will prevent erroneous level indication on RVLIS. ,

Figure 4.2-1 of the FSAR will need to be revised to show '

the new connections between 1(2)LT-447 and the RVLIS and the RCGVS. No other changes to the FSAR or the plant's Technical Specifications are required.

i 39 .

i

, The change does not pose'an.unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment ~important to safety

.is not increased. .The change does 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 is not' reduced.

3.4.71. MR 86-110 (Unit 2), ~ Reactor Coolant System. The modification adds new sensing points for reactor vessel level-transmitter (2LT-447). .This provides a reference leg tie-in to the pressurizer vent path in the

. reactor coolant gas vent system and a variable leg tie-in to a new reactor vessel level indicating system.

The new sensing points should improve the performance of 2LT-447, particularly while draining the RCS to 3/4 pipe.

A revised safety evaluation was presented because there was a change made in the final design for this.

modification.

Summary of Revised Safety Evaluation: .As stated in the original report, the tie-in for the new variable leg for 2LT-447 was to be at an existing vent valve at the RVLIS transmitter manifold. During the final design process, it was determined that the tie-in to the RVLIS would be at the sensing line prior to the transmicter manifold.

This would minimize the length of tubing.that must be

-installed to connect 2LT-447 to the RVLIS variable leg.

All materials and components used are consistent with original RVLIS installation. The tie-in was hydrostatically tested to 3106 psig (test pressure of the RVLIS sensing lines following original installation), or leak checked,.as appropriate. 2LT-447 will be isolated from the RVLIS by two normally-closed isolation valves installed at the tie-in. These two valves serve as the RVLIS QA boundary.

' The change will not increase the probability of a primary leak from the RVLIS nor will it decrease the reliability of the RVLIS. The change does not pose an unreviewed safety question. The probability of l

occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.

The change does 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 is not reduced.

40

3.4.72 86-114 (Unit 2), Component Cooling System. The modifi-cation replaced CC return check valve 2-755A in kind and reoriented the replacement valve s an upward angle.

It-additionally: installed a drain in the bottom of the check valve to improve valve seat drainage-and installed a drain in the low point just upstream of the _ valve.

(This was subsequently modified to remove the line slope and include use of a similar valve which has a'"built in-disc slope and no drain valve was' installed.")

Summary of Safety Evaluation: The modification.is a change to the facility as described in the FSAR. The addition of the drain valves and the reorientation of

-the valve and a portion.of the line to 15' from vertical will not affect the safe operation of the plant. This conclusion is based upon the following: (a) The work was done to the requirements of B31.1-1967; (b) The 15* slope'is within the tolerances of the NRC IE Bulletin 79-14 program; (c) the drains have been seismically evaluated; (d) The drain tap in the bottom of the valve and the sloping of the valve have been evaluated by the vendor and found not to be detrimental to the valve;-(e) The modification to the valve and line was made to resolve a particular problem with the configuration of this valve; (f) The changes made by this modification will have little impact upon the system flow parameters.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specificatiens is not reduced.

3.4.73 86-115 (Unit 1) & 86-116 (Unit 2), Containment Ventilation. The modifications improve the operability and performance of the steam generator channelhead ventilation system. New blower and filter units were installed and interlocked with the purge exhaust fans to prevent high airborne iodine concen-trations in the containment due to the channelhead .

ventilation system. The system was originally intended to be a temporary system, however, over the years has become permanent in nature.

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR. The channelhead blowers discharge into the purge exhaust ducting upstream of the containment isolation valves and following this modification are interlocked with the purge exhaust fans. Therefore, all releases from the channelhead blowers will be via the purge exhaust filters and fans and will be monitored by the containment purge exhaust radiation monitor.

41

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.The ducting.for the channelhead ventilation system is

.being replaced under MR 84-016(83-083). These.

modifications install permanent hard ducting from

,the blowers to the purge exhaust system. The new

~

ducting is being routed and supported such that it will '

~G -not~ affect the operability of-any. safety-related

. equipment during a design basis earthquake.

As a-result of the purge exhaust and supply valves'being locked shut,1the existence of the channelhead ventilation system will not result in an.anmonitored release during normal power operation. During an ~

outage,.the. operation of the purge exhaust system or the dependability of the purge exhaust fans will not be affected by the channelhead ventilation or the: proposed interlock.

The change does not pose an unreviewed safety question.

The probability of. occurrence'or.the consequences of an accident or malfunction of equipment important-to safety is not increased. The changes does 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 is not reduced.

3.4.74 86-124 (Unit 2), Reactor Coolant System. The modification removed the RCS "B" loop decon line (45*

fitting, piping, valve 2RC-544, and flanges) and capped the nozzle. (The Unit-1 decon line was previously' -

removed during. replacement of that unit's steam generators in 1983).

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR.

Capping of the "B" bot leg decon line will not affect the safe operation of the plant. The change was done to the requirements of B31.7-1969 and.the additional requirements found in Section 4 of the FSAR.

These requirements include evaluating the change of the original seismic qualification, inspection criteria, and hydrostatic testing requirements. The decon line had not been used in the plant and it served as a radiation trap. The change eliminated some mechanical pressure-retaining joints in the RCS and improve the system's integrity. ,

This change does not pose an unreviewed safety question. ,

The probability of occurrence or the consequences of an ,

accident malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

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3.4.75'86-134 (Unit 1) and 86-135 (Unit 2), Safety Injection-

. System. The modifications changed the safety injection-accumulator' level transmitters from a.

differential. pressure-detection system to a direct-reading capacitance-type. detecting system. The D

modifications required changes to the' detectors, transmitters, piping, conduit,-and support systems.

, Sununary of- Safety Evaluation: The modifications constitute a potential change to the facility or its operation'as de' scribed in Chapter 6.2.2 of the FSAR. In containment, piping changes will be-such that new piping was compatible with the existing accumulator system.

Also, all new piping and conduit was seismically mounted to ensure integrity of the SI accumulator system and to meet two over one criteria for safety-related materials and components per ANSI B31.1.

~The detectors are mounted in piping at the accumulators and were designed to meet the pressure and temperature and chemistry requirements of the accumulator system.

The level detection system is safety grade because determination of accumulator' injection is performed by the safety grade accumulator pressure detection system.

The level transmitters were located remotely in'the computer room auxiliary racks. They were mounted seismically to maintain the seismic qualification of the racks. The transmitter loops are powered from the instrument buses such that both transraitters on an accumulator will be powered by different instrument buses to maintain electrical independence. Also, the change in instrument bus loading is negligible since the instrument loops will be essentially equivalent to the existing circuitry.

3.4.76 87-019 (Unit 1) and 87-020-(Unit 2), Feedwater System.

The modifications removed the interlock which causes the main feedwater control valves to shut when the feedwater bypass reset pushbuttons are depressed.

Summary of Safety Evaluation: No evaluation is required. This modification does not change the feedwater isolation signals or actuation. It only affects feedwater bypass reset, which is considered an operator aid. The feedwater bypass reset is not mentioned in the FSAR or Technical Specifications since there is no safety aspect associated with it. The implementation of the the modification reduced the possibility of a transient or reactor trip due to a loss of normal feedwater since the possible closure of the main feedwater regulating valve (s) is removed.

43

j The installation of'a jumper on the bypass reset lock-in relay ensures.that_the actuation circuit for

feedwater bypass valve isolation is'not adversely affected by contact failure.

3.4.77 87-025 (Unit 1),' Main' Steam System. .The modification changed the worm-to-worm gear ratio from 45:1 to 76:1 for MOV-2019 and MOV-2020 (steam supply. valves to the steam-driven auxiliary feedwater pumps).

Summary of Safety Evaluation: lThe modification is'a potential change-to the facility or its operation as described in the FSAR. This modification improves the capability of the motor operators to deliver the design stem thrust for the valves. As a result of the modification, the valve stroke time will be increased from approximately 13 seconds to approximately 21 seconds.

Increasing the stroke time of:MOV-2019 and 2020 will not adversely affect tne presently installed low suction pressure trip or settings for the motor-driven and turbine-driven auxiliary feedwater pumps. Also, the condensate storage tank level will not be adversely affected by the increased closing stroke time of MOV-2019 and 2020.

The proposed modification will not affect the safe operation of the steam-driven auxiliary feedwater pumps or the plant because: (1) the increased valve stroke time will not prevent the steam-driven auxiliary feedwater pumps from delivering 200 gpm to each steam generator within the one-minute delay assumed in the FSAR. From experience with the old gear ratio, total auxiliary feed flow from the steamer is achieved in less than 30 seconds. The modification added a delay which is less than an additional 8-seconds. Thus, opening time is not a concern. (2) The present torque switch settings recommended by the manufacturer will still be used. The torque switch will protect the stem from excessive forces. (3) The new components are compatible with the motor operators and will not increase the possibility of motor operator failure. The change will not significantly impact seismic concerns (less than four ounces difference in weight) or environmental qualification concerns.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

44

3 In response'to-a number of staff questions concerning the status of MOV modifications and actions being taken

-in. response to NRC concerns relating to MOVs.

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3.4.78 87-035 (Unit 1) and 87-036 (Unit 2), Containment' Spray System. The modifications provide a safer, more convenient method to positively verify leakage past containment spray isolation valve 1(2)-863A&B.during.

performance of the leakage test by adding a' fitting and-cap to both air' test blank flanges inside containment.

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR. The modification provides more accessible. access ports to determine if significant valve leakage exists which eliminates personnel safety concerns which are encountered when using the present low point piping drain hole.

The existing drain holes (1/8" diameter) which are upstream of the proposed location, will be left in the line, but any significant leakage which would present a containment "spray dcwn" potential will still be seen at the proposed connection. Administrative / procedural controls are used to ensure replacement of the caps after each test.

Replacement of the cap mechanically seals the pipe boundary in order to prevent leakage during containment spray. The penetration is in an existing blank flange (air test connection) which also incorporates a mechanical seal.

The addition was designed per B31.1-1967 and materials used were in accordance with FSAR and piping class specifications. The additional mass is insignificant (less than one pound) with respect to the seismic Class 1 requirements of the system. The proposed modification had negligible effects on the spray system hydraulic characteristics. The addition was not hydrostatically tested or leak checked because it is a component of a system which discharges into a pressure vessel which is leak checked periodically.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

45

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3.4.79 87-046 (Common), Emergency Diesel Generators. The modification installed manual switches in the diesel generator #1 starting circuit to provide isolation of the automatic start portion of this starting circuit in the event of a fire in the cable spreading or 4 kV switchgear rooms which may result in a short to ground.

Summary of Safety Evaluation: No evaluation is required. The modification will not affect normal operation of the diesel generators from the control room or from the local control station in the diesel generator rooms.

The switches proposed for installation are Model OTZ switches, which will stay normally in a closed (normal) position and will be opened manually (bypass position) in accordance with AOP-10A, "Control Room Inacces-sibility."

A cover was provided and red-locked to administratively control the switch in the normal position; and additionally, a placard was affixed to identify the switch and the need to verify circuit continuity following use of the switch.

Installation of the switches in the #1 start circuitry will in no way affect the #2 starting circuitry or the manual starting of the diesel generators.

Installation of the modification was accomplished via a special maintenance procedure.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.80 87-047 (Unit 1) & 87-048 (Unit 2), Auxiliary Feedwater System. The modifications add a pressure gauge and associated tubing to monitor the cooling water supply to the steam-driven auxiliary feedwater pump and driver.

The basis for the addition is to accommodate Appendix "R" requirements. AOP-10A outlines the specific scenario.

Summary of Safety Evaluation: The modifications represent a change to the facility as described in the FSAR. The sensing line (3/8" stainless steel tubing) tapped into the existing cooling water supply header branch connection, originally provided for pump seal injection, that was capped off.

46

The resulting configuration meets or exceeds original design criteria. The tubing and gauge was seismically supported and had an insignificant impact on=the qualification of the existing piping. The tubing was run so it would not interfere with turbine / pump operation. The gauge was mounted on the gauge panel (1(2)RK-35) at the front of the turbine.

The modification did not impset the cooling system's ability to provide pump component cooling nor did it have an impact on pump operability. The addition of the sensing line does not affect the area's flooding analysis. The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.81 87-056, This modification consists of two changes: the addition of fuses in the remote (control room) start pushbutton circuit; and the removal of the cooling water supply solenoid valve SV-3755.

Surmary of Safety Evaluation: The addition of fuses to the remote pushbutton start circuit does not decrease the reliability or availability of the diesel-driven pump. The removal of a fuse will initiate an automatic start of the diesel; therefore, an inadvertent removal or failure of the fuse would not go undetected. If a short circuit were to occur in the pushbutton circuit, the fuses would blow and prevent the operation of the circuit breakers of the local control circuit. The fuses are sized to blow before the circuit breakers open. This feature is necessary to accommodate the control room / cable spreading room fire scenario coupled with a loss of AC for 10 CFR 50 Appendix R.

The removal of solenoid valve SV-3755 will have no

. adverse affect on the operation of the diesel. This solenoid valve opens (energize to open) when the diesel receives a start signal. Its purpose is to prevent the flow of cooling water to the engine coolant heat exchanger while the engine is shutdown.

Since the cooling water is supplied from the discharge of the diesel-driven pump (P35B), the function provided by the solenoid valve is unnecessary. Due to the specific piping configuration, there cannot be a situation where cooling water would flow while the diesel was shut down. The removal of this valve eliminates a component whose failure could adversely affect the operation of the diesel.

47

3.4.82 87-067 (Unit 2), Safety Injection System. The modification revises the SI test / recirculation line by' eliminating a tap and valve 883 and by modifying the configuration of valve 882. The purpose of the modification is to eliminate a leak point and to minimize future leakage.

-Summary of Safety Evaluation: The modification constitutes. a change to the facility as described in Sections 5.2 (containment isolation) and 6.2-(safety injection).

The configuration changes were made in accordance with the Code and system requirements. These minor configuration changes will not increase the probability of an SI system leak from the test / recirculation line nor will it. decrease the reliability'of the SI system.

Removal of valve 883 removes a potential recirculation boundary leak path, although of minor significance.

[ The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an

. accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accider.t or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced. e 3.4.83 87-067-01 (Unit 2), Safety Injection System. The addendum to the modification adds a test drain valve downstream of valve 2SI-897A. The drain will be used to test the 897A&B valves for leakage. The drain provides better indication of leakage and will be in a more accessible location than the existing vent valve.

