ML20151A827

Summary of Procedure Changes,Facility Changes & Special Test for 1987

ML20151A827
Person / Time
Site:	North Anna
Issue date:	12/31/1987
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
88-167, NUDOCS 8804070252
Download: ML20151A827 (160)

Text

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a PROCEDURE CHANGES REQUIRING A SAFETY EVALUATION DURING 1987 2

1 AND 2-ES-1.4: TRANSFER TO HOT LEC RECIRCULATION PURPOSE The Updated Final Safety Analysis Report (UFSAR) describes aligning two charging pumps only. During a design basis accident (DBA), only one charging pump will be operable. The procedure was revised to provide instructions for aligning a charging pump for hot leg recirculation when only one charging pump is operable.

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS An evaluation for Potential Unreviewed Safety Question was performed, and the following was determined.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The procedure revision provides guidance to align one charging pump since the UFSAR does not provide any guidance in this area. The procedure change was needed since only one charging pump would be operable during a DBA.

2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created.

One charging pump is required in a DBA. The operability of the pump is not adversely affected by this change.

3. The margin of safety as described in the basis of any technical specification was not reduced. The operability of flow paths and leak tightness of valves is not adversely affected.

SNSOC APPROVAL DATE April 15, 1987

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m 1 AND 2-EP-1: LOSS OF REACTOR OR SECONDARY COOLANT q PWMg To change the cold leg to hot leg safety injection recirculation swapover time 3

to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and the subsequent swapovers to every 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to prevent boron precipitation. This is due to tho increase in refueling water storage tank, casing cooling and accumulator boron concentration:.

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS An evaluation for a Potential Unreviewed Safety Question was performed, and the following was determined.

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The probability of boron precipitation is decreased by the more frequent swapover times.

2. The possibility for an accident or malfunction of a different type than was previously evaluated in the safety analysis report was not created.

, Swapover sequencing of valves remains unchanged. More frequent swapovers l will prevent boron precipitation.

3. The margin of safety as described in the basis of any technical i specification was not reduce 4 Operability of the emergency core cooling

] system has not been affected.

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SNSOC APPROVAL DATE April 15, 1987 l

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i HP-3.2.5: RADIOACTIVE MATERIAL CONTROL STORAGE OF CONTAMINATED TOOLS AND EQUIPMENT PURPOSE The change was to allow the storage of a contaminated rotor in warehouse facilities which are outside the security fence. t i

EVALUATION FOR UNREVIEVED SAFETY QUESTIONS

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The storage facility does not affect any safety related equipment.

2. The possibility for an accident or malfunction of a different type than -

was previously evaluated in the safety analysis report was not created.

Only the location of the contaminated rotor was changed. Increased  :

contamination levels in the new storage area are within limits specified by 10 CFR 20.  !

3. The margin of safety s described in the basis of any technical  ;

specification was not reduced. No safety related equipment will be ,

affected by the storage facility.

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SNSOC APPROVAL DATE l 1 .

i q May 19, 1987

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l l-TOP-11.3: PURCING "C" STEAM GENERATOR SECONDARY TO PROCESS VENTS VIA THE AERATED PRIMARY VENT SYSTEM  :

PURPOSE I

To reduce radiation levels in "C" steam generator prior to filling, draining and/or transferring.

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS

1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report was not increased. The arrangement will not change the process vent operation. Only suction of the vacuum pump is changed. Operation is similar to that of venting the reactor head. l

2. The possibility for an accident or malfunction of a different type than i

) was previously evaluated in the safety analysis report was not created. ,

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3. The margin of safety as described in the basis of any technical specification was not reduced. Process vent operation is not being changed nor is the margin of safety as described in the technical i specifications.

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' i SNSOC APPRCV.it DATE i

July 17, 1987 i

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SPECIAL TESTS REQUIRING A SAFTEY EVALUATION IN 1987 1-ST-58: DIESEL DRIVEN FIRE PUMP-FULL CAPACITY TEST PURPOSE To test the full capacity of the diesel driven fire pump (common to both Units) and to determine its head vs. flow curve.

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS An evaluation for a Potential Unreviewed Safety Question was performed, and the following was detetmined.

1. The performance of this Special Test,did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report because the motor driven fire pump was fully operable.

2. The performance of this Special Test did not create a possibility for an accident or malfunction of a different type than any evaluated previously in the saf(ty analysis report because the motor driven pump is capable of handling 100% of the fire protection needs.

3. The performance of this Special Test did not reduce the rergin of safety as defined in the basis of any Technical Specification because the action steps of Specification 3/4.7.14 were satisfied.

DATE COMPLETED:

2/3/87, 2/7/87, and 2/10/87 TEST RESULTS The measured (2/3/87) pump head curve wab less than the design curve.

Corrective maintenance was performed on the pump with assistance from the vendor. The test was reperformed on 2r7/87 and the measured pump head curve was again less than the design curve. Corrective maintenance was performed on a leaking relief va)ve and enhanced instrumentation was installed. The test was reperformed on 2/10/87 and the measured pump head curve exceeded the Technical Specification, UFSAR, and design head curves.

TEST RESULTS APPROVED BY SNSOC 2/11/87 5

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1-ST-69

DETERMINATION OF EMERGENCY DIESEL GENERATOR (EDC) LOAD LIMITER 3000 KW SETPOINT PURPOSE To permanently mark on the EDG load limiter the 3000 KW setpoint.

t EVALUATION FOR UNREVIEWED SAFETY QUESTIONS ,

An evaluation for a Potential Unreviewed Safety Questfor was performed, and the following was determined.

1. The performance of this Special Test did not increace the probabiliti of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety

, analysis report since this test was performed in conjunction with the normal surveillance.

l 1 2. The performance of this Special Test did not create a possibility for an accident or malfunction of a different type than any evaluated j previously in the safety analysis report since the test acquired i extra data in conjunction with the normal surveillance.

3. The performance of this Special Test did not reduce the cargin of safety as defined in the basis of any Technical Specification since  !

i this test acquired data during the performance of a no rmal i

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surveillance.

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l 12/16/86 (1J), 12/23/86 (2H), 1/7/87 (1H), and 1/12/87 (2J) l

! TEST RESULTS 1

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' This test was performed satisfactorily on all 4 diesel generators. l t

l TEST RESULTS APPROVED BY SNSOC I I

12/18/86 (1J), and 1/15/87 (2H, lH, and 2J).

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1-ST-72: LARGE BORE SNUBBER ACCUNULATOR SYSTD( TROUBLE SHOOTING -

PURPOSE To provide a systematic inspection of the snubbers and reservoir system connected to a 1crge bore snubber accumulator.

• Note: This test was performed three times. ,

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS An evaluation for a Potential Unreviewed Safety Question was performed, and the following was determined.

1. The performance of this Special Test did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report since this was only an inspection.

2. The performance of this Special Test did not create a possibility for an accident or a malfunction of a different type than any evaluated previously in the safety analysis report because this test did not manipulate any equipment. i

3. The performance of this Special Test did not reduce the margin of safety as defined in the basis of any Technical Specification since this test was not a required surveillance test.

DATE COMPLETED r 12/3/87 ,

t TEST RESULTS Sufficient fluid was found in the reservoir system to ' determine that all snubbers were operable.

TEST RESULTS APPROVED BY SNSOC 12/14/87 7

DATE COMPLETED i

8/21/87 i

TEST RESULTS Sufficient fluid was found in the reservoir system to determine that all snubbers were operable. The vent plug on 1-RC-HSS-005B was loose and determined to be responsible for the leakage. The plug wao tightened and the reservoir refilled and inspected for any further leakage. None was found.

TEST RESULTS APPROVED BY SNSOC 2/18/88 i

DATE COMPLETED 4/25/87 ,

TEST RESULTS Sufficient fluid was found in the reservoir system to determine that all snubbers were operable except 1-RC-HSS-005B. The valve block was empty

rendering that snubber inoperable. A engineering evaluation was ,

completed and determined that an over stress condition did not occur and "

the system was approved for continued operation.

TEST RESULTS APPROVED BY SNSOC l

2/18/88 l NOTE: THIS SPECIAL TEST WILL BE DESIGNATED AS A MAINTENANCE PROCEDURE IN 11tE F11TURE AND NOT A SPECIAL TEST.

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1-5T-73: DIFFERENTIAL PRESSURE TEST OF N011)R OPERATED VALVES PURPOSE To test Charging Pump Suction Valves 1-CH-MOV-1267A, 1269A, 1269B, and 1270A to determine the stem thrust required to open the valves under maximum design differential pressure and to validate the methods used to determine the required thrusts for IEB 85-03 notor operated valves (MOVs).

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS Au evaluation for a Potential Unreviewed Safety Question was performed, and the following was determined.

1. The performance of this Special Test did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis report since this test was performed to verify that the motor operators have sufficient thrust available to perform their safety functions.

2. The perfo.mance of this Special Test did not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report since the test was performed during a refueling outage.

3. The performance of this Special Test did not reduce the margin of safety as defined in the basis of any Technical Specification since all applicable limiting conditions for operation were satisfied during the performance of this test.

DATE COMPLETED 6/8/87 TEST RESULTS All valves open under design differential pressure of 180 psid.

TEST RESULTS APPROVED BY SNSOC 6/18/87 9

1-ST-74 "B" MOISTURE SEPARATOR REHEATER (MSR) REHEAT AND INTERCEPT TEST PURPOSE To determine if a cycle steam flow mismatch exists between the A and B MSR's due to a Reheat or Intercept valve being out of the expected position, and to verify that 1-IV-MS-1B and 1-RSV-MS-1B will shut when electrohydraulic control fluid is secured.

EVALUATION FOR UNREVIEWED SAFETY QUESTIONS An evaluation for a Potential Unreviewed Safety Question was performed, and the following was determined.

1. The performance of this Special Test did not increase the probability of occurrence or the consequences of an accident or malfunction of i

. equipment important to safety and previously evaluated in the safety analysis report since the valve freedom test, performed monthly, cycles the same values in the same manner.

2. The performance of this Special Test did not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report since the valve freedom test is an analyzed test, q 3. The performance of this Special Test did not reduce the margin of

safety as defined in the basis of any Technical Specification since

, the test was not used to satisfy any surveillance requirement.

DATE COMPLETED 7/7/87 TEST RESULTS Both the reheat and intercept valves closed when EHC was isolated. The "B" MSR Reheat Valve only opened approximately 1/8 of the normal travel.

The Reheat valve was declared inoperable on 7/8/87. The unit was shutdown and the valve was repaired. After shutdown, the intercept valve '

was also found to be inoperable. The subject of both the reheat and intercept valves being inoperable is addressed in North Anna Unit 1 LER 87-16.

I TEST RESULTS APPROVED BY SNSOC i

7/21/87 a

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FACILITY CHANCES REQUIRING A SAFETY EVALUATION DURING 1987 EWR 83-069 and EWR 83-069A DESCRIPTION EWR 83-069 allowed the motors of 2-CH-P-1B and 1-CH-P-1C to be interchanged.

The mounting holes in the base plate of the replacement motor for 2-CH-P-1B were elongated up to 1/2" to allow for proper alignment of the motor with the speed increaser. This was done with Westinghouse (motor manufacturer) concurrence. Elongation of the mounting holes removed insufficient metal to affect the strength of the motor base feet. To prevent axial movement during a seismic event, spacers were inserted in at least 2 holes at diagonal corners.

The spacers were fabricated from ASTM A-36 1/2" to 3/4" steel, and held in place by flat washers under heads of the bolts.

SAFETY ANALYSIS SUIStARY:

The modification does not affect the head curve of the pump or its performance as described in the UFASR. The pump was tested, and required surveillance performed, prior to being declared operable to verify performance and vibration levels. The seismic qualification of the motor mounting base was not affected.

This modification did not affect any pressure retaining components. The function and design basis of the charging pump was not changed by this modification. By allowing for proper motor alignment, this modification, maintains the current margin of safety.

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EWR 83-069B DESCRIPTION The base plate holes of the motor for 2-CH-P-1B were elongated to align motor to the speed increaser. EWR 83-069 & 83-069A were written to do this. The motor was returned to Westinghouse to be rebuilt. When it was returned, the bolt holes were no longer elongated.

SAFETY ANALYSIS

SUMMARY

Proper motor alignment will be ensured by this modification thus maintaining the safety margin. The ability of the charging pump to perform is unaffected. The seismic qualifications of the motor mounting is not affected.

Filler pieces will be added in diagonally opposed holes to assist in limiting axial movement.

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EWR 83-188 DESCRIPTION Blank flanges on drain lines from RHR heat exchangers were replaced with flanges containing nipples with threaded pipe caps. The affected drain lines were:

Unit 1: 3/4"-RH-33-602-Q2 Unit 2: 3/4"-RH-433-602-Q2 3/4"-RH-34-602-Q2 3/4"-RH-434-602-Q2 3/4"-RH-35-602-Q2 3/4"-RH-435-602-Q2 3/4"-RH-36-602-Q2 3/4"-RH-436-602-Q2 This modification significantly reduced the amount of time that personnel spent in high radiation areas when flushing the above lines. Installation and testing procedures were per NAS-1009 requirements. The weight changes on the lines due to this codification were negligible.

SAFETY ANALYSIS

SUMMARY

The seismic integrity, design basis, and operability of the RHR heat exchanger drain lines were not affected. The modification was performed in accordance with ASME Section III and NAS-1009 requirements. For these reasons, the probability that a new malfunttion, or already analyzed malfunction, could occur was not increased.

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EWR 84-287 84-288 DESCRIPTION In response to an NRC study concerning failure of Class IE safety-related switchgear circuit breakers to close on demand, Control Operations performed an evaluation of the Emergency Diesel Generator (EDG) output breaker relay. The evaluation confirmed the NRC study, i.e. incorrect operation of close permissive interlocks. The control circuits for EDG IH and IJ were modified by replacing the single phase voltage relay (type NRV manufactured by General Electric) with a new three phase SLV relay (also manufactured by General Electric) which operates on rising voltage.

SAFETY ANALYSIS

SUMMARY

Replacing the existing NRV relay with a new SLV relay will not alter the existing control circuit logic and it will enhance the reliability and availability of the EDG's. Modification is more reliable than original as this system will stand less chance of a common mode failure.

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EWR 84-316 DESCRIPTION In a study performed by Ecotech, it was determined that AC vital bus voltage degradation may occur due to non-safety equipment faults. To correct this for non IE connected loads, the study recommended the installation of current limiting, fast acting fuses in series with the existing circuit breakers. A procedure was provided to install fuses as required.

SAFETY ANALYSIS

SUMMARY

UFSAR system operation description is not affected by this EWR and installation of fuses will have no adverse affect on any systems or components. Fuses are added such that only one electrical supply is affected per fuse. Modification was performed to comply with requirerents of 10 CFR 50.49.

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EWR 84-463 DESCRIPTION The Unit #3 and #4 service water spray array system required removal due to upgrade of the Unit #1 and #2 spray array. The service water spray array for Unit #3 and #4 had no safety related function as these units were previously cancelled.

SAFETY ANALYSIS

SUMMARY

Unit #1 and #2 service water spray array will be isolated from the Unit #3 and

1. 4 spray array. Operation of the Unit #1 and #2 spray array will be unaffected by this modification.

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EWR 84-785 DESCRIPTION It was determined that the fuses in the secondary side of the control transformers utilized in tle emergency 480 VAC motor control centers (MCCs) are not adequately sized to protect the control transformers. Failure of these fuses to open under a short circuit condition may cause the insulation of the control transformer to ignite. New fuses of the proper size were installed.

SAFETY ANALYSIS

SUMMARY

Fuse replacements will provide greater protection against transformer fires.

The margin of safety is increased due to greater fire hczard protection.

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EWR 85-096 DESCRIPTION A 3-phase, 480 VAC circuit breaker VPNEliB3040 in our Emergency Generator and Diesel System was replaced with 3-phase, 600 VAC circuit breaker VPNFB3040. In eddition, circuit breaker PNFB3040 was upgraded from a Category III to a Category I component.

SAFETY ANALYSIS

SUMMARY

Installation of new breaker which meets or exceeds all standards required for the 480 VAC breaker will not alter the operation of the equipment as described in the UFSAR. The probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR has not been increased. No different type of accident or ea1 function previously evaluated in the UFSAR has been created. The margin of safety as described in the basis section of Technical Specifications has not been reduced by this modification.

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EWR 85-160 DESCRIPTION Isolation valves were installed between the flow gauge and recirculation header in each supply line for the Auxiliary Feedwater Pump Lube 011 Coolers. This allows for servicing of individual lube oil coolers and flow gauges without having to isolate the recirculation header, rendering all three Auxiliary Feedwater Pumps inoperable.

A S,FETY ANALYSIS

SUMMARY

The modification doesn't affect operability of the Aux Feedwater Pumps. A seismic analysis has been performed and no additional supports are required.

Testing was performed prior to declaring system operable which showed the capability to remove decay heat and RCS temperature reduction is unaffected.

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EWR 85-248 DESCRIPTION The spray arraya in the Service k'ater Reservoir were modified in order to meet the criteria for operable spray arrays. Some of the nozzels were replaced with caps in order to ensure that the design flow per nozzle was achieved.

SAFETY ANALYSIS

SUMMARY

The Modifications to the spray arrays do not affect heat removal capabil'. ties during worst case environmental conditions. No new type of accident has been created due to the capping of the nozzles. Margin of safety is still adequate as described in any Technical Specification basis.

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EWR 85-314 DESCRIPTION Unit 1 Safety Injection (SI) Accumulator drain valves (1-SI-122,1-SI-135, and 1-SI-156) were changed from 2" Velan globe valves to 2" Conval Y-globe stop clampseal valves with similar flow coefficients. The performance characteristics of the new valves meet the design requirements for the affected lines. A seismic analysis was perforced due to the increased weight of the new valves, and the lines were seismically requalified.

SAFETY ANALYSIS

SUMMARY

Changeout of the drain valves did not affect the ability of the SI Accumulators to deliver borated water to the Reactor Coolant System. The affected lines were requalified for operation during a design basis seismic event. The replacement valves conform to applicable ASME and ANSI specifications.

Applicable limiting conditions for operation were not affected by this modification. The design basis of the affected lines was not affected because the lines were seismically requalified.

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EWR 85-397 DESCRIPTION A restricting orifice was installed in the line in each bank of control room air bottles. The orifice was designed to provide a more even flow of air from the bottles during the first few seconds of operation. They were designed to still allow sufficient flow when the bottle air pressure drops to 200psig.

SAFETY ANALYSIS

SUMMARY

Modification will allows the system to function as described in l'FSAR and Technical Specifications. Installation of restricting orifices will minimize pressure transients when air bottles are dumped and provide a more even flow until pressure control valves can stabilize and properly modulate flow.

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EVR 85-424 DESCRIPTION A high radiation gate was installed in the entrance to the "C" Reactor Coolant Pump (RCP) tno tor cubicle at the 262 foot level of the Unit #2 reactor containment. Installation of door allows for access to seal table cubicle while restricting access to the "C" RCP motor cubicle.

SATETY ANALYSIS

SUMMARY

The door is constructed of metal and is designed to blev away from the RCP cubiele in the event of steam line break. Modification will not affect seismic integrity of structure.

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EWR 85-467 DESCRIPTION Control circuitry to Powers Operator Relief Valve block valses MOV-RC-1535, 1536, 2535 & 2536 was modified to prevent valve direction reversal until the valves have been fu12y stroked. This modification was performed in response to INP0-SER-84-84 which describes a stuck open valve condition. Circuit modification required the addition of jumpers in the block valve control circuitry located in the bench board of the control room.

SAFFTY ANALYSIS

SUMMARY

Isolation function of block valves is unchanged by this codification.

Modification will increase the ability of the component to perform its safety related function by preventing direction reversal during mid-stroke which will in turn prevents potential damage to the valve actuator.

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b EWR 85-611 DESCRIPTION Isolation valves were installed in sensing lines going to Unit #1 Refueling Vater Storage Tank (RWST) level transmitters L1-QS-100A,B, C&D. This installation was performed to allow the level transmitters to be replaced without draining the RWST. The existing isolation valves were reported to have leakage problems and could not provide adequate isolation.

3 FETY ANALYSIS

SUMMARY

Addition of second isolation valve on sensing line will not affect operation of level transmitter. The additional weight of the valve will not affect seismic qualification of tubing or level transeitter. Operability of the system was checked after installation to insure design is adequate and margin of safety is not reduced.

