Text

1996 Annual Results & Data Rept for Pbnp,Units 1 & 2
	ML20135D073
Person / Time
Site:	Point Beach
Issue date:	12/31/1996
From:	Dante Johnson; WISCONSIN ELECTRIC POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	PBL-97-0066, PBL-97-66, NUDOCS 9703050035
	Download: ML20135D073 (96)
	v • d • e

.- _ - _

i Wisconsin

'Electnc POWER COMPANY Point Beoch Nuclear Plant (414) 755-2321 6610 Nuclear Rd., Two RNers. WI 54241 Pill 97-0066 February 27,1997 Document Control Desk U. S. Nuclear Regulatory Commission Mail Station PI-137 Wa shington, DC 20555 Ladies / Gentlemen:

DOClui8 50-266 AND 50-301 ANNUAL RESULTS AND DATA REPORT- 1996 POINT BEACil NUCLEAR PLANT. UNITS 1 AND 2 Enclosed are eleven copies of the 1996 Annual Results and Data Report for Point Beach Nuclear Plant, Units 1 and 2. This report is submitted in accordance with Technical Specification 15.6.9.1.B pursuant to the requirements of 10 CFR 50.59(b). The report contains information regarding highlights of Point '

Beach Nuclear Plant operations during 1996 and includes descriptions of facility changes, tests and experiments, personnel occupational exposures, results of steam generator in-service inspections, and listings of reactor coolant system relief valve challenges.

Sincerely, Ts,@ h" Douglas F. Johnson Manager-Regulatory Services and Licensing hds Enclosures l.

l cc: NRC Regional Administrator, Region 111 i

NRC Resident inspector 9703050035 961231

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WISCONSIN ELECTRIC POWER COMPANY ANNUAL RESULTS AND DATA REPORT 1996 POINT BEACII NUCLEAR PLANT UNITS 1 AND 2 U.S. Nuclear Regulatory Commission Dockets Nos. 50-266 and 50-301 Facility Operating Licent e Nos.

DPR-24 and DPR-27 l _

PREFACE This Annual Results & Data Report for 1996 is submitted in accordance with Point Beach Nuclear Plant, Unit Nos. I and 2, Technical Specification 15.6.9.1.B and filed under Docket Nos. 50-266 and 50-301 for Facility Operating License Nos. DPR-24 and DPR-27, respectively.

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TABLE OF COMTENTS PAGE I. INTRODUCTION 3

11. lifGilLIGIIIS 3 Ill. AhilfNDMENTS TO FACILITY OPERATING LICENSES 4 IV. 10 CFR 50 59 & 10 CFR 72.48 SAFETY EVALUATIONS Procedure Changes 5 Modincations 30 Temporary Modincations 68 SPEEDS 74 Miscellaneous Evaluations 75 V. NUMBER OF PERSONNEL AND PERSON REM BY WORK GROUP AND JOB FUNCTION 90 VI. STEAM GENERATOR INSERVICE INSPECTIONS 91 Vll. REACTOR COOL. ANT SYSTEM RELIEF VALVE CilALLENGES Overpressure Protection During Nonnal Pressure and Temperature Operation 95 Overpressure Protection During Low Pressure and Temperature Operation 95 Vill. REACTOR COOLANT ACTIVITY ANALYSIS 95 O

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I. INTRODUCTION The Point !!each Nuclear Plant, Units I and 2, utilize identical pressurized water reactors rated at 1518 Mwt each.

Each turbine-generator is capable of producing 497 Mwe net ($24 Mwe gross) of electrical power. The plant is located approximately ten miles north of Two Rivers, Wisconsin, on the west shore of Lake Michigan.

II. IIIGIILIGilTS UNIT 1 liighlights for the period January 1,1996, through December 31,1996, included a 25-day refueling / maintenance outage. Major work items included: Completion of the emergency diesel generator tie-in work; modifications to atmospheric and ccndenser steam dump valves; boric acid and reactor makeup water flow transmitters; llP turbine exhaust moisture preseparators; and new containment isolation valves installed on the nitrogen supply line to the Si accumulators. A total of 21 modifications were complete.

In June and December, the unit was reduced to 70% power for atmospheric steam dump, condenser steam dump, stop valves, and crossover steam dump valves. Also in December, the unit entered a self-imposed power restriction to 90%.

Unit I operated at an average capacity factor of 94% (MDC net) and an electrical / thermal efficiency of 34%. The unit and reactor availability were 93.1% and 93.6%, respectively. Unit I generated its 87 billionth kilowatt hour on January 8,1996; its 88 billionth kilowatt hour on March 29,1996;its 89 billionth kilowatt hour on July 14,1996; its 90 billionth kilowatt hour on October 3,1996; and its 91 billionth kilowatt hour on December 23,1996.

UNIT 2 liighlights for the period January 1,1996, through December 31,1996, included the shutdown of the unit for its twenty-second refueling / maintenance outage. The outage began in October and the unit remained shut down through the end of 1996. Major work items included: Replacement of both steam generators; main control board wire separation resolution; replacement of 8 main steam condenser valves; 11P turbine crossunder piping; repair of service water west header; replacement of boric acid and reactor makeup water totalizers; replacement of X-02, X-03 and .X-04 transformer metering and relaying circuitry; and, exciter modifications, in May a turbine trip occurred because the turbine stop valves shut. In July the unit reduced power to 80% for repairs of 2P-27A heater drain tank pump and for the performance of atmospheric steam dump, stop vals es, and crossos er steam dump vah e tests. In August the unit was reduced to 48% power because of blocked circulating water traseling screens identified during chlorination of the intake structure.

Unit 2 operated at an average capacity factor of 69.2% (MDC net) and an electrical' thermal efficiency of 34%. The

, unit and reactor availability were both 75.8%. Unit 2 generated its 87 billionth kilowatt hour on February 14,1996; its 88 billionth kilowatt hour on May 13,1996; and its 89 billionth kilowatt hour on August i 1,1996.

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I' III. AMENDMENTS TO FACILITY OPERATING LICENSES During 1996 there were 2 amendments issued by the U. S. Nuclear Regulatory Commission to Facility Operating 1.icense DPR-24 for Point Beach Nuclear Plant Unit 1, and 2 amendments issued to Facility Operating License DPR-27 for Point Beach Plant Unit 2. The license amendments are listed by date ofissue and summarized below:

Amendment 168 to DPR-24. Amendment 172 to DPR-27. March 20.1996: The amendments extend the operation of both units with the current heatup and cooldown limit curves to 23.6 effective full power years.

Amendment 169 to DPR-24. Amendment 173 to DPR-27. Ortober 9.1996: The amendments incerporate the provisions of 10 CFR Part 50, Appendix J, Option B.

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IV.10 CFR 50.59 & 72.48 SAFETY EVALUATIONS PROCEDURE CIIANGES The following procedure changes were implemented in 1996:

1. CSP P.I. Unit I and Unit 2, Response to Imminent Pressurized Thermal Shock Condition, Revision 14 (Permanent)

The procedure isolates safety injection (SI) accumulators and verifies charging flow prior to securing Si pumps.

. Previous revisions secure Si pumps first, after Si termination criterion are met. This resequencing prevents accurhulator discharge from masking Si reinitiation criterion immediately after securing SI pumps.

The revision also adds an option to throttle SI pump discharge motor-operated valves (MOVs) as a method to maintain specified subcooling. This provides finer control of subcooling and limits Si pump restarts and associated pressure surge when Si reinitiation criteria are met.

Summarv of Safety Evaluation: The probability of an accident is not changed by this revision since no new accident initiators are created. Changing the order of accumulator isolation, verifying charging flow and securing Si does not create an accident since steps are performed afler SI termination criterion have been met.

Throttling Si flow to maintain RCS subcooling after si reinitiation criterion have been met still meets the purpose of the WOG procedure while limiting subsequent pressure surges from starting and stopping the Si pumps. The addition of pressure criteria for terminating depressurization has no effect on accident initiators since the pressure provided is a point from the pressure / temperature limit curve during the temperature soak requirements.

Throttling the Si discharge valves is expected to increase the wear on the seating surface. This does not increase the radiological consequences of an accident since the discharge valves are not relied upon to provHe a containment barrier in the event of a malfunction resulting in the loss of one train of SI. The Si containment boundary is the closed system with the SI suction from the RWST being the closed system boundary valves.

The possibility of equipment malfunction has not changed except for the wear on the seat on the Si discharge throttle valves as previously e suated. Since other equipment operation is under the same operating conditions as in the current revision the p :sibility of a malfunction of equipment is not changed.

There is no Technical Specification (1S) w hich applies to plant operation for the situation in w hich this procedure applies, therefore, the margin of safety defined within TS is not changed. The changes do not involve an unreviewed safety question. (SER 96-102)

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2. LOP-0, Unit I and Unit 2, Reactor Trip or Safety injection, Revision 20. (Temporary)

EOP-0.1. Unit I and Unit 2, Reactor Trip Response, Revision 15. (Temporarv)

LCA-0.0, Unit I, Loss of All AC Power, Revision 15. (Temocrary) liCMW, Unit 2, Loss of All AC Power, Revision 16. (IrmpaaIy)

CSP-5.1, Unit I and Unit 2. Response to Nuclear Power Generation /ATWS, Revision 13. (Temporary)

SEP-3.0, Unit I and Unit 2, Loss of All AC Power to a Shutdown Unit, Revision 13. (Temporary)

In support of emergency diesel generator (EDG) frequency issues, caution statements were included to account for current conditions. Standard EOP cautions that warn that motor-driven auxiliary feedwater pump (AFP) breaker may trip at Dows of >320 gpm were changed to Hows of >200 gpm. In EOP-0, this caution was also moved to an earlier location to ensure timely operator consideration of this potential problem.

Summary of Safety Evaluation: The revisions do not affect the probability of occurrence of an accident previously evaluated in the FSAR since the procedures are in effect post-accident. Although a caution does not dictate operator action (in accordance with NUREG-1358, Supplement 1), cautions provide information for ,

which their actions are based. The operation of the auxiliary feedwater system is not an initiator for an accident assumed in the FSAR.

The probabi!ity of occurrence of a malfunction is unchanged since required operator action remains the same.

Operator actions provided in the emergency operating procedures are required to mitigate the consequences of an accident as stated in FSAR Section 12.4.1. The auxiliary feedwater system automatically starts and delivers up to 800 gpm by design; however, only 200 gpm is required by the accident analysis. The auxiliary feedwater system requires " skill of the craft" control for a variety of postulated events to limit consequences. Operator action is required to limit auxiliary feedwater (AF) Dow to prevent excessive cooldown and steam generator overfill for the following events: Steam generato tube rupture; reactor trip; small break LOCA; and faulted steam generator. Other specinc operator actions include switchover to service water supply after depletion of the condensate storage tank. Evaluation of the emergency operating procedures, the time required to reach the cautions for the AF How, in actual experience and in evaluated simulator scenarios, provide assurance that the operators are able to adjust AF to 200 gpm prior to the expected time / current condition w hich could cause a P-38A AFP motor breaker trip. Operator actions do not negate the fact the plant design is based on assuming credible initiating accidents and that the protective and engineering safeguards systems are provided to limit the consequences of unlikely events.

The change improves operator knowledge and provides EDG frequency as another parameter to consider in operating the AF system above the design now of 200 gpm. This change does not involve an unreviewed safety question or TS change. (SER 96-027)

3. 11CP-02.015: IICP-02.015-1: 21CP-02 015-1: 21CP-02.01.5-L Reactor Protection System Permissives and Trip I.ogic Post Re~ueling Test. Resision O. (New Procedutt9 The procedures perform logic channel testing of the reactor protection system. This demonstrates of reactor -

trip bypass breakers prior to use if the reactor trip breakers are closed at the start of test. They further demonstrate operability of the P7, PS, and P9 permissives and reactor trip logic for source, intermediate and power range low setpoints. Steps proside the operator a list of prerequisite component and system conditions. .

Procedures were revised to be unit and channel specific. Steps list conditions necessary to verify the equipment is returned to service prior to obtaining signature. Acceptance criteria statements describe conditions and results necessary to satisfy TS requirements.

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Summary of Safety Evaludiml: Reactor protection (RPS) and engineered safeguards features (ESF) systems are not accident initiators. The procedures are required by TS, and the methodology is described in the FSAR.

The RPS or ESF systems are not degraded below the design basis function as described in FSAR Sections 7.2.2,

" Protective Systems, System Design," or Section 7.5.2," Engineered Safety Features Protective System Design." Test methods have demonstrated test effectiveness and are not altered by the procedure. The activity supports portions of RPS and ESF instrumentation of TS 15.4.1, The procedures do not reduce the frequency requirements for performance of TS Table 15.4.1-1 Items 1,2,3,5,6,11,15, and Table 15.4.1.-2 Item 25a.

This activity creates no other credible malfunctions than previously evaluated in FSAR Section 7.2 Protective Systems. The change does not involve an unreviewed safety question does not exist. (SER 96-007)

4. 21CP-04.002-1, Reactor Coolant Flow Transmitters, Ou' ;e Calibration, Revision 3. (Permanent)

The procedure changes the method used to select full flow differential pressure (AP) calibration data to achieve accurate reactor coolant system flow. Reactor coolant now calibration data is normally determined by reading the loop current at the test points of the reactor coolant flow transmitters near the end of cycle. The data are then used to calculate full flow difTerential pressure (AP) for each of the six transmitters for use in calibration.

The data are unique to each transmitter since each is tapped off of a different point on the RC loop elbow tap.