Summary of Safety Evaluation: The addendum to the modification is a change to the facility as described in .

the FSAR. The drain will not affect the safe operation of the plant because (1) it is not within the post-accident recirculation boundary; (2) it does not affect the operation or function of the SI recirc/ test line or decrease the reliability of the SI system; and (3) it meets the requirements of B31.1-1967 and the .

original system requirements. l The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

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3.4.84- 8'i-075A (Unit 1)- and '87-076A (Unit 2),s Safety Injection" System. These modifications changed;the closing.

circuits for the boric ~ acid tank-to aafety injection.

pump' valves 1(2)SI-826B and C by: removing the; .- >

. simultaneous' torque switch andl limit switch. actuation feature.

U -Summary of Safety Evaluation: The modification'is a change;to the facility or'itsloperation'as described in 7 the FSAR. The original circuit was designed for~

simultanec i actuation-of both the' torque:andl limit.

switches upon valve closure. Plant experience has shown-that-there is a potential for;these switches to, actuate independently. If valve closure is stoppe.d by.the limit switch without torquing shut, valve leakage is possible. -

Conversely, if the valve' torques out prior to actuating i E

the limit switch, the potential for valve hammering is created and the torque bypass in the opening direction - ,

would not be made up, which could prevent valve opening.

By removing the simultaneous torque switich and limit' -

switch-actuation feature,.the valves will be-torqued-shut following the receipt of an auto close (and manual-close) signal and the auto close signal will be blocked

~

following valve closure.

The changes eliminated the valve hammering and the valve leakage problem that have occurred. The.. .

modification.will not affect the operation of these valves. The change does not pose an:unreviewed safety question. The. probability of occurrence or~the consequences of an accident or malfunction of equipment  :

important to safety is not increased. The change does

~

not create the possibility for an accident or mal-

. ' function which has-not been previously evaluated. The -

margin ~of. safety as defined in the Technical ,

Specifications is not' reduced. '

b 3.4.85 87-076 (Unit 2), Various Systems. These mod!.fications :l changed the closing circuits' for MOVs 1(2)-112C (VCT -

. to charging pump suction), 1(2)-313 (seal water return ,

containment isolation), and 1(2)-427 (letdown ,

isolation). They also change the motor operator worm o gearing for valves 1(2)-112C and 1(2) '13 to make these  :

l valves self-locking. 1(2)-427 already is self-locking. f Summary of Safety Evaluation: No evaluation is l L required. The closing circuits for these valves were  ;

previously modified to provide for simultaneous  ;

actuation of both the shut' torque switch and a valve L closed limit switch upon valve closure. This was done  ;

! to eliminate the hammering problems associated with auto  !

closure of MOVs with nonself-locking gear ratios.

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h I Plant experience has shown that this simultaneous a Jactuation is difficult to consistently achieve and that these switches can actuate independently. .If valve closure'is stopped by the_ limit. switch without torquing

. shut,_ valve leakag'e can occur'. Conversely, if the valve torques out-prior to actuating the limit switch,?the potential for valve hammering is created and the torque open ' switch bypass switch would _ not be made up, which could prevent valve opening.

The modifications remove this simultaneous switch sctuation and install self-locking gearing. Following the modification, these valves operate in a manner similar to the majority of plant MOVs.

The modifications also replaced the gearing of valve 1(2)-1128 (reactor makeup water to charging pump suction) to make-it self-locking. This was done to prevent valve leakage due to the valve operator relaxing after the motor is deenergized.

The modifications will not significantly change the-stroke time of any of- the valves and will improve valve reliability. _The largest increase will be less than one second. The valves will close within the required times.

A safety evaluation is not required for these modifications because they do not_ change the facility or *

.its operation as described in the FSAR or other licensing bases; do not require a Technical Specification change, and do not change a commitment to the NRC.

The cha'nge does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.86 87-081 (Unit 1 & Unit 2), Primary sampling. The modifications replace the "cap" on the discharge side of 1(2)SC-954B with tubing, fitting (s) and a Whitey valve to allow relocation of the isolation valve test connections outside of the shield wall. The modifications will improve testing efficiency and will reduce radiation exposure.

50

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4-Summary'of Safety Evaluation: The. modifications changed the facility as described in the FSAR'..The configuration change does not affectithe containment boundary. lThe change is outside of the boundary valves (upstream of valve 1(2)-955). The change presents the potential for bringing the RCS pressure boundary outside' of the-shield wall only when valve 954 leaks.

.Thi .ts not a missile' concern as the potential missile is ' a ty small, normally isolated at power and is similar to other isolation valves test connections exposed to primary system pressures outside of the shield wall containment liner. The valve has been oriented to minimize direct trajectory exposure. ,

- The addition was performed in accor dance with B31.1 and will be Seismic Class I. Thus, it will not adversely impact the RCS pressure boundary. It should be noted that a failure of the line and 954B would only present an RCS leak due to the size of the sample line (within the capability of the charging system).

The change does not pose an unreviewed safety question.  ;

The probability of occurrence or the consequences of an accident or malfunction'of equipment important to safety ~

is not increased.

The change does not create the possibility for an 1 accident or malfunction which has not been previously evaluated. The niargin of safety as defined in the Technical Specifications is not reduced.

3.4.87 MR 87-085 (Common), Fuel Handling System. The modification involved the installatio1 of a bracket in the spent fuel pool to be used for sto.-ing both models of the new fuel baskets.

Summary of Safety Evaluation: The modification is a change to the facility or its operation as described in ,

the FSAR. The design of the storage bracket will ensure that the stored elevator basket will be in pla:e during an OBE. The installation or use of the bracket will not damage the spent fuel pool liner plate either in normal use or during an OBE. Control of the elevator basket elevation, while handling spent fuel, is by procedural controls. In the event that the manual winch contacts fail shut, a person stationed at the winch power panel will open the breaker, thus preventing a lift of the spent fuel beyond a safe elevation.

In addition to the changes made under this modification request, other issues relating to the handling of spent fuel in the new fuel elevator were considered in this safety evaluation. A summary of this review follows:

51

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'In the event' of an accident causing the. loss of water-fromfthe spent fuel pool,Lno. additional consequences are- ,

expected as a' result of'the spent fuel' assembly _ located.

-in the.new fuel elevator.

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The use of the'new fuel elevatorito hold' spent fuel- "

will add to-the -list of five .possible. locations.of

,  : spent l fuel, but will not change the ' conclusion. The .

' fuel assembly located on the. elevator' basket will~be the~ .'

closest to the water surface and;therefore, will require'

'the quickest _ operator. attention. 'It is procedurally-recommended that the elevator be lowered to its -lowest .

elevation while unattended during the fuel' assembly modification work.

The seismic qualification of.the new fuel elevator

. assembly is not known. .However,-based on'a static seismic analysis, the elevator, with the new basket and ,

fuel assembly, would withstand the peak OBE acceler- :t ations. -Therefore,'the probability of a fuel handling _

accident due. to a seismic event is not increased.

Gamma heating of the spent fuel pool wall will not exceed the design values. This is based on-the~

following:

em: a. Calculations assume storage of spent fuel'following~ -

three days of. decay. Fuel assemblies to be modified ,

in the new fuel elevator will have typically decayed by greater than three days.

b. Assemblies located in the new fuel elevator are located 10" away from the spent fuel pool wall. The i design values are-from 0.0-4.4". .;

c. The design wall temperatures are steady-state values. Although time required for temperature -

stabilization is not given, it is expected that i stable wall temperatures would not be reached in the approximate two days required to modify an assembly.

The change does not pose an unreviewed safety question. .

The probability of occurrence or the consequences of an  :

accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident er malfunction which has not ,

been previously evaluated. The margin of safety as defined in the Technical specifications is not reduced.

b 52 '

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3.4.88 87-086 (Unit 1) and 87-148 (UnitR2).. Main Steam System.

The modifications remove snubbers R-EB-2-7 (2R-EB-2-2)

,and EB-2-HY-(2R-EB-2-7) and replace snubbers R-EB-2-3,

R-EB-2-1, R-EB-2-6, R-EB-2-4, and EB-2-H17'(2R-EB-2-1, 2R-EB-2-3, 2R-EB-4-4, 2R-EB-2-5, and 2R-EB-2-6) with

- size EA4SS-5 energy absorbers. The energy absorbers will function as passive pipe supports with surveillance-included in our inservice inspection program.

Summary of Safety Evaluation: The modification is.a change to the facility as described in the FSAR.

Wisconsin Electric originally notified the NRC on7 May 13, 1986, of the intent to replace snubbers on the Unit 1 main steam bypass' system with energy absorbers-under the provisions of 10 CFR 50.59. The results of_

two analyses have demonstrated that system piping stresses and support loads meet existing plant criteria.

-Technical Specification changes need not be initiated or submitted to the NRC for these modifications.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change-does 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 is not reduced.

3.4.89 MR 87-097 (Unit 2) & 87-098 (Unit 2), Auxiliary F( 2dwater . System. The modifications will upgrade the turbine-driven auxiliary feedwater pumps. Based upon discussions with the manufacturer, the following changes to the original MR scopes were proposed.

a. Include the installation of an improved governor end/ thrust bearing.

b. Include replacement of the existing gear-type coupling with a disc-type ccupling.

The manufacturer believes the pump-to-turbine coupling is the source of the thrust loads whi,h are causing the thrust bearing to wipe. The modified thrust governor /

thrust bearing consists of a babbitted journal bearing similar to the present bearing with a separate bidirectional ball bearing which replaces the existing babbit face thrust bearing. This anti-friction thrust bearing has a bfyner load capability than the old thrust bearing and therefore, operates cooler and more easily handles thrust forces transmitted through the coupling.

The bearing is an improved design. No machining to our turbine was required for installation.

53

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Replacement of the existing gear-type coupling with a -

disc-type . coupling reduces . the maximur., axial force transmitted by the coupling from the pump to the ,

turbine. The reduced axial force will, in turn, reduce the thrust bearing load, which will lead to improved thrust bearing performance. The disc-type coupling is not as strong as the original coupling (the original coupling was over-designed). The new coupling has-adequate torque handling capability such that operability of the coupling, thus the pump. drive" system, will be assured out to the maximum possible load.- ' Seismic qualification of the turbine / pump set will be maintained by using a coupling with a mass very.

near to the existing coupling.

Summary of Safety Evaluation: The original safety evaluation is valid with respect to the turbine casing flex plate outboard support. The other. changes, coupling and bearing replacement, are not within the scope of the safety evaluation report; however, these changes do not present a change that falls within the scope of 10 CFR 50.59 applicability. Thus, these changes are not required to be addressed by the 10 CFR 50.59 analysis.

3.4.90 87-117 (Unit 1) and 87-143 (Unit 2), Waste Disposal System. The change removes the check valve 1/2-1713 (nitrogen to reactor coolant drain tank) from the containment boundary associated with P12. This is accomplished by cutting and capping the existing nitrogen tie-in to the reactor coolant drain tank (RCDT) gas vent header.

Summary of Safety Evaluation: The change enhances the containment boundary by .emoving an active isolation component (check valve 1713).from the boundary configuration.

The outside containment boundary configuration will instead include a manual valve and pipe cap (1/2-1793 and Swagelok-type cap) or just a cap if it is placed on the containment side of 1793. The resulting configuration is Seismic Class I, conforms to the applicable code requiremercs and is testable in accordance with 10 CFR 50 A).pendix J requirements.

The capability to supply nitroge cover gas to the RCDT will remain via a route through ths pressurizer relief tank (PRT) or the gas vent header. Thus, RCDT cover gas nitrogen for hydrogen control will not be affected. The only cperating limitation associated with this approach is the loss of the capability of supplying nitrogen to the RCDT from the PRT following an event which ruptures the PRTs rupture disc.

54 .

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m' This is not considered to be a significart limitation as the RCDT would not be a major concern at this timo and the~ gas vent header could be used to provide nitrogan.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or mallunction of equipment important to safety is not increased The change does not create the possibility for .i., accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.

3.4.91' 487-118 (Unit 1) and 87-127 (Unit 2), Instrument Air

' System. The modifications add a vent line in the instrument air system downstream'of the containment isolation valves. The proposed "blowdown"'line is similar in design to isolation valve test connections provided to accommodate 10 CFR 50 Appendix J.

Summary of Safety Evaluation: The modifications are a change to the facility as described in the FSAR. The-purpose of the line is to ensure that the PORVs are maintained in a failed close position during an Appendix -

R fire scenario. Use of the proposed blowdown line will be administratively controlled and does not represent a "release" problem since one does not have to combine a DBA with the Appendix R scenario (only time when need exists for use of the valve).

This change does not represent an unreviewed safety question since:

a. The resulting penetration will be configured within the criteria of the FSAR (two boundaries).

b. The resulting configuration will be Seismic Class I and QA-scope, s

c. The resulting configuration was designed, installed, and tested in accordance with the original code and ,

system design requirements.

d. The new boundary was leak tested (not directly) when the check valve ORTS (Operations Refueling .

Tests) are performed. I

e. Instrument air to the PORVs is not a safety-related h feature; the nitrogen and check valves serve an LTOP protection function. The nitrogen is not needed or valved in at power. LTOP is administratively placed

l. into service by shutting the bypass valve. >

l 55

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.S2 87-120 (Unit 1), Safety Injection System. The modification covers the replacement of IMOV-826C (boric acid to safety injection pumps).

Summary of Safety Evaluation: The modification is a change to the facility as described in the FSAR. The previous valve was a 150 lb, ANSI B16.5 class, motor-operated gate valve. The replacement valve is a 300 lb, ASME Class I, motor-operated gate valve. The new valve and resulting piping configuration are Seismic Class I.

The previous valve and pipe ends were Schedule 10. The weld ends for the new valve are Schedule 40. The ends of the new valve were remachined for Schedule 10 weld joint. The machining will not affect the strength or integrity of the SI system after the valve is installed.

FSAR Chapter 6.2 covers valves and fittings installed in '

the SI system. The new valve meets or exceeds the requirements listed in the FSAR and the specifications for the original valve.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.

The change does 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 is not reduced.

3.4.93 87-164 (Unit 2), Instrument Air System. The modification installed diaphragm valves in the instrument air header inside Unit 2 containment, upstream of existing gate valves. The gate valves, which are downstream of the containment boundary (penetration and containment isolation valves) presently demonstrate unacceptable seat leakage during Type "C" leakage tests of the CIVs. This limits the Type "C" test methodology.