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EWR 85-640 DESCRIPTION Concrete footings were fabricated for attachr.ent to service water spray array support structure. Footings were fabricated in accordance with approved construction drawings and specifications. Installation and justification of footings was provided under Design Change 84-31.

SAFETY ANALYSIS

SUMMARY

Justification and safety analysis for installation of footings were provided under Design Change 84-31. Fabrication of footings was perforned outside service water area and will not affect any safety related equipment.

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EWR 85-649 DESCRIPTION This Ek'R provides instructions for installation of mechanical cycle counters for reactor trip and bypass breakers including an estimate of breaker cycles to date to be used for the initial settings of the counters.

SAFETY ANALYSIS StDNARY This modification does not r.ffect the ability of the reactor trip and by pass breakers to open.

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EWR 85-656 DESCRIPTION Modification was performed to pressurizer level transmitter LT-RC 2000 to remove bellows sensor and capillary tubing and directly ccnnect tubing to transmitter. This modification is similar to modifications performed under Design Change 81-08A & B. This change was analyzed by Gilbert Coemonwealth and k'estinghouse and determined to be acceptable. Piping / tubing modifications were analyzed and determined to be acceptable.

SAFETY ANALYSIS

SUMMARY

Function of equipment will remain unchanged and availability will not be diminished by codification. Modification has been seismically analyzed and perforced in accordance with NAS-1011. UFSAR evaluates removal of bellows from sealed reference legs.

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EWR 85-672 DESCRIPTION This EWR authorized replaceeent of 3/4" Rockwell-Edwards (model #3624) globe valves with 3/4" Rockwell-Edwards (model #36124) globe valves. Valve weights were comparable (within 10%) and therefore a seismic reanalysis was not required. This EWR was for various locations on Unit 1.

SAFETY ANALYSIS SUMM.RY The new valves are identical to existing ralves with respect to pressure retaining ability and are supplied by the same vendor; therefore, integrity is not changed. The probability of failure is not increased nor will a different accident occur. The new valves have the same structural integrity as the existing valves.

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EVR 85-684 LESCRIPTION This Ek'R cuthorized replacement of 3/4" Rockwell-Edwards (model #3624) globe valves with 3/4" Conval Claepseal (model #12G2-316) globe valves. This was for various locations on ASME class 2 or 3 pipe only. Valve weights were compared and restrictions on applications were applied to ensure seismic requirec'ents were met.

SAFETY ANALYSIS

SUMMARY

Valve changeout will not affect system operation. Pressure retaining capabilities of the two valves are equivalent. Integrity of the new valves is equivalent to that of the old valves, and they will be purchased Category 1.

The probability of an accident is not increased nor is the possibility of a different type of accident created.

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EWR 85-715 DESCRIPTION Seven four-inch core holes were drilled in the auxiliary building wall in order to support piping for the gas stripper moisture separator. cas stripper coisture separation is accomplished by a knockout drum.

SAFETY ANALYSIS SUMMAR_Y Core drilling will not affect any fire barrier. Because no rebar was cut, the integrity of the walls is not compromised.

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i EWR 85-728 UNIT 1 DESCRIPTION Environmentally qualified equipment located in moisture harsh areas is required to be installed with environmentally qualified conduit seals to prevent moisture intrusion. Prior to this EWR, only Conax Seals were used. This EWR allowed Rosemount Seals to be usej on various instrumentation devices (e.g.,

transmitters.) The Rosemount Seals are enviromentally qualified to the designed environments.

SAFETY ANALYSIS

SUMMARY

Conduit to transmitter connector by Rosemount provides the same function as existing Conax connectors, therefore, operation does not change. Rosemount connectors are fully qualified for harsh environments.

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EVR 85-813 DESCRIPTION As a result of the temperature upgrade on l'nf t , it was discovered that additional weld metal was required on a ve**1ca? restraining hanger inside containment. This was needed because of LM additional stresses seen by the hanger. The weld size was increased to M s" at specifier locattor.s.

SAFETY ANALYSIS

SUMMARY

Since the modification will restora the seismic integrity of the hanger, the consequences of a calfunction or chances for a different type of accident will not be increased. The veld build-up will restore the structural integrity of the hanger, such that it will withstand a seismic event.

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EWR 85-835 DESCRIPTION This EWA increased the length of the thitable plug handling tool by four feet.

The tool wcs lengthened by replacing the inner and outer tuba se3vints with longer pieces.

SAFETY ANALYSIS

SUMMARY

Lengthening the tool does not affect operation of the tool or decrease its ,

integrity, therefore the probabild.ty of ftl lure is not increased. Op ration of the tool has not changed, therefore a different malfunction 1:111 not occur.

The increased length and weight of the tool does not affect its Integrity, i

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EWR 86-008 DESCRIPTION Reactor Coolant Hot Leg RTD Manifold Upstream Isolation Valve 1-RC-51 can not be opened due to interference between the handle and a nearby structural I-beam. The 14 1/2" long handle was removed. A modified handle was reinstalled. Weight changes due to handle modification were less than 10% of the original weight, so new seismic calculations were not needed. Weight changes due to the handle modification were determined and design documents updated to reflect the weight reduction.

SAFETY ANALYSIS SIDetARY Resistance Temperature Detector (RTD) operation is not affected by a modification to the valve handle. This modification facilitates RTD manifold isolation for maintenance, allowing the valve to perform its original design function. The valve weight change is within the bounding limits of existing seismic calculations. As the modification only affects the manual operation of an isolation valve, the RCS pressure boundary integrity is unaffected.

Sufficient leverage remains to operate the valve with the modified handle.

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EWR 86-176 DESCRIPTION During preparation for the installation of the reactor vessel head shielding, support #2RC41-076 was dismantled. Four (4) items listed on drawing 2RC41-076 could not be found. These items were replaced as specified in the materials list.

_ SAFETY ANALYSIS

SUMMARY

This EWR is not considered a modification. The design of the support has not been altered. The materials and testing requirements are the same as originally specified in Design Change Package 81-S01B.

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EWR 86-214 DESCRIPTION Replaced the packing cartridge on various Conval Clampseal 1/2", 3/4", and 1" forged carbon steel globe valves. These valves were ordered and supplied with bonnets constructed of stellite #21 (AMS 5385) and packing chambers constructed of 416 SS (ASTM A582), Applicable codes do not recognize these materials for pressure boundary applications. The packing chamber / bonnet of each valve was replaced with another made of SA-479 type 316 forged stainless steel. This material is recognized by applicable codes for pressure boundary applications.

SAFETY ANALYSIS

SUMMARY

The replacement packing chamber / bonnet assembly is designed for the valves and is an acceptable replccement per nuclear piping codes. The replacement of the packing / bonnet assembly will not affect seismic integrity of affected systems.

The replacement assemblies are designed for the existing valve bodies.

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EWR 86-252 DESCRIPTION In order to reduce the need for jumpers every time a thermocouple fails for the subcooling monitor, additional thermocouples have been identified and their inputs routed to the P250 computer for use in the subcooling program.

SAFETY ANALYSIS

SUMMARY

The modification performed will not affect the subcooling monitor system. No new type of accident or malfunction have been generated due to this modification. The basis in Technical Specifications has not been impacted.

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EWR 86-294 DESCRIPTION The fuel oil system for the Emergency Diesel Generators (EDGs) could o

• be vented and primed by loosening the fittings. This was considered unacceptable because fuel oil could spill. Vent valves were installed near the discharge of the fuel oil pump and on the duplex fuel oil filters, 1/4" tubing and valves were used. The discharge tubing from the valves was capped so that inadvertent opening of the vent valve would not cause a fuel spill.

SAFETY ANALYSIS

SUMMARY

Installation of these valves does not adversely affect operatien of the EDGs.

The additional weight is negligible, and so does not change seismic analyses.

The modification reduces the potential for fires near the EDGs, and so increases the margin of safety. Although this modification is performed on each EDG, Technical Specification for the EDGs will be satisfied during installation. Because the vent valves are capped when not in use, the modification does not increase the possibility of a new type of accident or malfunction.

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EWR 86-508 DESCRIPTION An additional isolation valve was installed in the sample system from the hydrogen contaminated liquid purge header to the volume control tank (VCT).

This additional valve was installed outside of the VCT Cubicle so that it could be operated without exposing personnel to high radiation. Previously, operators had to close a valve in the VCT Cubicle ( 1000MR at power) to purge sample to gas strippers. This modification was for ALARA and was performed on both units.

SAFETY ANALYSIS

SUMMARY

The operability of the system is not affected. Since the line is not seismic, no seismic analynis was performed. The new valve is being installed in accordance with applicable procedures and is designed for its intended application.

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EWR 86-520 DESCRIPTION This EWR authorized replacement of a 1 1/2" check valve (1-MS-71) with a 1" check valve. The old check valve was previously injected to stop leakage at an upstream trap. There were no 1 1/2" safety related check valves in stock.

Piping stresses were reanalyzed due to change in valve weight. Additional flow restriction was analyzed as being acceptable.

SAFETY ANALYSIS

SUMMARY

Probability of a main steam line break upstream of the trip valve is not increased by this modification. The valve is being replaced with a valve of the same pressure rating and ASME class as the original. The structural integrity of this ASME class 3 component is not affected.

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EWR 86-521 DESCRIPTION The emergency shower and eyewash station in the Service Water Pump House wac moved with the exception of the water storage tank 1-SW-TK-5. This tank is used by the Chemistry Department to add a corrosion inhibitor to the Service Water System.

SAFETY ANALYSIS

SUMMARY

UFSAR Figure 9.2.5 is to be revised to remove the eyewash station. This does not affect the integrity of the Service Water System since the eyewash station was not physically attached.

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EWR 86-545 DESCRIPTION The control room door (S76-26) that was damaged due to welding was replaced along with its frame.

SAFETY ANALYSIS

SUMMARY

The replacement door is High Power Rifle (Level IV) Bullet Resistant and Class A, fire rated. The associated frame assembly has been evaluated for seismic integrity. A leak test was performed on the door.

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EWR 86-550 DESCRIPTION The support for flex conduit 1 CH9380A, which is a feeder to the motor for 1-SW-P-4, had come unbolted. The support was reinstalled. The original holes for the support were repaired. QC verification was required for installation of the anchor bolts.

SAFETY ANALYSIS

SUMMARY

The support was reattached in accordance with approved specifications. The reinsta11ation of conduit support returned 1-SW-P-4 to its original design.

Therefore, operation of the Service Water System was not affected.

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EWR 86-654 DESCRIPTION Six indicator scales with a range of 0-200*F were replaced by scales with a 0-400*F range as required by Regulatory Guide 1.97 for the Qualified Resistance Temperature Detectors that were installed per DCP 83-34. Loops T-LM-100 182 and T-LM-200-l&2 were recalibrated to the new range of 0-400*F.

SAFETY ANALYSIS

SUMMARY

The change was in the range of the indicator scale only. The material is identical to the original.

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EWR 86-675 DESCRIPTION The broken operator linkage on 1-SW-6 is to be temporarily replaced with a previously repaired operator linkage bracket until a new linkage bracket is available. The broken locating pin for the valve stem operating a rm may be replaced with one fabricated fron 3/4" diameter or greater carbon steel round stock.

SAFETY ANALYSIS

SUMMARY

Should 1-SW-6 fail, alternate line-up capability provides supply redundancy so that Service Water will continue to be available as required by Technical Specification. The material used in this modification is adequate to withstand the worst expected environment to which it could be exposed. Addition of weight is negligible, therefore a stress reanalysis is unnecessary.

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EWR 86-688 DESCRIPTION Transmitter 01-FW-FT-100C was found to be inoperable. No replacement transmitters were in stock. The transmitter was replaced with an environmer. tally qualified component which meets or exceeds all design criteria.

The model was a Foxboro N-E13-DM.

SAFETY ANALYSIS

SUMMARY

Since the new transmitter is a better quality item, the probability of failure is reduced. The consequence of a malfunction remains unchanged. No basis as described in Technical Specifications has changed.

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EWR 87-070 D3SCRIPTION A differential pressure existed between the Unit #I & II Fan Rooms, the adjoining Instrument Rack Room within the Control Room ventilation envelope, and the Chiller Rooms. The block walls between the Fan Room and Instrument Rack Room were removed to decrease the restriction in the return air flow to the Fan Room. Reducing this restriction will cause the differential pressure between the Chiller Room and Fan Room to increase.

SAFETY ANALYSIS

SUMMARY

Removal of block wall was seismically analyzed and will not affect structural integrity. Block wall is not a fire barrier and therefore will not affect Appendix R analysis. Removing the wall will not create a different type of accident or malfunction than previously described in UFSAR.

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EWR 87-074 DESCRIPTION Six (6) 3/4" pressure taps were installed around the governor valves on each side of the High Pressure Turbine. Taps were installed in locations to allow Surveillance & Test Engineering to test and evaluate performance.

SAFETY ANALYSIS

SUMMARY

The main steam system is not nuclear safety related in this area. Failure of one of these lines would result in only a small steam line break.

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EWR 87-120,87-191, 87-215,87-631, 87-123 DESCRIPTION The limit switch wiring modifications in these EWRs set up each affected motor operated valve (MOV) for four (limit switch) rotor operation. This modification allows the torque switch bypass switches to be set nore conservatively without affecting the light indication or limit stops.

SAFETY ANALYSIS

SUMMARY

These modifications do not adversely affect the operation of systems described in the UFSAR but will instead provides increased MOV reliability.

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EWR 87-162 DESCRIPTION This EWR qualified the use of 3/4" and 1" VGS-60B (bolted bonnet) and VGS-60C (welded bonnet) gate valves to replace 3/4" and 1" VOS-60r globe valves for Service Water vents and drains. Cate valves are much easier to clean if clogging occurs than are the globe valves currently in service. Design documents are to be updated by revision to this EWR whenever a substitution is made. The substitute valves are permitted by the original piping specification NAS-174 and by ANSI B16.5. This EWR established procurement specification for the replacement valves, and requi*es review under the ASME Section XI Replacement Program / Safety Related Welding Procedure prior to the installation of any replacement. A seismic evaluation of the replacement valves added the following restrictions:

Maximum cantilever length from the pipe 0.D. to the valve centerline not to exceed 18".

The base pipe run shall not be less than 3" NPS.

SAFETY ANALYSIS

SUMMARY

The proposed replacement valves meet the same ANSI /ASME requirements as the valves currently in use. These valves meet the pressure ratings for the applicable lines and pipe classes. The probability of a common mode failure is not increased because the replacement valves have the same pressure retaining capability as the original vent and drcin valves. Seismic calculation CE-307, performed for EWR 85-678, qualifies the valves for use. The probability of a new, or previously evaluated, malfunction is decreased because the replacement valves are easier to clean than the original valves if plugging occurs.

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EWR 87-166 DESCRIPTION It was determined, based on engineering calculations, that the 1/8" thick diaphragms in the recirculation spray heat exchangers could yield when subject to external Type A test pressure conditions. Therefore these diaphrages were replaced with 1" thick diaphragm plates of the same material which will see stresses far below the maximum allowable stress valve for ASME III, Class 3 components.

SAFETY ANALYSIS

SUMMARY

Modification teets original design criteria of recirculation spray heat exchanger. Seismic evaluation was performed to insure seismic integrity of structure is unaffected.

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EWR 87-257 UNIT 2 DESCRIPTION The contact and thermal overload blocks for 2-SM-MOV-213A were damaged. These components are located in an motor control center in harsh environment.

Acceptable replacement parts were used to replace the damaged ones.

SAFETY ANALYSIS

SUMMARY

Replacement parts were exact replacements and environmentally qualified.

Electrical separation was maintained. No protective circuitry was involved.

Operability of SW System was not degraded dt' to the modification since the original installation configuration did not change.

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EWR 87-261 DESCRIPTION A study was performed to determine the acceptability of replacing obsolete Lonergan LCT 11 relief valves (RV) with Lonergan LCT 14 relief valves. The LCT 14 relief valve is made of the same material as the LCT 11. Also dimensions and weight were the same, therefore no modification to piping or additional supports are required. The orifice of the LCT 14 was slightly larger than the LCT 11, allowing for greater relieving capacity which is acceptable. The LCT 14 is an acceptable replacement for LCT 11.

SAFETY ANALYSIS

SUMMARY

The new RV's have same weight as original RV's, therefore seismic integrity of piping is not affected. Materials of construction conform to NAS-Specifications for original RV valves, are a direct replacement of improved design, and are rece mended by the manufacturer; therefore, probability of failure has not been increased.

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EWR 87-281 DESCRIPTION This EWR approved the following substitutions for C Steam Generator (S/G) blowdown sample isolation valves:

Mark # Old Valve Type New Valve Type 1-SS-576 Vought 2"-SW-12111, Vought 2"-SW-2801, 21.8 lbs (VGS-60B) 16.0 lbs (VGS-60C) 1-SS-216 Vought 1"-SW-12141, Vought 1"-SW-2821, 8.7 lbs (VOS-60C) 7.6 lbs (VOS-60D)

NAS-1009 Class 601 piping specification calls for both VGS-60B gate valves and VOS-60C glove valves on 2", and smaller, lines. These are 600 lb. bolted bonnet, forget carbon steel valves. The original plant piping specification (NAS-174) allows VGS-60C and VOS-60D valves in this application. These valves are 600 lb. welded bonnet, forged carbon steel. As a result, bolted bonnet and welded bonnet valves can be used interchangeably on 2" and smaller Class 601 piping. The replacement valves were procured as safety related, and met the QA and technical requirements in the purchase specifications for the original valves. Design document updates will be performed by revision to this EWR when the substitute valves are actually installed. A seismic analysis was done to verify that the weight changes would not have any adverse affects.

SAFETY ANALYSIS

SUMMARY

System function and operability are unaffected because the applicable piping specifications allow the valves to be used interchangeably in this size of Class 601 pipe. The seismic evaluation determined that the lighter replacement valvec would not adversely affect system operability during a seismic event.

As the valves are allowed by existing standards and installed in accordance with existing procedures, the probability of a new or previously analyzed malfunction has not increased.

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EWR 87-306 DESCRIPTION To assist in the alignment of the Casing Cooling Pump Motor, carbon steel plates were welded to the motor base plate. The additional weight of the plates will not affect the seismic calculations for the pump. The plates shall be welded to the motor base plate using a safety related weld procedure.

S/.FETY ANALYSIS

SUMMARY

The plates will only be used during alignment of the pumps and will therefore have no effect in the function or operation of the pumps. The seismic analysis of the pump is unaffected.

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EWR 87-331 DESCRIPTION It was determined that it is acceptable to upgrade the Reactor Coolant Pump (RCP) Seals as described in Westinghouse Technical Bulletins #NSID-TB-85-5 Rev.

I and #NSID-TB-85-15 and Westinghouse Letter JDF-84-72, dated 9-19-84. Their modifications will improve seal performance as well as provide increesed seal reliability.

SAFETY ANALYSIS

SUMMARY

All replacement components are providei by Westinghouse and designed for service conditions. The controlled RCP leakage and RCS leakage limits with the upgraded seals will still be in accordance with the technical specifications The overall RCP seal design is not being altered. Only certain components are being upgraded to improve seal reliability. The consequences of seal failure remain unchanged.

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EWR 87-337 DESCRIPTION A carbon steel deck plate fell into the reactor cavity. A visual inspection and liquid dye penetrate test was used to determine if there was any damage. A significant dent in the liner floor was the only rejectionable indication. A stainless steel plate was installed over the dent and welded all around the edges. A final dye penetrant test of the repair welds was done to verify a satisfactory condition.

SAFETY ANALYSIS

SUMMARY

This codification does not affect the design function of the liner since it is restoring it to its original condition. The weight of the plate added is insignificant and will not affect the seismic analysis, i

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EWR 87-338 DESCRIPTION An evaluation was performed to determine if a nearly identical replacement switch for operation of Reactor Containment Purge Air Supply MOV-HV100B could be used. The only differences between the two switches were minor differences in the casing molding. Switch was a standard unit used throughout industry and not specific to nuclear applications.

SAFETY ANALYSIS

SUMMARY

The replacement switch is identical in operation and function. It would require the additional failure of MOV-HV100A or a passive failure of vent line to allow significant release of radioactive material in the event of an accident.

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EWR 87-339 DESCRIPTION In order to increase the margin of differential pressure between the Unit #1 air conditioning room and the chiller room, supply fan 1-HV-F-24, which supplied the chiller room, was relocated to exhaust the chiller room. The fan was relocated to an opening at the opposite end of the same wall and a short duct with a balancing damper and adjustable discharge vanes was installed in its place.