Since Unit 2 steam generator replacement, the full-flow data obtamed (based on plugged SGs) at the end of the past cycle for Unit 2 is no longer valid it is expected that reactor coolant flow and the flow calibration data will be similar to that of Unit I which has essentially clean SGs.13ased on this expectation, the highest differential pressure (AP) number from Unit I calibration data obtained from 1990-1994 was selected. This number (317.6" water) is taken from 1-FT-411. Utilizing the highest dp is conservative for low flow trip since a lower actual differential pressure (AP) causes a lower indication and places it closer to the trip point. Recent historical data (1980 to present) shows that the Unit I dp for 100% flow tends to be slightly higher than Unit 2, so this method is conservative.

Summary of Safety Evaluation: The change causes the reactor coolant How transmitters to be calibrated in a more conservative manner by calibrating them to a wider range. This calibration results in a lower indicated flow which is conservative to the low flow trip. In addition, procedural steps are taken to evaluate and adjust RC How during startup. The change does not afTect capability of the fuel cladding, reactor coolant system, or containment to perform its safety functions. The low flow trip logic is not altered nor is the low How trip setpoint afTected. Therefore, the activity does not increase the probability of occurrence or the consequences of an accident previously evaluated in the FSAR.

The changes do not reduce the margin of safety defined in the flasis of TS. The changes to the reactor coolant flow calibration cause the indicated flow to move in a conservative direction in relation to TS setpoints. The actual trip setpoint is nct changed nor are TS setpoints, so the margin remains conservative. This change does not involve an unreviewed safety question. (SER 96-127)

5. 1&2lCP-04.036 anilMlCP-04.036-1, Containment liigh Range Radiation Monitoring Sy stem Channels

. RE-126, RE-127, and RE-128 Refueling Calibration, Revision O. (New Proce_durts)

The procedures calibrate high range radiation monitoring channels. Only one chan.nel is removed from service

, and calibrated at a time. Scaling of E/l convener feeding indication at control board C-20 is changed to produce a live zero indication. This enables operators to observe a channel failure.

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Summary ofSafety Evaluation: The activity is required by TS. Radiation monitoring systems (RMS) are not accident initiators or safeguards actuators but are used for post-accident monitoring. The change to a live zero indication enhances the ability for observance of system operability with no reduction in detector sensitivity due to the changes. The activity has no effect on accident probability or assessment. Detector range meets Regulatory Guide 1.97 guidance. These procedures do not restrict access to vital areas or impede actions to mitigate the consequences of an accident. If an accident occurred during the performance of these procedures, the minimum required 2/3 operable channels criteria of TS Table 15.3.5-5, item 7, is maintair'ed. Only one channel is out of service at a time. This activity enhances the operational performance of the equipment which decreases the occurrence of malfunctions. Previous testing methods used for this activity have demonstrated to be effective and are not altered. The activity creates no other credible malfunctions thn previously evaluated in FSAR Section 11.2, Radiation Protection. The change does not involve an unreviewed safety question.

(SER 96-097) -

6. l&2iCP-04.039-1. Turbine Valve Calibration and Functional Test of Trip Circuits, Revision 6 (New Procedurs) .

These procedures calibrate turbine governor and stop vah es and functionally test auxiliary goven.or, independent emergency overspeed protection and emergency turbine trip circuits during shutdown conditions.

Summary of Safety Evaluation: While this surveillance test is not described in the FSAR, it is a TS required activity as shown in Table 15.4.1 1, item 42 and Table 15.4.1-2, item 18. Turbine generator overspeed is analyzed in FSAR Section 14.1.12. Testing is performed on a shut down unit, but affects the operating unit because the activity opens sliders in both units. This makes both units ofIOPS inoperable while this test is performed. TS 15.3.4.F requires only one turbine overspeed protection system that trips the turbine stop valves or shuts the turbine governor valves to be operable far the operating unit. This test does not disable the mechanical or auxiliary governor trips of the openting unit. The change does not involve an unreviewed safety question. (SER 96-099)

7. IlCP-05.058A-1: llCP-05 ox:S-n 2iCP-CiO58A-1: 21CP-05.058B-1. Safeguards Timing Relays Calibration, Revision 0. (New Procedurn)

The four procedures reflect the requirements of service water (SW) and emergency diesel generator (EDG) operability as discussed in administrative restrictions, DCS 3.1.7," Service Water Pump Operability," and DCS R l 3.1.17," Emergency Diesel Generator Operability."

litunmary of Safety Evahnttion: The procedures implement changes to the method of dealing with SW and EDG operability during calibration of the SW auto start timers (on EDG breaker closure). During the calibration of Train A timers, conditions exist where all three SW timers are removed and/or not post-maintenance tested to fully return them to service. During this time period, automatic SW cooling cannot be assured to the EDG that is aligned to that unit and bus w here the calibrations are performed. T herefore. a conservatis e I CO is entered on both units prior to performing SW timer calibrations on the standby emergency power supply to l A-05 or 2A-05. This LCO is not required for Train B because that train's EDGs are not dependent upo, SW for cooling. Additional requirements are placed upon EDG operability w hen calibrations '

are performed on Unit 2 because of cross unit electrical alignments specified for outside air temperature and -

fuel oil supply. Additional alignment of SW pumps to ensure full system operability and to ensure cooling of \

potentially connected EDGs is specified to ensure that potential malfunctions are avoided. -

During the niay calibration, service water availability to the shutdown unit is the most limiting factor. Even though a SW pump will not respond to an auto start when its relay is remos ed for calibration, there is sufficient

, redundancy and diversity of signals to ensure that the service water system function is maintained and that with

('

alignment of the SW system there is adequate cooling to a potentially aiigned EDG. Additionally, with the Train A rest ictions that are entered, additional av.orance of cooling is obtained as the single failure cri.terion is '

relaxed As such the probability and consequences of presiously esaluated malfunctions is not increased paces

Since accepted alignments and LCOs are entered for these calibrations, and because the requirements for standby emergency power operability requires that all required support systems be operable, these procedure changes do not create the possibility of a different type of accident or malfunction and do not reduce the margin of safety speciGed in the Basis of U. The changes do not involve an unreviewed safety question.

(SER 96-123)

8. IICP-10.033: IICP-10.033RD-1: llCP-10.33Wil-1: IICP-10.033BL-1: IICP-10 033YL-!: 21CP.10.033:

21CP-10 033ED-1: 21CP-10 33Wil-1: 21CP-10.033BL-1: 21CP-10.033YL-1, NIS Power Range Calibration of New Axial Offset Correction Factor, Revision 0. (New Procedures)

These new procedures were developed frem existing procedures as part of the procedures upgrade project. The procedures provide a list of prerequisite component and system conditions and w ere developed to be unit and channel specific. Steps list conditions necessary to verify the equipment is returned to service prior to obtaining signature. Acceptance criteria statements to describe conditions and results necessary to satisfy TS requirements. The change removes the turbine runback feature due to a dropped rod.

Summary of Safety Evaluation: The activity is required by TS and testing methodology is described in FSAR Section 7.4.3. One power range channel is calibrated at a time, preserving required minimum number of operable channels. The minimum operable channel criteria listed in TS Table 15.3.5-2 is not violated and the activity controlled by this procedure supports testing requirements of TS 15.4.1.-1, Item 1. The procedures do not restrict access to vital areas or impede actions to mitigate the consequences of an accident since existing testing methods have been demonstrated effective and are not altered by this t,ctivity. Radiation doses to the general public are not increased as a result of this procedua. The power range ponion of NIS is not degraded beyond that described in FSAR Section 7.4.2," System Design." FSAR Section 7.4.2 states,"Online testing and calibration features are provided for in each channel." Components calibrated are returned to service and such return is verified. Nuclear instrumentation power range is supplied with suf6cient redundancy to provide the capability for channel testing at power. The change does not involve an unreview ed safety question.

(SER 96-008)

9. lWP 91-116*Zl. Connection of Unit i Train A UV, SI & SW Autostarts to G-02, Revision 0. (New Procedure)

IWP 91-116*Z2. Addition of Test Switches to Degraded Grid Follower Relays and PT Fuses G-03 Mini Loop to I A-06, Revision 0. (New Procedure)

The procedures control the emergency diesel generator (EDG) modification work performed during UI R23.

This work completed the 6nal Phase 3D of MR 91-116 which added two safety-related EDGs. The phase tied in G-02 EDG as an operable standby emergency power source to I A-05 safeguards bus.

Unit I safety injection (SI),1 A-05 bus undersoltage (UV) and Train A service water (SW) pump automatic start signals were connected to G-02 and tested. In addition, spare, unused Si trips to cubicles I A00-61 and 2A00-72 were disconnected (no breakers are in the cubicles), test switches were added to the degraded grid voltage follow er relay circuits for the I A-06 bus, potential transformer (PT) fuses were installed in the i A-06 bus synchronization circuit; and the G-03 EDG was tested from its alternate de supply.

Summary of Safety Evaluation: Previously evaluated accidents or transients applicable to the EDG tie-in activities are a loss of offsite power (LOOP), a loss of electrical load to Unit 2, and a Unit 2 trip. Train A work is done Grst followed by Train B work. Work begins after Unit 1 is in cold shutdown and single train residual heat removal (RilR) capability is demonstrated. During the Train A work, the Train B residual heat removal (RilR) and component cooling w ater is in service. The opposite is true during the Train B work.

The spent fuel pool cooling (SFP) pumps and valves are powered from Unit 1 Train B and Unit 2 Train A. As only one train of standby emergency power is affected at a time during the EDG tie-in work, standby emergency power remains as ailable to at least one train of SI'P coohng. The probabihty of an accidem or malfunction of equipment important to safety pres iously es aluated in the l'SAR is not increased Page 9

During the LCO timeframe for standby emergency power to Unit 2 Train B for G-04 and G-03, the single failure requirement for the emergency power system is relaxed. TS-required shared safeguards equipment is operable. This means that a failure of G-02 or G-01 does not have to be postulated to ensure an accident on Unit 2 can be mitigated. Thus, entering these LCOs does not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

The required redundant EDGs, either G-Oi,0-02, G-03, or G-04 are tested as required by TS before entry into a standby emergency power LCO of greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter. Compensatory measures ensure ofTsite power remains available to the safety systems required to mitigate the consequences of an accident or malftmetion of equipment important to safety previously evaluated in the FSAR. .

No new failure modes of equipment different than previously evaluated are created by the tie-in activities.

Failure of the G-02 breaker to I A-05 or G-02 are identical to those already evaluated in the FSAR. The operational testing of G-01, G-02, G-03 and G-04, including the G-02 output breaker to I A-05 (I A52-66), _

provides additional assurance that a malfunction does not occur. Therefore, the possibility of a malfunction of equipment in4portant to safety of a different type than previously evaluated in the FSAR is not created.

The margin of safety dermed in the Basis for TS is not reduced by the activities. The margin of safety for standby emergency power sources is met by entering the allowed LCOs for G-01, G-02, G-03 and G-04 and providing for required support functions. Under these conditions the required equipment redundancy is maintained and the margin of safety dermed in the Basis for TS is not reduced. These activities do not involve an unreviewed safety question. (SER 96-012)

10. lWP 96-022-2, Installation of New DC Control Power to 2B-03 and 1B-04, Revision 0. (New Procedure)

IWP 96-022d, installa' ion of New DC Contral Power to 2B-03 and 1 B-04, Revision 0. (&w Procedure)

IWPs 96-022-2 and 96-022-3 place the new de control power feed to 28-03 and 111-04, inservice. The new feeder cables and breakers were installed via IWP 96-022-1.

Summary of Safety Evaluation: DC control power to the safeguards buses is used to respond to accidents previously evaluated in the FSAR. The evaluations show the de control power to the 480 V switchgear buses do not act as an initiator of these accidents. Therefore, IWP 96-022-2 and 96-022-3 do not increase the probability of occurrence of an accident previously evaluated in the FSAR.

To minimize the impact of rendering safety features out of service, the work is performed with the associated unit reactor core defueled. With the core defueled, the only load required to support the shutdown unit is the spent fuel pool cooling pump for decay heat removal. The spent fuel pool cooling pumps are powered from the affected buses but the loss of de control power to these buses does not affect the operability or availability of these pumps because it is fed hom motor control (enters (MCC) for uhich the breakers on the suitchgear have no automatic safety function (e g., the breakers on 211-03 and 115-04 feeding the MCCs are not required to cither open or close to respond to previously analyzed accidents).

To support operation of the opposite unit (e.g., Unit I operation while the de control pow er to 2B-03 is -

swapped), the controlling procedures requires the service water pump powered off the bus to be running before commencement of work. Therefore, should a safeguards actuation occur w hile work is ongoing, de control power is not required to close the service water pump feeder breaker. The same holds true upon a LOOP scenario, the breaker w ill already be closed and awaiting re-energization w hen power is re-established.

l The 1B-04 and 28 03 buses supply emergency power to many safety-related components that have TS associated with them. The activities do not prevent the emergency power from being supplied to these loads (except for a short duration during the swapover process), the margin of safety of TS associated with these loads is not reduced. In the process of swapping the de control power from the existing to the new feed, necessary cr;uipment remains operable and therefore the margin of safety is not reduced. The activities do not involve ar unreviewed safety question. (SER 96-086)

11. NP S 6.6.. Applicability of 10 CFR 50.59 and 72.48 Screenings for I&C Procedures, Revision O.

(&w Procedure)

This procedure provides guidance for when a 10 CFR 50.59 or 72.48 screening / evaluation is required for specinc revisions to instrument and control (I&C) technical procedures. NP 8.6.6," Applicability of 10 CFR 50.59 and 72.48 Screenings for I&C Procedures," supported by this 10 CFR 50.59 safety evaluation and 72.48 screening, functions as a 10 CFR 50.59 and 72.48 screening for specine I&C technical procedure changes.