56

x 4

Summary of Safety Evaluation: The modification constitutes a change to the facility as described in the FSAR (Figure 5.2-12). The new valves added 5-10 pounds at each location. Pressure ratings of the new valves exceed system design pressure requirements. Operating system pressure drop across the new valves will be insignificant. The valves were designed, installed and supported per B31.1 or equivalent.

The instrument air system is non-seismic, and therefore, no seismic analysis is required. The valves are not within the containment boundary and will not present any significant stress at the boundar y. An inservice leak test was performed after installation and the valve leakage will be surveyed during the performance of the containment isolation valve ORTS.

This addition will not affect the functionality of the instrument air system or containment boundary, and will provide better sealing characteristics to aid in the performance of ORTS.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.4.94 MR 87-187 (Unit 2), Auxiliary Steam System. Snubber EB-8-H206 was added in 1981 to the auxiliary steam piping in the facade to accommodate the large thermal deflections at that point in the piping and the expected need for a rigid restraint during seismic loading.

However, it appears that removing the seismic constraint altogether was not considered at that time. This modification removed the subject snubber.

Summary of Safety Evaluation: The modification constitutes a change to the facility as described in the FSAR. A new stress analysis of the subject piping was performed. The analysis, documented via stress report 200044, verifies that snubber EB-8-H206 may be removed from the auxiliary steam line and still satisfy all applicable piping code requirements.

The FSAR will be amended to remove the EB-8-206 snubber information from the safety related snubber list contained on Table 6.2-13, Unit 2.

Technical Specifications do not mention individual snubbers, but reference them as a grcup back to the FSAR.

l l 57 1

, . +

a s 4

J No' NRC commitments are affected by the removal of this

. snubber.

This change ~will not' increase the probability of an- -

auxiliary, steam system failure nor will it decrease the' reliability of the auxiliary steam system, ~The change:

does not pose an unreviewed safety question.

The probability of occurrence or the consequences'of an accident or malfunction of- equipment important to safety t

. is not' increased. The change does 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 la 'not reduced.

13.4.95- 202 (Unit 2), Containment..s The modificaton consisted -

of drilling numerous small holes (~1/2" diameter) in the containmentifloor slab expansion joints; repairing the:

floor-slab' expansion joint seal at the hole locations. I with'a configuration slightly different than'the-original configuration; and core drilling a 6" diameter I hole in the floor slab. All of the' holes were down 3 to the containment-liner plate. The containment liner plate was not disturbed.-

This effort is being pursued to accommodate investigation (s) into the effect on the containment-liner from trapped stagnant water. The small holes in the expansion joints allow for potential measurements to identify galvanic corrosion cells.

The larger hole did provide a concrete sample for analysis to determine potential corrosion rates and provide visual inspection access to the containment liner.

l Summary of Safety Evaluation: The small holes in the -

expansion joints and the repair of the expansion joint l seal do not present a 10 CFR 50.59 concern. The  ;

' expansion joint design is not described in the licensing basis. The expansion joint seal repair material exposed i p to containment atmosphere was silicone, similar to the original material. The new sealant is slightly ,

different in that it remains more pliable than the i original. The new seal was not made as thick as.the old seal to ensure that the joint can accommodate  !

expansion / contraction cycles without losing a watertight -!

, seal. The barrier provided by the floor slab and i associated joint seals is not a containment boundary.

Thus, the resulting joint configuration is equivalent {

to, if not better than, the original expansion joint .

configuration. }

F f

58 f

. , _ . - . . , __ _ , _ _ , . . _ _ ~_ . - . _ _ _ _ _ . _ _ -

[ y .

4 h

, .r l-

, - The 6"-diameter hole is'being addressed by 10 CFR 50.59

~because of the'FSAR: implication that the containment liner is either' coated with a protective coating or is

'in contact with concrete, The liner at-theLbottom of-the 6"-diameter hole was

~

s

.left in the.as-found condition. Accelerated corrosion of.the liner at this location due to the removal.of the -

concrete core is not expected since the hole will.be.

~

covered with a plate and an airtight seal. .'Thus, the bare = liner will not.he exposed to post-DBA fluid.'

-Corrosion from the environment within the core hole will be minimal in-view of the minimal amount'of air and potential stagnatt fluid.

A bladder / plug device was replaced inLthe core hole'to ensure a minimal-amount of air. Thus, the resulting configuration meets the intent of the FSAR in that the

. liner at this location will.be protected fram accelerated corrosion.

The liner at this location will be-inspected on a periodic interval to ensure that accelerated corrosion is not occurring. This will be performed during the pre-ILRT containment inspection. The performance-of the ILRT will-also ensure that a containment-boundary problem does not go undetected.

The 6"-diameter core hole was placed in a location to'have negligible impact on the strength of the floor slab or tied-in structures. Th'e location was also selected such that internal containment missiles could

r not damage the liner. Missile protection will be provided if required.

The change does not pose an unreviewed safety question, The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

The following modifications were installed prior to 1987 but have not been included in an annual report:

3.4,96 M-088 (Unit 2), MSIV Backup Nitrogen Supply. This modification machined approximately one inch off the "open" stop of the flapper. In addition, three nitrogen bottles were piped into the instrument air supply line to the valve controller to raise the pressure under the piston of the controller to 150 psig versus the precent 95 psig.

59

9-The additional. force at the controller should be

~

-sufficient to counteract the closing force at the flapper caused by differential. pressures-above and below-the flapper at high steam flows. The closing time of the valve is not expected to be measurably affected by.

the increased controller pressure.

Summary of Safety Evaluation: The modification did not.

, , reduce the ability,of.the. valve.to close under~ trip conditions. Increasing the pressure in the control

- cylinder will counter balance the force on the ' valve flapper that is caused by pressure differences above and below the flapper caused by steam flow.

3.4.97 M-164'(Unit 1),. Containment Spray Testing Line. 'This ,

called for theLaddition of a test connection which will-permit utilization of the four inch service air header ,

when performing the test of the containment spray nozzles (required every five years). The original connections were only 3/4 inch and did not provide adequate flow?for a meaningful test. A-tell-tale connection downstream of valves 868A & B was included in ',

this modification to permit verification of possible valve leakage against the containment spray pump discharge head during normal surveillance spray pump ,

testing. ,

Summary of Safety Evaluation: The modification does not obstruct or reduce flow in the' containment spray line. The blank flange will only be removed under administrative control.

3.4.98 M-251 (Unit 1), R-19 Sample Monitor Discharge to Service Water. Modification to R-19 Rad Monitor Sample Piping ,

and Addition of Strap-On Monitor to Steam Generator

• Blowdawn Tank Liquid Discharge Pipe.

Summary of Safety Evaluation: The piping change to the original R19 monitor improved the reliability of flow to the monitor. The new strap on monitor on the steam generator blowdown tank will add a degree of redundancy to the steam generator blowdown isolation ,

function of R-19 and improve the response time. .

3.4.99 M-285 (Unit 2), Installation of Blowdown Evaporator Room  !

Vent. This modification ducted the evaporator. building to and from the auxiliary building.

• Summary of Safety Evaluation: This modification will prevent unmonitored gaseous releases during startup of the waste evaporator.  ;

60

e

- 3.4.100- E-052 (Unit 1)'and E-047 (Unit 2), "MSIV Shut" Relay to AST/ET Trip. This modification will trip the turbine-auto stop trip solenoid and the emergency trip solenoid

'of that unit's turbine whenever-a main' steam isolation valve leaves'the full open position.

Should a main steam isolation valve slam' shut, a turbine trip would immediately take place, reducing the forces on the second= isolation valve should that one also shut.

Summary of Safety Evaluation: Upon'the sudden closure of a main steam isolation valve, this circuitry will

-attempt to trip the turbine before high steam flow-

"wipes in" the second main steam isolation valve.

This modification will result in reducing the stresses in the steam piping that would result from a sudden closure of a main steam isolation valve under a high steam flow condition.

3.4.101 E-065, PAB & Containment Building Crane's Control Circuitry Change. This modification installed a fail safe relay system in the turbine and auxiliary building cranes which will stop.the cranes in the event of an operating function failure.

Summary of Safety Evaluation: The maintenance department 1 built and successfully tested a model of this device. The crane vendor has also approved the modification.

3.4.102 IC-261-07. Hardware Changes to RMS Computers. This addendum adds one-minute average history data capability to RMS, nonvolatile memory capability to DAMS and improved system software (EPRDMS). The software changes include a requirement for manual acknowledgement after an alarm condition has been cleared. The modification is necessary to improve operator interface to the system as well as to provide better system time response.

4 Summary of Safety Evaluation: The one-minute average i boards improve information available to the operator and result in no other significant changes to the RMS function or control. Installation was controlled by special maintenance procedure to ensure proper system operation after modification.

3.4.103 MR 84-260, Unit 2, Reactor Coolant System. The modification removed the internals of RTD manifold valve 2-564A and replaced the internals with a check valve cap of the same size and rating as the valve.

61 w% _ . - - . _ _ __. - -__ . _ _ _ _ - _ _ _ _

. Summary- of Safety Evaluation: .The modification did not affect the safe operation of the plant because (1) all work was performed to the appropriate standards and codes;-(2) the_ check valvo and cap are interchangeable with'the stop-valve for the same pressure / temperature rating (per telephone conversation with Rockwell-Edwards);-(3) little change in the flow rate occurred. Any-adverse change in the flow rate will be detected by the flow monitoring orifice. No Technical Specification changes are required.

3.4.104 Containment Fan Cooler Air Flow, Fan blade pitch-changes were made on-the Unit I and Unit 2 containment accident fans which reduced the overall airflow rates.

An analysis was completed with respect to containment heat removal capability of the accident fans. The analysis provided a graph which detailed the relationship between heat- removal rate and service water inlet temperature for each fan cooler unit at an air flow rate of 34,000 cfm. .The graph indicated-that the heat removal rate is reduced to 48.5 mBtu/hr at the design service water temperature of 75*F. This heat removal rate is only 1.5 mBtu/hr below the design rate. _Therefore, an 11.7 percent reduction in air flow rate results'in only a 3 percent heat removal rate reduction in the fan cooler performance. ,

Summary of Safety Evaluation: The analysis evaluat.d the impact of this reduced heat removal capability on the containment analysis.for the transients assumed in FSAR Section 14.3.4, "Containment Integrity Evaluation." The FSAR identifies the three scenarios of energy release that have the greatest potential to challenge containment integrity (i.e., the 60 psig design capability) as follows:

a. Blowdown from a large area rupture of the RCS.

b. Blowdown from a large area rupture with post-blowdown energy addition from a core residual energy and hot metal energy as steam (all available energy).

c. Blowdown from a large area rupture with post-blowdown energy addition from a worst case zirc/ water reaction. ,

t of the three cases considered, Case "a" is the most realistic scenario. Cases "b" and "c" were evaluated to demonstrate the capability of the containment to l withstand post-accident energy additions without credit i for core cooling. Case "b" is highly unrealistic since ,

the accumulators and SI system will act rapidly to' reflood and subcool the core.

l l

l 62 '

. . _ _ _ _ , . _ _ . _ _ ~ _ ,- _ __ ..

s te r 3

a For Case "c,"fsufficient energy necessary to cause a metal / water reaction was unrealistically stored in the core rather than transferred to the coolant.

, For' all of- the above analyses; three mechanisms were modeled for containment atmosphere heat-removal as-follows:

-- Containment structures

- One train of containment spray starting 60 seconds into the transient.

'One train of containment fan coolers (two~ fan units) starting at oO seconds into the transient.

For Case "a," a range of large area ruptures.was postulated for calculating the containment pressure.

The double-ended rupture, which is the. largest atea break, resulted in the maximum peak pressure of 53 psig.

Except for the 0.5 fta area break, the peak pressures reached by all of the postulated ruptures . occur prior to the initiation of the containment fan cooling system.

Although this system is initiated for the 0.5 fta break, it only removes 0.01 percent of the blowdown energy (reference FSAR Table 14.3.4-4), and thus provides a very limited amount of heat removal. -Therefore, the reduced performance of the containment fan cooling system does not affect the containment integrity for this transient.

The analysis provided a graphical representation of the design heat removal capability of each mechanism over the duration of the case "b" and "c" transients, as shown in FSAR Figure 15.4.5-12.- A dashed curve was added, however, to the figure to indicate the reduced capability of the containment fan cooling system-at the current air flow rate. Two additional graphs were provided which represent the heat removal addition to the containment from the case "b" and "c" transient energy release. The heat removal curve acts as a cap over the heat addition curve. That is, provided the heat addition curve remains below the heat removal curve, the containment design pressure of 60 psig will not be exceeded. Based upon analysis of these curves, it can be concluded that the reduced capability of the containment fan cooling system has a minimal effect on the containment capability margin. Therefore, the reduced air flow rate delivered by the accident fans is acceptable and is covered by the existing FSAR analyses for Cases "b" and "c."

63

m I: e-

~

e The three cases evaluated in the FSAR for-containment 4-l integrity are not significantly affected by reduced fan

. cooler air' flows. Therefore, the operating restriction of 66*F maximum inlet service water temperature to the inlet of the' coils may be removed.

In addition .to evaluating the effect'of the reduced system performance upon the accident analyses, the analyses studied the effect of system performance on containment ambient temperature during the summer months when service water inlet temperature to the coils is 70*F.

The design heat load the containment fan cooling system under normal operation was 3.8 mBtu/hr as established by' an early 1970s': analysis: however,. based upon thermal data measured during Unit 2 fan testing, it was found that the actual heat load is closer to 5.25 mBtu/hr.

Using actual heat load data, it was determined that the average ambient. temperature should not exceed 110*F with three fan cooling units (six fans) operating or 104*F with four fan cooling units (eight fans) operating.

Therefore,-the containment fan cooling system with the present air flow rate should be capable of maintaining the containment ambient temperature below the design value of 105*F.

In conclusion, the containment fan cooling system is capable of removing heat from containment at an acceptable rate under both post-accident and normal conditions with the present fan blade pitch setting and design service water inlet temperature. The reduced cooling capability, however, is a change to the facility as described in the FSAR, and thus requires an evaluation in accordance with the requirements of 10 CFR 50.59. The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.