SAFETY ANALYSIS

SUMMARY

The control room habitability system will still operate as designed. The design function and operation of the chiller room fan and emergency filter is unaffected. Chiller room fan failure would not cause failure of redundant systems.

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EWR 87-343 DESCRIPTION It was found to be acceptable to use ASTM A325 type 1 bolts for the Reactor Cavity Seal Ring in place of ASTM A307 Grade B bolts. These bolts are temporary and are removed upon the completion of refueling.

SAFETY ANALYSIS

SUMMARY

This modification does not affect the structural integrity or operability of the cavity seal ring. The replacement acterial is stronger than the original and will increase reliability. The sealing of cavity and therefore the water level in the cavity is unaffected.

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EWR 87-347 DESCRIPTION Pressurizer safety valve loop drain valves 1-RC-139 and 1-RC-140 required replacement. The valves replaced were 1" Rockwell figure 3624 -alves which weigh 18 lbs. The repic. cement valves are 1" Rockwell figure 36124 and weigh 15 lbs. New valves were basically a direct replacement except for valve weight.

This thange in valve weight was analyzed and fout to be acceptable without sny additional support requirements.

SAFETY ANALYSIS

SUMMARY

Installation of new valves will not affect system operation er pressurizer safety valves. The difference in valve weight was seismically analyzed and determined not to affect system integrity.

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EWR 87-362 DESCRIPTION When MOV-FW-100D was cycled by hand, the operator stem unscrewed from the coupling block rendering the valve inoperable. If the valve had been cycled from the control room and this had occurred, the control room operator would get indication that the valve was open when it was actually closed (indication comes off of a limit switch on the operator). A solution was to drill and pin the operator stem to the connecting block. This action had already been taken on MOV-FW-100A,B,C. This EWR addressed pinning MOV-FW-100D, MOV-FW-200 A.B.C.D.

SAFETY ANALYSIS

SUMMARY

The modification will not affect operability of the MOVS based on experience with MOV-FW-100A,B.C. Due to the small amount of material removed and the location, the MOVs are not expected to be adversely affected by a seismic event. Cycling of the valves after pins are installed shall prove operability.

The modification will increase the reliability of the MOV's. However, if the pin breaks and the stem becomes separated from the coupling block, another valve could be lined up to supply Auxiliary Feedwater.

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EWR 87-371 DESCRIPTION A 3/4" NPS Voght model SW-1211 bolted bonnet gate valve (valve 1-FW-75, S/G 1evel transmitter root valve) was replaced with a 3/4" Conval Clampscal model 11G3-105 globe valve. The Conval valve has a higher preasure rating than the original Voght valve. fhe weight difference between the two valves is less than 10%, so a seismic analysis is not required. The replacement valve has a higher flow coefficient than the original valve, but this is of no concern because the valve is used in a very low flow application. The replacement valve was purchased as safety related and exceeds the OA and purchase requirements of the original valve. The valve was installed per maintenance procedure MMP-C-W-1 (approved for safety related equipment). This work is exempt from the ASME repair / replacement program because the valve size is less than 1" NPS. Design documents were updated by MMP-C-W-1.

SAFETY ANALYSIS

SUMMARY

Steam Generator level indication and control system is unaffected by this modification because level indication does not depend upon dynamic flow through the valve. No seismic concerns exist since the weight difference between the two valves is less than 10%. The design temperature and pressure ratings of the replacement valve exceed the requirements for the original valvo.

Inspection and test requirements relating to valve installation are addressed by the installation procedure. As the replacement valve performs an equivalent function to that of the original valve, no unanalyzed condition is created by this modification.

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EWR 87-365 DESCRIPTION Valves 1-SI-189, -271, -272, -273, and -275 were changed from 3/4" (T58)

Rockwell-Edwards globe valves to 3/4" #12G2-316 Conval Clampseal globe valves.

The replacement valves met or exceeded the requirements and technical specifications for the original valves (P.O. NA-10, 2485 psig &200F). Being less than 1", the valves were exempt from the station ASME repair / replacement program. The valves were installed per maintenance procedure MMP-C-W-1. The original valves weighed 12 lbs. and the new valves weighed 10 lbs; calculations were performed to verify that the weight changes did not compromise structural integrity of connected piping, supports, or components. Design drawings were revised to reflect the valve replacement.

SAFETY ANALYSIS

SUMMARY

The service ratings of the replacement valves exceed the line design temperature and pressure. The valves were purchased as safety related, and inatalled in accordance with a procedure approved for use on safety related valves. The installation procedure included NDE requirements to verify proper installation. The valves were seismically qualified. After replaceeent, the required testing will be performed before the valves (and affected systems) are declared operable. If a single valve fails, the opposite train will still supply adequate flow. This modification did not increase the probability of an unanalyzed, or previously analyzed, malfunction.

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EWR 87-379 DESCRIPTION Declutch shaft bushings for Limitorque motor operated valve operators were fabricated to original dimensions using magnolia tronze (modified ASE 64).

Although this material is not the same as the original bushing material, it is used in similar applications, such as bearings. Since the declutch shaft is only used when placing the operator in manual operation and rotation of the shaft is 1/4 turn, the new material is considered to be acceptable.

SAFETY ANALYSIS

SUMMARY

Due to the size and material of the bushing it will not be affected by a seismic event or affect surrounding equipment. Wear of the bushing due to friction will not be increased based on manufacturing material and recommended uses. Operability of valve is not affected by this modification since the substitute material is of equal quality to the original. Previous evaluations in the UFSAR are not affected.

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EWR 87-385 DESCRIPTION The Hypoid GL-Gear on the drive sleeve of 1-CH-MOV-ll15D rotated on the drive sleeve. In doing so it destroyed the pin and elongated the hole in the drive sleeve. A new pin was fabricated of the same material as the original pin and to the original dimensions. A new hole was drilled in the drive sleeve 1r0*

from the original.

SAFETY ANALYSIS StM(ARY The modification does not affect the operation of valve and since the change in weight of operator is negligible the seismic analysis is unaffected. The purpose of the pin is to prevent the hypoid GL gear from rotating independent of the drive sleeve. Drilling a new hole in the drive sleeve did not affect its integrity due to the negligible amount of material removed. Installing the new pin allows the valve to operate as designed. However, if the valve failed to open, the alternate / parallel flow path through 1-CH-MOV-1115B would supply nuction to the charging pumps.

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EWR 87-395 DESCRIPTION The pump, 1-SW-P-10, was running but no flow was being achieved. It was discovered that the casing pin, which prevents rotation of the inner and outer rings, had become disengaged from the casing. This could cause the suction or discharge to be blocked.

A new slightly longer pin was fabricated. This will prevent disengagement of the casing pin and eliminate similar failures in the future.

SAFETY ANALYSIS

SUMMARY

The opera:ional design of tFe pump is unaffected. This modification does not compromise the pump's integrity. Failure of the pin could lead to pump failure and/or flow blockage, thereby preventing continuous monitoring of effluent.

There are however, alternate methods that are acceptable should failure occur.

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EWR 87-406 DESCRIPTION Two of the General Electric SLV relays experienced failure of power supplies while undergoing commission testing prior to installation for service in Surry Unit #2. The problem was determined to be thermal runaway and subsequently has been corrected by modification to the power supply circuitry. This EWR installed the modified power supplies for the SLV relays which are being used for safety related fuse undervoltage, overvoltage and other applications.

SAFETY ANALYSIS

SUMMARY

The modification increased the equipment reliability. The equiprant will not be exposed to any conditions beyond that to which the original equipment was exposed. Separation of electrical power supplies was not affected.

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EWR 87-410 DESCRIPTION A replacement bonnet and valve internals were needed for 1-BD-TV-100B. These parts were obtained from a replacement valve for 02-BD-TV-200C. The replacement valve was furnished without a backseat. The backseat was not required since they are containment isolation valves which are required to fail closed.

SAFETY ANALYSIS

SUMMARY

The replacement is identical except for the backseat. The lack of a back seat will not prevent repacking of the valve, since isolation can be made both upstream and downstream of the valve.

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EWR 87-414 UNIT 1 DESCRIPTION The subject EWR disconnected four pressurizer heaters due to bad insulation / connection. Adequate heating wattage is still available for RCS pressure control and unit heat up. The minimum heater wattage is still available as required by the Technical Specification.

SAFETY ANALYSIS

SUMMARY

RCS pressure control using heaters remains unchanged with adequate capacity remaining available. Taping ends of determinated cables will prevent inadvertent contact with other conductors. Greater than 125 watts of pressurizer heater capacity will be maintained as required by the Technical Specifications.

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EWR 87-415 DESCRIPTION Gland studs for Rockwell and Velan valves were required to be manufactured because of the long lead times for purchase. The material used was ASTM A-193GRB8 Class 2. A replacement gland stud of the appropriate size was obtained from stock to be used as a pattern.

SAFETY ANALYSIS

SUMMARY

The material used has the same chemical properties as the original material, but has improved mechanical properties. Therefore, the replacement studs are of a supegior quality to the original studs. The upgrade of bolt material will maintain the integrity of the component.

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EWR 87-429 DESCRIPTION Two (2) safety related hangers require that tapered washers be provided to allow full bearing on the baseplate. Engineering was performed to provide dimensions r,nd other details for fabrication of washers.

SAFETY ANALYSIS

SUMMARY

This modification will ensure proper contact of bolts and maintain eystem integrity. The modification will not increase the probability of occurrence or consequences of a malfunction.

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EWR 87-441 DESCRIPTION This EWR machined new motor operated valve stem nut lock nuts for use on 1-SI-MOV-1860A&B. New lock nuts could not be obtained from the manufacturer in a timely manner. Measurements were obtained from the manufacturer and a comparable material was used.

SAFETY ANALYSIS

SUMMARY

The modification meets or exceeds the original code requirements. Identically operations as before. The operation of the valve is not changed. Therefore the change does not increase the probability or consequences of a malfunction.

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EWR 87-442 DESCRIPTION The motor for 1-Sk'-MOV-103C was bad and there were no naw motors in stock. A motor from a Unit #4 motor operated valve was used. The specifications for the two motors were the same except for current and horsepower, but these differences were considered to be acceptable. The new motor was environeentally qualified and safety related as was the original. The new motor was directly compatible with the original and satisfies all previous codes and standardo. Modification doesn't affcet power supplies in such a way as to cause loss or degradation of electrical power supplies.

SAFETY ANALYSIS

SUMMARY

Since the motors are essentially the same, the consequences or probability of a malfunction have not increased, nor has the possibility of a different type of accident. The cargin of safety was not reduced because the basis section of the Technical Specifications was unaffected.

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EWR 87-502 DESCRIPTION Three sets of studs, lugs, washers and nuts on 1-HV-Fu-3A were missing or broken. The broken studs were rewelded onto the filter housing using the same lugs, washers, and nuts. The missing studs were replaced with threaded rod or bolts and attached to the housing by welding. Compatible nuts and washers were used and replacement lugs were fabricated.

SAFETY ANALYSIS

SUMMARY

The repair is not a modification except that a different type of stainless steel is being used for the components. Therefore, the repairs are equivalent to the original components. If all three studs were to fail, the affected filters would have three (3) remaining hold down devices. These would be sufficient to prevent bypass leakage.

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EWR 87-542 DESCRIPTION It was determined that for nuts in sizes up to 2 3/4" diameter that ASTM A-194 Grades 2H and 7 are interchangeable.

It was also determined that for bolts, studs and threaded rods in sizes up to 2 1/2" diameter that ASTM A-193 Grades B7 and B16 are interchangeable.

SAFETY ANALYSIS

SUMMARY

ASTM standards have the same mechanical requirements for tle size of fasteners referenced in this Ek'R. These standards do not change any failure analysis for equipment or systems. The standards for components must meet original design specification. The use of these fasteners does not affect system operations.

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87-559 DESCRIPTION 01-CH-MOV-1350 was found to have a high running load along yith a bad helical spring pack. Repeatability could not be achieved with this spring pack during testing. The spring pack was replaced with a pack which could achieve desired thrust band.

SAFETY ANALYSIS

SUMMARY

The consequences of a valve failure have not been changed due to this modification. The reliability and operability of the emergency borate motor operated valve will be increased, thus increasing the margin of safety. No new type of accident or malfunction has been created.

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EWR 87-569 & 87-570 DESCRIPTION These modifications involved the temporary installation of N-16 radiation monitors on Unit #1 & #2 respectively to provide early detection of primary to secondary leakage through the steam generator tubes. N-16 is produced by the activation of oxygen nuclei in the coolant as it passes through the reactor core. The N-16 detectors shall be located on the secondary side main steam manifold and provide digital indication in the Control Room. Permanent installation of N-16 radiation monitors will be performed under design change 87-27-3.

SAFETY ANALYSIS

SUMMARY

Installation of the N-16 radiation monitors does not affect any safety related equipment. Power for the N-16 radiation monitor is provided by a single semi-vital power supply with electrical isolation provided by circuit breakers and internal fuses. Therefore installation cannot cause loss of channels. The modification increases margin of safety in the event of a steam generator tube failure by providing leak detection prior to break detection.

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EWR 87-572 DESCRIPTION Downcomer flow resistance plates were installed in the Unit #1 steam generators to reduce the mass flow through the tube bundles in the steam generators. This modification was performed to reduce the potential for fluid elastic instability in the tubes. The plates are made of ASTM A-285. Grade C Carbon Steel and welded to the tube bundle wrapper in the downcomer annuals. This ERR was performed in conjunction with Design Change 87-26-3.

SAFETY ANALYSIS

SUMMARY

Installation o1 downcomer plates essentially returns the steam generators to their original condition which included flow resistance plates which were subsequently removed. Stress limits were obtained using ASME Boiler and Pressure Vessel Code,Section III, subsections NB and NG for guidance. If downcomer plates come loose, they will lodge in downcomer area and not adversely affect the steam generator.

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EWR 87-574 DESCRIPTION The actuator housing bolt holes for the upper bearing cap were stripped. A solution was to drill the holes out to the next larger size, recap them, and install new bolts.

SAFETY ANALYSIS

SUMMARY

The new bolts to be used are of greater strength than the actuator's maximum thrust. The additional material removed from the actuator will not affect system response nor actuator integrity during a seismic event. Sinec actuator is being returned to original conditions, the modification doesn't increase possibility of an accident nor does it create an unanalyzed type of accident.

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EWR 87-590 DESCRIPTION The spring pack for 1-HV-MOV-1000 had no preload and could not be adjusted and still maintain required wore travel. Since the original spring pack was no longer manufactured, an equivalent spring pack was installed. The new spring pack provided the same torque bands as the original and the spring pack and torque switch were tested to determine proper torque switch settings.

SAFETY ANALYSIS St2 MARY This motor operated valve is not used under adverse conditions. Valves are closed, locked shut and de-energized when unit is in modes 1-4 purge supply and exhaust valves (HV-MOV-100A, B C, D) are required to close in the event of high radiation in containment during refueling since new spring pack is same as original in form, fit, and f'inction. No increase in probability of failure or different type of accident as described in the UFSAR.

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EWR 87-599 DESCRIPTION After flux thimble eddy current inspections on Unit 2, it was found that the path used for calibration had a 50% wall loss and would have to be isolated.

This EWR provided a procedure to change the calibration path to a spare thimble.

SAFETY ANALYSIS

SUMMARY

Relocation of thimbles does not create a new accident or ea1 function. Leakage detection of incore thiebles reeains operable. Incore instrumentation system will operate as originalb designed. Relocating the thimble tubing will not change the seismic analyais.

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EWR 87-600 DESCRIPTION Limitorque (torque switch) spring pack #60-600-0021-1 is no longer manufactured. This EWR identified a replacement spring pack and specified that the operators affected be MOVATS tested to determine the required torque switch settings.

SAFETY ANALYSIS

SUMMARY

The original spring pack is no longer manufactured and spring pack #0301-111 is an acceptable replacement. The replacement spring pack will meet all requirements of the original spring pack.

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EWR 87-605 Fuses (FRN-1) in our Emergency Diesel Generator Circuits were replaced with a new type of fuses (FRN-R-1). The new fuses have a rejection feature. All specifications are the same.

SAFETY ANALYSIS

SUMMARY

The upgraded fuses do not increase the probability of occurrence or consequences of a malfunction. No new type of n.t1 function has been created.

The fuses are are upgraded. Therefore; margin of safety is not reduced.

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EWR 87-609 DESCRIPTION The installed spring pack for this motor operated valve resulted in steam thrust in excess of the valve manufacturer's recommendation. This EWR provided instructions for spring pack replacement to allow torque switch adjustment within the 2329-4473 # required bond.

SAFETY ANALYSIS

SUMMARY

The new spring pack will meet original specifications. System integrity is improved by replacement of spring pack to insure that thrust limits are met.

86

EWR 87-620 DESCRIPTION This EVR installed-new modified trim set and larger diameter valve stem on the Unit 2 feedwater regulating valves. Continued operational problems and stem failures necessitated thfs change. These changes will dampen vibrations and reduce stresses in the stem caused by flow induced vibration.

SAFETY ANALYSIS

SUMMARY

Main feed regulating valves will operate as originally designed. Feedwater isolation will occur as originally designed. Seismic analysis is not changed ,

since weight change is within 10% of original valve weight. Adverse conditions will not affect valve operation.

87

ENR 87-623 DESCRIPTION 1-RS-MOV-101A would not pass type "C" test when torqued shut but had a satisfactory leak rate when hand shut. This Ek'R provided instructions for use of a limit switch (LS-8) to reduce seating torque and enhance seating.

SAFETY ANALYSIS

SUMMARY

Closing the valve by liait vill allow it to stop in a position that will enable it to pass its type "C" test.

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EWR 87-625 DESCRIPTION The existing flex hose from the reservoir to 1-RC-IISS-008C was broken and was replaced using two hoses with connecting fittings. This was done since there was no hose of the correct length available. The new two piece hose enabled the snubber to operate properly.

SAFETY ANALYSIS

SUMMARY

Failure of the codification will be no more significant than if the un-modified snubber reservoir failed. Using a two piece hose does not increase the probability of ea1 function of safety-related equipment previously evaluated in the UFSAR nor does this codification create a new type of unanalyzed accident.

The hose allows the snubber to be operable and d..es not affect the basis section of the Technical Specifications.

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EWR 87.643 DESCRIPTION The gland leek-off lines on 2-SI-MOV-2863A8B (isolation valves to enarging pump suction header) were recov<d and caps installed at the valves. The original purpose of these lines was to provide indication tnat the packing was failing, and to direct packing leak-off to the Auxiliary Building Sump. An evaluation of the lines ind!cated that they were probably plugged with snric acid and not perf o rming their original design function. Changes in pa. kins configuration caused the leak-off lines to no longer be sealed 'sy valve packing. With the new packing configuration, the gland leak-off lines served no pr pose.

SAFETY ANALYSIS

SUMMARY

The gland leak-off lines do not affect the safe operation of the affected valves. In addition, the lines are not dercribed in the UFSAR. As operation of the valves is uaaffected, accident probabilities are unaffectes'. No new type of valve malfunction is expected as a result of removing the gland leak-off lines.

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EWR 87-686 DESCRIPTION A Nu11matic pressure regulator model' #40-100 was found to be an acceptable substitute for the original air set installed with 1-CH-TV-1204A, under design change 84-57.

The original air set incorporated an integral filter and the replacement does not. This will not affect the function of the replacement pressure regulator, but it was recommended that the original air set be repaired and installed, or a similar~one with an integral filter be installed at the next available

- opportunity.

SAFETY ANALYSIS S12 MARY Operation of the letdown system is maintained by this modification. The subject valve will operate as intended for normal operation and the safety functions, containment isolation.

The valve is safety related and the solenoid operated valve and limit switches are seismically and environmentally qualified. However, there are no special requirements for the actuator air pressure regulator.

A failure of this modification will result in a loss of normal letdown. Excess letdown will still be available for continued unit operation. The safety function of the valve remains unaffected in the event the air set fails.

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EWR 87-696 DESCRIPITON Unit 2 "B" safety injection accumulator relief valve (2-SI-RV-2858B) was leaking by. The relief valve was gagged and an identical valve from Unit I was placed on a nearby vent line flange connection. Pipe stresses were analyzed and found to be acceptable.

SAFETY ANALYSIS

SUMMARY

The consequences of a malfunction or a different type of accident has not been created or increased. The margin of safety is maintained since operation and design function are unchanged. By installing an operable relief valve, the operation and design function is unchanged.