Procedure revisions which do not require a 10 CFR 50.59 screening because they are covered by this safety evaluation are listed in Table I of NP 8.6.6. Changing a procedure to reflect a new setpoint does not require a 10 CFR 50.59 screening because tnere is a comment requiring the procedure revision to reference the safety evaluation report number on the procedure change form. The safety evaluation for the setpoint change addresses the safety issues regarding the new setpoint and the impact on the probability ofinitiating an accident or affecting any other equipment important to safety.

The types of procedure changes allowed by NP 8.6.6 have appropriate restrictions and guidance to ensure that the change does not adversely affect equipment important to safety used to mitigate the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. It does not create the possibility of r.ew malfunctions of equipment. It also does not affect the probability of new accidents.

Much of the instrumentation is used to maintain the margin of safety defined in the Basis for TS. The changes to I&C procedures allowed by NP 8.6.6 as covered by this safety evaluation are restricted to ensure it has no impact on the margin of safety. Therefore, an unreviewed safety question is not involved. (SER 95-091)

12. OI-15 A, I&2P-2A&B Charging Pump Alternative Modes Operation, Revision 0. (New Procedure)

The procedure transfers l&2P-2A charging pumps to an alternate shutdown / Appendix R power supply and subsequently restoring the transferred pump to its normal supply. This procedure aligns 1&2P-2B charging pump to the opposite unit's B-03 bus via the Kirk breaker system.

Summarv of Safelyhahldtion Equipment is not considered operable w hile it is powered from the alternate

. shutdown buses B-08%09, and applicable equipment LCO entries are required. Equipment pow ered from its Appendix R power socree does not have the same control system as it would from its normal supply. The loads are controlled from different locations (e.g., C-45 alternate shutdown control panel or the opposite unit

, control board for the P-2B charging pumps). Additionally, equipment powered from B-08/B-09 does not have the degraded voltage protection that it would from the normal 480/4160 vacuum source. B-08/B-09 provide loss of voltage protection for connected loads. Except the above differences, the equipment operates the same as if powered from normal powcr.

The procedure reduces the consequences of an equipment malfunction due to a fire in the 4 kV switchgear room, fire in the cable spreading room, or by increasing the availability of power to the equipment in the event of a failure of the normal power supply.

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Charging pumps 1&2P-2B may be considered operable when powered from the opposite unit via the Kirk breaker system. Design provisions (such as Appendix R cable routing) allow the pump to be operated from an equivalent, safety-related power source and controlled in the control room via the opposite unit's control board.

An operability test must be performed prior to considering the pump operable. Note that when a pump is aligned to the opposite unit, then the opposite unit Train B charging pump is inoperable. This may require an LCO on the opposite unit, depending on the number of charging pumps w hich are operable.

l l Design interk>cks prevent the normal and ahernate power sources from being tied together. Return to service testing operates the equipment from its normal power supply prior to declaring the equipment operable. The change does not involve an unreviewed safety question. (SER 96-039)

13.01-100, Adjusting SI Accumulator Level and Pressure, Revision 5. (Permanent) -

STPT I1.2, Setpoint Document - Relief Valve, Revision 3. (Permanent)01-100 ensures that the Si accumulator nitrogen fill piping is not subjected to excessive pressure. The -

procedure contains a caution to open SI-834A&B prior to opening SI-846 to prevent subjecting the lower-rated piping to full 12-pack pressure. Additional limitations ensure that overpn nurization cannot result from valve leakage during periods w hen an SI accumulator is not being filled. The controls include changing the order in which the nitrogen 12-packs are aligned and include a provision for venting the 12-pack side of the CIV SI-846 valves after the gas space is filled.

The change deletes the requirement for stationir.;,i dedicated operator to shut containment isolation valves SI-834A or SI-834B should containment isolation occur or be required. The dedicated operator requirements were eliminated because SI-834A&B are no longer containment isolation valves. .

STPT 11.2 deletes the 900 psi setpoint for HCV-957. Valve testing and analysis show that HCV-957 is not '

capable of providing a significant relief function until nitrogen pressure is well above 900 psi. A pressure relief device is not required in this application by USAS B31.1, Power Piping Code.

Summary of Safety Evaluation: The piping associated with the nitrogen fill lines is of two different ratings.

The piping between the nitrogen 12-pack in the primary auxiliary building truck access and containment isolation valve SI-846 is SI-2503N piping (2500 psig at 650*F). The nitrogen 12-packs are pressurized to approximately 2500 psig when delivered. The connection of the 2500 psig 12-packs creates a concern that the lower-rated piping could be overpressurized, either during valve lineups or due to leakage through valves used to isolate the 12-packs from the nitrogen piping. Additional limitations ensure that overpressurization cannot result due to valve leakage during periods w hen an Si accumulator is not being filled.

The function of the Sl accumulator is not changed. The FSAR does not describe the method by which the Sl accumulator nitrogen pressure is adjusted. This change provides assurance that penetration P-14C is not os erpressurired by valse leakage. In order for valve leakage to create an merpressure condition for the lowet-rated piping, there would have to be leakage past three valves in-series. Additional administrative controls remove the source of the high pressure when an accumulator gas space is not being filled.

During Si accumulator gas space fills, the pressure increase is controlled by the operator and terminated prior to pressure reaching the high alarm setpoint of 760 psig. In addition, when the accumulator is filled, it is also protected from overpressurization by Si accumulator relief valve SI-830A or SI-830B, which has a setpoint of 800 psig. An engineering review ofline pressures ensure the entire SI-903R line class piping is protected by the accumulator relief valve. This review considered the large capacity of the relief valve, the maximum expected nitrogen flow from a 12-pack, conservative component and piping pressure drops, and the B31 1 Code allowables.

Implementing administratis e contrals to prevent m erpressurizing the Si accumulator nitrogen till lines and piping associated uith containmeni penetration P-14C does not constitute an unresiewed safety question.

[SER %-OlN Ps. e 12 ,

14.01-112, Aligning Equipment To Appendix R Power Supply, Revision 0. (New Procedure)01-112 transfers equipment to its alternate power source and restores the equipment to its normal power source.

Equipment within the scope of this procedure includes 1&2P-10A&B residual heat removal (RHR) pumps, P-328, C, E, and F service water (SW) pumps,1&2P-2A charging pumps, and B-08/B-09 alternate shutdown buses, This procedure also aligns 1&2P-2B charging pumps to the opposite unit via the Kirk breaker system.

S.ummary of Safety Evaluation: During this procedure, equipment is not considered operable while it is powered from the alternate shutdown buses B-08/B-09, and applicable LCO entries are required. Equipment powcred from its Appendix R power source does not have the same control system as it would from its normal supply. Automatic start signals are not functional when aligned to its alternate source. The loads are also controlled from different locations (e.g., C-45 alternate shutdown control panel, or the opposite unit's control board for the P-2B charging pumps). Additionally, equipment powered from B-08/B-09 does not have the degraded voltage protection that it would from the normal 480/4160 Vac source. B-08/B-09 do provide a loss of voltage protection.

This procedure reduces the consequences of an equipment malfunction due to a fire in the 4 kV switchgear room, or increases the availability of power to the equipment in the event of a failure of the normal power supply.

Charging pumps I&2P-2B may be considered operable when powered from the opposite unit via the Kirk breaker system. Design provisions (such as Appendix R cable routing) allow the pump to be operated from an equiva' lent, safety-related power source and controlled in the Control Room via the opposite unit's control board. An operability run must be made prior to considering the pump operable. Note that when a pump is aligned to the opposite unit, then the opposite unit's "B" charging pump is inoperable. This may require an LCO depending on the number of operable charging pumps. This procedure does not involve an unreviewed safety question. (SER 96-001)

15.01-114, Returning CCW to the CCW System, Revision 0. (New Procedure)01-114 allows the connection to the radwaste portion of the Unit 2 component cooling water system for the purpose of returning collected component cooling water to the system using a pump and drum; an arrangement used in the past during performance of ORT 67,68, and 69 (component cooling water to RCPs).

Summary of Safety Evaluation: Return rig failures, failures that could result in a loss of component cooling water system inventory and spraying of component cooling water, could occur. The components used in the return rig meet the system design pressure and temperature requirements of 200 psi at 150 F, except for the tygon tubing. The tubing has a maximum recommended working pressure of 100 psi at 73 F. This is adequate for this application, returning ambient temperature component cooling water to the suction side of the Unit 2 component cooling water pumps at a pressure of about 20 psig.

To prevent a loss of component cooling water system inventory and to minimize spraying of component cooling w ater due to a return rig failure, a check vahe is installed at the manual isolation to the component

,, cooling water system and an operator is stationed at the return rig when the return rig pump is running or the manual isolation valve is open. Thus, if a return rig failure occurred during use of the temporary connection, the Class 1 portion of the system could be isolated from the Class til portion and the temporary connection.

This mainuins the system integrity as required by the FSAR.

The design and w ork controls ensure the component cooling water system is not adversely affected while the temporary connection is made to the system. The change does riot involve an unreviewed safety question.

(SER 96-1M)

Page 13

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16.01-115, SFP Service Water Cooling Isolation for Maintenance, Revision 0. (New Procedure)01-115 provides requirements for isolating service water to the spent fuel pool (SFP) cooling system. 'l his allows maintenance to be performed on the service water portion of M system which is common to both SFP heat exchangers or which cannot be isolated without isolating service water to both heat exchangers.

Summary of. Safety Evaluation: Service water shall not be isolated by utilizing SW-650 service water inlet to SFP cooling. A failure of this valve to reopen would not allow either train of SFP cooling to be returned to service. The service water shall be individually isolated to each heat exchanger utilizing SW-652 to isolate llX-13 A and SW-653 to isolate liX-138. This climinates the possibility of a single failure (SW-650 stuck shut) from returning at least one heat exchanger to service.

An accident for the spent fuel pool cooling system was not speci0cally denned by FSAR Chapter 14. FSAR Section 9.3 mentions a loss of SFP cooling accident and that adequate time exists to restore cooling. The FSAR also allows intermittent use of the SFP cooling system and loss of cooling allows Dre water to maintain SFP -

inventory. The use of Gre water along with other sources of makeup to the SFP is addressed in AOP-8F," Loss of Spent Fuel Pit Cooling " The isolation of service water to SFP cooling does not affect the ability to perform actions via AOP-8F. Fuel handling operations shall not be performed during this procedure so there are ra impacts on fuel handling accidents.

Isolation of service water to the SFP cooling system to perform maintenance on the service water portion of the system does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility of an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the TS is not reduced. The change does not involve an unreviewed safety question. (SER 96-090)

17.01-115, Service Water isolation of Spent Fuel Pool, Revision 1. (Permaren0 The change allows SW-650, llX-13A&B spent fuel pool (SFP) heat exchanger supply, to be shut due to leakage ofindividual SFP heat exchanger isolation valves. It also allows the removal of the valves for repair.

Summary of Safety Evaluation: The change requires that contingency plans for closing the SW system be in place by the time the SFP reaches 120*F. This means that under worst case conditions attempts to reopen SW-650 would be made at a SFP temperature of 120'F. Prior tests performed indicated heatup times of 1.5'F/hr. This allows approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months if the valve cannot be reopened to make up Danges, install hoses, and restore temporary cooling to the SFP to maintain SFP temperature below 145'F. This amount of time is judged to be sufficient to allow this alternate Dow path to be established. An NRC SER transmitted April 4,1979 evaluating the SFP cooling system indicated that temperatures up to 145'F are acceptable. In the event that this contingency cannot be completed, AOP-SF would be entered for maintaining SFP water inventory .

Isolation of service water to the SFP cooling system to perform maintenance on the service water portion of the system does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment imponant to safety is not increased. The change does not create the ,

possibility of an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the TS is not reduced. (SER 96-109)

18. OP-l A, Cold Shutdown to Low Power Operation, Revision 57. (Permanent)

CL-LD,lleatup, Revision 11. (Permanent)

The procedure changes maintain the second safety injection (SI) pump disabled until RCS temperature is

>370 F. Procedure and TS requirements are to have the second S,, imp disabled when >275 F. Recent LTOP system design resiew s identified that the current TS s alue (275'F) for second Si pump availability is outside the design basis of the 1, TOP system.

Page 14 w _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

Summary of Safetv Evaluation The procedure changes limit one Si pump to be operable at all times below the minimum temperature allowed for the primary system pressure test. This is consistent with the design basis for LTOP. Therefore, the probability and consequences of a design basis mass input transient of one SI pump discharging into RCS is not increased, because only one SI pump is operable, precluding the scenario of exceeding the discharge capability of the PORVs. The PORVs and RH-861C maintain overpressure protection capability as described in the FSAR. The LTOP system also remains armed until the RCS is >370*F. As such, the probability and consequences of a previously evaluated accident are not increased.

The actions taken to maintain the second Si pump disabled are consistent with TS (motor breakers removed or discharge valves shut and power removed). This practice does not increase the probability or consequences of a spurious SI signal (or other occurrence which attempt to initiate two SI pump operation) occurring below the

, minimum temperature for the inservice pressure test because only one Si pump is capable of providing flow to the RCS. This single Si flow is within the relieving capacity of the PORVs (and RH-816C if available as a backup) to maintain pressure, and is consistent with the design basis of the LTOP system as described in the

~

FSAR and TS. The probability or consequences of other equipment malfunctions which lead to an overpressure condition are not increased because the volume / mass increase associated with them is unaffected; the relief capacity and setpoint is unchanged, and LTOP is maintained armed until the RCS is >370'F. The change does not involve an unreviewed safety question. (SER 96-108)

19. OP-3C, Hot Shutdown to Cold, Revision 65. (Permanant)

The primary-to-secondary leakage integrity test at 1900 psid for Unit 2 was removed. An NRC commitment was made in 1982 prior to Unit 2 SG sleeving, for a 1900 psi primary side hydrostatic test to be repeated during subsequent inservice inspections of the SGs. The intent of the 1900 psid primary to secondary hydro is to dislodge any possibly entrapped impurities in the sleeve portion which might mask rejectable indications during eddy current testing. The commitment was implemented during each subsequent Unit 2 steam generator scheduled inservice inspection. The sleeved Unit 2 SGs were replaced during U2R22, therefore no further inservice inspections are performed on sleeved SGs.