The margin of safety as defined in the Technical

, Cpecifications is not reduced.

64

10 3.5 Temporary Modifications 3.5.1 87-005 (Unit 2), Rod Position Indication. The temporary modification jumpered out the rod Ill rod bottom relay (installed January 30, 1987; removed November 6, 1987).

Summary of Safety Evaluation: The temporary modification is a change to the facility as described in the'FSAR. TS 15.3.10.D.3.a~ allows for the failures of' the RPI(s). The balance of the RPI system was' operable.

The NIS rod drop system was also operable.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.5.2 87-007 and 87-016 (Unit 2), Rod Position Indication System. The temporary modification installs a jumper on the rod bottom relay of rod K7 (87-007 installed February 7, 1987; removed April 3, 1987;87-016 installed April-5, 1987; removed November 6, 1987).

Summary of Safety Evaluations _ The. temporary modification is a <;nange to the-facility as described in the FSAR. TS 15.3.10.D.3.a allows for the failures of the rod position indicators. The balance of the rod position indication system,was operable with the exception of the rod bottom bistable for Ill (per temporary modification 87-005). The NIS rod drop system also was operable.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.5.3 87-018 (Unit 1), CRDM Cooling System. The temporary modification disconnected the Unit 1 CRDM cooling fans 1W3A&B from their motor feeders in order to provide 480 V AC power for the upflow modification equipment (MR 86-058) from the supply breakers for 1W3A&B. The power requirements for the upflow equipment are less than that for a CRDM cooling fan so 1B01 (1B02) loading will not be aff'ected (installed April 22, 1987; removed May 9, 1987).

65

?

• P

-? e >

ISummary'of Safety-Evaluation: 'Theitemporary modification is a- potential = change to the facility as described in the FSAR.-- Unit l' was in a . refueling shutdown' condition with the1 reactor vessel head removed and the cavity' flooded when this temporary. modification

-is required. The CRDM cooling fans are primarily

~

. intended to cool.the CRDMs during operation;'therefore,-

they are not'necessary during-thisLplant condition.

LThis temporary. modification was; removed'and IW3A&B restored to their normal configuration prior'to; replacing the reactor. vessel head and' system heatup.'

~

The change does notl pose an 'unreviewed safety; question.

The probability of occurrence or .the consequences of an z accident or malfunction of' equipment 'important to safety.

is not increased. The change-does 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 is not . reduced.

- 3.5.4 87-019, ICV-466 MFRV Manual Gag. On September 14, 1986, the Unit 1 "A" main feed reg valve was placed on a-manual' gag in order to perform emergency repairs to the valve. controller air supply line (installed September 14, 1987; removed September 14, 1987).

Summary of Safety Evaluation: The' temporary .

modification was a potential change to the facility or its operation as described in the FSAR; The evolution gagged valve ICV-466 open to the as-found, throttled position on a manual gag.

This action removes the automatic closure function of-the valve. Automatic closure is specified'during a safety injection and following a reactor trip (after cooldown) with Tavg >550*F and following steam generator overfill. The only portion of the valve closure function which falls into a protection area from the safety analysis standpoint is the safety injection trip. 'This function is redundant in that an SI signal will also trip both main feed pumps (each of which, in turn, isolates its associated discharge valve).

L To provide redundancy of actuation, an individual was assigned to trip the main feed pumps and shut the discharge valves should a safety injection on the unit occur or, if required, following a reactor trip or steam generator overfiil.

l 1

l 66 m

The change did not pose an unreviewed safety question.

The probability of occurrence of 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.

3.5.5 87-021 (Unit 1), Reactor Coolant System. The temporary modification replaced RTD bypass isolation valves 1-562A&B, 1-563A&B, and 1-559A&B with temporary loop isolation covers which are welded to the reactor coolant pipiag. The temporary modification was needed as part-of the installation process for the removal of the RTD bypass valves, vents and drains with the installation of integral flow control devices per MR 84-264 (installed April 14, 1987; removed May 21, 1987).

Summary of Safety Evaluation: The temporary modification is a change to the facility as described in the FSAR. The covers provided a boundary to the RCS while the installation work was proceeding with MR 84-264. During this time frame, the reactor was

• in refueling shutdown. The reactor coolant pumps did not run. The covers were of comparable stainless steel material and had a thickness greater than the associated piping and the minimum thickness requirements for-associated fittings. The welds were sized to provide adequate strength and were compatible to the piping material and Code requirements. The welds were visually inspected and leak tested.

The lowest elevation cover (for 1-563A at El. ?6.26')

was only a few inches lower than the minimum elevation i considered for a steam generator nozzle dam rupture.

The evaluation of that accident was found to be acceptable and is documented in MR 86-045. The largest diameter cover (3") is significantly smaller than that posed by the rupture of a nozzle dam. In addition, this elevation is above the half-pipe level of the >

reactor coolant system. Therefore, the temporary covers provided an adequate boundary to the reactor coolant system.

The manifolds were not needed for Tavg during this mode.

The PORV and associated block valve were locked open until the RCS was open to ensure no potential existed for significant pressurization and stressing of the cover welds.

The covers were removed when the final tie-ins of  ;

MR 84-264 are made. The piping affected by the plate '

weld was removed and replaced to remove the heat-affected zone.

67 -

I l-The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

, 3.5.6 87-037, Incore Thermocouples. The temporary modifica-tion cut and capped leaking incore' thermocouple F12 at location 27A (installed May 29, 1987). -

Summary of Safety Evaluation: The temporary modifica-tion is a potential change to the facility of its operation as described in the FSAR. The thermoc'.>uple was damaged on May 7, 1985, during CRGT split pin modifications.

The thermocouple developed a leak through the thermocouple connector. The thermocouple assembly was removed on the reactor vessel head side of the seal collar adapter. A 316 SS 1/8" connector with cap was installed to seal off the thermocouple assembly.

From an instrumentation standpoint, this does not create a problem because:

a. Thermocouple F12 is not used to provide an input into the reactor protection system, the reactor vessel level indication system, or for the subcooling margin.

b. This will be the only one of the 39 incore thermocouples out of service and will not affect the ability to meet the Technical-Specification requirements of Table 15.3.5-5 to have four thermocouples per quadrant.

From a pressure-retaining standpoint, all materials used were compatible with the existing components. The proposed Swaged connection is similar to those used for the thermocouple connections. All material in contact with primary coolant are stainless steel and will not be affected by contact with boric acid. If this seal were to leak, it would not exceed the capacity of one charging pump (see FSAR Section 14.3.1).

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

68

< _ . 4 g - .

\:

.. 3.5.7 039 (Common), NVAC. The_ temporary modification

. secured ventilation to the new battery rooms and disabled'the lowLair~flowfinput to'the white / yellow bus ventilation' alarm on 2C20.- Alarm indication.for'

• high temperature in'any' battery = room or electrical equipment room was still provided (installed

'May 28, 1987).

Summary of Safety ' Evaluation: =The temporary modification is a potential change to the facility orf

~

'its operation as described in 'the FSAR. The temporary modification was performed'to allow battery room temperature to increase to a value suitable for performing an equalizing battery charge. The previous arrangement resulted in the battery room temperatures being too cold.

The original system was designed to remove heat generated by the inverters'and battery chargers and to prevent accumulation of hydrogen in the battery rooms.-

This - temporary modification will not reduce air flow to the electrical equipment-rooms where the inverters and battery chargers are located and will not result in

~

the electrical' equipment rooms becoming overheated.

Even with the' supply dampers to the battery. rooms shut,

<there should not be a hydrogen accumulation greater than

'1% in the battery rooms. An air flow rate of,12 ft /3 min was calculated to be required to prevent the hydrogen concentration from exceeding 1%. This value is less than the expected leakage across the supply dampers, approximately 70.ft 3/ min.

This temporary modification will not affect the

. operability or reliability of'the white or yellow batteries orJtheir associated inverters and battery

~ chargers.- The change does not pose an unreviewed safety

, question.- The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.5.8 87-062, Condensate /Feedwater Systems. This temporary modification gagged ICV-476 open to the as-found, throttled position on a manual gag (installed September 11, 1987; removed September 11, 1987).

69

1

, , Summary of Safety Evalua' tion: This temporary

. modification was a potential, change to the-facility or -

.its' operation as described-in the FSAR; This action removed the' automatic closure function _of the _ valve.

Automatic closure is;specified during a safety injection' '

1 ,and following a reactor trip (after cooldown) with Tavg ,

greater than 550*F and.following steam generator overfill. The only portion of the valve closure; function'which falls into a' protection area fron'the safety analysis standpoint isithe safety injection trip.

This function is redundant in that an SI signal will' also trip both main feed pumps (which;'in turn,Lisolate their associated discharge valve).

The valve operator. function was provided by an independent individual who was assigned as "designated:

operator" to manually remove the gag.(close the valve)

~

upon notification that'a reactor trip, safety injection or steam generator, overfill has occurred.

~

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an

-accident or malfunction of cquipment important to safety is not increased. The change does 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 is not. reduced.

__ 3.5.9 '87-076 (Unit 2), Safety Injection System. The temporary.

modification installed a stem clamp on valve 2SI-8718 to clamp the valve closed when the motor operator was removed for maintenance. Valve 2SI-8718 is the interface valve between the containment spray system and the RHR system. This crossconnect is used during the recirculation phase when water is supplied to the suction of the containment spray pumps via the RHR pumps (installed October 14, 1987; removed October 18,1987).

Susunary of Safety Evaluation: This temporary modification constitutes a change to the facility as described in the FSAR. The clamp was designed not to damage the valve shaft and to provide adequate force to keep the valve closed at s'ystem design pressures.

The valve was only clamped shut during the refueling outage when the operability of the containment spray system was not required. The fact that the valve was clamped closed did not affect the operability or.

function of the inservice RHR system. Leakage was checked after installation of the clamp. The seismic qualification of the systems was not aversely affected because the removal of the motor operator will reduce stresses in the system.

70

,O The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.5.10 87-077 (Unit 2), Service Water System. The temporary modification installed a stem clamp on valve 2MOV-2907, containment vent cooler service water return, to prevent it from opening while its motor operator was removed for maintenance. Valve 2MOV-2907 and 2908 provide redundant, full-flow bypasses around the Unit 2 containment fan cooler service water return orifice.

The valves are normally closed during power operation; opening automatically upon receipt of an SI signal (installed November 3, 1987; removed November 5, 1987).

Summary of Safety Evaluation: The temporary modification is a change to the facility as described in the FSAR. TS 15.3.3.B.l.c requires four accident fan cooler units to be operable in order to make a reactor critical. This TM did not affect the Technical Specifications for fan cooler operability since the coolers still were capable of performing their function as analyzed in the FSAR via 2MOV-2908. However, it could be construed as an unreviewed safety question if the reactor was started with the clamp on 2MOV-2907 installed since the probability of a malfunction of equipment important to safety (the fan coolers) would be increased. This issue was precluded by performing the TH while the reactor was in the refueling shutdown or cold shutdown condition.

FSAR Section 6.3 (Page 6.3-9) describes the containment fan cooler service water return header as having, "two parallel bypass lines, each with an independent, full-flow isolation valve which opens automatically in the event of an ESF actuation signal to bypass the orifice." This TM changed the facility in this description to a single, automatically opening bypass.

This change does not pose an unreviewed safety question as long as it is limited to the refueling shutdown and cold shutdown conditions as described above.

Technical Specification 15.4.5.I.C.1 requires that the service water bypass valves be tested at each refueling.

2MOV-2907 had this test performed to verify operability after the TM was completed.

The valve is Seismic Class I per FSAR Appendix "A." The weight of the clamp is negligible compared to the motor operator so the seismic qualification of the valve will not be adversely affected.

71

The clamp is designed to hold the valve shut against a pressure greater than the maximum system pressure (service water pump shutoff head) of 125 psig. If.the clamp failed and the valve opened unexpectedly, service water flow through the fan coolers would increase but no serious consequences would result to the service water system.

In summary, this temporary modification was performed during refueling shutdown or cold shutdown. The valve was tested for operability after maintenance. The change did not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is 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 is not reduced.

3.5.11 87-078 (Unit 2), Safety Injection System. The temporary modification installed a clamp on the stem of valve 2SI-852B to keep the valve closed while the valve operator was removed for maintenance (installed November 3, 1987; removed November 3, 1987).

Summary of Safety Evaluation: The temporary modification is a change to the facility as described in the FSAA. The evaluation that was done for TM 87-076 is applicable to this TM. The one difference is that in this case there was no concern about leakage. Any leakage that might-occur would not be of any concern to the safe operation of the plant.

The change does not pose an unreviewed safety question.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does 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 is not reduced.

3.6 Core Reloads NOTE: THE 1986 CORE RELOAD SAFETY EVALUATIONS INCLUDED BELOW WERE INADVERTENTLY OMITTED FROM THE 1986 ANNUAL REPORT AND ARE INCLUDED HERE.

3.6.1 Unit 1 Cycle 14 Reload, The reload core contains 28 fresh optimized fuel assemblies (OFAs), 28 once-burned 0FAs, and 65 twice- or thrice-burned standard fuel assemblies.

Summary of Safety Evaluation: This evaluation is required because the reload core is a change to the facility as described in the FSAR.

72

The: Unit 1 Cycle 14 des gn and safety analysis is 4- acceptable and falls within earlier analyses discussed in the OFA RTSR and indicates that the operation of the

-Cycle 14 core does not involve a significant increase in the probability of accidents previously considered, does not involve a significant' decrease'in safety margin, and does not involve a significant hazard _

consideration. No additional Technical Specification changes are required beyond those already approved by  ;

the NRC for OFA transition cores.

3.6.2 Unit 2 Cycle 13 Reload, The Unit 2 Cycle 13 reload contains 32 fresh Region 15 optimized fuel assemblies (OFAs) at 3.4 w/o, 28 once-burned Region 14 0FAs, 32 .

=twice-burned Region 13 0FA, and 29_ Region 9, 11A, and 12 standard design assemblies. The five Region 9 assemblies and one Region 11 assembly from the spent fuel pool were previously discharged in WIS Cycles 9 and 11, respectively. The cycle 13 reload is the third reload containing a full region of 0FA fuel for Point Beach Nuclear Plant (PBNP) Unit 2.