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EWR 87-697 DESCRIPTION The original relief valve on "B" accumulator tank needed to be replaced, but it was no longer manufactured. Manufacturer suggested a suitable replacement which was the same size and same set pressure. New valve met or exceeded E-specifications for original valve. Therefore, it was acceptable for use.

Replacing the existing relief valve with a different model which has the same set pressure, capacity, and size does not affect operability of the system.

New valve does not affect integrity of piping. Previous analysis includes loss of accumulator in single failure criteria. The action statement of technical specifications will ensure that should an accumulator be inoperable, the time during which a concurrent LOCA can occur is minimal.

S_AFETY ANALYSIS

SUMMARY

Since the new relief valve meets the original valve's specifications, there is no increase in probability of malfunction of the relief valve (RV). Failure of the RV would cause accumulator to be inoperable, which has been previously analyzed. The margin of safety is not reduced.

93

EWR 87-831 DESCRIPTION The stem and disc had separated in 1-RC-88 ("C" cold leg loop RTD bypass valves), requiring the valve (2" seal-welded bonnet Reckwell-Edwards globe stop valve) to be replaced. The in-stock replacement Rockwell Edwards valve had been purchased to meet the requirements of ASME BPVC Section III Class 2 components rather than Class 1, and was 10 lbs. lighter than the original valve. However, 1-RC-88 is a Class I component. ASME BPVC Section III allows a Class 2 component to be upgraded to a Class 1 component either by performing additional NDE or by reducing the stress intensification factors and allowable amplitude in the design and fatigue analyses for the line. The valve was qualified for use by a reanalysis of 2"-RC-146-1502-Q1 for a weight change from 50 lb. to 40 lb. for 1-RC-88 in accordance with the requirements of ASME BPVC-III-1-NB, paragraph NB-3673.1, and installed.

SAFETY ANALYSIS

SUMMARY

The RCS is not adversely affected because the new valve meets all applicable AMSE Code requirements for Class I components, after stress and seismic analyses were performed for qualification. The probability of an unanalyzed, or previously analyzed, malfunction occurring has not increased because the valve replacement complies with AMSE BPVC requirements. As a result the margin of safety is unchanged.

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DC 79-71 SAFEGUARDS AND VALVE PIT SUMP PUMPS MODIFICATION DESCRIPTION The safeguards area sump pumps (1-DA-P-01A&B) and valve pit sump pumps, (1-DA-P-06A&B) automatically start when the sumps contain sufficient water levels. The pumps discharge water, including possibly contaminated water, into the auxiliary building sumps. Since the pump controls are not located in the main control room, the control room operator is not able to effectively control the operation of the sump pumps.

The operating conditions of the safeguards valve pit sump pumps are extremely severe due to the lime deposits and abrasion from the water which comes from inleakage from the containment mat area. Pumps with high reliability in severe operating conditions should therefore be provided for the service.

The modification to the control circuitry for the above-mentioned sump pumps includes adding ("start-stop") pushbutton pump controls on the main control board and deleting the automatic start feature. A white indicating light above the valve pit sump pump pushbuttons will show which sump requires pumping, as the valve pit sump pumps each take suction on their own respective sump. Level

. switches will be added to supply a "start" permissive signal to the respective i sump pumps and annunciate on the main control board on sump hi/hi-hi level.

The modification will allow the operator to decide if and when a particular sump should be pumped. -

Because of the severe operating problems and poor reliability experienced with submersible electric driven sump pumps, air operated pumps will be used for the safeguards valve pit sump. These pumps have been proven extremely reliable in a field test of approximately six months and have therefore been selected for this service.

SUMMARY

OF SAFETY ANALYSIS a) The operation of this system is not assumed in any FSAR accident analysis.

b) The possibility for an accident or malfunction of a different type from any evaluated in the safety analysis report is not increased. Neither the purpose nor the effectiveness of the system is altered.

c) The margin of safety as defined in the Technical Specifications is not changed.

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DC 80- S76 THROTTLE VALVE ADDITION TO AUX OIL PUMP DESCRIPTION A high pressure is observed in the auxiliary lube oil system. This is due to the installation of new oil coolers in the system. The new coolers have less flow restrictions than the old coolers. The decreased resistance alters the system characteristic curve and therefore, results in a higher pressure resulting down stream of the cooler. This high pressure is reflected in a higher pressure at the bearings.

Installation of a throttle valve in the auxiliary lube oil pump discharge line will enable adjustments to be made so that the pressure is brought down to the correct value.

SUMARY OF SAFETY ANALYSIS The probability o' an already evaluated accident or nalfunction is not increased because ne intended operation of the system is not changes.

The margin of safety as described in the basis is not reduced because installation of the throttle valve in the output line of the auxiliary lube oil pump will provide the correct pressure in the line.

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DC 81-19 TECHNICAL SUPPORT CENTER-ELECTRICAL INSTALLATION DESCRIPTION-In accordance with the requirements of NUREG 0696 and 10CFR50, Appendix E, '

Article IV.E.8 a Technical Support Center (TSC) shall be established.

In order to support the electrical power requirements for the TSC the following equipment must be added:

a) 480V Motor Control Center.

b) Uninterruptible Power Supply System including batteries, distribution panel, regulated bypass transformer and battery disconnect switch, c) Emergency lighting panel.

d) Normal lighting panel with associated dry type transformer.

e) Lighting fixtures and receptacles.

f) Non-essential and miscellaneous distribution power panels and associated dry-type transformers.

g) Cable raceways, cables and wiring.

h) Station grounding and isolated grounding.

I) Smoke Detection System.

J) Filter Unit Heat Detection System (Wiring only, Mechanical Installation in accordance with the TSC Facilities Installation Procedure DC-81-17), t

SUMMARY

OF SAFETY ANALYSIS Tnis Design Change does not involve modifications to safety related equipment 7 previously evaluated in the Safety Analysis report, systems that may in any way l

create accidents or malfunctions of a different type than previously evaluated j in the Safety Analysis report nor systems whose margin of safety is defined in  ;

' the Technical Specifications.

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DC 81-S20B NUREG-0696 SHORT TERM I&C PROJECT REMOTE MULTIPLEXER INSTALLATION DESCRIPTON In order to provide all plant data, equipment status, and information regarding equipment operation to the emergency response facilities, it is required, per NUREG-0696, that a complete, fully integrated and independent Data Acquisition System (DAS) be installed. The system should be capable of collecting all plant information and transmitting it to the emergency response facility and display systems.

This package provides instruction for the mounting, installation and wiring of the front-end remote multiplexer units for the DAS.

The Validyne DAS is qualified in accordance with IEEE-323-1974 and IEEE-344-1975 and the remote multiplexer front-end unit is qualified as a Class lE isolation device. The remote multiplexer will receive all input data and system parameters for complete interface with the Emergency Response Facilities and displays, and will provide information for improved emergency response per the requirements of NUREG-0696. Vital bus power will be supplied to all remote multiplexer units receiving safety related inputs.

S14 WARY OF SAFETY ANALYSIS This modification installs the front-end multiplexer units of the Data Acquisition System which will collect and transmit reliable plant data and information to the emergency response facilities. The front-end multiplexer units, mounting arrangements, cabling and wiring to the units will be Q.A.

Category I Class 1E. the multiplexer units have been qualified as Class 1E equipment and circuits will not degrade the operation of safety-related components or protection circuits and will not interfere with the operation of control room monitoring devices and plant system functions.

98

DC 81-32B MSR LINE MODIFICATIONS DESCRIPTION Moisture Separator Reheater (MSR) Main Steam tubes are susceptible to premature heating during start-up resulting in possible tube failure. This is due to MSR steam supply lines being too large and the 8" control valves (2-MS-FCV-204A, B, C, D) leaking.

The proposed resolution to the Main Steam Supply Line shall replace the existing automatic 1" purge control line with a manual 3" control line, which bypasses the present 8" control line. The proposed manual 3" control line will be bypassed with a new manual 1" purge control line, thus allowing for a smooth start-up transition and reduce the chances of premature heating and possible bowing.

SUMMARY

OF SAFETY ANALYSIS The modification is not safety-related and does not effect any safety equipment. The codified portions of the system will improve the control of the MSRs and will not change the Design Basis as described in the UFSAR. The margin of safety as defined in the basis for any Technical Specification is not reduced as this modification has no effect on the Technical Specification.

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DC-81-S54B REPLACEMENT OF REACTOR COOLANT RTD's DESCRIPTION The wide range and narrow range resistance temperature detectors (RTD's) presently installed in the reactor coclant bypass manifold have not been tested and qualified to meet the requirements set forth by NUREG-0588 and IEEE-323-1974. The present RTD's will be replaced by environmentally qualified RTD's. The RTD's will meet requirements for IE electrical installations. The electrical connections at the RTD heads have been redesigned to provide proper environmental seals for the replacement RTD's.

SUMMARY

OF SAFETY ANALYSIS The resistance temperature detectors replacements are designed consistent with codes and standards of the existing systems. This modification does not change the characteristics of any system in operation during reactor operation. This modification does not alter the basis for any Technical Specification.

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DC 82-14A and 82-14B CLASS 1E - TRANSMITTER REPLACEMENT UNIT 1 and 2 RESPECTIVELY DESCRIPTION Regulatory Guide 1.97 (RG-1.97) "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following and Accident" requires that sensing instrumentation which provides key plant variables to operations personnel in the control room and designated Category I shall be seismically and environmentally qualified. Instrumentation required to satisfy RG-1.97 ide..tified instruments which require replacement with environmentally qualified equipment was reviewed and the instruments identified are as follows:

Charging Flow to Regen. Ht. Ex.

Low Pressure Letdown Line Flow Hot Leg Flow Ind.

Boron Inj . Tk. Hdr. Flow Ind.

Low Head SI Hdr. Flow "A" Ind.

Flow in LPI Recirc. Air Cooling Coils Flow Volume Control Tank Level Volume Control Tank Level Steam Generator Wide Range Level Steam Generator Wide Range Level Steam Generator Wide Range Level RCS Pressure RCS Pressure Containment Pressure Containment Pressure Containment Pressure Containment Pressure Recire. Spray Pump Discharge Recirc. Spray Pump Discharge The RG-1.97 instrumentation which has been identified shall be removed and replaced with instruments which are seismically and environmentally qualified in accordance with IEEE-344-1975 and IEEE-323-1974 respectively. the replacement components are similar in design to those being replaced and shall be installed in the same locations. The transmitters will be mounted on the wall, on existing instruments racks and pedestals, and seismically installed in accordance with manufacturers instructions. The transmitters will be roterminated utilizing qualified cable seal connectors.

SUMMARY

OF SAFETY ANALYSIS The replacement transmitters are seismically and environmentally qualified, and provide Category I indication in the main control room. Their installation does not affect safety equipment or the operation of safety systems previously evaluated. The replacement transmitters provide the operator with Category I indication to assess the operation of safety systems and does not change safety limits or set points defined in the Technical Specifications.

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DC 83-10 SPENT FUEL CASK LIGHTING AND V/CUUM DRYING MODIFICATION UNIT 1 AND 2 DESCRIPTION The vacuum drying system (VDS) is an auxiliary system provided by Transnuclear, '

Inc. for use in conjunction with the'TN-8L spent fuel shipping cask. The VDS is utilized during spent fuel cask loadlng and unloading operations for gas  ;

sampling, leakage testing and interior cavity drying. The VDS has a flow rate capacity of 70 cfm. The exhaust from the VDS is potentially contaminated. The VDS exhaust will be connected to an existing hose connection in the !

Decontamination Building ultrasonic hood exhaust ductwer.k. The ultrasonic hood exhaust is connected to the ventilation vent stack and can be bypassed through the Auxiliary Building iodine filters, if required. Therefore, exhaust from the VDS may be monitored if it is contaminated. In oTder to use the VDS, electrical power, lighting end coamunications must be provided.

Electrical 480V power for the VDS will be provided from an existing power receptacle located in the Decontamination Building. The 480V power cable will be routed through conduit which terminates in the vicinity of the VDS in the NE corner of the upper level. The power feed to the VDS will be provided with a mating plug that will match with the power receptacle. This plug may be  ;

disconnected when the power receptacle needs to be utilized. Fluorescent lighting will be provided to illuminate both the upper and lower working platform levels. Additional mercury vapor lighting will be installed in the outside loading slab area. These fixtures will be located approximately 30 ft. ,

over the slab areas. The power supply for the VDS lighting will be from an existing lighting panel in the Decontamination Building. A wall counted telephone jack and Gaitronics unit will be provided at the upper working platform level.

SUMMARY

OF SAFETY ANALYSIS This design change does not modify safety related equipment nor reduce the margin of safety of systems as defined in the Technical Specifications.

All materials used are totally compatible (relative to design, quality, and functional requirements) with original equipment, therefore the possibility for a malfunction of a different type than previously evaluated in the UFSAR does not exist.

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DC-83-19A i ELECTRICAL SEPARATION OF RM-P- 159A & B/ NORTH ANNA / 1 DESCRIPTION The motors for Radiation Monitoring System Sample Pumps (RM-P-159A & B) are supplied from the same power source and share a common motor starter. As a result, a loss of.the single' power source or a starter failure will render both pumps inoperable. As these pumps are part of the Containment Atmosphere Particulate Radioactivity and Gaseous Monitoring systems, at least one pump is required to operate continuously per Technical Specifications.

In addition, one radiation monitoring pump and motor is located inside the RM-RMS-159A&B cabinet. Several years ago, due to the high vibration caused by the motor, a high flow relay switch became disconnected and caused an erroneous high flow alarm.

A modification was performed to allow the power to the pump motors to be supplied from one of two sources. Both motors are supplied by the same power source at any one time; however, it is possible to switch to a second source if it is required. In addition, separate motor starters with overload protection were provided for each motor. This allows maintenance on one motor while the other motor is in operation.

The vibration problem associated with the pump motor was eliminated by relocating the pump and motor (1-RM-P-159A) to a location outside the radiation monitoring cabinet, adjacent to the second pump.

SIM(ARY OF SAFETY ANALYSIS The modification implemented by this Design Change will enhance the existing system by making it more reliable. All new equipment will be equal or superior to the equipment presently installed. This modification decreases the system's down time by adding redundancy to the available power source for the system.

The basic function of the systems remains unchanged, and the probability of failure has been reduced.

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DC-83-19B ELECTRICAL SEPARATION OF RM-P-259A & B/ NORTH ANNA 2 DESCRIPTION The motors for Radiation Monitoring System Sample Pumps (RM-P-259A & B) are supplied from two power sources and share a common motor starter. As a result, a starter failure will render both pumps inoperable. As these pumps are part of the Containment Atmosphere Particulate Radioactivity and Gaseous Monitoring Systems, at least one pump is required to operate continuously per Technical Specifications. Also, the existing configuration for transferring the pumps from the normal supply to the alternate supply (via power receptacles) is not considered a good design practice for a permanent installation.

A modification was performed to allow the power to the pump motors to be supplied from one of two sources through a transfer switch. Both motors are supplied by the same power source at any one time; however, it will be possible to switch to a second source if it is required. In addition, separate motor starters with overload protection are provided for each motor. This allows maintenance of one motor while the other motor is in operation.

SUMMARY

OF SAFETY ANALYSIS The modification implemented by this Design Change will enhance the existing system by making it more reliable. All new equipment will be equal or superior to the equipment presently installed. This modification decreases the system's down time by adding redundancy to the available power source for the system.

The basic function of the systems remains unchanged, and the probability of

failure has been reduced.

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DC 83-23 AND 83-24 APPENDIX R EMERGENCY DIESEL GENERATOR ISOLATION UNITS 1 AND 2 REGPECTIVELY DESCRIPTION During the Appendix "R" Safe Shutdown Analysis it was determined that a fire in the control room could damage the safe shutdown system to the point that safe shutdown could not be assured. This was due to the loss of both diesel generators via their control circuits. This design change involves:

1. Relocation of existing Train H relays from the Control Room to the Emergency Switchgear Rooms and addition of current transducers for the ,

Train H circuit breakers.

2. Installation of a transfer switch, lockout relays, metering, and control and indication for circuit breakers and diesel engines in the Train H Diesel Generator Room.

3. Addition of meters in the Diesel Generator Room that are normally used for test and Calibration. Some of these will be used when the diesel generator is controlled from the Diesel Generator Room and will measure bus voltage, frequency, kilowatts, kilovar, and emergency supply breaker amperes.

4. Cable associated with this modification will be routed outside the Control Room and, with the exception of the Emergency Switchgear Room, outside plant fire areas containing Train J diesel cables.

S'DMARY OF SAFETY ANALYSIS This modification will enhance the availability of the Train H Emergency Diesel Generator in the event of a fire in the Control Room.

LOCA mitigating equipment is not degraded as a result of this design change.

The addition of a transfer switch and alternate controls increases the

. availability of the control circuits for the diesel generator. In addition the Diesel Isolation Switch, which is a permissive for the alternate controls is key locked and transfer of control to the Diesel Generator Rocm is alarmed in the Control Room to prevent unauthorized actuation. Contact to contact separation is provided between safety and control portions of the circuit.

The modification does not create a possibility of an accident or malfunction of a different type than any evaluated perviously in the UFSAR since the modification cannot cause damage to or misoperation of any equipment required to mitigate an accident.

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i DC-84-17 AND 84-18 VALVE MONITORING SYSTEMS PORY AND PRESSURIZER SAFETY VALVES, UNITS 1 & 2 RESPECTIVELY DESCRIPTION The Valve Monitoring System (VMS), presently installed for the Pressurizer Power Operated Relief Valves (PORV's) and the Safety Valves (SV's), is not seismically and environmentally qualified as required by USNRC Regulatory Guide 1.97 Rev. 3. Presently installed position switches on the PORV's and accelerometers on the discharge piping of the SV's and PORV's are not qualified to the requirements of IEEE-323-1974 and IEEE-344-1975. The R.G. also requires that the range of closed-not closed position be monitored for these valves.

The existing system provides position indication for each valve and a common alarm for valves in the not closed position in the control room.

The modification on PORV position monitoring systems shall comply with Regulatory Guide 1.97, Category 2, IEEE-323-1974, IEEE-344-1975 requirements and VA Power requirements for QA Category 1 components.

At present, each PORV, has two (2) (non-qualified) limit switches and two (2)

(non-qualified accelerometers to monitor valves open/ closed position. These accelerometers and limit switches shall be removed and in their place four (4) new qualified limit switches for redundant open/ closed position monitoring of PORVs shall be installed. To assure environmental integrity of the limit switches, qua11fied electrical seal connectors shall be used.

The existing (non-qualified) accelerometers on the discharge piping of the PORV's and also preamplifiers shall be removed to eliminate unnecessary surveillance , testing and maintenance. The associated valve monitoring instrumentation on acoustic monitoring cabinet will be deleted, including signals for alarms and indication. Existing control board indicators for these acoustic monitors shall be removed and their associated computer alarm points shall be spared.

The modification on Pressurizer SV's position monitoring shall meet Regulatory Guide 1.97, Category 2 and VA Power requirements for QC Category II. The replacement equipment shall meet the qualification requirements of IEEE-323-1974 and IEEE-344-1975.

On the discharge header of the SV's are located (non-qualified) acoustic monitors (3 active and 3 passive). These (non qualified) monitors and also preamplifiers associated with these monitors shall be removed. In their place, new qualified acoustic monitors and preamplifiers for SV position, shall be installed as required by R.G. 1.97. The existing cable, conduit, computer alarms, and annunciator window assignments for this system will be reused anI additional cables will be added.

The existing control board indicators for the safety valve acoustic monitors shall be removed.

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DC-84-17 AND 84-18

SUMMARY

OF SAFETY ANALYSIS The replacement limit switches and acouatic monitors are environmentally and seismically qualified to provide indication to the operator to monitor station operation. Their installation does not affect safety equipment or the operation of safety systems.

The replacement limit switches and acoustic monitors are environmentally qualified, capable of surviving the accident environment and, therefore, will provide more reliable ir.dication of station operation during a design basis event.

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DC 84-31-3 RESERVOIR SPRAY AND BYPASS SYSTEM, NA UNITS 1 & 2 DESCRIPTION The corrosive nature of the lake water at North Anna has resulted in degradation of the spray array support system. This degradation has required excessive maintenance of the system and has required an extensive inspection program. Previously employed chemical treatments (prior to sodium molybdate) designed to control the corrosive water have had an adverse effect on the reservoir clay liner. Several areas of concern need to be addressed and are listed below:

1. The aggressive water has caused corrosive of the existing carbon steel pipe supports. This corrosion is a result of the carbon steel components.