Summarv of Safety Evaluation: Removing the Unit 21900 psid pressure test does not increase the probability J

of occurrence, nor the coraequences, of a malfunction of equipment important to safety previously evaluated in l the FSAR. Also, eliminating this pressure test does not create the possibility of an accident of a different type, I I

or the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR. This pressure test is not mentioned in the TS, therefore deleting it does not reduce the margin of safety defined in the basis for TS. The change does not involve an unreviewed safety question.

(SER 96-107)

20. OP-4D. Part 5, Draining the RCS to a CVCS IlUT Without Entering Reduced inventory and Without Draining Steam Generator Tubes, Revision 0. (New Procedure)

OP-4D Part 5 drains the reactor coolant system (RCS) without draining the steam generator (SG) tubes.

During refueling or cold shutdowns when there is no need to drain the RCS to mid-loop, the SG tubes do not

. require draining. This evolution is a standard industry practice. As long as the RCS inventory is maintained above the top of the reactor vessels nozzles and no RCS openings are made below the water levels, the SG tubes will not drain and may be maintained as such for the duration of the shutdown.

Summarv of Safety Evaluation: OP-4D, Part 5 places the reactor vessel (RV) level at 70% w hich is about 19" below the RV flange and 41" above the top of the hot legs. The expected RV level changes due to SG temperature and containment pressure are such that they do not pose a threat to decay heat removal capability; inadvertently elraining the tubes, or over-filling the reactor vessel. The changes in the reactor vessel level may be compensated for by various means ofinventory make-up and letdow n which are maintained during a shutdown.

Page 15

l Prior to securing the last reactor coolant pump, the boron concentration is verified via sampling to be greater 'l than the required level for either a cold or refueling shutdown (as applicable). Therefore, no pockets of underborated water is present in the tubes, so if the tubes drained, shutdown margin would be maintained. The initial conditions of RP-1 A," Preparation For Refueling," were revised to verify a portion of the RCS including SG tubes and intermediate legs are borated to refueling shutdown concentration. Not meeting this initial condition prevents TS refueling operation until the entire RCS is adequately borated.

Inadvertent draining of the SG tubes resulting in a spill would be contained within the containment sump and could be processed by the waste handling systems. This draining would not pose a threat to decay heat removal or other equipment important to safety and ifit resulted in a shutdown loss-of-coolant accident. It would provide additional inventory for mitigation.

Not draining SG tubes results in eventually having to start a RCP without a pressure absorbing volume in the RCS. Starting a RCP with a solid RCS is acceptable provided the secondary water temperature of each SG is ,

<50 F above the RCS temperature. Not draining SG tubes does not result in operation of equipment important -

to safety in a manner outside of normal design, and does not adversely affect DHR operation or SG integrity.

Draining the RCS without draining the SG tubes and leaving the SG tubes filled, does not create the possibility of an accident or equipment malfunction of a different type or reduce the margin of safety. Not draining SG tubes during a shutdown does not involve an unreviewed safety question. (SER 96-015)

21. OP-6A, Operation of Component Cooling System, Revision 19. (Permanent)

The revision provides guidance on starting a component cooling water pump following fill and vent of the pump to prevent lifting a relief valve. Attachment A establishes flow at approximately 3600 gpm to reduce the perturbation on the system caused by pump start.

Summarv of Safety Evaluation: During the initial start of a component cooling water pump, the component cooling water system lineup is changed from the normal at-power lineup as described in the FSAR during the initial start of the pump. A Level 3 dedicated operator is stationed with guidance to realign CC-824A&B to normal if a safety injection occurs.

In the case of a large break loss-of-coolant accident, the dedicated operator would have sufficient time to reposition CC-824A&B to their normal at-power positions and not delay the transfer to ECCS containment sump recirculation. The duration of having CC-824A&B out of the normal alignment is the time needed to start the component cooling water pump after which the valves are returned to their normal position. The change does not involve an unreviewed safety question. (SER 96-089)

22. ORT-3A, Safety injection Actuation With Loss of Engineered Safeguards AC Unit 1, Revision 29.

Termanent) l ORT-3 '. was revised with Train A safeguards equipment being tested as the lead train. During UlR22,

&.i-3B was performed with Train B safeguards equipment tested as the lead train. The philosophy of l performing lead and lag train testing was previously reviewed and approsed. The order of testing by train and j methodology is similar to Unit 2 ORT 3 previously performed. -

l I

I Page 16

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

3 Summarv of Safety Evaluation: The changes to ORT-3 do not increase the probability of occurrence of an ,

accident previously evaluated. The test is conducted during cold or refueling shutdown in accordance with TS !

I

{ requirements and applicable LCOs. Due to the restrictions associated with the test (no other LCOs entered, no

, fuel motion, no other safeguards systems work or testing, shared safeguards systems operable) tnd because Unit 1 is in cold or refueling shutdown, there are no previously evaluated accidents whose probability could be ;

directly affected by the changes. The availability of offsite power is not affected by these changes. These tests j do not initiate accidents previously evaluated for the at-power unit. G-02 and G-04 EDG are normally aligned !

j to Unit 2 safeguards buses. This test requires removing G-02 and G-04 from service at different times, ,

requiring that emergency power LCOs be entered for Unit 2. The shared safeguards loads powered from IB-03 i

) remain available to Unit 2 except for the short period that IB-03 undervoltage load stripping is verified.

i Therefore, shared safeguards equipment LCOs are also entered for Unit 2. !

ORT-3 does not increase the probability of an equipment malfunction previously evaluated. This test is , ,

required by TS and as such is part of normal equipment use. During the G-02 EDG test, a single largest load ,

rejection test is performed. During the test, the SI pump is stopped, and after a 15-second coastdown is

~ restored. This tests the EDG capability to handle the load swing. The 15-second coastdown is a safety margin ,

which ensures no damage to the Si pump motor can occur. This test, as a quali6 cation for G-02 to the 1 A-05 I

bus, is consistent with the testing requirements used to qualify the other EDGs. The testing is consistent with l 4

FSAR EDG loading csiteria and does not affect EDG operability. l

The changes do not increase the probability of occurrence of an accident of a different type. The changes do {'

not create new accident initiators. During the test, both trains of RHR are operable if the core is not offloaded.

The test is performed by train and set up so that RHR cooling is not interrupted. The opposite train of spent i fuel pool cooling is in operation during the test. Offsite power is available during the test and tested _

components are available with manual or automatic action. Equipment is used within its design basis. The test j

> does not involve an unreviewed safety question. (SER 96-016) j i

!' Summarv of Safety Evaluation: The revision does not increase the probability of occurrence of an accident 1 previously evaluated in the FSAR. SER 96-016 states the Unit 2 Train B G-04 EDG is removed from service, f

the testing is performed one train at a time, and all four EDGs are loaded to the Unit I buses. These statements do not apply to this procedure change since the scope is limited to testing the G-02 EDG only. SER 96-016 i

also states the backup EDG is available to supply the bus being tested in the event any of the EDGs are lost. ;

This statement remains accurate even though G-03 EDG will not be available to backup G-04 on l A-06. Since !

the only bus being tested by this change is 1 A-05, the assumptions and conclusion of SER 96-016 are !

applicable to this procedure change (except as noted). )

i The revision does not increase the probability of occurrence of a malfunction of equipment important to safety 1 previously evaluated in the FSAR for the same reason stated in SER 96-016. The initial conditions for the l revised portion of the test are the rame as the original Train A portion of the test except the Train A F generatoi is starting from a higher level (approximately $10"). The procedure change includes auxiliary feedwater pump operating restrictions to prev'ent overfill. The restrictions are limiting the flow to 230 gpm fo-l l

no greater than 30 minutes which represents approximately 50% of the volume required to reach overfill conditions (7000 gal versus 12000 gal). (SER 96-016_Q.l.)

~

23. ORT 3A. Unit 2, Safety injection Actuation with Loss of Engineered Safeguards AC, Revision 30,(Permang11)

ORT 3B. Unit 2, Safety injection Actuation with Loss of Engineered Safeguards AC, Revision 29, (Permarf.ni)

ORT 3A and ORT 3B address new testing equipment and filtering circuits to be used to monitor the EDG loading sequence test points. The new test equipment is a Tektronix Testlab data analyzer Model 2510 or 2505.

Summary of Safety Evaluation: The changes do not affect the performance of ORT 3. Monitoring of the test points still takes place when speciGed in ORT 3 so performance of the test is not affected. The filtering circuits that are temporarily installed during ORT 3 testing do not affect the operation of any system. structure, or component described in the FSAR. As in previous testing, monitoring takes place at the ORT 3 test racks, Page 17

)

i RK-59A&B. Test po' 1ts 3%se racks are wired to spare relay and breaker contacts and to contacts in the ATWS mitigation sys cv actukion circuitry which are isolated by procedure. The temporary filter circuits cannot affect any SSR ech.ed with the AMSAC, spare relay, or spare breaker contacts as they are not part of the active circuit. Sina k performance of ORT 3 is not changed and no plant circuitry is afTected, the probability of occurrence and the cesequences of an accident or malfunction of equipment previously evaluated in the FSAR are not increased The change does not involve an unreviewed safety question.

(SER 96-131)

24. QRT 3C, Auxiliary Feedwater System and AMSAC Actuation Unit 1. Revision 0. (New Procedure)

ORT 3C is the same test that was previously performed as a part of ORT 3A or B. The test primarily requires ,

Operations support for the auxiliary feedwater (AF) testing and does not require the extensive system alignments of ORT-3A&B.

Summarv of Safety Evaluation: This test is performed with Unit I in cold or refueling shutdown. Motor starting duty limits are incorporated into the procedure. Other than setting these precautions and limitations and adding the steps for a dedicated operator because there are no LCOs, the procedure is the same as that performed as a part of ORT-3A&B.

Testing of the auxiliary feedwater system and AMSAC actuation is not an initiator for accidents previously evaluated in the FSAR. The motor-driven auxiliary feedwater pumps (AFPs) remain operable to Unit 2 during l the test. During a portion of he test, the discharge motor-operated valves (MOVs) to Unit 2 for the motor-driven AFPs, are opened and verified to shut on a Unit I actuation signal. A caution verifies that the valves to Unit 2 are shut and provides operator action to shut them if necessary to limit the discharge of cold auxiliary feedwater to the Unit 2 steam generators. This testing does not increase the probability of a -

malfunction of equipment important to safety. The testing does involve repeated starts of safety-related equipment. Appropriate restrictions are incorporated into the procedure for motor starting duty times to ensure the AFPs are not adversely affected. The testing is routine within the normal design limits of the equipment.

The AFPs remain operable to Unit 2 throughout the test. They are not required to be operable to Unit I while it is in cold or refueling shutdown. 'Ihe margin of safety defined in the Basis for the TS is not reduced. The testing does not involve an unreviewed safety question. (SER 96-020)

25. PBTP-041. flydrostatic Test of Unit 2ilX-1 A Steam Generator, Revision 1. (Permanent)

PBTP-042, Ilydrostatic Test of Unit 211X-1B Steam Generator, Revision 1. (Permanent) i The change adds testing for P-38A auxiliary feedwater discharge path to 2ilX-l A and testing for P-38B auxiliary feedwater discharge path to 2ilX-1B for surveillance testing as required by TS 15.4.8. It also adds testing for main feedwater check valves seat leakage; adds guidance for controlling condentate storage tank temperatures during the test; adds motor starting duty limits of P-38A&B; and changes the hydrostatic test point fill connection from main feedwater piping to auxiliary feedwater piping.

Summarv of Safety Evaluation: The tests are performed at the appropriate step in the procedure so that the tests -

reflect previous test parameters and satisfy ASME Code Section XI testing criteria. In addition, the discharge path check valves for 2P-29 to 211X 1 A and 2HX-1B are seat leakage tested to satisfy ASME Code Section XI requirements. The tests satisfy the requirements for surveillance testing of the P-38A discharge flow path to 2ilX-1 A and P-38A discharge flow path to llX-1B as required by TS 15.4.8. Also, while P-38A or P-38B are utilized during the test, an LCO is not required for Unit I since the Unit I discharge path MOVs are in l automatic and the Unit 2 discharge path MOVs are in manual. With the discharge path MOVs in automatic control for the Unit I the s alves properly align to Unit 1 (open) and Unit 2 (shut) if required to respond to Unit 1. In adJition. the temperature of the north condensate storage tank (CST) during the test is maintained

~

at s I10"F. this temperature limit maintains the suction water of the auxiliary feedwater pumps within the design hasis udues as es aluated in the accident analy ses. It ensures that the auxiliary feedwater pumps suction l

Paue 18 l

1 l

supply does not deviate from the temperature limits required for the suction of the auxiliary feedwater pumps, ;

which is 32 F to 120 F. l 1

The hydrostatic test rig installation point is changed from the main feedwater line to the auxiliary feedwater line. The change has no efTect on the performance of the test. Meving the hydrostatic test rig allows for easier transition from utilizing P-38A and P-38B to the hydrostatic test rig. A procedural temporary modification is utilized to cover the connection to valve 2AF-63 which is the make-up point for the hydrostatic pump. This valve is on the suction piping of 2P-29 which is out of service during the test, but the valve is returned to service as needed for this test and then taken out of service. In addition, the main steam line supports in containment are not required to be pinned when filled with water for the test due to the installation of j MR 96-026 which upgrades the supports. This does not alter or affect an SSC since it is within the design l capabilities of the SSCs utilized during the test.