Summary of Safety Evaluation: The reload core is a change to the facility as described in the FSAR. No additional Technical Specification changes are required beyond those already covering 0FA transition cores. No  :

special environmental considerations are involved and, therefore, evaluation of environmental effects by the  ;

NRC staff and issuance of environmental impact or assessment statements are not involved.

The Unit 2 Cycle 13 reload design and safety analysis is acceptable and falls within the scope of earlier l analyses discussed in the OFA RTSR, and indicates that

-operation of the Cycle 13 core does not involve a ,

l significant increase in the probability or consequences ,

of accidents previously considered, does not involve a  !

significant decrease in safety margin, and does not  :

l involve a significant hazard consideration. ('

3.6.3 Unit 2 Cycle 13 Revised Reload, This was a revision to l the original safety evaluation report for the Unit 2 Cycle 13 reload core. The revision resulted from the f

, discrepancy between two Region 11 fuel assemblies in I the original reload design, and the two Region 11 i assemblies which were actually loaded. l The as-loaded core contained two Region 11 fuel  ;

assemblies not in the original Westinghouse loading l pattern. Baffle locations A6 and H1 contained ,

assemblies L60 and LSI rather than L76 and L61.  ;

l 73 l

< t

w Revised Safety Evaluation Summary: The Unit 2 Cycle 13 reload design and safety analysis is acceptable and falls within the scope of earlier analyses discussed in the OFA RTSR and indicates that operation of the cycle 13 core does not involve a significant increase in the probability or consequences of accidents previously considered, does not involve a significant decrease in safety margin, and does not involve a significant hazard consideration.

3.6.4 Unit 1 Cycle 15 Reload, The Unit 1 Cycle 15 reload contained 16 fresh Region 17A optimized fuel assemblies (OFAs) at 3.2 w/o, 16 Region 178 0FAs at 3.6 w/o, 28 ,,

once-burned Region 16 0FAs, 28 twice-burned Region 15 0FAs, and 33 Region 11, 12B, and 14 standard design assemblies. The eight Region 11 assemblies and one Region 12B assembly from the spent fuel pool were previously discharged in WIS Cycle 11 and WEP Cycle 12, respectively. The Cycle 15 reload was the third reload containing a full region of 0FA fuel for Point Beach Nuclear Plant (PBNP) Unit 1.

Summary of Safety Evaluation: The Unit 1 Cycle 15 reload core was a change to the facility as described in the FSAR. The reload did not_ affect FSAR Chapters 1, 2, 4-13, and appendices; the FSAR Safety Analysis (Chapter 14); the Technical Specifications; or NRC commitments as found in SERS. Chapter 3 of the FSAR was affected only in that the current operating cycle of the plant was stated and must be changed every year.

The only Technical Specification change required beyond those already covering 0FA transition cores was the fuel rod substitution change awaiting NRC approval. No special environmental considerations were involved, and therefore, evaluation of environmental effects by the NRC staff and issuance of environmental impact or assessment statements were not involved.

The Unit 1 Cycle 15 reload design and safety analysis was acceptable and fell within the scope of earlier analyses discussed in the OFA RTSR, and indicated that operation of the Cycle 15 core did not involve a significant increase in the probability or consequences of accidents previously considered, did not involve a significant decrease in safety margin, and did not involve a significaat hazard consideration.

74

.3.6.5 Unit 2 Cycle 14 Reload,' The Unit 2 Cycle 14 reload contained 16 fresh Region 16A optimized fuel assemblies

'(OFAs) at 3.2 w/o, 16 fresh Region-16B 0FAs at 3.6 w/o, 32 once-burned Region 15 0FAs, 28 twice-burned' Region 14 0FAs, 20 thrice-burned Region'13 0FAs, and-9 Region 11A and 12 standard design assemblies. The four Region 12 assemblies from~the spent fuel pool were previously discharged in Unit 2 Cycle 12. The Unit 2

Cycle 14 reload was the fourth reload containing a full ~

region of 0FA fuel for Unit 2.

Summary of Safety Evaluation: The Unit 2 Cycle 14 reload core was a change to the facility as described in the FSAR. The reload did not affect FSAR Chapters 1, 2, 4-14 and appendices; the FSAR Safety Analysis (Chapter 14); the Technical Specifications; or NRC commitments as found in SERs. Chapter 3 of the FSAR was affected only in that the current operating cycle of the plant was stated and must be changed every year.

No additional Technical Specification changes were required beyond those already covering 0FA transition cores. No special environmental considerations were involved, and therefore, evaluation of environmental effects by the NRC staff and issuance of environmental impact or assessment statements were not involved.

The Unit 2 Cycle 14 reload design and safety analysis was acceptable and fell within the scope of earlier analyses discussed in the OFA RTSR, and indicated that operation of the Cycle 14 core would not involve a significant increase in the probability or consequences of accidents previously considered, would not involve a significant decrease in safety margin, and would not involve a significant hazard consideration. Therefore, provided the startup physics testing would not result in any discrepancies with the analysis assumptions, the operation of Cycle 14 was acceptable based on its reload design and safety analysis.

75

'l p.

4.0 NUMBER OF PERSONNEL AND MAN-REM BY WORK GROUP AND JOB FUNCTION POINT BEACH NUCLEAR PLANT 1987 JuefUAL JOB FUNCTION SUI 9tARY Greater Total Reactor Than rem For Operations & Routine Special Waste Group 100 mrem Work Group Surveillan:e_ Maintenance Inspections Maintenance Processing Refueling PJint Beach Employees Operctions 39.650 28.900 7.240 3.510 63 ----- ----- -----

Maintenance and Pu k Maintenance 86 101.160 -----

.67.660 5.990 9.270 -----

18.240 Chemistry and Health Physics 35 24.240 23.060 ----- ----- -----

0.920 0.260 Instrumentation tnd Control 15 7.850 -----

2.600 0.440 2.010: -----

2.800 R=ctor Engineering 2 6.780 ----- -----

0.180 -- - -----

0.600~

Achainistration and Engineering, Quality

& Regulatory Services 9 3.860 1.110 -----

2.720 ----- ---- -

0.030 Contract Workers and Others 329 354.470 0.070 -----

14.710 329.520- 10.170 -----

GRAND TOTALS 545 532.010 53.140 70.260 31.280 340.800 '11.090- 25.440-76

n a

5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION The results of the findings from steam generator tube inspections are as follows:

5.1 Unit'l-Inspection Plan During the Unit 1 Refueling 13 outage, eddy current inspection was not performed due to the two consecutive indication-free inspections performed in 1985 and 1986. This is allowed per Technical Specification 15.4. A 4.C and may be extended through 1988 if desired.

5.2 Unit 2 5.2.1 Inspection Plan During the Unit 2 Refueling 13 outage, eddy current testing was performed from October 8 to October 22, 1987. The extent of the inspection programs in each steam generator is as follows:

Eddy Current Inspection Plan Extent of Inspection Number of TuSes Inspected "A" SG "B" SG Hot Leg (Cold Leg) Hot Leg (Cold Leg)

Full Length 120 3 (118)

Sleeves 60 51 To Top of Sleeve (63) .72)

U-Bend 18 19

1. 1 TSP 1439 (27) 1568 (2279)

1. 2 TSP 25 7 (23)

1. 3 TSP 3

1. 5&6 TSP 2 TOTALS 1662 (90) 1653 (2692) l j 77 l

t

7

1 -

• w 5.2.2 Inspection.Results The following is a summary of the results of eddy-

current inspection showing the number of tubes with indications in the ranges. listed:

~

' Eddy Current Inspection Results: :c Hot Leg (Cold Leg) ,

"A" SG "B" SG

<20%- 15(14) 2(221) 20-29%. 4(28)- 3(291)-

30-39% 5 -(4) 2(252) 40-49%. -1 (9) 50-59% 0 (2) 60-6% 1 .0-70-79%- 1- 2 80-89%- 7 1 90-100%- 0 1

• UDI -- 19 9

• • DI- 5 (2) 5 (21)

TOTALS 58(48)- 25(796)

• UDI indications are those whose quantitative analysis-has not been possible but in previous instances have necessitated repair.-

• • DI - indications whose quantitative analysis'has not been possible but in previous instances have not necessitated repair.

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 eddy current inspection performed in 1987.

Plugged Tubes in the "A" Steam Generator Row - Column Indication % Location Origin 8-3 80% 8.7" ATE IIL OD 23-8 41% #1 TSP HL OD 33-18 UDI 6 G" ATE HL OD 41-33 UDI 4.8" ATE HL OD 7-36* 22% 0.8" ATS CL OD 42-37 UDI 5.4" ATE HL OD 37-47 UDI 7.1" ATE HL OD 35-60 UDI 4.4" ATE HL OD 37-60 UDI 4.4" ATE HL OD 33-62 UDI 4.2" ATE HL OD 35-62 UDI 3.5" ATE HL OD 35-63 UDI 8.0" ATE HL OD 40-63 86% 7.8" ATE HL OD 78

7 .

f'.- '

i

.- j Row - Column Indication % Location Origin 41-63 67% 6.3" ATE HL' OD I 33-64 75%- 4.1" ATE HL OD L

.37-64 82% 3.8" ATE HL OD 40-64. UDI 5.1" ATE HL OD 41-64 UDI- 7.2" ATE HL. OD 31-66 88% 8.7" ATE HL-

. OD 34-77 UDI 3.6" ATE HL- OD -

36-68' UDI 9.6" ATE HL OD 35-73 '85% 3.2" A1E HL OD 80 -89% 10.1" ATE HL OD 23 UDI 4.8" ATE HL OD-1-83 .

UDI 4.0" ATE HL OD  ;

16-83** --

1-84 UDI 3.4" ATE HL OD 21-85 88% 9.9" ATE HL OD 1-86 UDI 4.4" ATE HL OD ,

13-87 UDI 4.3" ATE HL OD 1-89 UDI 3.7" ATE HL OD ATE -'Above Tube End * - Plugged Due to Tube Pull [

HL - Hot Leg ** - Plugged Due to Restriction  !

UDI - Undefined Indication Plugged Tubes in the "B" Steam Generator  !

l Row - Column Indication % Location Origin 10-2 43% #1 TSP CL OD 6-4 UDI 5.2" ATE HL OD  :

6-5 UDI 5.7" ATE HL OD 1-6 UDI 8.2" ATE HL OD 6-6 UDI 3.7" ATE HL OD  !

6-8 UDI 3.8" ATE HL OD 23-10 UDI 2.7" ATE HL OD 17-12 70% 12.5" ATE HL OD 1-25 91% 3.8" ATE HL OD 9-32** -- --- --

1-40 UDI 2.7" ATE HL OD 29-46* -- --- --

25-55** -- --- --

23-75 80% 7.0" ATE HL OD 23-77 73% 5.5" ATE HL OD ,

23-79 UDI 5.9" ATE HL OD  :

23-81 UDI 5.0" ATE HL OD l

1 i

ATE - Above Tubesheet CL - Cold Leg HL - Hot Leg ATE - Above Tube End UDI - Undefined Indication I

• - Plugged Due to Tube Pull

• • - Plugged Due to Sleeving Rolldown [

l i

t I

79 [

t

g v

3

, 5.2.4~ Tubes with Indications Not Plugged-TheffolloUing.is a_ list of tubes'which had indications but were not plugged as a. result of eddy current

' inspections performed in 1987.

"A" Steam Generator l

Row - Column Indication Location ~ 0rigin 7-l' DI #1 TSP CL OD  :

13 DI #1 TSP HL OD .j 18-5 31% #1 TSP HL OD 18-6 39% #1. TSP HL OD

.25-8 34% #1: TSP HL OD 21-9 DI- #1 TSP HL 0D 11-25 21% #1. TSP CL OD 21% 'O.9" ATS CL OD 11-26 22% 0.4" ATS CL OD 40-26 26% 5.4" ATS HL OD 11 23% 0.6" ATS CL OD 13-27 33%~ 0.6" ATS CL OD-14-27 28% 0.7" ATS CL OD 16-28 29% 0.6" ATS CL OD-16-29 32% 0.7" ATS CL .0D 14-30 24% 0.6" ATS CL OD 17-30 21% 0.7" ATS CL OD.

17-31 27% 0.8" ATS CL 10D 1: '18-31 29% 0.7" ATS CL OD 13-32 29% 0.6" ATS CL OD 32 35% 0.6" ATS CL .0D 19-32 22% 0.8" ATS CL OD 20-32 25% 0.8" ATS CL OD 19-33 20% 0.9" ATS CL. OD 20-33 23% 0.7" ATS CL oD 19-34 22% 0.8" ATS CL OD 20-34 20% 0.9" ATS CL- "0D 11-35 24% 0.6" ATS CL OD

, 35 22% 0.9" ATS CL OD

( 23% 0.5" ATS CL OD 19-35 24% 0.7" ATS CL OD 18-36 22% 0.6" ATS CL OD l, 19-36 22% 0.5" ATS CL OD 20-36 23% 1.0" ATS CL OD 43-36 27% 0.3" ATS HL OD 17-37 20% 0.5" ATS CL OD 37 22% 0.5" ATS CL OD 10-46 27% 0.8" ATS CL OD 12-46 30% 0.8" ATS CL OD 14-46 DI 0.5" ATS CL OD 16-46 24% 1.2" ATS CL OD 10-47 21% 0.8" ATS CL OD 33-61 DI 5.1" ATE HL OD 80

4 .