2. These previous chemical treatments have caused dispersion of the clay composing the select soil liner resulting in suspension and softening of the liner and increased turbidity. This interaction caused the treatments to be discontinued and also limits the types of chemical treatments which can be used.

3. The existing spray system design does not use the surface area of the reservoir to its full potential in providing heat rejection under design meteorological conditions. While the spray system has been analyzed and found to be capable of rejecting design basis heat, significant improvements in the thermal performance of the service water reservoir and spray system appear possible for both design basis accident conditions and normal operation.

4. The ice build-up on the existing fiberglass piping and steel guy wires during the winter months has necessitated a great deal of maintenance work.

5. The fiberglass piping has not proven to be durable. Repair work is required to keep the nozzles and risers in an operable condition.

To resolve this problem a new reservoir spray and bypass system will be installed. The final design concept consists of a new spray system which covers the central and western portions of the reservoir. The partially constructed North Anna Units 3 and 4 pump house will be completed and serve as a valve house for the spray and winter bypass headers.

The various stages of this project will be accomplished by seven (7) design changes and several Engineering Work Requests.

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DC 84-31-3 This design change (84-31) installs the new reservoir spray and bypass piping and support system. The system is to be installed in the central and western portions of the service water reservoir and replaces the existing spray system.

The piping installed will connect to the piping sections installed by design change 84-36, which consists of eight 18 inch spray headers and two 24 inch bypass headers. NUREG-0733, Analysis of Ultimate Heat Sink Spray Ponds, dated August 1981, was used to design a system with adequate thermal performance and water loss characteristics. Carbon steel pipe will be used throughout.

The new service water spray system consists of four pairs (eight total) of individual controllable spray arrays which form the two headers. Each pair of arrays is capable of handling 100 percent of the flow and heat load generated by one unit during normal operation and design basis accident (DBA) conditions.

Additionally, either of the two headers can handle 100 percent of the flow and heat load generated by both units during normal operation and DBA conditions.

Therefore, even with the loss of a complete header, 100 percent capacity remains available.

SUle(ARY OF SAFETY ANALYSIS This design change does not place any new equipment into operation; however the new design and equipment meet or exceed all of the original safety-related requirements of the existing service water system. The new service water system merely uses a new safety related flow path ane spray arrays to accomplish the cooling process. The new spray array will be more efficient than the existing one and the new valve arrangement provides more flexibility.

The new service water design uses the same basic technology as the present design. It consists of the reservoir, pumps, piping (some existing and some new) and a new, more efficient spray array. The winter bypass is a new feature, but it will not be used during an accident. The same accident types will exist for the new system configuration. The new system will be as good as, or better than, the existing system in terns of preventing and/or mitigating these postulated accidents.

The margin of safety is increased with this new system since the new valve arrangement will minimize the effect of one valve failing closed, and the new spray array will be more efficient and more durabic than the existing system.

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DC 84-35 SERVICE WATER RESERVOIR IMPROVENENTS VALVE HOUSE - STRUCTURAL, NORTH ANNA UNITS 1 AND 2 DESCRIPTION The corrosive nature of the lake water at North Anna has resulted in degradation of the spray array support system. This degradation has caused L excessive maintenance of the system and has required an extensive inspection program. Chemical treatments designed to control the corrosive water have had an adverse effect on the reservoir clay liner. Several areas of concern need to be addressed and are listed below: ,

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1. The agressive water has caused corrosion of the existing carbon steel  !

pipe supports. This is a result of the previous chemical treatments being inadequate in inhibiting corrosion of the carbon steel components.

2. These previous chemical treatments have tended to suspend the upper layers of the compacted clay liner. This interaction caused the -

treatments to be discontinued and also limits the types of chemical treatments which can be used.

3. The existing spray system design does not use the surface area of the reservoir to its full potential in providing heat rejection under design meteorological conditions. While the spray system has been analyzed and found to be capable of rejecting design basis heat loads, significant improvements in the thermal performance of the service i water reservoir and spray system appear possible for both design basis i accident conditions and normal operation.

4. The ice build-up on the existing fiberglass piping and steel guy wires during the winter months has necessitated a great deal of mair.tenance  !

work. i

5. The fiberglass piping has not proven to be durable. Repair work is required to keep the nozzle and risers in a operable condition.

r To resolve this problem a new reservoir spray and bypass system and liner covering will be installed. The final design concept consists of a new spray system which covers the central and western portions of the reservoir. The partially completed North Anna Units 3 and 4 pump house will be completed and ,

serve as a valve house for the spray and winter bypass headers. l The various states of this project will be accomplished by six (6) design changes and several Engineering Work Requests. l This design change (84-35) installs only the remaining floor sections, walls,  !

roof, and HVAC penetration missile barriers for the valve house mentioned previously. These will be made of reinforced concrete and will be built up from the existing North Anna Units 3 and 4 pump house floor. Openings for ,

pipink penetrations and doors will be provided, but final architectural details 3 will be added by a subsequent design change (84-36). >

3 r

110 ,

e i

DC 84-35

SUMMARY

OF SAFE 1Y ANALYSIS The new design and equipment meets or exceeds all of the original safety-related requirements of the existing service water system. The additional steel and concrete added by this design change will be seismically designed and will provide missile protection for equipment.

i l

l 111

DC 84-36-3 SERVICE WATER RESERVOIR IMPROVEMENTS VALVE HOUSE MECH /ELEC AND FINAL STRUCTURAL, NA UNITS 1 & 2 DESCRIPTION The corrosive nature of the lake water at North Anna has resulted in degradation of the spray array support system. This . degradation has caused excessive maintenance of. the system and has required an extensive inspection program. Previously employed chemical treatments (prior to currently used sodium molybdate) designed to control the corrosive water have had an adverse effect on the reservoir clay liner. Several areas of concern need to be addressed and are listed below:

1. The aggressive water has caused corrosion of the existing carbon steel pipe supports. This ic a result of the previous chemical treatments being inadequate in inhibiting corrosion of the carbon steel components.

2. These previous chemical treatments have caused dispersion of the clay composing the select soil liner resulting in suspension and softening of the liner and increased turbidity. This interaction caused the treatments to be discontinued and also limits the types of chemical treatments which can be used.

3. The existing spray system design does not use the surface area cf the reservoir to its full potential in providing heat rejection under design meteorological conditions. While the spray system han been analyzed and found to be capable of rejecting design basis heat loads, significant improvements in the thermal performance of the se.vic e water reservoir and spray system appear possible for other design basis accident conditions and normal operation.

4. The ice build-up on the existing fiberglass piping and steel guy wires during the winter months has necessitated a great deal of maintenance work.

5. The fiberglass piping has not proven to be durable. Repair work is required to keep the nozzle and risers in an operable condition.

This design Change will install the piping, mechanical equipment, electrical equipment and the final structural items in the Valve house. The installed piping will provide the connecting piping between the buried piping installed by DC 84-37 and the reservoir spray array piping installed by DC 84-31.

To resolve these problems a new reservoir spray and bypass system will be installed. The final design concept consists of a new spray system which covers the central and western portions of the reservoir. The partially constructed North Anna Unit 3 and 4 pump house will be completed and serve as a valve house for the spray and winter bypass headers.

112

DC 84-36-3

SUMMARY

OF SAFETY ANALYSIS This design change does not place the new service water system into operation; however the new design and equipment meet or exceed all of the original safety-related requirements of the existing service water system. The new service water system will merely use a new safety related flow path and spray arrays to accomplish the cooling process. The new spray array will be more efficient than the existing one and the new valve arrangement provides more flexibility. The new 480V MCC's that are energized in this design change are safety related, seismic Class 1.

The new service water design uses the same basic technology as the present design. It consists of the reservoir, pumps, piping (some existing and some new) and a new, more efficient spray array. The winter bypass is a new feature, but it will not be used during an accident. The same accident types will exist for the new system configuration. The new system will be as good as, or better than, the existing system in terms of preventing and/or mitigating these postulated accidents.

The margin of safety is increased with this new system since the new valve arrangement will minimize the effect of one valve failing closed, and the new spray array will be more efficient and more durable than the existing system, the margin of safety with the new system is not decreased when the system is operating in the winter bypass mode because the system automatically switches from bypass to spray upon receipt of a Safety Injection signal, 113

DC 84-53 REGULATORY CUIDE 1.97 DIERGENCY VENTILATION DAMPER POSITION INDICATION / NORTH ANNA / UNIT 1 DESCRIPTION The following modifications were made to provide position indication of emergency ventilation dampers and satisfy R.G. 1.97 requirements:

MOV's-HV-104-1/2 were supplied with spare limit switches in the MOV operator supplied by Rotork. Also, a set of limit switches were wired to indicator lights on the Vent. Panel in the Control Room. Position indication will be utilized from the spare Limit Switches and wired into the Validyne Data Acquisition System.

MOV's-HV-118-1/2 were supplied with limit switches in the MOV operator supplied by Rotork. Also, a sec of limit switches were wired to indicator lights on the Auxiliary Shutdown Panel in the Emergency Switchgear Room. Position indication will be utilized from the spare Limit Switches and wired into the Validyne Data Acquisition System.

A0D's-HV-128-1/2/3/4 were supplied without limit switches. Environmentally qualified limit switches will be mounted such that the original qualification and operation of the damper is not degraded. The limit switch (open and closed) will be wired into the Validyne Data Acquisition Systen for position indication.

A0D's-HV-107A-1/2/3/4 and 107B-1/2/3/4 have unqualified limit switches installed for position indication of the dampers. Environmentally qualified limit switches will be installed to provide the present position indication (open and closed) and to provide a signal to the Validyne Data Acquisition System.

The existing and new limit switches will provide signals to the nearest multiplexer system for CRT display in the Control Room ard Technical Support Center. The multiplexer system is being installed under NUREG-0696 (DC-81-S20A). All conduit is to be seismically restrained near safety-related equipment.

Conax qualified seal connectors will be used to environmentally seal equipment located in harsh environment areas.

l

SUMMARY

OF SAFETY ANALYSIS l The Emergency Ventilation Damper Position Indication modification is designed i consistent with codes and standards of the existing systems. This modification

( does not change the characteristics of any system in operation during reactor operation or during construction implementation of the modification.

i 114

DC 84-57 REGULATORY GUIDE 1.97 CONTAINMENT ISOLATION VALVE POSITION INDICATION / NORTH ANNA / UNIT 1 DESCRIPTION Regulatory Guide 1.97 Rev. 3 "Instrumentation for Light-Water-Cooled Nuclear Power Plant and Environs Conditions During and Following and Accident" requires that Containment Isolation Valve position indication be classified as a Regulatory Guide 1.97 Category 1 item. As such, seismically and environmentally qualified position indication to the requirements of IEEE 344-1975 and 10CFR 50.49 respectively must be provided to the Control Room Operator.

Item B-14 of the North Anna Regulatory Guide 1.97 review identified that 1-HCV-1936, the Accumulator Vent Line Control Valve, does not have adequate position indication to meet the Category 1 requirements of Regulatory Guide 1.97. Also, it is located inside containment below the submergence level which requizes that any limit switches installed on the valve be qualified for submergence.

In addition, at the 30% meeting for this modification it was requested by Station Engineering that a new Control Valve (1-TV-1204A) be installed in the Letdown Line inside of containment to perform the Containment Isolation function of 1-HCV-1200 A,B,and C.

The present containment isolation valve, 1-HCV-1936, will be relocated above the flood level in the pipe penetration area. This relocation is necessitated by the fact that there are no limit switches available that are qualified for submergence.

NAMCO EA180 limit switches which are environmentally and seismically qualified to IEEE-323-1974 and IEEE 344-1975 will be installed on 1-HCV-1936 to provide control room valve position indication. The addition of there limit switches requires that the main Control board be mod!fied to add open and closed indicating lights for 1-HCV-1936.

A new valve, 1-TV-1204A vill be installed inside containment in the Letdown Line between 1-HCV-1200A, B. & C and penetration 28. It will assume the Containment isolation function of 1-HCV-1200 A,B, & C. however, the Phase A containment isolation trip of 1-HCV-1200 A,B & C will be retained in order to prevent the lifting of the relief valve 1-RV-1203 upon a containment isolation signal.

115

DC 84-57

SUMMARY

OF SAFETY ANALYSIS The modifications to the Letdown Line and the Accumulator Vent Line are designed consistent with the Design Basis Criteria for Containment isolation Valves and the requirements of Regulatory Guide 1.97. The relocation of 1-HCV-1936 does not functionally change the operation of the Safety Injection System in any mode. The installation of 1-TV-1204A does not functionally change the operation of the Letdown System beyond requiring the opening and closing of 1-TV-1204A at the same time as 1-TV-1204B, its redundant valve outside of containment. 1-TV-1204A assumes the inside Containment Isolation function of 1-HCV-1200A, B, and C. however, these valves will still close on a Containment Isolation Signal. This modification does not affect any existing accident analyses.

116

~ _ . _

DC-84-69 AND 84-70 PRESSURIZER SAFETY VALVE LOOP SEAL INSULATION OVENS UNITS 1 AND 2 RESPECTIVELY D_KSCRIPTION The reactor coolant system safety valves must be qualified for design basis ,

transients and accidents. NUREG-0737 Section II.D.1 provfded qualification and documentation requirements and stateo in part that "valve quall'ication shall '

include qualification of associated control circuitry, piping and supports, as well as the valves themselves."

There are three safety valves provided to protect the reactor coolant system from over pressure. Each safety valve is connected to the steam space of the "

, pressurizer with 6-inch piping in a loop-seal arrangement to maintain subcooled water (rather than steam) at the valve seats.

If the reactor coolant pressure increases to the safety valve setpoint (2485 psig) all three valves will lift and system pressure will accelerate the water slugs through the valves and discharge piping to the pressurizer relief tank.

The original piping analyses did not consider the fluid transient loads associated with a water slug passing through the system. EPRI studies (Review of Pressurizer Safety Valve Performance as observed in the EPRI Safety and Relief Valve Test Program, dated June 1982 WCAP-10105 - Westinghouse Corporation) indicated significantly Higher loads than those postulated in 1  ;

earlier analyses.

Subsequent analyses of safety valve transients indicate unacceptable fluid transient loads on piping and supports downstream of the safety valves assuming ,

4 "cold" loop-seal water temperatures upstream of the valves. Two modifications l 1 are necessary to qualify the piping for the Design Basis Transient and accident conditions expected: (1) modify existing supports and provide additional (

supports on the discharge of the safety valves, and (2) maintain the loop seal l water temperature 400'F. A design change package (DC84-71) "Pressurizer Safety Valve Discharge Pipe Support Modification," has been prepared to address Item (1). This design change package addresses Item (2) and details a method 7 of maintaining the loop real water temperature greater than or equal to 400'F. l The proposed resolution is to install metal reflective thermal insulation boxes  !

! which will enclose the safety valve loop seal piping, exposing the loop to an uninsulated portion of the pressurizer. In alation boxes will then be j installed enclosing the exposed pressurizrc areas and the loop seal piping.  !

The insulation boxes will be designed to maintain the water in the loop seals j at an elevated temperature (greater than or equal to 400'F). l i

t 117 r

DC 84-69 AND 84-70

SUMMARY

OF SAFETY ANALYSIS The installation of insulation boxes around the pressurizer safety valve loop-seal piping will not affect the operation of the pressuriter or the safety valves. The function of the insulation boxes is to maintein the water in the loop-seal piping at an elevated temperature to minimite fluid transient forces on the valve discharge piping.

Original piping analyscs did not consider the fluid trrnsient load as,ociated with a water slug passing through the system. This is a requirement of NUREC-0737 II.D.I. The instsliation of the insulation boxes will reduee fluid transient loads to ensure the piping and eJpports are quclified for the design basis transients.

118

4 DC 84-76 MECHANICAL CLEANING OF SERVICE WATER PIPING LUBE OIL, CEAR AND SEAL COOLERS SUPPLY AND RETURN HEADERS / NORTH ANNA / UNIT 1 (COMMON)

DESCRIPTION The service water system piping at North Anna Power Station has undergone severe corrosion attack due to the aggressive nature of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting. Some small bore piping has required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and thereby avert or postpone any requirement for further pipe replacements, a corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate-based chemical that has been effective in significantly reducing the corrosiveness of the service water.

However, the treated service water cannot presently come in contact with the internal surface of a majority of the service water system piping due to the layer of corrosion product, silt and slime that has developed over the years of plant operation. In order for the corrosion inhibitor to be effective on service water piping, the internal surfaces must be cleaned to bare metal.

Until the piping is thoroughly cleaned, corrosion will continue to occur beneath the buildup in spite of the presence of corrosion inhibiting chemicals in the service water.

The desired pipe surface preparation will be accomplished by chemical and mechanical cleaning methods. The amount of corrosion product on the pipe internal surfaces makes it impractical to clean thn piping using chemical methods alone. In order to reduce the amount of chemicals required, reduce the time required for final chemical cleaning and insure the effectiveness of the chemical cleaning process, the piping will first be cleaned using a mechanical technique referred to as hydrolasing.

For mechanical cleaning purposes, the service water system is divided into manageable segments for isolation. The segment to be mechanically cleaned by this Design Change Procedure is:

10"-WS-20-151-Q3 from Line 24"-WS-95 to Valve 1-CC-272 10"-WS-22-151-Q3 from Line 24"-WS-101 to Valve 1-CC-309 4"-WS-46-151-Q3 from Removable Spool to Line 10"-WS-20 4"-WS-56-151-Q3 from Removable Spool to Line 10"-WS-22 4"-WS-C52-151-Q3 from Removable Spool to Line 10"-WS-22 3"-LW-185-151-Q3 from Y re 10"-WS-20 to Valve 1-LW-248 Se rvice water system valves MOV-SW-113B and MOV-SW-113A will be replaced to insure tight shut off.

The hydrolaser is a 'igh pressure, low G.P.M. water blaster; therefore, the

' amount of water generated is not significant enough to pose any problems.

119

-, -- , , _-..,-.-~-m - - - . , , - - ,

DC 84-76

SUMMARY

OF SAFFTY ANALYSIS Water supply will be maintained to the charging pump coolers by an al ernate supply in the event of failure of the on-line service water loop, and flooding of the auxiliary building basement is prevented by administrative control by locking closed valves 18 inches and greater. In the event of a LOCA, the off-line system will be placed in an emergency serviceable status by installing prefabricated spool pieces in place of valves removed for cleaning and repair.

This header could then be used as an emergency services water supply for critical equipment.

Implementation of this DCP results in temporary modifications to safety related equipment, but the system will be in its original configuration when the procedure is completed; therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.

The service water system consists of two (2) 100% capacity piping loops shared by both Units 1 and 2. The Technical Specifications (3/4.7.4) require that both be operable. If one is inoperable for a period greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then both units must begin to shut down. The construction period will require that one loop be removed from service in order to perform the necessary cleaning, but this will be performed within the prescribed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period.

Therefore, based on the Technical Specification criteria, no threat of increased probability of accident will occur.

120

DC 84-77-3 MECHANICAL CLEANING OF SERVICE WATER PIPING LUBE OIL, CEAR AND SEAL COOLERS SUPPLY AND RETURN HEADERS / NORTH ANNA / UNIT 1 (COMMON)

DESCRIPTION The service water system piping at North Anna Power Station has undergone severe corrosion attack due to the aggressive natore of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting. Some small bore piping has required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and thereby avert or postpone any requirement for further pipe replacements, a corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate-based chemical that has been effective in significantly reducing the corrosiveness of the service water.

However, the treated service water cannot presently come in contact with the internal surface of a majority of the service water system piping due to the layer of corrosion product, silt and slime that has deveJoped over the years of plant operation. In order for the corrosion inhibitor to be effective on service water piping, the internal surfaces must be cleaned bare metal. The short and long term consequences of not thoroughly cleaning the pipe internally will be that corrosion will continue to occur beneath the build-up in spite of the presence of corrosion inhibiting chemicals in the service water.

The desired pipe surface preparation will be accomplished by chemical and mechanical cleaning methods. The amount of corrosion product on the pipe internal surfaces makes it impractical to clean the piping using chemical methods alone. In order to reduce the amount of chemicals required, reduce the time required for final chemical cleaning and insure the effectiveness of the chemical cleaning process, the piping will first be clean d using a mechanical technique referred to as hydrolasing.