The changes do not alter the original intent of the procedure or adversely affect an SSC. The margin of safety as defined in the Basis of the TS is not reduced. The change does not involve an unreviewed safety question.

~

(SER 96-132)

26. PBTP-043. Verify Selected I A05 Loads at increased Frequencies, Revision 0. (New Procedure)

The procedure verifies proper operation of P-38A motor-driven auxiliary feedwater pump at >60 Hertz operation. The test data is used to ensure Train A EDG loading sequence and the resulting Train A bus frequency is acceptable for P-38A AFP operation.

Summary of Safety Evaluation: Unit I is in refueling shutdown during the test. An LCO for normal offsite ,

power is entered for the 1 A-05 bus. The probability of occurrence of a loss of offsite power is not increased. l Equipment operated at elevated frequencies are operated within its design capacity. Procedure controls are in place for switching operations. The probability of a malfunction of equipment important to safety is not increased.

Required equipment is operable to mitigate the consequences of an accident or malfunction of equipment previously evaluated.

No new accident initiators are created by this test and no new failure mechanisms are created. Train B decay ;

heat remova'(DHR)is in service. The possibility of an accident or malfunction of equipment of a different I type is not created. l Appropriate LCOs are entered when required equipment to support Unit 2 is inoperable. The margin of safety

is not reduced. This test does not invoh e an unreviewed safety question. (SER 96-022)

27. PBTP-043. Use of Dedicated Operator for P-38A Motor-Driven Auxiliary Feedwater Pump Discharge Vahe AF-4012 to Control Discharge Flow, Revision 0. (New Procedure)

A dedicated operator is utilized to control the discharge flow of the Train A P-38A motor-driven auxiliary l feedwater pump (AFP) per procedure. This prevents the pump tripping on overcurrent during a high flow j condition with the emergency power supply operating at a high frequency. The use of the dedicated operator allows the pump to be returned to service and the LCO exited. The dedicated operator ensures the flow is limited to 200 gpm which is the design flow assumed in the FSAR accident analysis.

i Summary of Safety Evaluation: The use of the dedicated operator allows P-38A AFP to be returned to service and the LCO exited. The dedicated operator ensures the flow is limited to 200 gpm which is the design flow assumed in the FSAR accident analysis. The dedicated operator has written instruction to operate the P-38A flow control valve AF 4012 to limit total flow (including recirculation flow) after an automatic auxiliaiy feedu ater actuation to 200 ppm after loads have sequenced on the emergency power source (approximately l

Page 19

I 2 minutes) to prevent the potential trip of the AFP should it be running at higher flow with the emergency l l

power supply operating at a high frequency. After the AFP recirculation valve has shut (after approximately 3 minutes), the auxiliary feedwater flow is reverified to be s200 gpm. )

I This requirement for a dedicated operator remains in effect until resolution of the P-38A AFP motor breaker tripping on overcurrent is complete.

, Operation of the auxiliary feedwater system is not an initiator for an accident assumed in the FSAR. The use of ;

the dedicated operator to limit flow allows the return of the AFP to service. The probability of occurrence of an accident previously evaluated in the FSAR is not increased.

Use of the dedicated operator to limit flow precludes any potential tripping of the AFP due to being supplied by .

an emergency power source operating at a high enough frequency to cause a high enough load on the AFP causing it to trip. The use of the operator decreases the probability of the pump tripping when required to operate under these conditions. Therefore, the probability of a malfunction of equipment important to safety is ,

not increased.

The use of the dedicated operator allows the return of P-38A AFP to service to aid in mitigating the effects of an accident previously evaluated in the FSAR. The dedicated operator ensures that when the AFP is returned to '

service, the discharge flow is limited to the design flow assumed in the accident analysis and this precludes the possibility that the AFP may trip when delivering a higher flow while operating off an emergency power supply which may be operating at a high enough frequency to cause the AFP to attempt to deliver a higher than i required flow and trip on overcurrent. ' Die operator limits total flow (including recirculation flow) to approximately 200 gpm afler actuation of auxiliary feedwater and is not replacing automatic actions required to respond to an accident. This is completed after loads have sequenced on the emergency power source and the frequency stabilized (approximately 2 minutes). The AFP will not trip on overcurrent due to high flow /high frequency conditions during this time because the protection on the breaker for the AFP that will trip the breaker under the conditions identified and anticipated will not trip the breaker until at least 250 seconds. Flow of 200 gpm has been shown to be adequate to mt trip the P-38A AFP at the elevated frequencies. Afler the AFP recirculation valve has shut (after approumately 3 minutes), auxiliary feedwater is reverified to be approximately 200 gpm. These instructions do not supersede the requirements of the emergency operating procedures.

The margin of safety as defined in the Technical Specifications is not reduced as the AFP is restored to complete operable status within the use of a dedicated operator. This does not create an unreviewed safety question. (SER 96-023)

28. RMP 9043-1, Emergency Diesel Generator G-01 Train A Preventive Maintenance, Revision 0.

(New Proculurs)

RMP 9043-11, Emergency Diesel Generator G-012-Year Electrical Inspection, Revision O. (New PIEEnh!rs) i 141P 9043-12. Emergency Diesel Generator G-01 Safety-Related Protective Relay Calibration. Revision O.

(New Procedure)

EMP 9043-13. Emergency Diesel G-012-Year Mechanical Inspection, Revision 0. (New Procedurs)

RMP 9043-14. Emergency Diesel G-016- and 12-Year Mechanical and Electrical Inspection, Revision 0.

(New Procedure)

RMP 9043-16. Emergency Diesel Mini-Power Pack Inspection, Revision 0. (New Procedurs) !

The RMP 9043 series are the G-01 liDG maintenance procedures. They replace RMP 43 Few of the ac'tual phy sical maintenance acin ities associated with this procedure changed Many programmatic enhancements Page 20 E_______ _ _ _ _ _

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were incorporated as well as a complete technical review of TS, FSAR, drawings and CH AMPS with format upgrade and human factoring steps including a list of prerequisite component and system conditions. ,

Procedures were made unit and train specific. RMP 9043-16 is common to all the EDGs and covers power !

pack inspections.

Safetv Evaluation Summary: T'e activity is required by TS. Testing methodology is described in FSAR Section 15.4.6 TS 15.3.7.B.I.f scecifies the required normal and standby emergency power requirements for 1 A-05/1B-03 and 2A-06/28-04 and actions to be taken in the event that the specified power supplies are unavailable. RMP 9043-1 lists this TS and ensures compliance with it. The procedures do not restrict access to vital areas or impede actions to mitigate the consequences of an accident since existing testing methods have demonstrated its effectiveness and have not been altered via this activity. Radiation dose to the general public

. is not increased by this procedure. The changes do not involve an unreviewed safety question. (SER 96-005) ,

29. RP-I A, Preparation for Refueling, Revision 42. (Temocrarv)

~

RP-1B. Recovery from Refueling, Revision 32. (Temnorarv) 1 The procedures were revised for use during the U2R22 steam generator replacement refueling outage during the defueled phase. The reactor vessel upper internals and head are installed in/on the vessel for storage, then removed to its normal storage locations to allow core on-load. The procedures are used for the movements.

Refueling operations TS requirements were deleted; steps that were accomplished prior to core unload were modified to ensure required plant conditions exist, and steps that are not to be performed until after core reload were deleted.

Summarv of Safety Ev.eluation: TS 15.3.8 refueling requirements apply only during refueling operations.

There are no refueling operations during performance of these temporarily changed procedures. No operations are conducted nor do conditions exist that are contrary to FSAR.

The accident of concern is a fuel handling accident in containment. Since the reactor does not contain fuel

! during the performance of these modified procedures, and the RCS is in refueling shutdown condition and open to the containment atmosphere, procedure use does not increase the probability of occurrence, nor the consequences of an accident previously evaluated in the FSAR nor does it create the possibility of an accident of a different type than previoudy evaluated in the FSAR.

Since the core is off-loaded and the normal procedure and equipment for the movement of the head and upper internals is used, the use of these modified procedures does not increase the probability of occurrence nor the consequences of the malfunction of equipment important to safety previously evaluated in the FSAR. It does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.

The use of the modified procedures during the defueled phase of U2R22 SGRP refueling outage does not involve an unreviewed safety question. (SER 96-116)

30. RP-IC, Refueling Revision 36. (Permanent)

The revision to RP-lC adds steps to require analysis of sufficient shutdown margin during core shuffle and analysis to verify sufficient heat removal capacity from the spent fuel pool (SFP) cooling system. The changes meet commitments made to the NRC in response to Bulletin 89-03," Potential Loss of Required Shutdown Margin during Refueling Operations" and verifies bounds of the analysis for re-racking the SFP. The change also removes a requirement for core shuffles that apply to core offload and reload.

a! e

Summarv of Safety Evaluation: The changes do not aerect the operation of the chemical and volume control system (CVCS) or other system that delivers water to the reactor coolant system (RCS). There is no increased probability that dilute water can be delivered to the RCS. The changes do not change the manner in which fuel I is handled. Therefore, the probability of the accidents previously evaluated in the FSAR are not increased because the probability of the initiating events have not changed.

The two accidents evaluated in the FSAR that pertain to refueling activities include dilution during refueling and fuel handling accidents. The results of the dilution accident show that operators have sufficient time to respond to a dilution to prevent criticality. The changes continue to maintain the proper accident initial conditions such that the response time for the operators is not decreased. The changes do not alter the manner in which fuel is handled. Therefore, the consequences of accidents previously evaluated in the FSAR are not .

l increased because the results of the accident analyses have not changed. -

l The change does not increase the consequences of a malfunction of equipment important to safety. The function of equipment necesary to mitigate a fuel handling accident is not affected by this change. The i function of the nuclear instrumentation system and CVCS is not affected by this change nor does the change affect the ability of these systems to mitigate a fuel handling accident or affect the ability of operators to respond to a dilution accident.

The change las no effect on equipment important to safety except to ensure that only a minimum number of RCCAs are removed from the core. This ensures shutdown margin greater than 5%. No other systems are affected by these changes. No new accident scemtrios, failure mechanisms or limiting single failures are

introduced by the changes.

TS 15.3.8 requires refueling boron concentration >l800 ppm. The Basis for this requirement maintains j shutdown margin greater than 5% in the cold conditions wita rods inserted. The Basis also states that this I

boron requirement maintains the core subcritical with RCCAs removed from the core. The change requires that l

additional analysis be performed prior to refueling to ensure that the 5% shotdown margin is maintained for intermediate fuel and RCCA moves during the shuffle sequence. Therefore, this change ensures that the margin l of safety as defined in the TS is maintained during refueling. The changes do not pose an unreviewed safety question. (SER 96-019)

31. RE-2. Dry Cask Loading and Storage, Revision 3. (Temnorary) l RP-7 Part 6, Preparing a Multi-Assembly Scaled Basket (MSB) for Storage in a Ventilated Concrete Cask j (VCC), Revision 5. (Temporarv) red. Unloading the Multi-Assembly Scaled Basket. Revision 2. (Temnorarv)

RP-8 Part 4. Placing the MSB Transfer Cask (MTC)into the SIP. Revision 3. (Ismporary) .

The procedures were revised to safely place the fuel assemblies in the Multi-Assembly Sealed Basket (MSB) back into the spent fuel pool.

i

! Summary of Safety Evaluation: The changes include steps to remove Hammable vapor from under the MSB, to

! remove the shims, and restore the MSB shield lid to its proper orientation of the support ring. The MSB L Transfer Cask (MTC) was not impacted by this event and continues to be used as intended in its original design.

l l The intended design and physical configuration of the MSB and MSB shield lid are not changed as a result of the change to the procedures or implementing the procedure changes. The procedure changes restore the MSB

. shield lid to its original location and does not affect any of the shielding characteristics of the MSB. Therefore. ;

the changes do not create an unreviewed safety question or ensironmental in. pact. $FR 96-044) )

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l- l l

Pee 22 l

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

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32. Eb8, Unloading the Multi-Assembly Sealed Basket (MSB), Revision 3. (Permanent) .j

~

RP-8 Part 4. Placing the MSB Transfer Cask (MTC) into the Spent Fuel Pool, Revision 2. (Pcrmanent)

I The procedure changes provide direction during the unlikely event that the MSB has to be returned to the SFP and maintained in the SFP for an extended period of time and eventually unloaded. The changes relate to returning the MSB/MTC to the SFP for an extended period of time while the Unit 2 core is ofnoaded during U2R22. .

Summary of Safety Evaluation: The only fuel handling accident that applies to activities that occur in the SFP is when a fuel assembly or control rod cluster is dropped onto the floor of the reactor activity or spent fuel pool.

Placing the MSB/MTC into the SFP does not require handling fuel assemblies. Therefore, there is no .

possibility that a fuel assembly would be dropped to the floor of the SFP.

Since the MSB/MTC has no impact on the fuel handling equipment function, operation or design, placing the [

MSB/MTC into the SFP does not increase the probability of occurrence of an accident previously evaluated in

' the FSAR. ;

When fuel is to be moved from the MSB to the SFP, the worst case accident would be if the fuel assembly struck a sharp object. This is the accident analysis in the FSAR. The activities assumed in the analysis are those found in the highest specific power fuel element in the core removed from the reactor vessel after a 100 -

hour decay period following operation at 1518 Mwt for a full core cycle. All of the elements in the assembly ,

were assumed to break. The analysis concluded that doses at the site boundary control center are well below the 10 CFR 100 limits.