R'sw - Column Indication % Location Origin 4-35-64 DI 6'.4" ATE HL .0D 77 30*d 0.3" ATS HL OD g 4 DI 0.2" ATS HL .0D-8-77 27% 0" ATS HL OD

'7-78 35% 0.6" ATS HL OD

, 21 24% 18.0" ATS HL~ OD "B" Steam Generator Row - Column Indication % Location ~ Origin 3-1 37%- #1 TSP CL OD 6-1 36% #1 TSP CL OD 23% #1 TSP HL OD 1-2 <20% #1 TSP CL OD 12-2 39% #1 TSP CL OD 10-3 DI #1 TSP CL OD 20-6 20% #1 TSP CL OD 29-16' 36% #1 TSP CL OD 5-18 36% 0.4" ATS HL OD 17-18 39% #1 TSP CL 00 28-23 DI 0.4" ATS HL- OD 11-24 23% 0.7" ATS CL OD

'28-24 .DI 0.3" ATS HL OD 8-25 33% 0.3" ATS CL OD 9-25 31% 0.4" ATS CL OD

- *10-25 23% 0.7" ATS CL OD 49% 0.4" ATS CL OD 11-25 32% 0.8" ATS CL OD DI 0.4" ATS CL OD 12-25 33% 0.8" ATS CL OD DI 0.1" ATS CL OD

• 13-25 37% 0.6" ATS CL OD 9-26 29% 0.3" ATS CL OD 11-26 30% 0.6" ATS CL OD

• 12-26 29% 0.3" ATS CL OD 38% 0.8" ATS CL OD 13-26 33% 0.5" ATS CL OD 14-26 29% 0.7" ATS CL OD 9-27 23% 1.8" ATS CL OD 11-27 29% 0.6" ATS CL OD 12-27 23% 0.2" ATS CL OD 35% 0.6" ATS CL OD 30% 1.1" ATS CL OD 14-27 30% 0.7" ATS CL OD 15-27 29% 0.5" ATS CL OD 6-28 29% 0.3" ATS CL OD 23% 0.8" ATS CL OD

• 9-28 37% 0.4" ATS CL OD 34% 1.1" ATS CL OD 28% 1.3" ATS CL OD 10-28 24% 1.3" ATS CL OD 11-28 23% 1.5" ATS CL OD 81

Row - Column -Indication % Location' Origin 12-28_ 34% 0.7" ATS CL OD

• 3-28 32% 0.6" ATS CL OD 36%, '1.1" ATS CL OD;

• 14-28 28% 0.3" ATS CL OD.

38% '0.8" ATS CL OD 15-28 32% 0.6" ATS CL .0D

• 16-28 -39% 0.6"'ATS CL: OD 17-28 ,32%. 0.4"'ATS CL OD 5-29 DI 0.4" ATS CL OD

• 6-29 39% 0.9" ATS CL OD 37% 0.3" ATS CL OD 7-29 34% 0.4" ATS CL OD 8-29 23% 1.1" ATS CL OD 29% 0.7" ATS CL OD 9-29 25% 1.1" ATS CL OD

, 24% 0.7" ATS CL -OD 13-29 28% 0.9" ATS CL OD 14-29 29% 0.6" ATS CL OD 15-29 32% 0.5" ATS CL OD 16-29 26% 1.2" ATS'CL OD 35% 0.6" ATS CL OD 23-29 25% #1 TSP CL OD 4-30 13% 0.3" ATS CL OD 5-30 ,

2% 1.0" ATS CL OD 0I 0.1" ATS CL OD 6-30 27% 1.0" ATS CL OD 20% 0.7" ATS CL OD 0.3" ATS CL 29% OD 9-30 23% 2.1" ATS CL OD

, 32% 0.9" ATS CL OD 12-30 24% 0.9" ATS CL OD 13-30 35% 0.4" ATS CL OD 15-30 23% 0.9" ATS CL OD 17-30 23% 0.8" ATS CL OD 18-30 31% 0.7" ATS CL OD 26% 0.3" ATS CL OD 19-30 33% 0.5" ATS CL OD 5-31 23% 1.1" ATS CL OD DI 0.3" ATS CL OD l *6-31 38% C.4" ATS CL OD I 7-31 26% 1.8" ATS CL OD 25% 0.8" ATS CL OD 25% 0.4" ATS CL OD

• 8-31 23% 1.9" ATS CL OD 36% 1.1" ATS CL OD 9-31 25% 2.0" ATS CL OD 35% 1.2" ATS CL OD 10-31 23% 0.9" ATS CL OD 13-31 23% 1.3" ATS CL OD 15-31 22% 1.3" ATS CL OD 17-31 20% 0.6" ATS CL OD 18-31 25% 1.1" ATS CL OD 28% 0.6" ATS CL OD 82

> 1 k

, Row - Coldmn Indication % Location Origin 36 24%1 #1 TSP CL OD

• 5+32' 36% 0.9" ATS CL OD

'DI. 0.4" ATS CL OD 6-32. 29%- 1.0" ATS CL OD 33% 0.5" ATS'CL OD 7-32 25% 2.0" ATS CL OD 34% 0.8" ATS CL: 'OD-

-31% 1.2" ATS CL OD 8-32. 25% 1.1" ATS CL OD~ l 9-32 39% 1.0" ATS CL OD

.: 10-32 23% 1.1" ATS CL OD

.*11-32 45% 1.7" ATS CL OD 29% 1.1" ATS CL OD 12-32 24% 1.5" ATS CL OD 13-32 23% 2.4" ATS CL. OD 29% 1.1" ATS CL OD 17-32' 25% 1.0" ATS'CL OD 19-32 '30% 0.6" ATS CL_ OD 21-32 DI 0.2" ATS CL o0D 22-32 DI #1 TSP CL OD

4-33 32% 0.6" ATS CL OD DI 0.4" ATS CL OD 5-33 33% 0.8" ATS CL OD

'32% 0.4" ATS CL OD 34% 1.3" ATS CL OD 6-33 27% 1.2" ATS CL OD 23% 0.5" ATS CL OD 1.0".ATS.CL 24% OD 7-33 29% 1.3" ATS CL' OD l' 21% 0.6":ATS CL .OD 29% 2.1" ATS CL OD 8-33 23% 1.1"-ATS CL OD 9-33 25% .1.9" ATS CL OD 29% 1.3" ATS CL OD

~* -

33% 1.1" ATS CL OD 10-33 21% 2.0" ATS CL OD 32% 1.1" ATS CL OD 25% 0.8" ATS CL OD 12-33 33% 1.8" ATS CL OD 23% 1.0" ATS CL OD 13-33 26% 0.8" ATS CL OD 27% 1.4" ATS CL OD 16-30 27% 0.9" ATS CL OD 18-33 22% 0.8" ATS CL OD DI 0.4" ATS CL OD 19-33 23% 0.8" hT3 CL OD DI 0.3" ATS CL OD 20-33 DI 0.3" ATS CL OD 21-33 DI 0.3" ATS CL OD 27% 0.9" ATS CL OD 5-34 27% 1.1" ATS CL OD 6-34 28% 1.1" ATS CL OD 29% 0.6" ATS CL OD 83 l . . . . . . . . - , . . . .

Row - Column Indication % Location Origin 34 24*A~ 1.1" ATS CL OD s

9-34 21% 1.4" ATS CL OD

, .10-34 27%. 1.2" ATS CL OD 11-34 <

-21% 1.0" ATS CL OD

13-34 23% 0.9" ATS CL- OD 18-34 23% 0.8" ATS CL OD 19-34 21% 0.8" ATS CL OD 35 24% 1.1" ATS CL OD

35. 27% 1.0" ATS CL OD 7-35 23% E2.0" ATS CL' OD

, 32% 0.9" ATS CL OD

!. 24% 0.4" ATS CL OD l 8-35 29% 0.8" ATS CL 0D 9-35 28% 0.8" ATS CL OD 10-35 20% 1.0" ATS CL- OD 22% 1.9" ATS CL OD l

12-35 24% 1.0" ATS CL OD 16-35 27% 0.9" ATS CL OD.

l t

18-35 32% 0.8" ATS CL OD 20-35 DI 0.3" ATS CL OD 21-35 DI 0.3" ATS CL OD 5-36 23% 0.6" ATS CL OD 6-36 29% 0.3" ATS CL OD 8-36 29% 0.9" ATS CL OD

• 9-36 20% 1.8" ATS CL OD 37% .1.2" ATS CL ~OD 10-36 28% 1.8" ATS CL CD 23% 1.0" ATS CL OD 11-36 34% 1.4" ATS CL OD 31% 0.9" ATS CL OD 12-36 23% 0.8" ATS CL OD 13-36 20% 0.9" ATS CL OD 14-36 20% 1.2" ATS CL OD 16 22% 0.7" ATS CL OD 20-36 20% 0.5" ATS CL OD 21-36 DI 0.5" ATS CL OD 5-37 26% 0.6" ATS CL OD 39% 0.7" ATS CL 00 6-37 26% 0.4" ATS CL OD 7-37 32% 0.4" ATS CL OD 39% 0.6" ATS CL OD 8-37 28% 1.1" ATS CL OD 29% 1.2" ATS CL OD 9-37 27% 0.8" ATS CL OD 10-37 24% 0.6" ATS CL OD 22% 1.1" ATS CL OD 11-37 35% 0.7" ATS CL OD 12-37 25% 0.8" ATS CL OD 13-37 29% 0.9" ATS CL OD 27% 0.7" ATS CL OD 16-37 21% 0.6" ATS CL OD 24% 0.4" ATS CL OD 84

. 6 Row - Column Indication % Location Origin 21-37 27% 0.7" ATS CL OD 31% 0.7" ATS CL OD 22-37 DI 0.4" ATS CL OD

• 23-37 38% 0.6" ATS CL OD .

5-38 34% 0.3" ATS CL OD

• 6-38 38% 0.5" ATS CL OD 7-38 22% 1.8" ATS CL OD '

30% 0.5" ATS CL OD 8-38 30% 2.2" ATS CL OD l 9-38 25% 0.7" ATS CL OD 10-38 32% 0.6" ATS CL OD 11-38 21% 1.0" ATS CL OD 12-38 27% 1.3" ATS CL OD 30% 0.6" ATS CL OD 0

13-38 28% 0.9" ATS CL OD 16-38 27% 0.6" ATS CL , 0D 17-38 25% 0.6" ATS CL OD 18-38 24% 1.3" ATS CL OD j 30% 1.0" ATS CL OD 20-38 26% 0.6" ATS CL OD 26% 1.2" ATS CL OD 21-38 26% 0.6" ATS CL OD 22-38 27% 0.6" ATS CL OD 23-38 DI 0.5" ATS CL OD 4-39 30% 0.8" ATS CL OD

• 5-39 39% 0.9" ATS CL OD 38% 0.5" ATS CL OD 7-39 30% 2.0" ATS CL OD 31% 0.6" ATS CL OD 23% 0.3" ATS CL OD -

8-39 20% 0.4" Als CL OD 9-39 23% 2.0" ATS CL OD 24% 0.8" ATS CL OD 10-39 22% 0.6" ATS CL OD 11-39 27% 1.4" ATS CL OD 28% 0.6" ATS CL OD 12-39 25% 1.3" ATS CL OD 34% 0.8" ATS CL OD 13-39 29% 0.7" ATS CL OD 16-39 31% 0.7" ATS CL OD 22% 0.5" ATS CL OD 17-39 23% 1.1" ATS CL OD 32% 0.6" ATS CL OD 20-39 23% 0.5" ATS CL OD 21-39 29% 0.5" ATS CL OD 23-39 24% 0.3" ATS CL OD

• 5-40 39% 0.6" ATS CL OD 6-40 24% 0.9" ATS CL OD

• 7-40 35% 2.1" ATS CL OD .

30% 2.2" ATS CL OD -

DI 0.3" ATS CL CD 39% 0.5 ATS CL OD 85

1 9_ , ,-

T s;g * .

, s.

' Row - Column' Indication;% . Location Origin 2.2" ATS CL 8-40 ~ 23%_ OD 26% 0.6" ATS CL OD' 9-40 29% '0.7" ATS CL OD 10-40 28% 0.7" ATS CL _0D-11-40 25% 0.6" ATS CL OD-12-40 33% 1.4" ATS CL' OD 27% _0.7" ATS CL 0D- 3

.13-40 34% 1.0" ATS CL OD 14-40 29% 1.0" ATS CL- OD 22%. 0.7" ATS CL OD 15-40 26% 0.7"'ATS CL OD DI 0.4" ATS CL OD 16-40 33% 0.7" ATS CL OD-17-40 23% 0.6" ATS CL OD 19-40 25% 0.5" ATS CL OD

• 21-40 34% 0.6" ATS CL OD DI- 0.2" ATS CL OD 22-40 -

27% 0.5" ATS CL OD 23-40 22% 0.8" ATS CL OD 5-41 30% 0.9" ATS CL OD DI 0.3" ATS CL OD

• 6-41 39% 0.6" ATS CL OD 7-41 29% 0.5" ATS CL OD 8-41 25% 0.4"_ATS CL OD

• 9-41 39% 0.7" ATS CL OD

• 10-41 25% 1.4" ATS CL OD 37% 0.7" ATS CL OD

• 11-41 38% 0.9" ATS CL OD 26% 0.6" ATS CL OD 13-41 34% 0.9" ATS CL OD

'14-41 33% 1.2" ATS CL OD 34% a0.5" ATS CL OD 15-41 32% .0.5" ATS CL OD 16-41 29% 0'.5" ATS CL OD:

I 18-41 31% 0.8" ATS CL OD

[ *19-41 23% :1.1" ATS CL OD

{ 32% 0.6" ATS CL OD I 20-41 30% 0.5" ATS CL- OD l 22-41 23% 0.7" ATS CL OD 5-42 23% 0.9" ATS CL OD f

l 6-42 33% 0.6" ATS'CL OD l DI 0.3" ATS CL OD 7-42 34% 0.6" ATS CL OD DI 0.3" ATS CL OD 8-42 34% 0.7" ATS CL OD 10-42 23% 1.9" ATS CL OD 34% 0.8" ATS CL OD 11-42 31% 1.4" ATS CL OD 12-42 27% 1.0" ATS CL OD 30% 0.5" ATS CL OD 13-42 30% 1.6" ATS CL OD 30% 0.9" ATS CL OD 38% 1.5" ATS CL OD 86

( _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. o' ~

Row -' Column Indication'% Location . Origin 14-42 21% 1.3" ATS CL OD 27% 0.5" ATS CL OD

=15-42 28% 0.7" ATS CL OD 23%- 0.8" ATS CL OD 18-42 32% 0.6" ATS CL OD

• 19-42 .32% 0.5" ATS CL -OD- .

38% 0.8" ATS CL OD 21-42 26% 1.0" ATS CL OD.