For mechanical cleaning purposes, the service water system is divided into manageable segments for isolation. The segment to be mechanically cleaned by this Design Change Package is:

10"-WS-21-151-Q3 from Line 24"-WS-96-151-Q3 to Line 10"-WS-20-151-Q3 10"-WS-23-151-Q3 from Line 24"-WS-102-151-Q3 to Line 10"-WS-22-151-Q3 4"-WS-47-151-Q3 from Removable Spool to Line 10"-WS-21-151-Q3 4"-WS-57-151-Q3 from Removable Spool to Line 10"-WS-23-151-Q3 4"-WS-C53-151-Q3 from Removable Spool to Line 10"-WS-23-151-Q3 Ten inch (10") butterfly valves MOV-SW-213A and B will be removed from the piping, and replaced with refurbished valves from the se rvice water project valve pool.

121

DC 84-77-3 Upon completion of this DCP, the system will be in its original configuration, with the exception of the addition of the corrosion monitoring probe.

The hydro laser is a high pressure, low G.P.M. water blaster; therefore, the amount of water generated is not significant enough to pose any problems.

A 2 inch branch connection with valve will be installed in order to facilitate a corrosion monitoring probe which will be added at a future date.

SUMMARY

OF SAFETY ANALYSIS Water supply will be maintained to the charging pump coolers by an alternate supply in the event of failure of the on-line service water loop, and flooding of the auxiliary building basement is prevented by administrative control through locking closed all isolation valves 18 inches and greater. In the event of a LOCA, the off-line system

• ill be placed in an emergency serviceable status by installing prefabricated spool pieces (if necessary) in place of valves removed for cleaning and repair. This header could then be used as an emergency service water supply for critical equipment.

The only modification created by this design change is the addition of 2" non-high energy piping. Non-high energy piping, as defined by UFSAR Appendix 3C.2.2.1, is piping in which the maximum operating pressure is less than 275 psig and the maximum operating temperature is less than 200'F. Failure of non-high energy piping has not been postulated in the Updated Final Safety Analysis Report. Failure of this small bore piping would not cause a loss of adequate service water flow to any safety related component.

The service wate.- system consists of two (2) 100% capacity piping loops shared by both Units 1 and 2 The Technical Specifications (3/4.7.4) require that both be operable. If one v. inoperable for a period greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then both units must begin to shut down. The construction period will require that one loop be removed from service in order to perform the necessary cleaning, but this will be performed within the prescribed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period.

Therefore, based on the Technical Specification criteria, no threat of increased probability of accident will occur.

Flow will be maintained to the charging pump lube oil, gear and seal coolers at all times through the on-line service water loop or the alternate supply via the fire protection system. The flow rate provided by the hase from the fire protection system is sufficient to provide water to the various coolers of four (4) charging pumps (2 per units) which is the minimum required by Technical Specifications 3/4.5.2.

122

t DC 84-84-3 PIPE PRESERVATION AUXILIARY SERVICE WATER SUPPLY I

AND RETURN / NORTH ANNA / UNIT 1 AND 2

~

DESCRIPTION The service water system piping at North Anna Power Station has undergone severe corrosion attack due to the aggressive nature of the water and the presence of sulfat. reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting. Some small bore piping has required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and thereby avert or postpone any requirement for further pipe replacements, a corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate based chemical that has been effective in significantly reducing-the corrosiveness of the service water.

However, the treated service water cannot presently come in contact with the i internal surface of a majority of the service water system piping due to the ,

layer of corrosion product, silt and slime that has developed over the years of plant operation. In order for the corrosion inhibitor to be effective on service water piping, the internal surfaces must be cleaned to bare metal. The short and long term consequences of not thoroughly cleaning the piping internally would be that corrosion will continue to occur beneath the buildup

1. in spite of the presence of corrosion inhibiting chemicals in the service i

! water, and piping through-wall leaks or restricted flow would develop. ,

, The desired pipe surface preparation will be accomplished by chemical and mechanical cleaning methods. The amount of corrosion product on the pipe  !

internal surfaces makes it impractical to clean the piping using chemical methods alone. In order to reduce the amount of chemicals and time required [

for final chemical cleaning, and to ensure the effectiveness of the chemical cle: Ling process, the piping will first be cleaned using a mechanical technique referred to as hydrolasing.

This design change covers the cleaning of portions of the Auxiliary Service Water piping.

Two butterfly valves must be repaired as part of this design change. Both valves are for 24 inch nominal piping and are identified as MOV-SW-215A and MOV-SW-220A.

The pipe cleaning will be accomplished by a hydrolasing process. The hydrolaser is a high pressure, low C.P.M. water blaster; therefore, the amount i of water generated is not significant enough to pose any problems.  :

1 123 i

s . _ _ _ _ _ _ _ - . _

DC 84-84-3

SUMMARY

OF SAFETY ANALYSIS Water supply will be maintained to the charging pump coolers by an alternate supply in the event of failure of the on-line service water loop, and flooding of the auxiliary and turbine building basements are prevented by administrative control by locking closed valves 18 inches and greater. In the event of a LOCA, the off-line system will be placed in a backup status which could be utilized to provide service water to critical equipment by installing blind flanges in place of valves removed for cleaning and repair.

Implementation of this DCP results in temporary modifications to safety related equipment, but the system will be in it original configuration, other than the implementation of stainless butterfly valve disc, when the procedure is completed.

The service water system consists of two (2) 100% capacity piping loops shared by both Units 1 and 2. the Technical Specification (3/4/7.4) requires that both be operable. If one is inoperable for a period greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then both units must be shut down. The construction period will require that one loop be removed from service in order to perform the necessary cleaning, but this will be performed within the prescribed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period.

Therefore, based on the Technical Specification criteria, no threat of increased probability of accident will occur.

In addition, contingency plans have been developed to:

a. Provide backup water supply to the charging pump lube oil, gear and seal coolers in the event of loss of the on-line header; and

b. Return the isolated service water header to operation n the event of a LOCA

c. Prevention of flooding of critical equipment.

124

DC 84-85-3 SERVICE WATER PIPE PRESERVATION COMPONENT COOLING HEAT EICHANGERS SUPPLY AND RETURN HEADERS / NORTH ANNA / UNITS 1 AND 2 i

DESCRIPTION r The service water system piping at North Anna Power Station has undergone severe corrosive attack due to the aggressive nature of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general  ;

wall thickness reduction and wide spread pitting. Some small bore piping has required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and thereby avert <' postpone any requirement for further pipe replacements, a corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate based chemical that has been effective in significantly reducing the corrosiveness of the service water.

However, the treated service water cannot presently come in contact with the internal surface of a majority of the service water system piping due to the layer of corrosion product, silt and slime that has developed over the years of plant operation. In order for the corrosion inhibitor to be effective on service water piping, the internal surfaces must be cleaned to bare metal. The short and long term consequences of not thoroughly cleaning the piping internally will be that corrosion will continue to occur beneath the buildup in spite of the presence of corrosion inhibiting chemicals in the seraice water.

The desired pipe surface preparation will be accomplished b; chemical and mechanical cleaning methods. The amount of corrosion product on the pipe internal surfaces makes it impractical to clean the piping using chemical methods alone. In order to reduce the amount of chemicair and time required for final chemical cleaning, and to ensure the effectiveness of the chemical cleaning process, the piping will first be cleaned using a mechanical technique referred to as hydrolasing.

For mechanical cleaning purposes the service water system is divided into manageable segments for isolation. The segment to be mechanically cleaned by this Design Change Package consists of portions of the Component Cooling Heat Exchanger supply and return headers for Units 1 and 2.

Service water system valves 1-SW-MOV-108A, -108B, -208A, -2088, 1-SW-247 and 1-SW-250 will be replaced with refurbished valves form the service water valve pool to ensure tight shut off capability.

The hydrolaser is a high pressure, low G.P.M water blaster; therefore, the amount of water generated is not enough to pose any significant problems.

While portions of the Service Water System are drained during these procedures, 2 inch branch connections with valves vill be installed in order to facilitate a corrosion monitoring system which will be added at a future date.

125

DC 84-85-3 St20tARY OF SAFETY ANALYSIS This modification does not permanently modify any safety-related equipment with the exception of the addition of the four (4) 2 inch corrosion monitor branches, nor reduce the margin of safety of the systems as defined in the Technical Specifications, no un-reviewed safety question is created.

Water supply will be maintained to the charging pump coolers by an alternate supply in the event of failure of the on-line service water loop, and flooding of the auxiliary building basement is prevented by. administrative control through locking closed all isolation valves 18 inches and greater and temporarily sealing the charging pump cubicles to a flood level of 44 inches-

-above the Auxiliary Building Basement floor. In the event of a LOCA, the off-line system will be placed in an emergency serviceable status by installing prefabricated spool pieces (if necessary) in place of valves removed for cleaning and repair. This header could then be used as an emergency service water supply for critical equipment.

Implementation of this DCP results in modifications to safety related equipment, but the system will be in its original configuration with the addition of four (4) 2 inch corrosion monitor branches and stainless steel discs in place of cast steel in refurbished butterfly valves when the procedure is completed; therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.

The service water systes consists of two (2) 100% capacity piping loops shared by both Units 1 and 2. The Technical Specifications (3/4.7.4) require that both be operable. If one is inoperable for longer than a specific time period, then both units must begin to shut down. The construction period will require that one loop be removed from service at a time in order to perform the necessary cleaning, but this will be performed within the prescribed time period. Therefore, based on the Technical Specification criteria, no threat of increased probability of accident will occur.

J

j. 126

DC 84-86-3 PIPE PRESERVATION COMPONENT COOLING HK AND PIPING BE'lVEEN ISOLATION VALVES / NORTH ANNA / UNITS 1 AND 2 DESCRIPTION The service water system piping at North Anna Power Station has undergone severe corrosive attack due to the aggressive nature of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting. Some small bore piping has required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and thereby avert or postpone any requirement for further pipe replacements, a corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The

. corrosion inhibitor is a molybdate based chemical that has been effective in significantly reducing the corrosiveness of the service water. However, the treated service water cannot presently come in contact with Lhe internal surface of a majority of the service water system piping due to the layer of corrosion product, silt and slime that has developed over the years of plant operation. In order for the corrosion inhibitor to be effective on service water piping, the internal surfaces must be cleaned to bare metal. The short and long term consequences of not thoroughly cleaning the piping internally will be that corrosion will continue to occur beneath the buildup in spite of the presence of corrosion inhibiting chemicals in the service water.

The desired pipe surfece preparation will be accomplished by chemical and mechanical cleaning methods. -The amount of corrosion product on the pipe internal surfaces makes it impractical to clean the piping using chemical methods alone. In order to reduce the amount of chemicals and time required for final chemical cleaning, and to ensure the effectiveness of the chemical cleaning process, the piping will first be cleaned using a mechanical technique referred to as hydrolasing. Two valve connections are to be installed to facilitate future installation of a corrosion monitoring station as a means to determine the effectiveness of the chemical treatment program in this section of piping.

This design change covers hydrolasing the internals of four Component Cooling Heat Exchanger (HX)'s exposed to service water corrosion and the service water piping between the HX's isolation valves. In addition, the isolation butterfly valves need repair. The cleaning necessary for each HX and piping, including repairing the isolating butterfly valves, requires removing from service the 24" redundant supply and return headers that feed the four HXs.

The hydrolaser is a high pressure, low G.P. M. (approximately 30 G.P.M.) water blaster; therefore, the amount of water generated is not significant enough to pose any problems.

127

DC 84-86-3

SUMMARY

OF SAFETY ANALYSIS Each DCP entering a Technical Specification Time Period will address the following:

a. Providing water supply and return to maintain service water to critical equipment, if the on-line loop fails;

b. Re-establishment of the off-line redundant supply and return line during a LOCA; and

c. l'revention of flooding of critical equipment.

The component cooling service water supply lines are isolated from the HXs by double isolation valves. Double isolation will be maintained during most of the completion of this design change by the installation of a blind flange in place of one of the isolation valves. Only one of the two component cooling service water supply / return lines will be out of service at any one time and it can be available within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to re-establish redundant cooling to the component cooling system which prevents overheating of critical equipment.

The flooding of the auxiliary building basement is prevented by administrative control by locking closed valves 18 inches and greater. In the . vent of losing the on line piping or a LOCA, the off-line system will be re-established to provide service water to critical equipment by installing blind flanges in place of valves removed for cleaning and repair.

The only modification created by this design change is the addition of two small bore branches with valve. Failure of small bore piping has already been evaluated in the Updated Final Safety Analysir Report. Failure of the small bore piping would not cause a loss of adequate service water flow to any safety related component.

The service water system consists of two (2) 100% capacity piping loops shared by both Units 1 and 2. The Technical Specifications (3/4.7.4) require that both be operable. If one is inoperable for a period greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then both units must begin to shut down. The construction period will require that one loop be removed from service in order to perform the necessary cleaning, but this will be performed within the prescribed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period.

Therefore, based on the Technical Specification criteria, no increased probability of accident will occur.

128

h DC 84-91-2 [

PIPE PRESERVATION - SERVICE WATER SUPPLY AND RE'tIIRN [

FOR RECIRC. SPRAY HI/ NORTH ANNA / UNIT 2 ,

DESCRIPTION I The Service water system piping at North Anrta Power Station las undergone severe corrosive attack due to the aggressive nature of the water and the presence of sulfate reducing bacteria. The corrosien rt in the fora of general i wall thickness reduction and wide spread pitting. Some email % ore pipipg has I

required replacement due to through wall leaks or rearrictli flow; In order to preserve the integrity of the service wat9t f system piping, and

~

thereby avert or postpone any requirement for lirther gipe replacewents, a  ;

corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate based cht\1 cal . hat has been eftactive in significantly reducing the corrosiveness of ;he ser ice'veter, n

r However, the treated service water cann0t presti.'tif come in centact with the internal surface of a majority of the service water systds piping due to tie .

layer of corrosion product, silt and s114a trat has devel; ped over the years of ,

plant operation. In order for the Sorrnston 1stbibiter to be effect.ve on i service water piping, the internal rforfaces aust be ,'eaned r6 bare metal. The short and long term consequences of hot thoruughly nieani.13 the piping ,

internally will be that corrosion will continue ti occur beneaty (Se lealdup in  :

spite of the presence of corrosion inhibitinq chtmlerlr 11 Jhs service water. -

The desired pipe surface preparation will be sceoxplishtd by chemical and mechanical cleaning methods. Aie occune of cry'tosion p*odact on the pipe internal surfaces makes it impractical to clean th4 pr?ing using chemical i methods alone. In order to reduce the amount cf chemicals and time required -

for final chemical cleaning, and to errare ths effectivenett of the chemical cleaning process, the piping will first be cleaned using s meththical technique ,

referred to as hydrolasing. (

The service water system is divided into manageable segments fer ut,thanical cleaning purposes. the segment to be cleaned under this DCY includes portions  ;

of the service water supply and return piping for the Unit 2 leci culation Spray Heat Exchangers. l l

Sixteen butterfly valves are removed as part of this desigt cangs in order to  ;

be repaired. i Two check valves must be replaced as part of this design change. Both valves are for 24 inch nominal piping. i The pipe will be mechanically cleaned by a hydrolasing process. Tne hydrolast-  !

is a high pressure, low flow (30 C.P.M.) water blaster, shere the water is  ;

supplied form the Lake Anna Reservoiri therefore, the amount of wJ sr generated  ;

is not significant enough to pose any problems. j f

129

DC 84-91-2

SUMMARY

OF SAFETY ANALYSIS Implementation of this DCP results in temporary modifications to safety related equipment; but the system will be in its original configuration, other than the implementation of stainless butterfly valve disc and check valve replacement, when the procedure is completed. During the implementation period, a temporary water supply will be maintained to the charging pump coolers by an alternate supply in the event of f ailure of the on-line service water loop, and flooding of critical equipment has been prevented by locking closed all isolation valves 18 inches and greater and sealing the charging pump cubicles to a flood level of 44 inches above the auxiliary basement floor. In the event of a LOCA, the off-line system will be placed in a backup status which could be utilized to provide service water to critical equipment by installing removed valves or by installing temporary blind flanges in place of the valves removed for cleaning.

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

The service water system consists of two (2) 100% capacity piping loops shared by both Units 1 and 2. The Technical Specification 3/4.7.4 requires that both be operable. If one is inoperable for a period greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then both units must begin to shut corn. The construction period will require that each loop be remos d form service at separate times in order to perform the necessary cleaning, but each of these isolations will be performed within the preJeribed 72 hour time period. Therefore, based on the Technical Specification criteria, the margin of safety is not reduced.

J 130 a- ,

DC 84-99-1 PIPE PRESERVATION OF MAIN HEADER FIPINC FOR THE RECIRC.

SPRAY HEAT EXCHANGERS / NORTH ANNA / UNIT NO. 1 DESCRIPTION The service water system piping at North Anna Power Station has undergone

~

severe corrosive attack due to the aggressive nature of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general

vall thickness reduction and wide spread pitting. Some small bore piping has a required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and I

thereby avert or postpone any requirement for further pipe replacements, a

corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate based chemical that has been effective in significantly reducing the corrosiveness of the service water.

However, the treated service water cannot presently come in contact with the i internal surface of a majority of the service water system piping due to the layer of corrosion product, silt and slime that has developed over the years of plant operation. In order for the corrosion inhibitor to be effective on 1 service water piping, the internal surfaces must be cleaned to bare metal. The j short and long tere consequences of not thoroughly cleaning the piping internally will be taat corrosion will continue to occur beneath the buildup in spite of the presence of corrosion inhibiting chemicals in the service water.

The desired pipe surface preparation will be accomplished by chemical and mechanical cleaning methods. The amount of corrosion product on the pipe internal surfaces makes it impractical to clean the piping using chemical i methods alone. In order to reduce the amount of chemicals and time required 1 for final chemical cleaning, and to ensure the effectiveness of the chemical l cleaning process, the piping will first be cleaned using a mechanical technique l referred to as hydrolasing.

1 The service water system is divided into manageable segments for mechanical cleaning purposes. Four segments are to be cleaned under this DCP, which

! consist of a portion of the 24" main header supply and return pipe lines for j the U-l Recirculation Spray Heat Exchangers. In addition to cleaning, lines

! 18"-WS-326-151-Q3 and 18"-WS-324-151-Q3 vill have a corrosion monitoring I

station installed in their manway covers, and line 24"-WS-26-151-Q3 vill have a second station installed in its run piping. Each of the corrosion monitoring stations consist of two 2" branch connections, one to accommodate the future installation of a probe and the other a test coupon.

The pipe will be mechanically cleaned by a hydrolasing process, which utilizes high pressure water blasts to dislodge the corrosion product adhering to the inside pipe surface.

131 E

DC 84-99-1

SUMMARY

OF SAFETY ANALYSIS Since this procedure does not permanently modify any safety related equipment, nor reduce the margin of safety of the systems defined in the Technical Specifications, no unreviewed safety question is created. Time Periods are required to implement the overall system cleaning plan. Accordingly, this DCP provides contingencies for the following:

1. Providing an alternate water supply to maintain service water to critical equipment if the on-line loop fails;

2. Re-establishment of the off-line redundant supply during a LOCA; and

3. Prevention of flooding of critical equipment.

The flooding of the Auxiliary and Turbine Building basements is prevented by administrative controls and by locking closed valves equal to or greater than 18 inches. In addition, the charging pump cubicles will be temporarily scaled to a level of 44 inches above the floor to protect critical equipment. This added precaution would allow sufficient time for taking the necessary measures to terminate flooding, if the unlikely flooding of the Auxiliary Building basement does occur. Implementation of this DCP results in temporary modifications to safety related equipment; but the system will be in it original configuration when the procedure is completed. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.

The only modification created by this design change is the addition of small bore piping.

132

DC 84-100-1 PIPE PRESERVATION - SERVICE WATER SUPPLY AND RE'11JRN FOR RECIRC. SPRAY HX / NORTH ANNA / UNIT 1 DESCRIPTION The service water system piping at North Anna Power Station has undergone severe corrosive attack due to the aggressive nature of the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting. Some small bore piping has required replacement due to through wall leaks or restricted flow.

In order to preserve the integrity of the service water system piping, and thereby avert or postpone any requirement for further pipe replacements, a corrosion inhibitor chemical treatment program has been initiated. Corrosion inhibitor chemicals were added to the reservoir on July 13, 1984. The corrosion inhibitor is a molybdate based chemical-that has been effective in significantly reducing the cor'.osiveness of the service water.