De selection of the fuel assemblies stored in the MSB is limited by the requirements found in the Certificate of Compliance. Since these limits require that the fuel in the MSB be a minimum of 5 years old, and have a .

maximum initial enrichment of 4.2 wt% U-235, and only one assembly be moved at a time, the consequences I of a fuel handling accident w hile moving that fuel assembly are bounded by those analyzed in the FSAR.

Therefore, placing the MSB/MTC into the SFP is acceptable and does not increase the consequences of an accident that was previously evaluated. He changes do not involve an unreviewed safety question. -

t 10 CFR 72.48 Evaluation Summarv: During the unloading procedure, equipment important to safety is handled or used as designed and in accordance with the VSC-24 SAR. The handling of the MSB, and MTC and its various components is done in accordance with existing procedures based on the VSC-24 SAR. The intended operation of equipment important to safety is not affected by placing the MSB/MTC into the SFP. Based on this evaluation, the evolution of placing the MSB/MTC into the SFP does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the ISFSI licensing basis.

l Placing the MSB'MTC into the SFP does not affect the ability of the MSB to perfonn its intended function.

L The events that hase been postulated to occur do not impact the ability of the MSB to perform its function.

. Therefore, placing the MSB/MTC in the SFP does not affect the margin of safety as defined in the ISFSI licensing basis documents. This evolution does not require license conditions to be changed because the I j Certincate of Compliance is not affected. l l .

l The potentici for an increase in occupational exposure is minimal as a result of placing the MSB/MTC into the !

[ SFP, Therefore, the placing of the MSB/MTC into the SFP does not create the possibility of a signincant ;

i increase in occupational exposure than previously evaluated in the ISFSI licensing basis document.

Placing the MSB/MTC into the SFP for an extended d of time, and its associated procedure changes do not present an unreviewed safety question, signincantly increase occupational radiation exposure, create a signiGeant unreviewed environmental impact, or change the license conditions contained in the Certi0cate of l Compliance. GER 46-115) i e

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l l'. ige 21 I

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Eummarv of Safety Evaluation: The revision addresses the need to place new fuel into the SFP prior to placing it in the reactor vessel. SER 96-115 had new fuel going directly from the new fuel vault to the upender and into the reactor vessel. Eight (8) of the new fuel assemblies are required to be placed into the SFP during the core reload so that the core can be loaded with rod cluster control assemblies (RCCA). Additionally, more than 8 r

new fuel assemblies may be placed in the SFP, as necessary, to facilitate core reload. If the need arises to p; ace the MSB/MTC into the SFP during the core reload, the new fuel that has been placed into the SFP is returned to the new fuel vault prior to placing the MSB/MTC into the SFP. The new fuel will be .; turned to the vault as contaminated prior to placing the MSB/MTC into the SFP. (SER 96-115-01) l

33. RP-8 Part 4. Placing the MSB Transfer Cask (MTC) into the Spent Fuel Pool (SFP), Revision 3. (Temnorarv)

This temporary change weighs the MSB shield lid prior to placing the . 'UMSB back into the spent fuel pool -

(SFP) for unloading the MSB. It is necessary to know the correct lid weight to ensure that the correct slings, with appropriate safety factors, are used to remove the lid.

Summary of Safety Evaluation: The PAB crane is used to support the weight of the shield lid during these evolutions. The power to the crane is removed during the lift in accordance <ith the danger tag procedure, providing positive control over the crane power supply. The shield lid is manually raised using chain falls only, a maximum of 1" The crane operator is in continuous communication with a person stationed at the crane power disconnect.

The intended design and physical configuration of the MSB and MSB shield lid are not changed as a result of the change to the procedures or implementing the procedure changes. Therefore, the changes do not create an unreviewed safety question or environmentalimpact. (SER 96-045)

34. RP-b Part 4, Placing the MSB Transfer Cask (MTC)into the Spent Fuel Pool (SFP), Revision 3. (Temnorarvi i

This temporary change adjusts the !cngth of the slings and levels the shield lid prior to placing the MTC/MSB back into the SFP for unloading the MSB. These slings ensure heavy load requirements are met.

Summarv of Safety Evaluatign: The PAB crane is used to support the weight of the shield lid during these evolutions. The power to the crane is removed during the lift in accordance with the danger tag procedure, providing positive control over the crane power supply. The shield lid is manually raised using turnbuckles only, a maximum of 1" The crane operator is in continuous communication with a person stationed at the crane power disconnect. ,

The intended design and ability of the MSB and MSB shield lid to perform its intended functions is not changed as a result of the change to the procedures or implementing the procedure changes. Therefore, the changes do not create an unreviewed safety question or significant environmental impact. .'SER 96-046)

35. SEP-2.1. Unit l&2, Shutdown LOCA with RllR Aligned for Low Ilead injection, Revision 2. (Permanent) !

SEN2 Unit l&2, RilR Aligned for Decay 11 eat Removal, Revision 0. (New Procedure) l l

SEP-2.1 considers more detail in the RilR system lineup for the expected entry conditions of Sl accumulators -

isolated, RCS temperature <400'F, and RHR aligned for low head injection. The previous revision considered RiiR aligned for either low head injection or decay heat removal. The option of RiiR aligned for decay heat removal was moved to procedure SEP-2.2.

Si termination criteria was changed as a result of recalculating EOPSTPTs [E.16) and E.15. These values changed from [270 FJ 310"F to [280*F] 325*F as the result of a calculation updating the Westinghouse Owners Group Si reduction computer code using errors from PBNPs new setpoint methodology.

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Page 24

Manual starting of an Si pump has been changed to preferentially start the Train B SI pump first and the Train A Si pump with discharge through the discharge cross-connects as a contingency. This provides two flow paths to the core which is provided by the Train B SI discharge. This reflects shutdown guidance provided in OP-3C.

Another change in RHR system configuration is the split mode operation to allow the technical support center an option to separate RilR trains for one to be used for RiiR and the other to be used for low head injection.

This is similar to steps in OP-7A.

Summarv of Safety Evaluation: The change to split Train A and B RilR system operation allows for using one RiiR train for injection and one for RilR provides for further cooldown while maintaining core cooling. This may reduce radiological consequences of an accident. Leakage from both RiiR trains is maintained in accordance with TS 15.4.4.IV.

The RHR system is operated with cross-connects shut. Since either one or both RiiR trains may provide core cooling by design, flow through the core is not compromised by this change. Train A, placed on RilR, will

~ '

control cooldown, and Train B will control inventory while in the injection mode. This change does not involve an unreviewed safety question. (SER 96-017) l

36. SLP 1, Unit 1, items Lifted by Containment Polar Crane, Revision 8. (Permanent) [

SLP 2. Unit 2, items Lifled by Containment Polar Crane, Revision 8. (Permanent)

SLP 10, Load Weight Listings, Revision 8. (Permanent)

The safe load paths document moving and lifting the RV jib crane mast (but not the crane itself), personnel platform, and base assembly within containment.

Summary of Safety Evaluation: he probability of an accident previously evaluated in the FSAR is not increased. For example, the probability of a design basis LOCA is not increased because the safe load paths are only used when the reactor is in cold shutdown and the RVjib crane is only used during shut down periods.

The changes do not affect any aspect of the accident scenarios identified in the FSAR. Furthermore, the movement of the RV jib crane mast assembly is in accordance with the heavy load movement criteria and NUREG-0612. The safe load path selected for the assembly is the same as one of the safe load paths proscribed for the reactor vessel missile shield plug and was designed to minimize possible imoact to other systems. The plug weighs considerably more than the mast assembly. The changes do not increase the probability or severity of an accident. ,

1 The change only affects activities undertaken during refueling outages. The load path selected does not pass over equipment important to safety during outage periods and therefore does not affect the margin of safety in l the Basis 08 is This change does not involve an unreviewed safety question. (SER 96-125)

37. SMP 1181 and SMP 1182,4 kV Under Frequency Channel Functional Test, Revision 0. LNew Procedures)

The procedures perform channel functional tests of 4kV underfrequency relays per TS Table 15.4.1-1, Item #12, for Unit I and Unit 2.

. \

Summary of Safetv Evaluation: The testing places the A-01/A-02 underfrequency relay channels into the trip condition, one at a time, to ensure that the channel is operable for reactor coolant pump (RCP) trip initiating action. Since the underfrequency trip contacts are wired in parallel with undervoltage trip contacts, and since the undervoltage contacts are verified to physically trip respective RCP breakers, performing voltage checks verifies underfrequency channel input to RCP breaker trip coils are sufficient to meet the required TS channel functional test.

Pay _M J

It is not anticipated that this testing will trip an RCP, however, PBNP is evaluated for a loss of reactor coolant flow should one or both RCPs inadvertently trip during this testing. Voltage checks are in place to preclude an inadvertent RCP trip from occurrmg. Only one relay channel will be tested at a time. This testing does not constitute an unreviewed safety question. (SER 96-014)

38. SMP 1183 and SMP 1184. Change in G-01/G-02 EDG Governor Settings, Revision 0. (New Procedures).

The evaluation reviews the safety impacts of changing the governor settings on emergency diesel generators (EDGs) G-01 and G-02. The changes reduce the generator frequency when lightly loaded, while still maintaining the frequency above 60 liz when loaded to 2850 kW. This improves the situation in which an auxiliary feedwater pump (AFP) tripped on overcurrent during performance of ORT-3. liigh frequency was one of the contributing factors and this change reduces the frequency effect. An engineering evaluation -

assesses the amount this change will improve the condition and evaluate whether additional changes are necessary. Changes are as follows:

Set the droop characteristic to 3% from 0 to 2850 kW (i.e., the speed will" droop" of 3% when the load is increased from zero to 2850 kW). The existing droop characteristic is 4% from 0 to 2500 kW.

Change the no load speed setpoint range to 927-936 rpm. The present setpoint range is 940 to 950.1 rpm.

Summary of Safety Evaluation: These changes result in the initial no load speed of each of the EDGs, when operated in an isolated mode, to less than that which would exist with the present governor settings. The frequency of the ac power produced by the EDGs would, however, remain above the nominal value of 60 Hz for loads >2850 kW on each of the EDGs. Since the maximum EDG loading which would be present on either of the EDGs under design basis conditions (as described in FSAR Section 8.2.3) is 2728 kW, the minimum frequency supplied to all safeguard equipment supplied from the EDGs (as defined in FSAR Table 8.2-1) would be greater than or equal to the nominal value of 60 liz. The maximum frequency of the power present at each of the loads is less than that which would exist with the present settings. Maximum EDG loading is determined in Calculation N-91-016, Revision 1. These changes have no effect on the ability of the EDGs to start and accelerate the required safeguards loads (as described in FSAR Section 8.2.3). Calculation N-93-097 documents the capability for EDGs G-01 and G-02 to stan and accelerate the required loads with a nominal EDG output frequency of 60 liz. With the new settings, EDG frequency remains at or slightly higher than 60 liz for all postulated loading scenarios. Starting of these loads with EDG output frequency slightly higher than the nominal value of 60 Hz does not negatively affect this capability.

Changes to the settings result in the EDGs being more sensitive to changes in the EDG speed controller w hen an EDG is operating in parallel with the offsite power source. This results in larger load changes for a given change in the speed controller. It is not anticipated that this change significantly affects the ability to control the EDG load levels when operating in this manner. In addition, the EDGs are only operated in this manner when conducting testing (at w hich time they are considered out of sers ice) or for a s ery short duration w hen recovering from the loss of normal supply to the associated safeguards buses.

Changes in the governor may have an effect on voltage regulation for the EDG. This effect is minimal and I

does not affect operation of the EDGs. -

The changes do not increase the probability of, nor the consequences of, an accident previously evaluated in the l FSAR. It does not increase the probability of, nor the consequences of, equipment important to safety. It does ;

not create the possibility of an accident of a different type or a malfunction of equipment of a different type l than that previously evaluated in the FSAR. Lastly, the changes do not reduce the margin of safety defined in j the Basis of Technical Specification. The changes do not involve an unreviewed safety question.

(SliR 96-021)

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39. SMP 1185. Replace CVCS liUT Outlet Valves BS-Il13, BS-1126, BS-ll27, Revision 0. (New Procedure) l The procedure replaces the chemical and volume control system (CVCS) holdup tank (HUT) outlet valves on all three tanks. The butterfly valves are replaced with ball valves. The new ball valves are welded in plare. 1 Summary of Safety Evaluation: Replacement of these valves with welded-in ball valves eliminates the gasketed Dange connection and eliminates leakage. A freeze seal is used to isolate the valves from its associated HUT The freeze seal is used rather than draining the tank to prevent off-gassing of the liquid in the l tank. It is difGcult to purge and vent these tanks to allow for a safe atmosphere due to radioactive gas and the potential of hydrogen to be present in the area of welding.

Flooding in the primary auxiliary building (PAB) as a result oflosing the freeze seal could affect important to safety equipment. The RHR pumps and heat exchangers are directly below the work area. There is no possibility of the water Dooding on the PAB El 8' as there are Door drains and grating in the direct area of the valves. The RHR pumps are within its own cubicles and are protected from Dooding. Contingencies include having pipe plugs in the area and the limited amount of water in the tanks. NDE inspections, use of a contractor with cxtensive experience in freeze sealing, and adequate pipe support were implemented to minimize the risks of this work.