35% 0.5":ATS CL OD 6-43 24% 0.6" ATS CL OD 8-43 35% 0.6" ATS CL OD

• 9-43 26% 2.0" ATS CL OD .l 36% 0.5" ATS CL OD l

• 10-43 '29% 0.8" ATS CL OD l DI 0.4" ATS CL OD 37% 1.7" ATS CL OD 12-43 35% 1.1" ATS CL OD i 29% 0.6" ATS CL OD l 13-43 24% 1.6" ATS CL OD 31% 0.8" ATS CL OD 16-43 26% 0.6" ATS CL OD 17-43 DI 0.4" ATS CL OD 18-43 33% 0.7" ATS CL OD

• 19-43 33% 0.7" ATS CL OD 21-43 22% 1.1" ATS CL OD 29% 0.6" ATS CL OD 22-43 22% 1.0" ATS CL OD 23-43 25% 0.9" ATS CL OD 4-44 DI 0.4" ATS CL OD 8-44 21% 0.8".ATS CL OD DI 0.5" ATS CL OD

• 9-44 53% 0.6" ATS CL OD 10-44 22% '1.0" ATS CL OD 11-4.4 33% 1.2" ATS CL OD 34% 0.6" ATS C' OD 12-44 29% 1.0" ATS CL OD 13-44 33% 0.8" ATS CL OD 14-44 29% 0.8" ATS CL OD 15-44 28% 0.9" ATS CL OD 29% 0.3" ATS CL OD 16-44 31% 0.8" ATS CL OD 22% 0.3" ATS CL OD 17-44 28% 0.9" ATS CL OD 18-44 31% 0.7" ATS CL OD 20-44 21% 0.9" ATS CL OD

• 21-44 34% 1.0" ATS CL OD 22-44 29% 1.2" ATS CL OD 23-44 26% 0.9" ATS CL OD 1-45 DI 1.1" ATS HL OD

• 4-45 30% 1.1" ATS CL OD 37% 0.4" ATS CL OD

• 9-45 45% 1.7" ATS CL OD 39% 0.7" ATS CL OD 87

-e

' Row - Column- Indication % Location -Origin

~12-45 ! 30% 1.9" ATS CL' '0D

^ DI 0.6" ATS CL- OD 13-45 31% 2.0" ATS-CL OD 29%= 1.5" ATS CL OD 27% 0.8".ATS CL OD 45 21% 2.1" ATS CL OD DI 0.4" ATS CL OD 15-45 23% 0.7" ATS CL OD DI 0.3" ATS CL OD 16-45' 23%

~1.4" ATS CL OD DI 0.3" ATS CL OD

• 17 38% 0.9" ATS CL OD DI 0-5" ATS CL

. OD

• 18-45 37% 0.6" ATS CL OD

• 19-45 38% 0.9" ATS CL OD 20-45 DI 0.3" ATS CL OD 22-45 20% 1.1" ATS CL OD 23-45. 35% 0.9" ATS CL OD' 3-46 22% 0.8" ATS CL OD 4-46 35% 1.2" ATS CL OD 35% 0.5" ATS CL OD.

'13-46 27% 1.6" ATS CL OD 35% 0.8" ATS CL OD 15-46 22% 1.8" ATS CL 0D-25% 0.6" ATS CL OD

• 16-46 23% 2.5" ATS CL OD 39% 1.9" ATS CL OD 35% 1.0" ATS CL OD 37% 0.5" ATS CL OD

• 17-46' -28% 1.6" ATS CL OD 43% 0.9" ATS CL OD 25% 2.3" ATS CL OD

• 18-45 37% 1.5" ATS CL OD 39% 0.9" ATS CL OD-47% 0.5" ATS CL OD

• 19-46 123% 2.1" ATS CL 0D 41% 0.8" ATS CL OD 20-46 29% 1.0" ATS CL OD 29-46 29% #1 TSP CL OD 33-46 29% #1 TSP CL '0D 3-47 21% 0.5" ATS CL OD

• 4-47 29% 1.5" ATS CL OD 50% 0.9" ATS CL OD 37% 0.4" ATS CL OD 13-47 30% 1.9" ATS CL OD 31% 0.8" ATS CL OD 14-47 23% 1.9" ATS CL OD 35% 0.8" ATS CL OD 17-47 25% 2.7" ATS CL OD 27% 2.0" ATS CL OD 35% 1.2" ATS CL OD 88

.. - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I

Row - Column Indication % Location Origin 18-47 27% 2.0" ATS CL OD 33% 1.1" ATS CL OD 21% 0.6" ATS CL OD

• 19-47 38% 1.2" ATS CL OD 31% 0.6" ATS CL OD 20-47 23% 1.0" ATS CL OD 21-47 ?S% 2.4" ATS CL OD 25% 1.8" ATS CL OD 25% 1.1" ATS CL OD 22-47 27% 1.3" ATS CL OD 23-47 31% 1.9" ATS CL OD 25% 1.0" ATS CL OD

• 7-48 37% 1.7" ATS CL OD 34% 0.7" ATS CL OD

• 8-48 22% 2.0" ATS CL OD 38% 1.5" ATS CL OD 29% 0.9" ATS CL OD 10-48 35% 1.3" ATS CL OD

• 12-48 25% 1.7" ATS CL OD 39% 0.9" ATS CL OD

• 13-48 25% 2.0" ATS CL OD 39% 1.0" ATS CL OD

• 14-48 25% 1.9" ATS CL OD 39% 0.9" ATS CL OD

• 15-48 27% 1.9" ATS CL OD 38% 0.8" ATS CL OD 33% 0.5" ATS CL OD 16-48 24% 0.6" ATS CL OD 31% 0.7" ATS CL OD 17-48 29% 0.6" ATS CL OD 22% 1.9" ATS CL OD

• 19-48 39% 1.1" ATS CL OD

• 20-48 26% 2.1" ATS CL OD 33% 1.7" ATS CL OD 38% 1.2" ATS CL OD 37% 0.8" ATS CL OD 22-48 23% 0.3" ATS CL OD

• 23-48 23% 1.7" ATS CL OD 34% 0.9" ATS CL OD 38% 0.3" ATS CL OD

• 24-48 38% 0.9" ATS CL OD l DI 0.4" ATS CL OD 30-48 27% #1 TSP CL OD 33-48 28% #1 TSP CL OD 7-49 29% 0.5" ATS CL OD 8-49 25% 1.2" ATS CL OD ~

30% 0.4" ATS CL OD

• 9-49 37% 1.8" ATS CL OD 32% 1.3" ATS CL OD 10-49 24% 1.3" ATS CL OD 11-49 35% 1.2" ATS CL OD 12-49 23% 1.3" ATS CL OD 89 g

\

o Row --Column Indication % Location Origin

.13-49 29% '1.6" ATS CL 10D -

25% .0.9" ATS CL OD 14-49 28% .0.9" ATS CL OD 15-49 24% 0.6" ATS CL OD 19-49; 34% '1.0" ATS CL 0D 21-49 34%. 1.2" ATS CL OD 28% o0.6" ATS CL OD 23-49 31% 1.1" ATS CL OD.

• 24-49 33% 0.8" ATS CL OD DI 0.4" ATS CL OD

• 25-49 33% 0.3" ATS CL OD 7-50 22% 2.1" ATS CL OD 28% 0.5" ATS CL OD 8 22% 0.4" ATS CL- OD

• 9-50 25% 1.2" ATS CL OD 37% 0.3" ATS~CL OD 10-50 24% 1.2" ATS CL OD 11-50 20% 1.1" ATS CL~ OD 12-50 25% 1.1" ATS CL OD 13-50 28% 1.0" ATS CL OD 14-50 30% 0.7" ATS CL o0D 15-50 22% 0.7" ATS CL OD 17-50 24% 0.8" ATS CL OD 20-50 23% 0.6" ATS CL OD 21-50 30% 0.9" ATS CL OD 22-50 32% 1.1" ATS CL OD 33%. 0.6" ATS CL OD 23-50 29% 1.0" ATS CL~ OD 31% 0.6" ATS CL OD 24-50 26% 0.5" ATS CL OD

.4-51 35% 0.7" ATS CL OD 6-51 23% 2.0" ATS CL OD 21%. 1.6" ATS CL OD 33% 0.8" ATS CL OD 7-51 21% 2.1" ATS CL OD 22% 0.9" ATS CL OD 24% 0.3" ATS CL OD 9-51 23% 0.3" ATS CL OD 11-51 28% 0.8" ATS CL OD 13-51 29% 1.0" ATS CL OD 29% 0.5" ATS CL OD 15-51 34% 0.8" ATS CL OD 16-51 35% 0.6" ATS CL OD 17-51 22% 0.7" ATS CL OD 19-51 30% 0.9" ATS CL OD 23% 0.4" ATS CL OD 22-51 32% 0.7" ATS CL OD 24-51 30% 0.6" ATS CL OD

• 25-51 44% 0.4" ATS CL OD 5-52 DI 0.3" ATS CL OD 6-52 30% 0.9" ATS CL OD 7-52 29% 0.9" ATS CL OD DI 0.4" ATS CL OD 90

.. _ -_- _ _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ J

1 - -

9

.s Row - Column; -Indication % Location Origin.

• 8-52 ~39% 0.5" ATS CL OD

• 9-52 -39% 0.5" ATS CL 0D 10 26% 0.4" ATS CL' OD 12-52 23% 0.5" ATS CL OD 13-52 26% 0.5" ATS CL OD 15-52 34% 0.6" ATS CL 0D l 16-52 26% 0.8" ATS CL- OD l

.18-52 23% 0.9" ATS CL OD

• 19-52 _ 32% 1.4" ATS CL OD.

36% 0.9" ATS CL OD 5 23% 0.8" ATS CL OD 6-53 30% 1.0" ATS CL OD 23% 0.7" ATS CL OD 7-53 24% 0.6" ATS CL OD DI 0.3" ATS CL OD

• 8-53 36% 0.6" ATS CL OD 10-53 22% 1.2" ATS CL OD 31% 0.3" ATS CL OD 11-53 35% 1.3" ATS CL OD 13-53 '23% 1.1" ATS CL OD 14-53 24% 1.0" ATS CL OD 17-53 25% 0.8" ATS CL OD 18-53 21% 1.4" ATS CL OD 23% 0.5" ATS CL OD 19-53 35% 0.8" ATS CL OD 20-53 31% 1.0" ATS CL OD 21'-53 34% 0.7" ATS CL OD

• 23 33% 0.6" ATS CL OD 25-53 22% 0.4" ATS CL OD 3-54 23% 0.3" ATS CL OD 6-54 33% 0.9" ATS CL 10D c 7-54 22%- 2.3" ATS CL OD 30% 1.0" ATS CL 0,D '

DI 0.3" ATS CL OD l *8-54 32% 1.1" ATS CL OD

! 33% 0.6" ATS CL OD l- 36% 0.3" ATS CL OD

• 9-54 23% 1.8" ATS CL OD 35% 1.3" ATS CL OD 37% 0.5" ATS CL OD

• 10-54 37% 0.4" ATS CL OD 11-54 31% 1.4" ATS CL OD 14-54 23% 1.0" ATS CL OD 15-54 30% 0.9" ATS CL OD 16-54 32% 0.9" ATS CL OD 17-54 30%

1.0" ATS CL OD 18-54 31% 1.0" ATS CL OD 19-54 29% 1.3" ATS CL OD 21% 1.0" ATF CL OD 20-54 34% 1 X ATS CL OD 21-54 31% 1.1" ATS CL OD 24% 0.7" ATS CL OD 22-54 23% 0.7" ATS CL OD 91

e-Row - Column Indication % Location Origin

• 24-54 31%1 0.8" ATS CL :OD

.35% 0.4" ATS CL OD 7-55 22% 0.8" ATS CL OD

'DI 0.2" ATS CL OD-8-55 29% 1.2" ATS CL OD 27% 0.6" ATS CL OD 0.3" ATS CL DI. OD 9-55 30% 0.4" ATS CL OD 10-55 31% 1.5" ATS CL. OD 11 31% 1.4" ATS CL OD 14-55 30% 1.2" ATS CL OD 16-55 23% 1.0" ATS CL OD 19-55. 30%- 1.1" ATS CL OD

• 22-55 39% 1.2" ATS CL OD 38% 0.6" ATS CL OD

'*23-55 32% 0.9" ATS CL OD 0 9" ATS CL

~

25-55 35% OD 20% 0.3".ATS CL OD 7-56 34% 0.7" ATS CL OD DI 0.2" ATS CL OD 8-56 DI 0.2" ATS CL 0D 9-56 DI 0.2" ATS CL OD 10-56 22% 1.6" ATS CL OD DI 0.3" ATS.CL OD 11-56 26% 1.3" ATS CL OD 13-56 24% 1.2" ATS CL OD 14-56 22% 1.1" ATS CL OD 15-56 29% 1.3" ATS CL OD 20-56 29% 1.0" ATS CL OD 22-56 26% 0.8" ATS CL OD

• 23-56 22% 1.4" ATS CL OD 33% 0.7" ATS CL -

OD

• 24-56 23% 1.3" ATS CL 0D 22% 0.8" ATS CL OD 36% 0.6" ATS CL OD 25-56 22% 0.8" ATS CL OD 21% 0.3" ATS CL OD 3-57 33% 0.4" ATS CL OD 6-57 DI 0.3" ATS CL OD 7-57 DI 0.3" ATS CL OD 8-57 DI 0.2" ATS CL OD 9-57 DI 0.4" ATS CL OD 16-57 23% 1.1" ATS CL OD 17-57 28% 1.2" ATS CL OD 18-47 28% 1.2" ATS CL OD 19-57 21% ~ 0.9" ATS CL OD 22-57 24% 0.8" ATS CL OD 3-58 26% 0.4" ATS CL OD 4-58 32% 0.5" ATS CL OD 33% 0.3" ATS CL OD 8-58 31% 0.4" ATS CL OD 18-58 35% 1.0" ATS CL OD ,

19-58 31% 1.0" ATS CL OD I i

92

s y

Row - Column Indication % Location Origin 21-58' 23% 0.7" ATS CL OD 58 22% 0.6" ATS CL' OD 23-58 26% 1.1" ATS CL OD 25%- 0.5" ATS CL. OD 58 30% 0.6" ATS CL- OD 33-58 21%- .#1 TSP CL OD 8-59 22% 0.2" ATS CL OD l 11-59 28% 0.8" ATS CL OD 12-59 24% 1.2" ATS CL OD 15-59 28% 1.2" ATS CL OD 24% 0.9" ATS CL OD 16-59 20% 0.8" ATS CL OD  !