However, the treated service water cannot presently come in contact with the internal surface of a majority of the service water system piping due to the layer of corrosion product, silt and slime that has developed over the years of plant operation. In order for the corrosion inhibitor to be effective on service water piping, the internal su :es must be cleaned to bare metal. The short and long term consequences of not thoroughly cleaning the piping internally will be that corrosion will continue to occur beneath the buildup in spite of the presence of corrosion inhibiting chemicals in the service water.

The desired pipe surface preparation will be accomplished by chemical and mechanical cleaning methods. The amount of corrosion product on the pipe internal surfaces makes it impractical to clean the piping using chemical methods alone. In order to reduce the amount of chemicals and time required for final chemical cleaning, and to ensure the effectiveness of the chemical cleaning process, the piping will first be cleaned using a mechanical technique referred to as hydrolasing.

The service water system is divided into manageable segments for mechanical cleaning purposes. The segment to be cleaned under this DCP includes portions of the service water supply and return piping for the Unit 1 Recirculation Spray Heat Exchangers.

Sixteen butterfly valve are removed as part of this design change in order to be repaired.

Two check valves must be replaced as part of this design change. Both valves are for 24 inch nominal piping.

The pipe will be mechanically cleaned by a hydrolasir.g process. The hydrolaser is a high pressure, low flow (30 G.P.M.) water blaster, where the water is supplied from the Lake Anna Reservoir; therefore, the amount of water generated is not significant enough to pose any problems.

133

DC 84-100-1

SUMMARY

OF SAFETY ANALYSIS Implementation of this DCP results in temporary modifications to safety related equipment; but the system will be in its original configuration, other than the implementation of stainless butterfly valve disc and check valve replacement, when the procedure is completed. During this temporary period water supply will be maintained to the charging pump coolers by an alternate supply in the event of failure of the on-line service water loop, and flooding of critical equipment has been prevented by locking closed all isolation valves 18 inches and greater and scaling the charging pump cubicles to a flood level of 44 inches above the auxiliary basement floor. In the event of a LOCA, the off-line system will be placed in a backup status which could be utilized to provide service water to critical equipment by installing removed valves or by installing temporary blind flanges in place of the valves removed for cleaning.

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

The service water system consists of two (2) 100% capacity piping loops shared by both Units 1 and 2. The Technical Specification (3/4.7.4) requires that both be operable. If one is inoperable for a period greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then both units must begin to shut down. The construction period will require that each loop be removed from device at separate times in order to perform the necessary cleaning, but each of these isolation will be performed within the prescribed 72 hour time period. Therefore, based on the Technical Specification criteria, no increased probability of accident will occur.

In addition, contingency plans have been developed to:

a. Provide backup water supply to the charging pump lube oil, gear and seal coolers in the event of loss of the on-line header, and

b. Return the isolated service water header to operation in the event of a LOCA.

c. Prevention of flooding of critical equipment.

A 134

r l

DC 85-05 a 85-06  ;

REACTOR VESSEL HEAD SHIELDING UNITS 1 AND 2 RESPECTIVELY DESCRIPTION At each refueling of the North Anna Power Station, many operations are performed in the reactor vessel head area, which is a relatively high radiation '

field (200 mr/hr or higher).

In order to reduce the exposure from these operations, permanent reactor vessel 4 head shielding was installed. This shield would protect the maintenance l personnel from all major sources of gamma radiation originating at the reactor vessel head and its appendages. It does not interfere with the head flange work, thermocouple column work, or any other regular maintenance on the head.

SUMMARY

OF SAFETY ANALYSIS f

The head shielding is installed to protect station personnel during construction and maintenance activities in the reactor cavity and does not i affect operation of the Reactor Coolant System or any other system. Also the i head shielding is seismically supported in order to protect safety related i equipment in the area during a postulated seismic event. This design change f does not change the basis of any Technical Specifications.

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q I 135

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DC 85-07 AND 85-08 INADEQUATE CORE COOLING MONITOR SYSTM UPCRADE UNITS 1 & 2 RESPECTIVELY DESCRIPTION To satisfy R.G. 1.97, Revision 3 and NUREG-0737 Item II.F.2 (including ), the current non-redundant non-Class IE incore (core exit) thermocouple system will be upgraded to a redundant Class IE Core Exit Thermocouple (CET) system. In addition to satisfying the above requirements, the existing redundant RVLIS and Core Cooling Monitor (CCM) systems will be integrated with the upgraded CET system into a common system called the Inadequate Core Cooling Monitor (ICCM) system. The existing electronic and display equipment for the RVLIS, CCM, and CET systems will be removed and replaced with a single redundant ICCM electronics and display system. The new ICCM electronic system will receive all redundant Class 1E CCM and CET field inputs as direct digital data from the station Multiplexer system. The redundant Class 1E RVLIS inputs will be hardwired inputs. Reactor Coolant Pump (RCP) running status contacts are non-safety related and will be hardwired to the new ICCM cabinet through built-in isolators.

To satisfy the requirements of NUREG-0696, the emergency response facilities (ERF) computer will receive 50 top mounted Type K thermocouples inputs via the same multiplexer system used to input the thermocouple inputs to the ICCM system.

SUMARY OF SAFETY ANALYSIS This modification upgrades the incore thermocouple and core cooling monitoring systems to Class IE systems. The new safety related top mounted incore thermocouples ant cabling system replaces the existing top mounted non-safety related thermocouple and cabling system.

This modification provides r,afety related information to the control room operators on approach to an inadequate core cooling condition in a way which is easier to interpret that that currently provided.

This modification will ensure that an approach to an inadequate core cooling condition can be detected.

136

DC 85-24 AND 85-25 CONTAlleENT WITROCEN SUPPLY NEADER VALVE ADDITION UNITS 1 4 2 RESPECTIVELY DESCRIPTION The containment nitrogen supply header provides high pressure nitrogen gas for use in the safety injection (SI) accumulators, the pressurizer relief tank and the pressurizer power operated relief valves (PORVs).

Periodic charging of the gas storage system is required to maintain the PORV nitrogen reserve tanks pressurized above 1950 psig. System pressure surges during charging' of the storage bottles result in lifting of a relief valve (RV-1894) in the nitrogen supply line to the SI accumulators. Lifting of the relief valve discharges nitrogen into the containment and makes it difficult to charge the system to 2000 psig as required for the PORV reserve tanks.

This modification installs a remote manual isolation valve upstream of the pressure control valve (PCV) supplying nitrogen to SI accumulators. This remote manual isolation valve will be normally open but could be shut from the Control Room prior to charging the nitrogen storage system and PORV reserve tanks. Shutting of this isolation valve while charging the system would isolate SI-PCV-1846 and SI-RV-1894 (Unit #1) or SI-PCV-2846 and SI-RV-2894 (Unit #2) (in supply line to SI accumulators) from pressure surges which cause the relief valve to lift. The isolation valve would be returned to the normally open position after system charging.

Variable valve position control will be provided to allow the operator to minimize system pressure surges when opening the valve.

SU)StARY OF SAFETY ANALYSIS The nitrogen supply piping is safety related since it is connected to the pressure boundary of components (accumulators, PORV nitrogen reserve tanks) required to function to mitigate an accident. The supply piping is classified as Category I - Seismic, and the new isolation valve to be installed by this design change will meet Category I seismic criteria. The isolation valve will "Fail Safe" in the open position to assure recharging if the Safety Inj ec tion accumulators if required.

The3 nitrogen supply piping is used during normal operation to recharge the Safety Injection Accumulators Pressurizer PORY nitrogen reserve tanks and to supply a nitrogen blanket for the Pressurizer Relief Tank (PRT). The supply i header is not required (other than pressure bouadary) during or after an I accident. The new valve to be installed in the supply piping will be Category l I Seismic.

137 i

i

DC 85-24 and 85-25 Nitrogen pressure supplied by the nitrogen supply header is required to ensure the operability of the Safety Injeciton Accumulators and the Pressurizer PORV nitrogen reserve tanks. Low pressure alarms are provided to ensure nitrogen pressure is sufficient. There are no Technical Specification requirements for the nitrogen supply header availability or operability. Installation of the new remote manual isolation valve will allow operators to recharge the Pressurizer PORY nitrogen reserve tanks without lifting the relief valve in the supply to the SI accumulators.

t 138 i

DC 85-29 AND 85-30 REPLACDGDfT OF STATION BATTERIES UNITS 1 & 2 RESPECTIVELY DFSCRIPTION The existing 125V DC Station Batteries (excepting 1-IV) are approaching the end of service life and require replacement. The new calculations indicate that larger capacity batteries are required to provide an adequate safety margin.

This modification removes existing batteries, battery racks inter-rack cabling and conduit from the battery rooms. It installs the new batteries (with a larger capacity), battery racks (2 tier), inter-rack cabling and conduit in the battery rooms. The new batteries shall be Exide "2GN23" and consist of thirty 2 cell units per battery room.

SIDE (ARY OF SAFETY ANALYSIS These design changes remove the exir, ting Station Batteries from the battery rooms and replaces them with new and larger capacity batteries which will coordinate with the balance of the present 125V DC system. The new batteries will be designed, fabricated, and installed to meet or exceed the requirement sections of the original Design Basis Document.

The materials used in the modifications are totally compatible (relative to design, quality, and functional requirements) with original equipment.

Therefore, the possibility for an accident or malfunction of a different type than previously evaluated does not exist.

139

DC 85-31 AND 85-32 CONTAlletENT INSTRUMENT WELL AND MAT SUMP PUMP IMPROVEMENTS UNITS 1 & 2 RESPECTIVELY DESCRIPTION Because of the poor reliability of the air driven pumps in the containment instrument well, due to calcium scaling, it is proposed that a chemical addition system be added to the containment mat sump pumps.

A chemical storage area will be installed next to the instrument well shaft with its cover slightly above grade level A chemical storage tank, metering pump and the existing air station will be located in the ehemical storage area.

To solve the liquid waste conformance problem, the new sump pumps will diecharge from the containment mat sump to the clarifier holdup tank in the waste disposal building. Mixing with other water through the clarifier, the drain water from the containment eat sumps would be diluted and neutralized such that the clarifier fluid will be within discharge limits.

SUletARY OF SAFETY ANALYSIS This modification does not contain equipment important to safety that has been previously analyzed. The installation work and operation of the containment mat sump pump will not affect equipment important to safety.

The piping modification involves small low pressure non radioactive water lines running from a sump located outside to the waste disposal building. On this basis, no new accident situation is created.

140

DC 85-34-3

, SERVICE WATER RESERVOIR IMPROVENENTS TIE-IN VAULT - STRUC11JRAL / NORTH ANNA / UNITS 1 & 2 1

DESCRIPTION l

I The corrosive nature of the lake water at North Anna resulted in degradation of i the spray array support system. This degradation has caused- excessive maintenance of the system and has required an extensive inspection program.

Chemical treatments designed to control the corrosive water have had an adverse effect on the reservoir clay liner. Several areas of concern need to be

! addressed and are listed below:

, 1. The aggressive water has caused corrosion of the existing carbon steel l pipe supports. This is a result of the previous chemical treatments j being inadequate in inhibiting corrosion of the carbon steel components.

! 2. These previous chemical treatments have tended to suspend the upper i layers of the compacted clay liner. This interaction caused the j treatments to be discontinued and also limits the types of chemical 1 treatments which can be used.

3. The existing spray system design does not use the surface area of the i reservoir to its full potential in providing heat rejection under i design meteorological conditions. While the spray system has been analyzed and found to be capable of rejecting design basis heat loads.

I significant improvements in the thermal performance of the service i

water reservoir and spray system appear possible for both design basis accident conditions and normal operation.

4. The ice build up on the existing fiberglass piping and steel guy wires during the winter months has neceseitated a great deal of maintenance work.

I l 5. The fiberglass piping has not proven to be durable. Repair work is required to keep the nozzle and risers in an operable condition.

i To resolve this problem, a new reservoir spray and bypass system will be 1

installed. The final design concept consists of a new spray system which covers the central and western portions of the reservoir. The partially 1 constructed North Anna Units 3 and 4 pump house will be completed and serve as a valve house for the spray and winter bypass headers.

I l The various stages of this project will be accomplished by seven (7) design j changes and several Engineering Work Requests.

This design change (DC-85-34-3) includes the construction of a tie-in vault. 31 ft. x 30 ft. x 27 ft. high, around the existing .erating service water lines.

This reinforced concrete structure will be constructed in such a manner to provide continuous protection of these service water lines against tornado missiles.

141

DC 85-34-3 This design change (DC-85-34-3) includes the installation of an electrical distribution system in the tie-in vault, electrical ducts and a manhole to provide a means of routing cables to and from the tie-in vault.

SUMMARY

OF SAFETY ANALYSIS The new design and equipment meets or exceeds all of the original safety related requireeents of the existing service water system. The tie-in vault will provide continuous missile protection to the service water piping during construction.

The argin of safety is not reduced since the new piping added by this design change serves the same purpose as the existing underground piping and does not reduce any safety margins.

142

DC 85-37 AND 85-38 CIRCULATING WATER SYSTEM MODIFICATIONS UNITS 1 & 2 RESPECTIVELY DESCRIPTION The North Anna Units 1 and 2 condensers have experienced many failures including tube degradation and tube-to-tubesheet joint leakage requiring frequent waterbox entries to locate leaks and plug tubes. A Virginia Power Condenser Task Force noted that the incidence of condenser failures (tube leaks) was much greater during vinter months. The Task Force concluded that high condenser vacuum resulting from cold circulating water temperatures (during winter months), was causing tube-to-tube vibration damage from high exhaust steam velocities.

At the request of Virginia Power, Stone & Webster Engineering Corporation performed a study which addressed the following methods of raising condenser pressure during winter months to prevent high vacuum and reduce the potential for tube failure due to excessive tube vibration:

Reducing the number of operating circulation water pumps Taking one of the four condenser waterboxes out of service Lowering the water levels in the condenser waterboxes Recirculating water from the condenser outlet to the condenser inlet, and Bypassing cold water around the condensers.

As noted in the Condenser Refurbishment study, SWEC recommended reducing the number of operating circulating water pumps and throttling of pump discharge valves to raise condenser pressure during winter months.

Based on a review of the circulating water system control logics, the following modifications were performed:

Install two pushbuttons (throttle, open) and an indicator for each discharge valve on the intake structure panel in the Control Room.

Modify the circulating water pump discharge valve control circuit to allow throttling to the 45 degree open position.

Modify the turbine trip steam dump pe rmissives to allow two circulating ' ater pump operation.

Adjust the No. 3 rotor to the 45 degree open position and add a potentiometer to each discharge valve.

Remove existing red and greea indicating lights for discharge valve position indication on the intake structure panel and replace with red and green pushbuttons/ lights.

143

DC 85-37 AND 85-38 St4 MARY OF SAFETY ANALYSIS This modification allows the circulating water pump discharge valves to be throttled from full open to 45 degree open and revises the steam dump permissive to require only two ,f four circulating water pumps operating. The constructional and post construction phases of this modification does not impact the operation of existing safety related equipeent or systems.

The constructional and post construction phases of this modification does not affect the availability of safety related systems.

This modification will al'iew operators to remote manually throttle circulating water pump discharge satisfied with only two of four circulating water pumps operating. The circulating water system is non-safety related and is not a Technical Specification item.

144

1 o

DC 85-42-3 l 4

PRIMARY GRADE WATER STORACE TAIEC MODIFICATION / NORTM ANNA /  ;

3 UNITS 1 & 2  !

?

I i 4 D'4SCRIPTION h i

J A survey of the primary grade water storage tank was performed in 1980 and it  !

! was observed that the aluminum flotation devices attached to the floating deck  !

had corroded and portions of the buoyancy material had become water logged. I

{ Although continued deterioration of the aluminum components will most probably (

not cause technical specifications to be exceeded, the constant egress of '

l aluminum into the primary grade water system could lead to aluminum deposits in  !

the primary system.  !

l A economic review of the cost for a flexible membrane, a new metal deck, and l

4 repairing the old deck was made and it was concluded that the deck should 'u- i

! repaired and that the following modifications would be madet  !

) I l A. Replace the aluminum pontoons filled with polyurethane eith stainless  ;

steel tubular floats.  ;

i; ~ B. Replace the peripheral seal with a new design seal which cannot become

] water logged. .

i  !

r C. Replace the aluminum legs with new stainless steel legs.  !

i D. Add a stainless steel ring skirt which will extend into the water to j minimize contact between the air which exists above the floating deck and the main water surface.

I i S1304ARY OF SAFETY ANALYSIS l l This modification does not contain or affect equipment important to safety that '

has been previously analyzed. The repair work and operation of the primary l grade water storage tank floating decks will not affect equipment important to  !

safety. I 1 Since this modification involves no equipment important to safety and does not increase potential for radioactive release, no technical specification limits i l

will be exceeded either during construction or operational phases of the  :

l modification. (

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145 j

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DC 85-47-2  ?

SAFECUARD VALVE FIT StBtF PUNF MODIFICATION  !

4 NORTM ANNA / UNIT 2 j

o i I

l plubCRIPTION f

t The safeguards valve pit sump is presently equipped with two electric sump l y pumps (2-DA-P-6A&B). Because of the severe operating problers and poor i reliability experienced with electric driven sump pumps, air operated sump i

, pumps will be used for the safeguards valve pit sump pumpt 2-DA-P-6A&B. These (

j pumps have proven extremely reliable in a field test for approximately. siy i

monthe and have, therefore, been selected for this service.

! t St20lARY OF SAPETY ANALYSIS j s  ;

!e 'ihis modification does not centain equipment important to safety that has beep -i

! previously analyzed. The construction vork and post construction oseration of  !

the safeguards valve pit sump pumps will not affect equipeent important to  !

l safety.

~

The piping modification involves small low pressure water lines ryneing from a i

, sump located outside the containment an this basis, no new accident situation  ;

} is created either during construction or operational phases of this }

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As this mcdification involves no equipment important to safe..y and has ont j increased potentis' for radioactive release, no technicul specification limits  !

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DC 85-49 AND 85-50 EMERGENCY BUS U.V. RELAY REPLACEMENT UNIT 1 AND 2 RESPECTIVELY DESCRIPTION Rochester instruments Systems (RIS) undervoltage relays type RP-2035 have exhibited a tendency to drift from the designated setpoint and/or an inability to maintain deadband adjustment. These relays are presently installed and are in service at North Anna Power Station in the emergency bus degraded voltage protection scheme. The degraded voltage relays are all located in the 4160V emergency switchgear (one set for "H" bus and one set for "J" bus) in the Emergency Switchgear Room. There are three single phase PR-2035 degraded voltage relays associated with each bus (4160V "H" and 4160V "J"). The existing PR-2035 degraded voltage relays also provide overvoltage monitoring for each 4160V emergency bus.

The existing three (3) single phase degraded voltage RIS type PR-2035 relays for each 4160V emergency bus, 1H and 1J, will be removed and replaced with General Electric (GE) three phase type SLV relays. The GE type SLV relay is a static voltsge type with a drop-out to pick-up ratio of 99 percent which is

necessary to meet North Anna's degraded voltage protection setpoint requirement of 90 2 1 percent of nomimal (4160) volts.

The SLV relays are physically larger than the PR-2035 relays and there is not sufficient space on the instrument compartment doors of any emergency switchgear cubicles for installing the new relays. Therefore, the new degraded voltage SLV relays will be located on the rear door of the emergency switchgear Potential Transformer (PT) cubicles.

The GE type SLV relays are three phase relays consisting of three independent voltage level detectors and a common de power supply. Therefore, only one SLV relay is required to replace the existing three (3) single phase degraded voltage relays. However, to minimize the consequence of a failure of the common power supply, which affects all three Phases, and to allow for removal of a SLV relay for bench checking r.nd/or calibration, two relays will be installed in parallel, per 4160V emergency bus, for sensing degraded voltage, both degraded voltage SLV relays will be normally energized to establish a conservative "fail safe" logic design for inputs to the existing degraded voltage protection actuation scheme.

A separate 125V de voltage source is required to power SLV relays. This source is available internal to each 4160V emergency switchgear bus. This 125V de power source is the same source which energizes the auxiliary relays which perform the two out of three logic and initiate all subsequent loss of voltage and degraded voltage protection.

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o DC 85-49 AND 85-50

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SUte(ARY OF SAFETY' ANALYSIS The implementation of this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Final Safety Analysis Report.