In the unlikely event of the loss of the freeze seal or a tube rupture at the freeze seal location, there is the potential to cause an unscheduled release of radioactive gas within the PAB. If a release occurred, it would be monitored by the PAB ventilation. A pipe rupture at the freeze seal was evaluated. No efi%ts are planned to plug a rupture at the freeze seal within the HUT cubicle should one occur. There is potential for oxygen depletion due to the nitrogen pplied for the freeze seal and the potential for a radgas release from the ruptured pipe. No one shall enter the cubicle until the atmosphere allows entry without SCBA. The tank will drain into the cubicle and be contained within the cubicle. Radgas will escape the tank once the water has drained. The cubicles are kept at a negative pressure relative to the surrounding areas. The release shall be drawn i N the ventilation which is in the cubicle and monitored by the PAB ventilation system. The tank overpressure shall be limited to less than I psig prior to freeze sealing to limit the potential release in the unlikely event of a tube rupture or loss of freeze seal. The waste gas system shall be isolated from the tanks. A gas sample shall be taken prior to initiating the freeze seal to determine the potential release and any HP posting requirements in the event of tube rupture. The tank can be purged with nitrogen and processed through the waste gas system.

The process does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The process does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety in ;

the TS is not reduced. (SER 96-038) l hmmary of Safety Evaluatiom: Sampling the CVCS IlUT gas spaces prior to the planned activity is not necessary. In the unlikely es ent of a gas leak, a sample taken before work has started would not reduce the

- consequences of a release nor provide information regarding postings. The PAB vent stack RMS monitor would trend the release and subsequent vent stack sampling would assist in quantifying the release. Also, the j work area would be evacuated and subsequent entry would be based on current air samples and surveys. l

. Therefore, sampling the CVCS HUTS prior to the planned maintenance is not necessary. (SER 96-038-01)

40. SMP 1186. IX-03 Transformer Bushing Repair, Revision 0. (New Procedme)

SMP 1186 removes IX-03 transformer from service to facilitate repair to one ofits bushings and to return j IX-03 back to service.

f hmmarv of Safety Evaluation: Normal offsite power for both units is altered to perform maintenance on iX-03. Pow er for both units comes through 2X-03. In addition, G-05 combustion turbine is operating to proside the other source of power required by TS 15.3.7.A.I.b. The probability of a previously evaluated accident is not increased Pqe27

l I

h if G-05 becomes inoperable while IX-03 is out of service, Unit I must be shut down in accordance with TS 15.3.0. If a design basis accident occurred on Unit 2 during the shutdown of Unit I,2X-03 is adequate to

! carry the shutdown loading for one unit and the accident loading on the other unit.

! If 2X-03 becomes inoperable while IX-03 is out of service, one of the units is shut down and the other unit's power is reduced to 50% in accordance with TS 15.3.7.B.I.b.

The consequences of a previously evaluated accident are not increased. Removal of IX-03 from service for i maintenance does not increase the probability of a malfunction of equipment or the consequences of a )'

malfunction of equipment as long as G-05 and 2X-03 remain operational.

The electrical system configuration during the procedure is within that allowed by TS. The possibility of a j malfunction of equipment important to safety is not increased. The margin of safety is not reduced. The 1 change does not involve an unreviewed safety question does not exist. (SER 96-052) ,

41. IS-33, Surveillance Testing: Containment Accident Recirculation Fan-Cooler Units (Monthly), Unit I, Revision 13. (Permanal)

TS-34, Surveillance Testing: Containment Accident Recirculation Fan-Cooler Units (Monthly), Unit 2, Revision 16. (Permanent) 01-70, Service Water System Operation, Revision 10. (Permanent)

The procedures associated with containment fan cooler (CFC) flow rates are changed to set the CFC flowrates and low flow alarms as described in Calculation 96-0117. These changes ensure the fluid flashing concerns identified in Condition Report 96-026 do not result in flashing of service water in CFC coils and that the necessary flow is maintained through these coils.

Summarv of Safetv Evaluation: The change in CFC flow rates has an innuence on the containment pressure ,

and temperature profiles resulting from an accident, but has no influence on the probability of accident !

initiation. The service water (SW) system is considered accident mitigating equipment and the reliability of the equipment assumed to function in the safety analysis is not affected by this change.

The reliability of equipment assumed to function in the safety analysis is not affected by reducing the CFC heat removal capability to the equivalent of 1.5 CFCs. The procedure changes are based upon Calculation 96-0117 which takes credit for the reduced CFC performance previously evaluated and accepted by SER 96-055 and utilizes the 3 operating SW pumps which was evaluated and accepted by SER 96-056. This calculation also ensures that setting the Cl C How rates in this manner does not negatively impact other equipment important to '

safety that is cooled by service water. The calculation assures that the necessary Cow is available to the CFC.

G-01 and G-02 EDG cooling heat exchangers, component cooling water heat exchangers, PAD battery room heat exchangers and the auxiliary feedwater pump bearing coolers. The procedure changes ensure that the throttle valves are set so actual flowrates through the CFCs are maintained within the flow rate band established by Calculation 96-0117. The procedures specify a narrower setting band for Hows considering instrument -

inaccuracies so actual now is maintained within the band provided by Calculation 96-0117.

The change results in positiening the throttle valves in a more tightly throttled position. This new throttling position does not negatively impact the system. The valves are designed for throttling so this new position is not expected to has e a degrading effect on the valves. The new throttle positions are very similar to previous settings which were maintained in the past (e.g. prior to 1992) when the valves were set for 1000 gpm with 3 SW pumps running and therefore it is not considered to be of concern with system overpressurization due to the new throttle sats e settings. Silting concerns due to the lower throttle valse settings are not considered to be a concern since normal now through the CFCs is not significantly changed (e.g. logs still specify flow to be "50n gpmi .md since the new now with a CFC return line motor-operated valve (MOy) open is not Page 28

signincantly different from past values. In addition, silting in the CFC supply and r monitored and addressed by the SW radiography program.

TS 15.3.3 and 15.4.5 specify operability requirements and testing frequency for the CFCs.

changed the containment heat removal requirement for each individual CFC and effec operability requirement for CFC heat removal (e.g. 75% of original requirement). The ensure that operability is met at the frequency required by the TS by ensuring the CFC flow rates a This change does not reduce the margin of maintgined within the band specified by Calculation 96-0117.

safety as defined in the Basis of the TS. The change does not involve an unreviewed safet (SER 96-057)

42. Work Package WP 1080A. Temporary Gantry Crane Equipment Chase, Revision 0. (hrmanent)

WP 1080A installs and removes a 5-ton rated temporary gantry crane at El 66' w! thin Unit 2 contain vicinity of the equipment chase during the steam generator replacement project. This crane is

~ handle equipment in the equipment chase and to lift equipment between the equipment chase El 2 operating deck.

j Summarv of Safety Evaluation: WP 1080A is limited to cold shutdown, refueling shutdown, refueling operations, or the Unit 2 defueled condition.

The use of the temporary gantry crane and expected handled loads does not result in adverse effects upon th El 66' slab and its supporting stmetures. The temporary gantry crane is qualified seismically for use during plant fueled conditions including qualification of the containment structure for seismic effects of the total m of the temporary gantry crane and its lifted loads. Temporary gantry crane assembly /use/ disassembly is qualified by design and by implementation requirements such that the temporary gantry crane component d not increase the probability of an accident previously evaluated in the FSAR.

The capability of plant equipment to perform its safety function as described in the Basis of TS is unaffected by WP-1080A. Calculations demonstrate that the containment structural integrity is not compromised. Other than ,

the containment structures, no plant systems or components are affected as a result of the implementation of this work package. There are no adverse effects on safety limits, seismic requirements, setpoints or operating l

parameters as a result of WP-1080A. In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded by WP-1080A. WP-1080A does not reduce the margin of safety as defined in the Basis of TS. The change does not involve an unreviewed safety l question. (SER 96-110)

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l MODIFICATIONS 1 he following modifications were implemented in 1996:

1.

MR 90-015 and MR 904)l5'A. (Common), Emergency Diesel Generators (EDGs). MRs relocate G-01/G-02 EDG sump tank level switches G-01/02-FTS-N and .

m G-01/02-FTS-H their current location on top of the sump tank to a standpipe locatedsociated off the sump tank and its as pad.

Sy.mmary of Saferv Evaluation:

Failure of the level switches could prevent the associated EDG from ~

performing its safety-related function. Moving the level switches to a standpipe does introdu mechanisms. If the isolated valve was left shut, the standpipe would not reflect actual tan prevent starting the fuel transfer pumps when tank level drops below the required point. Likew valves to an operator checklist. The risk of valve misalignme introduce a significant risk of common mode failure of both G-01 a The EDG room Appendix R design basis includes withstanding a fuel o oil fire. There is Also, if the Gre occurred, it would be well w ithin the analyzed d .

The modifications to the G-01/G-02 EDG fuel oil level switches do n.not involve an unr (SER 96-077) 2.

MR 90-048'B,(Unit 2), RCS, MR 90-048*B replaces the existing boric acid flow recorder flow indicator and a reactor makeup water (RMW) flow indicator. The r old totalizer pred were replaced with new digital preset counters. The existing 4-position e selector control s with a new 5-position selector control switch to add a blend function toowthe existing contro cutout circuit was installed to prevent the RMW totalizer from receiving How signals at negli levets. The existing How controllers were refurbished and the existing control system is fine t Summary of Saferv Evaluation:

totalizer counters. and installation of the low Howy tocutout improves reliability a the control system by providing the most accurate flow indication. The improvements to the reactor ma system provide better control of core reactivity and provides better indication if an accident w improves the ability of an operator to respond to an accident. The modi 0 cation does not incre ,

consequences of an accident previously evaluated in the FSAR.

The addition of the blend function improves the ability to perform a blending operation

. The same pumps and there are pressure changes at the RMWsae,control W flow valve 2CV-I valve does not allow undesired RMW flow difference a eup betwee water storage tank does not allow undesired RMW flow to the reactor coolant pumps system when no RMW at both shutdown and at power to indicate aopotential ,

aa e valve is dilution is not alTected by this modification. The new digital preset counters doa unction not introduce not a failure or m lf Page 30

already been evaluated by the FSAR. The same power supplies are used for the new totalizers, and the installation is seismically qualified.' he modification does not increase the probability of occurrence of a malfunction ef equipment important to safety previously evaluated in the FSAR.

During the modification, the chemical and volume control system blender is'out of service as a reactor makeup water flow path or a boric acid flow path. This modification does not affect the abihty to borate or dilute to the Unit I reactor coolant system or to blend to the spent fuel pool via the Unit i blender. The modification does not involve an unreviewed safety question. (SER 96-120)

3. MR 91-116. (Common), EDGs. MR 91-116 installs two additional Class 1E emergency diesel generators (EDGs). This safety evaluation covers the final configuration as described in the diesel project design dated August 1993. Interim configurations and connection to existing systems were evaluated separately. This revision clarifies statements from prior safety evaluations prepared for MR al-116.

Summary of Safety Evaluation: FSAR Tables 8.2-1 and 8.2-2 are revised to reflect the results of Calculation

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N-91-016, Revision 1. All EDGs can be shown to have sufficient capacity without using the 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months load rating. This does not mean that the load limit cannot revert to the 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months load rating if needed. This only shows that sufficient capacity exists on all EDGs at a lower load rating for the injection phase and at the same load rating for the recirculation phase. The NRC safety evaluation for the EDG project states:

"The worse-case loading for any of the four EDGs would occur on the Train B EDGs during the injection phase of a 12 CA on one unit, with the other unit in cold shutdown. This loading is 2902 kW lasting for less than 1/2 hour, which is less than the 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months rating (2952 kW) for the new EDGs.

After the first half hour, the worse case loading is reduced to 2581 kW, which is less than the 2848 kW continuous (2,000 hour0 days 0 hours 0 weeks 0 months ) rating."

The revised maximum loadings still meet the fundamental conclusion of the NRC SER, which is that the

injection phase is within the 200 hour0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months load rating, and the recirculation phase is within the 2,000 hour0 days 0 hours 0 weeks 0 months load rating limits.

When standby emergency power to 1 A-05 is inoperable, G-02 EDG will not be declared inoperable based on the following:

The LCO for I A-05 power is TS 15.3.7.B.I.f, which states,"The normal power supply or standby emergency power supply to Unit 1 A05/B03 or Unit 2 A06/B04 may be out of service for a period not exceeding 7 days provided the required redundant engineered safety features are operable and the required redundant standby emergency power supplies are started within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before or after entry into this LCO and every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter. If the normal power supply is out of service, an operable emergency diesel generator is supplying the affected 41f>0/480 Volt buses. After 7 days, both units will be placed in hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ."

Whether or not G-02 is considered operable, the appropriate LCO requirements are invoked. The use of G-02 operability tojustify the situation ofinoperable required Train B ESF equipment for Unit 2 is not allowed by TS because required ESF equipment in Train B for both units is operable.

Based on this revised safety evaluation of the interdependence of G-02 and 1 A-05 emergency power, if during previous entries into I A-05 emergency power LCO, G-02 was or was not declared inoperable is inconsequential. his is because the applicable LCO for I A-05 emergency power (TS 15.3.7.B.I.f) contains the appropriate requirements fbr the dual unit effects of the inoperability of emergency power to engineered safety features for both units.

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Be conclusion of this safety evaluation that this modi 6 cation is not an unreviewed safety question for the purposes of 10 CFR 50.59 and that the new EDGs and the reconnguration of the emergency power system for PUNP is in accordance with appropriate FSAR, design basis, and licensing basis requirements, remains valid.

(SER 93-021-26)

4. MR 91-116*U. (Common), EDGs. MR 91-116*U provides a more accessible isolation valve at HX-66 auxiliary feedwater pump (AFP) room cooler for maintenance activities. A new valve, SW-105A, is installed in the line downstream of SW-105. The piping and components downstream of SW-105 are presently classined as non-QA Scope, however, the installation must meet seismic Class I requirements.