17-59 23% 0.8" ATS CL OD 'l

• 18-59 37% 1.0" ATS CL OD 19-59 32% 0.9" ATS CL OD 20-59 DI 0.7" ATS CL- OD 21-59 26% 0.6" ATS CL OD 22-59 23% 0.6" ATS CL OD 23-59 27% 0.7" ATS CL OD 25-59 DI 0.2" ATS CL OD

• 6-60 39% 0.7" ATS CL OD DI 0.3" ATS CL OD 7-60 29% 0.7" ATS CL OD 8-60 DI 0.3" ATS CL OD 31% 1.1" ATS CL OD 12-60 24% 1.3" ATS CL OD 15-60 25% 1.1" ATS CL OD 16-60 24% 1.0" ATS CL OD

• 18-60 36% 0.9" ATS CL OD 19-60 30% 0.8" ATS CL OD 20-60 DI 0.8" ATS CL OD 22-60 26% 0.9" ATS CL OD

• 24-60 41% 1.1" ATS CL OD 31% 0.7" ATS CL OD DI 0.3" ATS CL OD 25-60 27% 0.6" ATS CL OD DI 0.2" ATS CL OD 4-61 29% 0.3" ATS CL OD 9-61 22% 0.9" ATS CL OD 11-61 34% 1.6" ATS CL OD 15-61 25% 1.3" ATS CL OD 16-61 23% 1.0" ATS CL OD 17-61 23% 1.0" ATS CL OD 19-61 23% 0.6" ATS CL OD 21-61 20% 0.9" ATS CL OD

• 22-61 36% 0.9" ATS CL OD 23-61 26% 0.9" ATS CL OD 4-62 23% 0.4" ATS CL OD 9-62 23% 0.7" ATS CL OD 27% 1.1" ATS CL OD 27% 1.8" ATS CL OD 10-62 27% 1.8" ATS CL OD 11-62 25% 1.3" ATS CL OD 93

~ '

p ,

L K .

0- Location

. Row'- Column Indication % . Origin 16-62 30% 0.8".ATS CL' OD-O.4" ATS CL OD

>23-62 23%'

~*24-62 37% 1.0".ATS CL OD 38% 0.5" ATS CL. OD 3-63 DI. 0.3" ATS CL OD E6 -63 27% 0.8"LATS'CL OD .!

63 30% 0.2" ATS CL OD 30% 0.7" ATS CL- OD 29% 1.4" ATS CL OD' 9-63 32% 0.5" ATS CL OD 10-63 27% '1.9" ATS CL OD I 11-63 25% 1.6" ATS-CL OD 15-63 23% 0.9" ATS CL OD 16-63 27% 0.8" ATS CL OD 18-63 28% 0.7" ATS CL OD 19-63 32% 0.7" ATS CL OD 31% 1.1" ATS CL OD 21-63 20% 0.8".ATS CL OD 22-63 DI 0.5" ATS CL OD:

23-63 22% 0.7" ATS CL OD 24-63 32% 0.4" ATS CL OD 6-64 25% 0.5" ATS CL OD 8-64 DI- 0.3" ATS CL OD 24% 0.5" ATS CL OD 22% 1.8" ATS CL OD 15-64 31%. 0.6" ATS CL OD 23% 1.1" ATS CL OD 16-64 34%' O.8" ATS CL OD 17 32% 0.7" ATS CL -OD

• 18-64 37% 0.7" ATS CL OD

t. 23% 1.4" ATS CL OD 21-64 27% 0.6" ATS CL OD 29% 1.4" ATS CL OD 23-64 DI 0.4" ATS CL OD-29% 1.0" ATS CL 0D

'24-64 29% 0.9" ATS CL OD DI 0.4" ATS CL OD 4-65 23% 0.9" ATS CL OD 5-65 22% 1.0" ATS CL OD

• 7-65 25% 1.2" ATS CL OD 38% 0.4" ATS CL OD 8-65 DI 0.7" ATS CL OD 23% 1.7" ATS CL OD 10-65 27% 0.6" ATS CL OD 11-65 24% 0.4" ATS CL OD 12-65 23% 0.5" ATS CL OD 13-65 22% 1.2" ATS CL OD 23% 0.9" ATS CL OD 15-65 29% 0.7" ATS CL OD 16-65 30% -

0.8" ATS CL OD 17-65 23% 0.8" ATS CL OD 94

3 e.

Row - Column Indication % Location- Origin

~

19-65 23% 0.6".ATS CL OD 65 DI 0.7" ATS CL OD-22-65 31% 1.3" ATS CL OD

-DI 0.6" ATS CL OD 24-65 20% 0.8" ATS CL OD j 66 DI 0.2" ATS CL OD 4-66: 30% 0.9" ATS CL OD 33% 0.4" ATS CL .'

OD 5-66 24% 0.4" ATS CL OD l 6'-66 32% 1.1" ATS CL OD l 29% 0.5" ATS CL OD 7-66 30% 1.2" ATS CL OD 32% 0.4" ATS CL OD 8-66 31% 0.5" ATS CL OD 9-66 32% 0.6" ATS~CL OD 11-66 24% 1.0" ATS CL OD DI 0.3" ATS CL OD 12-66 23% 0.5" ATS CL OD

• 13-66 38% 0.8" ATS CL OD DI 0.4" ATS CL OD 14-66 24% 0.8" ATS CL- OD 15-66 27% 0.7" ATS CL OD 16-66 30% 0.7" ATS CL OD 18-66 33% 1.2" ATS CL OD 25% 0.6" ATS CL OD 19-66 32% 1.3" ATS CL OD 22% 0.6" ATS CL OD-20-66 28% 0.8" ATS CL OD 21-66 28% 1.5" ATS CL OD 29% 0.7" ATS CL OD l 22-66 28% 0.8" ATS CL OD

!. 23-66 25% 0.8" ATS CL OD L 4-67 25% 0.9" ATS CL OD l 6-67 33% 1.1" ATS CL OD 7-67 -35% 0.5" ATS CL OD 8-67 27% 0.6" ATS CL OD

• 9-67 39% 0.9" ATS CL OD 28% 0.5" ATS CL OD g 10-67 22% 0.8" ATS CL OD 34% 0.4" ATS CL OD 11-67 30% 0.5" ATS CL OD 13-67 29% 0.5" ATS CL OD

• 15-67 39% 0.7" ATS CL OD 16-67 24% 0.7" ATS CL OD

• 18-67 36% 0.6" ATS CL OD

• 19-67 38% 1.2" ATS CL OD 39% 0.6" ATS CL OD 21-67 29% 0.6" ATS CL OD 6-68 24% 1.3" ATS CL OD 28% 0.6" ATS CL OD 7-68 33% 0.6" ATS CL OD 95

r s

s Row - Column Indication % Location Origin A9-68 39% 0.8" ATS CL OD 10-68 31% 0.9" ATS CL OD 12-68 32% 0.5" ATS CL OD 13-68 -29% 0.3" ATS CL OD 14-68 35%~ 0.7" ATS CL OD 25% 0.4" ATS CL OD 15-68 28% 0.6" ATS CL OD-16-68 31% 1.1" ATS CL OD-

, 27%- 0.6" ATS CL OD

• 17-68 36% 1.0" ATS CL OD 35% 0.5" ATS CL OD-18-68 24% 1.2" ATS CL OD-35% 0.5" ATS CL OD 21-68 27% 0.4" ATS CL OD 6-69 29% 0.6" ATS CL OD S-69 27% 0.6" ATS CL OD 9-69 33% 0.8'I ATS CL 0D 10-69 24% 0.9" ATS CL OD 27% 0.4" ATS CL 'D O

11-69 33% 0.8" ATS CL OD 22% 0.4" ATS CL OD 12-69 31% 0.7" ATS CL OD

• 14-69 39% 1.0" ATS CL OD 32% 0.6" ATS CL OD 15-69 23% 0.4" ATS CL OD

• 16-69 39% 0.4" ATS CL OD 23% 1.1" ATS,CL. OD o *17-69 38% 0.4" ATS CL OD 19-69 27% 0.5" ATS CL OD 4-70 27% 1.1" ATS CL OD 6-70 23% 0.5" ATS CL OD 7-70 24% 0.5" ATS CL OD 8-70 34% 0.8" ATS CL OD 9-70 '34% 0.8" ATS CL OD 10-70 34% 0.9" ATS CL. OD 11-70 24% 0.9" ATS CL OD 12-70 31% 0.8" ATS CL OD 13-70 22% 0.9" ATS CL OD 29% 0.4" ATS CL OD 14-70 27% 0.9" ATS CL OD 35% 0.4" ATS CL OD 15-70 39% 0.4" ATS CL OD 16-70 22% 1.0" ATS CL OD 26% 0.4" ATS CL OD 17-70 23% 1.2" ATS CL OD DI 0.4" ATS CL OD 5-71 30% 0.4" ATS CL OD 10-71 28% 0.8" ATS CL OD 11-71 27% 0.7" ATS CL OD 14-71 24% 0.4" ATS CL OD 15-71 DI 0.4" ATS CL OD ,

96

3 l

Row - Column Indication % Location Origin 16-71 27% 0.6" ATS CL OD 9-72 22% 0.8" ATS CL OD 10-72 20% 0.6" ATS CL OD 11-72 23% 0.5" ATS CL. OD 12-72 DI 0.3" ATS CL OD 10-73 22% 0.5" ATS CL OD 33-73 26% #1 TSP CL OD 4-74 DI 0.4" ATS CL OD 5-74 DI 0.3" ATS HL OD 10-76 24% 0" ATS HL OD 1-77 DI 4.4" ATE HL OD 33-77 32% #1 TSP HL OD 27-83 24% 1.6" ATS HL OD

• These tubes were sleeved in the cold leg during Unit 2, Refueling 13.

5.2.5 Additional Inspections Based on the North Anna tube rupture event, Westinghouse issued a notification to some of its steam generator owners, regarding susceptibility to similar type failures, requesting that testing and modeling be performed to analyze for this type of failure. An additional eddy current examination sample was considered to be the most expedient and informational approach to answer the questions posed. As a result, all of the tubes in rows eight through thirteen were examined from the last support plate before the U-bend

(#6HL) to the first support plate past the U-bend

(#6CL). This was an attempt to positively identify the anti-vibration bar location and the local denting present at the tube support plates. A total of 500 tubes in the A steam generator and 536 tubes in the B steam generator were inspected. The A steam generator inspection revealed 471 cold leg #6 TSP dents and 521 hot leg #6 TSP dents. The B steam generator inspection revealed 484 cold leg #6 TSP dents and 486 hot leg #6 TSP dents. In both instances, no degradation was identified at either of the hot or cold leg areas of interest and no preventive maintenance was required.

5.2.6 Limited Sleeving Program During the fall outage, an attempt was made to sleeve a small portion of our Unit 2 B steam generator cold leg tubes using the Westinghoi.se SM10WS system, This project was necessary due to cold leg wastage problems which were discovered during recent eddy current examinations. A total of 200 candidate tubes were selected based on eddy current results. The criteria used for selection of these tubes were: first, the location of the tube; second, the location of the defect; and third, the severity of the degradation.

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.c s-Due to time constraints imposed on the completion of-these candidate tubes and the mechanical difficulties involved in the SM10WS sleeving process, the number of ~

installed sleeves was only 89. Of these~89 sleeves, two were plugged due to a "rolldown" condition which was deemed unacceptable by Westinghouse engineering. A list of-the successfully sleeved cold leg tubes is as follows.

Row-column Row-Column Row-Column Row-Column 10-25 11-41 13-48 23-55 13-25 19-41 14-48 23-56 1 12-26 19-42 15-48 24-56 9-28 9-43 19-48 18-59 j 13-28 10-43 20-48 6-60 l 14-28 19-43 23-48 18-60 16-28 9-44 24-48 24-60 6-29 21-44 9-49 22-61 q 6-31 4-45 24-49 24-62 l 8-31 9-45 25-49 18-64 5-32 17-45 9-50 7-65 11-32 18-45 25-51 13-66 9-36 19-45 8-52 9-67 23-37 16-46 9-52 15-67 6-38 17-46 19-52 18-67 5-39 18-46 8-53 19-67 5-40 19-46 23-53 9-68 7-40 4-47 8-54 17-68 21-40 19-47 9-54 14-69 6-41 7-48 10-54 16-69 9-41 8-48 24-54 17-69 10-41 12-48 22-55 5.2.7 Tube Pull Two tubes were pulled during the fall outage (one from each steam generator) for inspection purposes. The reasoning was to inspect the crevice region and the first tube support plate region in two "model" tubes having eddy current indications in these areas. These tube samples were sent to the Westinghouse inspection facility and destructively examined for verification of eddy current signals and information regarding any active modes of possible tube degradation. The results in each instance verified the eddy current signals and typified the indications witnessed during recent eddy current examination. Each of these tube holes was plugged with a welded plug.

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o 6.0 REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES 6.1 Overpressure Protection During Normal Pressure & Temperature Operation 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 1987.

6.2 Overpressure Protection During Low Pressure & Temperature Operation There were no challenges to the Unit 1 power-operated relief valves during low pressure and temperature operation.

During low pressure and temperature operation, the low temperature overpressure protection circuit momentarily opened a Unit 2 power-operated relief valve on November 11, 1987, at 1137 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.326285e-4 months <br /> following an inadvertent connection of the safety injection accumulators to the reactor coolant system.

7.0 REACTOR COOLANT ACTIVITY ANALYSIS There were no indications during operation of Unit 1 and Unit 2 in 1987 where reactor coolant activity exceeded that allowed by Technical Specifications.

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WISCONSIN Electnc m coww L 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE, W153201 (414)221-2345 VPNPD- 88-12 2 NRC- 8 8-18 February 29, 1988 l U. S. Nuclear Regulatory Commission Document Control Desk l Washington, D. C. 20555 Gentlemen:

DOCKETS NOS. 50-266 AND 50-301 ANNUAL RESULTS AND DATA REPORT POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Enclosed are ten copies of the Annual Results and Data Report for the Point Beach Nuclear Plant, Units 1 and 2, for the year  ;

1987. This report is submitted in accordance with Technical '

Specification 15.6.9.1.B and pursuant to the requirements of 10 CFR 50.59(b). The report contair.s information regarding operational highlights of Units 1 and 2, descriptions of facility changes, tests and experiments, personnel occupational exposures, steam generator inservice inspections, reactor coolant system relief valve challenges, and reactor coolant activities in excess of specified limits.

Very truly yours,

s . .J Y

fU ll

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C. W. Fay Vice President Nuclear Power Enclosure Copies to NRC Regional Administrator, Region III g NRC Resident Inspector tlto