This modification will not change the existing degraded voltage protection circuit actuatien logic, setpoints or function. It will, however, use two relays in parallel to sence a degraded voltage condition on any phase of each 4160V emergency bus. Therefore, this DCP will' change the input configuration for the existing degraded voltage protection actuat:lon circuit for each 4160V emergency bus.

The implementation of this modification does not .:reate a possibility for an accident or a malfunction of a different type than any evaluated previously in the Final Safety Analysis Report.

This modification will ensure the availability of safety related power systems as designed and currently analyzed in the UFSAR. This modification will not change the existing degraded voltage protection circuits logic, setpoints or functions.

This modification will also ensure that the margin of safety as defined in the Technical Specifications will be maintained.

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DC 86-04 AND 86-05 5th POINT FEEDWATER HEATER TUBE BUNDLE AND EXTERNAL DRAIN COOLER REPLACEMENT, UNITS 1 & 2 RESPECTIVELY DESCRIPTION The major concern is that the tubes in the 5th point feedwater heaters and the external drain coolers are made of copper alloy (Arsenical Admiralty). Present industry practice tends toward removal of copper alloys from the feedwater system in order to protect the steam generators and turbines. Thus, a replacement of the 5th point feedwater heater tube bundles and the external drain coolers is desired.

An additional benefit of the replacement will be the removal of original heater tube bundles and drain coolers, which have various levels of wear, tubes plugged and leakage.

Since the 5th point feedwater heaters are located in the condenser neck, they will not be totally replaced but will be rebundled with 304SS tubes. The tubeside relief valves will also be replaced as part of this modification.

The 5th point heater external drain cooler will be replaced complete with new drain coolers having ASME SA-688, TP 304 stainless steel tubes with a 0.05%

maximum carbon content. The tube-to-tubesheet joints will be rolled. Also, the tubeside and she11 side relief valves for the drain coolers will be replaced.

S_UMMARY OF SAFETY ANALYSIS The replacement tube bundles and drain coolers are essentially one for one replacements of an enhanced design and will not affect any of the operation or ability of equipment important to safety to perform their safery functions.

The tubes in both the replacement tube bundles and the new drain coolers will be 304 SS (the existing tubes are arsenical admiralty copper alloy).

Impingement plates, air vent manifold, air vent chamber, air vent piping, and spacers in the replacement tube bundles will be stainless steel in lieu of commercial steel, all of which contribute to an improved bundle design. These

, changes are designed consistent with the affected systems design basis and will I

be installed using existing design criteria. This modification will not change any of the affected systems operations and the Unit 1 5th point replacement feedwater heater tube bundles and enternal drain coolers will not and do not perform any safety function.

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DC 86-06-2 CONTAINMENT INSTRUMENT AIR ISOLATION / NORTH ANNA / UNIT 2 DESCRIPTION The containment isolation valves for the containment instrument air supply on Units 1 and 2 are closed on different containment isolation phase signals.

Unit 1 instrument air isolation valves TV-IA102A and TV-IA102B automatically close on a Phase B containment isolation signsl. Unit 2 instrument air isolation valves TV-IA202A and TV-IA202B automatically close on a Phase A containment isolation signal. NUREG-0737,Section II.E.4.2, states that containment instrument air is listed as an essential Level 2 system which will isolate on a Phase B signal.

The Unit 2 containment instrument air isolation valves TV-IA202A and TV-IA202B will be changed from the containment isolation Phase A signal to the containment isolation Phase B signal. This design change will provide consistent operation between Unit 1 and Unit 2 on a containment isolation Phase B actuation, and puts Unit 2 in conformance with NUREG -0737.

SUMMARY

OF SAFETY ANALYSIS This design change will provide consistent operation between North Anna Units 1 and 2. Both units will isolate the containment instrument air isolation valves on a containment isolation Phase B signal which is required for compliance with NUREG-0737.

This modification will ensure the availability oi? instrument air on a Phase A isolation signal.

This design change will not add or delete any equipment. Therefore the margin of safety is not reduced.

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DC 86-07 AND 86-08 PRESSURIZER HEATER STATUS INDICATION, / UNITS 1 & 2 RESPECTIVELY DESCRIPTION Regulatory Guide (R.G.) 1.97, Revision 3, requires the monitoring of the following variable for the listed range and category:

Pressurizer Heater Status (for heaters fed from Emergency Power Sources)

Range 0 - 110% of rated power Category 2 Type D The above Type D variable is identified as Category 2 by R.G. 1.97. Category 2 requires equipment qualification in accordance with R.G. 1.89, Qualification of Class lE Equipment for Nuclear Power Plants. New equipment required for this modificaticn will be seismically installed.

Presently, total power consumption for pressurizer heaters on buses A, B, C, H, and J is provided to the plant computer. This modification will retain the existing total power indication (for heaters on buses A, B, C, H, and J) as well as provide individual power consumption indication for the heaters fed from the H and J buses. The instrumentation for the H and J buses is being provided so that the operator can avoid overloading an emergency power source.

To comply with R.G. 1.97 requirements, a pressurizer heater status system will be added. Pressurizer heater status will be monitored by adding Rochester Instrument Systems power (watt) transducers to the instrument compartments for 480 V Emergency Buses located in the Rod Control Room.

SUMMARY

OF SAFETY ANALYSIS This modification provides instrumentation to allow the operator to monitor the power consumpcion of the backup pressurizer heaters. The installation of the instrumentation within this design change does not impact the operation of existing safety related equipment or systems. The new instrumentation is environmentally and seismically qualified and provides independent indication to the operator to allow him to assess equipment operation. Consequently, the instrumentation installation does not affect safety equipment or the operation of safety systems.

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DC 86-11 AND 86-12 R.C. 1.97 ACCUMULATOR TANKS PRESSURE TRANSMITTER UPGRADE UNITS 1 & 2 RESPECTIVELY DESCRIPTION North Anna Power Station utilizes a Safety Injection System (SI) containing three (3) accumulator tanks per unit. The SI system is used for safe shutdown of the reactor during or after an emergency. The proper function of the SI system depends on the pressure in the accumulator tanks; hence, monitoring of this pressure is essential. To accomplish this task and to meet the requirements of R.G. 1.97, Rev. 3, qualified pressure transmitters are required to replace those presently in service.

To comply with R. G. 1.97 requirements, this Design Change Package will install six (6) new qualified pressure transmitters per unit above flood level to replace those non qualified transmitters which are now in service and mounted below flood level in the reactor containment.

The replacement transmitters wil?. be Rosemount Pressure Transmitters Model 1153GD8RA, Series D. They are seismically and environmentally qualified, and have been qualified for use at North Anna by QDR File No. 8.5.

SUMARY OF SAFETY ANALYSIS This modification provides environmentally qualified instrumentation to allow the operator to monitor the pressure of the Safety Injection Accumulator Tanks.

The replacement of the instrumentation does not impact the operation of existing safety related equipment or systems. The new instrumentation is environmentally and seismically qualified and provides indication to the operator to allow him to assess equipment operation. Consequently, the instrumentation installation does not affect safety equipment or the operation of safety systems.

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DC 86-13 AND 86-14 R.G. 1.97 - STEAM GENERATOR WIDE RANGE LEVEL INDICATION UPCRADE, UNITS 1 & 2 RESPECTIVELY DESCRIPTION Steam generators (S.G.) wide range level indication loops are presently on non-safety related circuits. R.G. 1.97 requires S.G. wide range level instrumentation loops be independent channels, electrically independent and physically. separated from each other and from equipment that is non-safety related. Therefore, these loops must be upgraded to meet the requirements of R.G. 1.97 for Type D, Category 1 variables.

The R.G. 1.97 S.G. wide range indication modification will upgrade the three loops on each unit from a non-safety to safety related bus. Each loop will be powered from a different safety related bus and separation will be maintained from the level transmitter, through safety related penetrations, to the safecy related process cabinets. Each S.G. wide range level indication loop will be powered from a different safety channel than the corresponding auxiliary feedwater flow indication loop. The existing non-safety loops output signals will be maintained as isolated non-safety output signals from the upgraded safety related loops and as such will not be upgraded.

SUMMARY

OF SAFETY ANALYSIS This modification provides environmentally qualified instrumentation to allow the operator to monitor the wide range steam generator level during and after and accident. The replacement of the cabling and installation of new loop power supply and isolation cards within this design change does not impact the operation of existing safety related equipment or systems. The new cabling and printed circuit cards are environmentally and seismically qualified and provide indication to the operator to allow him to assess equipment operation. During implementation of this package, wide range steam generator level indication will not be available to the operators.

New printed circuit cards are to be added to Westinghouse protection racks and implementation of a more reliable power supply for these loops will increase the margin of safety associated with these loops. During implementation of this package, wide range steam generator level indication will not be available to the operators.

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DC 86-16-1 SMALL BORE SNUBBER ELIMINATION / NORTH ANNA / UNIT 1 DESCRIPTION The Design Change Package has been developed to address the inherent problems created by the use, maintenance and testing of snubbers. Because of the potential problems associated with a snubber's proper operation, the high cost associated with snubber maintenance and testing, and the significant ALARA problems associated with the maintenance, inspection, and testing of the snubbers, there are significant benefits to be gained by reducing the snubber population to the extent possible.

The resolution to this problem was to eliminate existing snubbers or to replace snubbers with rigid supports to the extent that such a change is practical.

The elimination of certain existing snubbers or their replacement with rigid struts will eliminate the periodic maintenance and testing which they presently require, thereby reducing man-Rem exposure, spare parts inventory, and testing costs. It will assure that improper operation of the eliminated snubbers can no longer occur and thereby reduce the opportunity for improper snubber operation to overstress the line. The effort is based upon analytical evaluations performed by E&C - Engineering Mechanics, as qualified by calculations in accordance with Nuclear Design Control Procedures and in conformance with the present seismic Design Basis for the affected piping systems. The loadings for remaining supports and stress levels in all lines are within allowable limits with the designs implemented.

SUMMARY

OF SAFETY ANALYSIS All loadings and postulated conditions analyzed earlier, were analyzed after the reduction of snubbers, hence the probability or the consequence of occurrence of an accident or malfunction of equipment important to safety and previously evaluated in the UFSAR is not increased. In fact by reducing the number of snubbers, the probability of malfunction of these snubbers and the effects of those malfunctions on every other safety related system are eliminated.

The margin of safety was evaluated. Stresses in components are still within allowables. The structural margin of safety defined by the applicable code is maintained. This modification does not reduce in any way the margin of safety as defined in the basis of Technical Specifications.

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DC 87-01 AND 87-02 l MAIN STEAM VALVE HOUSE INSTRUMENT RELOCATION UNITS 1 & 2 RESPECTIVELY DESCRIPTION There is a possible failure mode of the SOVs (SOV-MS111A&B and S0V-MS211A&B) which is to change state in the superheated steam environment. The short and long term consequences of not resolving the problem is that a failure of this type could result in the main trip valves controlling steam flow to the turbine driven auxiliary feedwater pump to close. Therefore, the SOVs which are required to function during and after a Main Steam Line Break are not qualified to meet the anticipated temperature environment associated with the release of superheated steam in the upper elevations of the Main Steam Valve House (MSVH) and are to be relocated.

The MSVH Instrument Relocation modification will replace the S0V's with new equipment of the same type, and relocate them to an environment the equipment is qualified for.

SUMMARY

OF SAFETY ANALYSIS The relocation of the equipment within this design change does not impact the operation of existing safety related equipment or system.

The SOVs and the instre-/ " tubing between the solenoid valve actuators and the trip valves will be e- ally installed as not to degrade the seismic qualification of the <- i trip valvec.

The relocation of instrumentation is implemented to satisfy the environmental qualification of the equipment, and thus will provide a reliable means of operation during and after a main steam line break.

The relocation of the SOVs for operating the steam supply to the turbine driven auxiliary feedwater pump trip valves will increase the reliability of the trip valves to function during and after a main steam line break.

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DC 87-05 AND 87-06 VITAL BUS MODIFICATIONS FOR REACTOR TRIP PROTECTION UNITS 1 & 2 RESPECTIVELY DESCRIPTION To ensure North Anna is not subject to spurious tripping due to loss of a single vital bus power supply, the following modifications will be made to the RCP breaker position input to the SSPS, (Solid State Protection System) the vacuum breakers on the condenser outlet waterbox air separating tank, and the power supplies to the Westinghouse controllers for the feedwater control valve.

A. RCP Breaker Position To prevent a spurious reactor trip on loss of one channel power, a normally closed RCP breaker position contact will be connected to the existing input relay. The input relay will be energized via the compartment channel power when the RCP is tripped. The output contact of the existing input relay will now be a normally open contact instead of a normally closed contact. Therefore, when the input relay is energized, the RCP trip logic will be actuated.

B. Condenser outlet Waterboxeo Vacuum Breakers Each of the eight (8) vacuum breakers per unit have both a Train A and Train B control solenoid valve, and all eight Train A solenoid valves share the same electrical circuit as do the eight Train B solenoid valves. The operation of each of the eight vacuum breakers is identical. Currently, when required to open the vacuum breakers, the solenoid valves are de-energized which indirectly causes the Circulating Water pumps to trip which results in a turbine trip which trips the reactor.

At the point in the circuit where the solenoids are currently connected, two relays will be added. A contact from each of these relays will be used as an extension of this same circuit and wired in series to energize the rolenoid valves.

This modification will require replacing the existing normally clo ed 3-way solenoid valves located in the Turbine Building with 3-way solenoid valves which are designed for normally open operation.

C. Feedwater Regulating Valves Controllers Each of the four (4) 7300 series primary plant process control cabinets per unit is equipped with a primary and backup power supply to feed all equipment within each cabinet. Both power supplies of each cabinet are currently fed from the same source, either Vital Bus I, II, III, or IV. Therefore, in the e;ent a vital bus is lost, the output voltage of the power supplies from the affected cabinet will be "zero," and, for control cabinets 6, 7, and 8, the lack of voltage will cause the feedwater regulating valves to close which will indirectly trip the reactor due to low SG 1evel.

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DC 87-05 AND 87-06 This modification will provide an alternate reliable power source for the backup power supply in process control cabinets 5, 6, 7, and 8. The existing semi-vital buses do not have adequate spare capacity or spare breakers to power the process control cab. nets backup power supplies.

SUMMARY

OF SAFETY ANALYSIS A. RCP Breaker Position The energize-to trip design to trip the reactor via the reactor coolant pump interlock is an anticipatory trip. No credit was taken for this trip in the station safety analysis and no accident analysis needs to be re-evaluated.

GDS-23 tor protection system failure modes allows "the protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis". Since the RCP trip is anticipatory and provides a departed and diverse backup to the 2 out of 3 reactor coolant low flow trip, the GDC-23 criteria is satisfied.

Consequently, the change in logic function of the RCP trip to energize-to-trip does not affect safety equipment or the operation of safety systems.

B. Condenser Outlet Waterboxes Vacuum Breakers This modification changes the operation of the vacuum breakers from de-energize to trip open to energize to trip open. This modification does not affect safety system equipment or operation.

The construction and post-constructic1 phases of this modification do not result in a possibility for an accident or malfunction.

Energize-to-trip operation of the solenoid is being installed to reduce the possibility of unintended equipment operation.

C. Feedwater Regulating Valve Controllers This modification constitutes an operational enhancement since the FSAR assumes the total loss of the cabinet. Therefore, this modification does not increase the probability of an accident or malfunction of equipment important to safety because the analog electronics will still be available after the loss of a vital bus.

j Operation of the reactor protection system remains unchanged. The addition of the backup AC power supply does not constitute a challenge to the protection system.

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7-3 DC 87-020'-3 SiEAk GEM 7.RATOR L%'NtDG?R ILOW RESISTANCE PLATES' INSTALLATION / NORTH ANNA /

UNirS 1 & 2

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DESCRIPTION The oriclnal design of the series 51 steam generators at North Anna 1 and 2 incluteJ flow resistance plates. They were located in the downcomer annulus between the tube bundle wrapper and the steam generator shell and consisted of perforated plate welded to the tube bundle wrapper. The plates added a resistance to the flow path of the water traveling down the downcomer. These plates were subsequently removed to increase flow through the tube bendic. The increased tube bundle flow increased the crossflow into the bundle and across the tubesheet at the bottom of the wrapper. This increased crossflow was considered desirable to reduce the area of the tubesheet with the potential for sludge deposition. Presently, sludge accumulation is not as great a concern as was the case when the plates were removed, due to changes in steam generator water chemistry, improved tubesheet cleaning methods, more frequent tubesheet cleaning and increased attention to generating of corrosion products in the secondary system.

In the present configuration, the steam generator recirculation flow through the boiler section causes some excessive fluid hydrodynamic excitation of some of the 3388 tubes in the tube bundle resulting in flow induced stresses.

Creation of better flow characteristics in the tube bundles will reduce the field hydrodynamic excitation of the tubes.

The proposed resolution for each of the steam generators at Units 1 and 2 is the installation of twenty downcomer flow resistance plates in each steam generator. This will reduce the flow induced stresses on some of the tubes by creating better flow characteristics in the downcomer annulus and tube bundle.

The purpose of the modification is to reduce the potential tube vibration due to fluid-elastic excitation. An evaluation of the fluid elastic stability of the tubes not supported by the anti-vibration bars (AVB's) has shown that the reduction in flow will result in improved stability of the tubes to fluid elastic excitation in the U-band region of the tube bundle. An evaluation of the steam generator hydraulics has shown that the installation of the resistance plates will result in a significant reduction in the circulation ratio and mass flow through the tube bundle.

The relationship between reduction in flow area at the downcomer flow resistance plate and reduction in circulation ratio was determined by a thermal hydraulic analysis. The flow area includes the holes in the plate segments, the gap between the plate and the shell, the gap between subassemblies, and the gap between the plate and the wrapper. The determination of the design flow area included consideration of the sensitivity of circulation ratio to variations in the installed flow area. The design of the codifications has provisions for adjustment of location to control the gap variations. The tolerance of the flow area with the plates installed will result in a recirculation ratio meeting the design requirements.

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DC 87-026-3 SUlWARY OF SAFETY ANALYSIS The implementation of this DCP does not constitute an "Unreviewed Safety Question" as defined in 10 CFR 50.59 except for the steam generator tube rupture accident.

An assumption made for determining the off site doses for the previous steam generator tube rupture accident analysis is affected by the mass change on the secondary side of the steam generator. That analysis assumed that the steam generator tube rupture location was covered by either water or a steam water mixture. This permitted the analyst to assume that much of the iodine in the break flow remained in the liquid water mass on the secondary side of the steam generator thus limiting the dose available for release. Current calculations show that a mass of approximately 113,500 pounds of water would be needed to provide tube bundle covery. Westinghouse calculations for steam generator mass with the modifications are below this value for unit power levels above 59% of rated thermal power.

An analysis of the impact that this difference has on the calculated doses has been performed. As a result, the thyroid dose for the case in the current UFSAR increases from 0.355 rem to 1.77 rem. This is well below the 10CFR100 acceptance limit. Additional cases have been performed following the guidelines in USNRC Standard Review Plan 15.6.3. For the case with a preaccident iodine spike, the thyroid dose is 26.7 rem compared to the acceptable limit of 300 rem. The case with a concurrent iodine spike has a calculated dose of 1.52 rem compared to the acceptance limit of 30 rem. As shown, all three cases result in calculated doses well below the respective acceptance criteria.

By reducing flow induced stress on certain steam generator tubes, the modification should reduce the probability of a steam generator tube failure.

The installation of the flow restriction plates may cause the steam generator tube wrapper to deform near the resistance plate welds during a main steam line break but this results in no degradation of the tubes and was analyzed when the original flow restrictor plates were in place.

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VIHOINIA ELECTHIC AND PowEn COMPANY Ricianown,Vinoxx1A 2020I

• L 8 2 ""^" March 31, 1988 Vaca Passinsur NilCLsAa ormaATIONS United States Nuclear Regulatory Commission Serial No.88-167 Attention: Document Control Desk NAPS /JHL Washington, D. C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

SUMMARY

OF PROCEDURE CHANGES, FACILITY CHANGES AND SPECIAL TESTS PURSUANT TO 10 CFR 50.59 Pursuant to 10 CFR 50.59(b)(2), enclosed is a summary of procedure changes, facility changes and special tests that were completed for North Anna Units 1 and 2. A summary of the safety evaluation for each procedure change, facility change and special test is included.

Very truly yours, W. L. wart Attachments cc: U. S. Nuclear Regulatory Commission 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 l

l Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station l

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