Summary of Safety Evaluation: The FSAR does not show that the service water system causes or affects the probability of an accident. In the event of a major malfunction in the service water system,it is possible to -

isolate the ponion of the system affected and maintain service water to essential services of the plant. In addition, each component is provided with an individual isolation valve. The installation of a new isolation valve downstream of the existing isolation valve, SW-105, in a location that is more readily accessible for .

maintenance evolutions does not change this. Therefore, implementation of MR 91-116*U does not increase the probability of the occurrence of an accident or malfunction of equipment important to safety.

The consequences of accidents are determined by the results of analyses that are based on the initial conditions of the plant, the type of accident, transient response of the plant and operation and failure of equipment and systems. The service water system is designed to prevent a component failure from curtailing normal station operation. Eech component has individual isolation valves to permit removing any piece of equipment from the system. Adding an additional isolation valve does not change this capability. The new valve materials and design is compatible with the existing system piping and valves. He piping con 6guration with the new valve is seismically adequate. Service water flow to HX-66 is isolated at SW-105 during this modiGeation. Welding is visually inspected and an inservice test of the new valve is performed to ensure that system integrity is maintained. Therefore, the consequences of an accident are not increased by the implementation of MR 91-il6*U.

The margin of safety as defined in the TS Basis is not reduced by the implementation of MR 91-il6*U. It does not a.Tect the number of operable service water pumps, its power supplies nor does it change the ability of the service , 4r system to prevent a component failure from curtailing normal station operation. In addition, each compon a ri o provided with an individual isolation valve. An additional isolation valve does not change this capability. Therefore, the installation of a new isolation valve in a more readily accessible location does not reduce the margin of safety as defined in the TS Basis. MR 91-116*U does not create an unreviewed safety question. (SER 96-002)

I 5, MR 91-116* Al,(Common), EDGs. MR 91-ll6* Al recon 0gures controls for G-01 emergency diesel generator (EDG) on main control board C-02 to match those previously installed for G-02, G-03 and G-04 EDGs installs a new test switch on C-02 MCB rear to facilitate the testing of G-01 metering, and replaces the I&2 A52 73 breaker control switch on C-34A with a key lock switch.

Summary of Safetv Evaluation: Installation of the new G-01 EDG subpanel, G-01 metering test switch and C-34A breaker control switch replacement takes place w hile G-01 is out of service so its safety functions are -

not affected. Installation is performed according to an approved installation work plan to ensure that no other ,

systems or components in panel C-02 main control board are effected. Functional testing is conducted at the l completion of the installation to ensure that the G-01 control circuitry functions as designed. The installation i process and design does not involve an unreviewed safety question. (SER 96-073)

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6. MR 91-134. (Unit 1), Main Steam. MR 91-134 replaces both Unit I atmospheric steam dump valves :

(1MS-02015 on Train B and IMS-02016 on Train A).

Summary of Safety Evaluation: The valves are installed during the Unit I refueling outage with main steam out j of service and the appropriate equipment isolated during installation. A 3/4" connection is installed between l the upstream isolation valve and the steam dump valve to facilitate post-installation hydrostatic testing of the valve.

The atmospheric steam dump valves are accident mitigators, not an FSAR accident initiator. For a general steam line break outside containment but prior to the main steam isolation valves, the upstream isolation valves

. (MS-227, MS-244) are the containment isolatioa valves identified in the FSAR. Ther Ntation valves are not

, altered as a result of this modification. Equipment isolation during installation provides co. 'ainment closure required for a fuel handling accident scenario. The 3/4" hydrostatic tap is installed in accorda. Se with applicable Codes and standards and is downstream of the containment isolation valve. This modification does not increase the probability of occurrence of an accident and does not create the possibility of an accident of a

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different type previously evaluated in the FSAR.

The new valves will flow over the seat, combined with the new trim design that provides an ANSI Class V shut off, and the new valves tighter sealing characteristics than the existing valves. In addition, the new trim design j and material decreases the possibility of steam cutting which can increase steam leakage and reduce seal j tightness. Thus, the modification does not increase the probability of a malfunction of the valve such j inadvertent opening.

The new valves have the same capacity as the existing valves (each pass 5% of flow). Thus, inadvenent I opening of the new valves would not allow more steam to escape; thus, no larger load increase.

The modification does not increase the probability of occurrence or the consequences of a malfunction of I equipment important to safety previously evaluated in the FSAR. The modification does not create the possibility of a malfunction of a different type than previously evaluated in the FSAR. The modification does not involve an unreviewed safety question. (SER 96-013)

7. MR 92-120. (Common), ISFSI. The outside diameter of the MSB shell was specified to be 62.5"+0.0/-0.5" During final assembly, the actual shell outside diameter was measured at two localized areas as being 62.527" and 62.542". Other areas measured were within specification. A spider is used in the MSB top to control diameter during shipment. Installation of welding shims during MSB loading is expected to bring the shell into tolerance.

10 CFR 72.48 Safety Evaluation Summarv: The deviation to the MSB des gn during fabrication does not affect the form, fit or function of the MSB. The principle function of the MSB i, to serve as a high integrity. l leak-tight container. This function is not affected. The deviation from the specified tolerance was addressed by

, the manufacturer to reduce the likelihood of reoccurrence in subsequent MSB fabrications.

The VSC-24 safety analysis report was reviewed to ensure that the structural, thermal and shielding analysis conservatively bound the deviation from the design. The vendor reviewed the deviation and concurred that the design was conservative and the deviation is acceptable.

The change does not pose an unreviewed safety question, significantly increase occupational exposure, create a significant unreviewed environmental impact, or change the license conditions as contained in the VSC-24 Cenificate of Compliance. (SER 46-029)

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8. MR 92-124,(Common), ISFSI. Deviations in the design dimensions of the MSB storage assemblies for Containers 2 and 3 are as follows: The outside diameter per the drawing is 59.2" 10.4. The dimensions after fabrication for Container 2 is 59.82" worst case; 3 worst case is 59.69". The minimum gap between the storage sleeve assembly and the shell wall, to allow for thermal expansion of the storage sleeve assembly, per the SAR is 0.20". The minimum gap between the sleeve assembly and the shell observed during inspection of Container 2 was 0.25" with Container 3 at 0.31" A deviation in the perpendicularity of the storage sleeve assembly was also noted. The perpendicularity of Contained exceeds the requirement of 0.1", by 0.275" j maximum over the MSB sleeve length of 163.6". Each of the sleee assemblies successfully passed the use of l a go-no-go gauge, which simulates the insertion of a fuel assembly. !

10 CFR 72.48 Safety Evaluation Summarv: he minimum gap between the storage sleeve assembly and the shell wall, to allow for thermal expansion of the storage sleeve assemb:y, per the safety analysis report is 0.20". l This minimum gap is sufricient to allow unhindered storage sleeve thermal expansion relative to shell during the off-normal, severe environmental condition noted in Section i1.1.1 of the safety analysis report. Other accidents or postulated events are not affected by the deviation in the MSB storage sleeve diameter or -

perpendicularity. The minimum gap between the sleeve assembly and shell observed during inspection of Container 2 was 0.25". The minimum gap observed during inspection of Container 3 is 0.31" The perpendicularity of container 3 exceeds the requirement of 0.1", by 0.275" maximum over the MSB sleeve length of 163.6". Each of the sleeve assemblies successfully passed the use of a go-no-go gauge, which simulates the insertion of a fuel assembly. He deviation in perpendicularity of Container 3 does not affect the ,

accident scenarios presented in the safety analysis report. The deviations do not affect the structural strength, fundamental geometry, leak-tight integrity or the thermal design characteristics of the MSB. Here are no adverse effects expected as a result of these deviations in the MSB storage sleeve assembly. He change does not involve an unreviewed safety question. (SER 96-030)

9. MR 92-120. (Common), ISFSI. ECR 95-0380 changes the storage sleeve support bar dimensions from 1.45" in '

thickness to 1.35" The support bar runs the length of the support plate,28" There are three support plates located on the multi-assembly storage basket (MSB) sleeve assembly. The change in dimension to the support bar ensures that the overall dimension of the storage sleeve assembly is maintained at 59.2"10.4" 10 CFR 72.48 Safety Evaluation Summarv: The change to the MSB storage sleeve assembly does not affect the leak tight integrity of the MSB assembly. The change to the storage sleeve support bar does not change the criticality analysis, heat transfer ability, or the ability of the MSB to provide support for the fuel.

The consequences of accidents in the VSC-24 safety analysis report are the amount of radiation exposure that could be incurred should an associated accident occur. The VSC-24 safety analysis report accidents were reviewed to determine if this change could cause an increase in the radiation exposure for these accidents. The MSB is the confinement boundary to prevent radiological release. The MSB storage sleeve support bar does not affect leak tight integrity of the MSB. The MSB internak continue to provide appropriate heat transfer, prevent criticality and provide support for the fuel. ,

The change to the storage sleeve assembly support bar has been evaluated and determined that no licensing !

analysis is affected. The safety analysis report was reviewed to ensure that the structural, thermal and shielding analysis conservatively bound the engineering change request from the design. The design is conservative and the change is acceptable.

This change does not pose an unreviewed safety question, significantly increase occupational exposure, create a significant unreviewed environmental impact, or change the license conditions as contained in the VSC-24 Certificate of Compliance. (SER 96-031) l t

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10. MR 92-120. (Common), ISFSI. The evaluation addresses 4 non-conforming conditions identined with MSB 2:

e The structural lid extended beyond the shell edge by a maximum of 0.097" and a minimum of 0.050".

The requirement is that the structural lid be flush with the top of the shell. The disposition required that the shield lid have a recess machined 0.130" deep by a minimum diameter on the bottom of the shield lid such that the shield fits into the shell 0.130".

The shield lid-to-shell assembly did not have the correct gap to allow for the shims to be assembled per the drawing requirements. The shield lid had was machined to 0.125" x 1.25" to provide room for the shim thickness and root gap.

e The shell-to-shim gap was not within the required micrances. The drawing requires that the gap between the shim and shell be 0.030-0.090" The gap was exceeded by 0.045" in some localized areas. The shims were built up in the localized areas to ensure the appropriate gap is maintained.

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. The structural and shield lid were manufactured prior ta the shield lid support ring weldment. This resulted in both of the lids being slightly oversized and the resultant gaps between the lids and shell being too small. This variance applies to MSB 2 and 3.

1.(LCFR 72.48 Safety Evaluation Summary: The changes facilitate fabrication and document minor deficiencies encountered during fabrication. The changes made to the MSB shield lid, structural lid, and shims do not effect the strength, fundamental geometry of the MSB structure, or leak tight integrity of the MSB. The weldjoint configurations have been prequalified to allowable tolerances on the critical dimensions and as long as those configurations are maintained within the tolerances, the weld joint is acceptable. The changes noted result in a configuration that is within the prequalified weld joint and therefore are acceptable. The final weld joints meet the requirements of ASME Section lit, with the shield lid weld joint as a seal weld and the structural lid weld as a full penetration weld.

The changes do not pose an unreviewed safety question, significantly increase occupational exposure, create a significant unreviewed environmental impact, or change the license conditions as contained in the VSC-24 Certificate of Compliance. [SER 96-032)

11. MR 92-120. (Common), ISFSI. During the construction process to produce multi-assembly storage basket storage sleeves, the flat stock material is bent at 90* angles to form the required components for welding into the box shape to store the fuel assemblies. During the bending process, some sleeves have experienced cracking along portions of the outside radius of the bend. The extent of the cracking varies between individual sleeves, but has ranged from no cracking up to several inches along the length of the 163.6" sleeve. The option to repair or discard the sleeve material allows the fabricator to evaluate the economics of completing the repair or discarding the material and replacing it with new material.

. 10 CFR 72.48 Safety Evaluation Summary: VSC-24 safety analysis report Table Il.0-1," Design Basis Off-Normal and Accident Events," provides a list of the events for which the VSC-24 was analyzed and the components affected by each of the events. The component addressed in this safety evaluation is the MSB slave and a change to the sleeve material has the potential to affect the performance of the MSB internals. The events which could affect the MSB internals are listed as (1) Off-normal environmental conditions (100*F w/ solar load and -40*F, no solar load);(2) Blockage of one-half of the air inlets;(3) Off-normal handling I

load-irnpact at 2ft/sec crane speed;(4) Maximum anticipated heat load,125'F ambient temperature and full solar load,(5) MSB drop in a future shipping cask;(6) Earthquake; and (7) Complete blockage of air vents.

The safety analysis report states that the gap between the basket assembly and the wall of the MSB "is sufficient to allow unhindered storage sleeve thermal expansion relative to the shell." Again, because the repair restores the sleese to meet its original design strength, the original analy sis remains s alid.

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The repair ofithe identified cracks returns the sleeve to its original strength, original dimension, and replaces removed material with the same material. This ensures that the structural integrity, the heat transfer

- characteristics, criticality analysis, and the shielding characteristics used in the original design and licensing basis are maittained. Therefore, the probability of occurrence and the consequences of an accident contained in the design t nd licensing basis are not increased, the probability of occurrence and the consequences of a l malfunction orequipment important to safety are not increased. A new accident type or new malfunction is not

. created. The nargin of safety is maintained. There is no increase in occupational exposure, and there is no l unreviewed environmental impact because of this change. None of the requirements contained in the VSC-24 !

Certificate of Compliance are modified because of this change. l

- The change does not pose an unreviewed safety question, significantly increase occupational exposure, create a signincant unn viewed environmental impact, or change the license conditions as contained in the VSC-24 -

I Certificate of Compliance. (SER 96-033) j

12. MR 92-120. (Common),ISFSt. The evaluation addresses 2 non-conforming conditions identified with MSB 3: .

The shield lid support ring is locally oversized in one area. The drawing requires that the shield lid l support ring measure 2.00* 10.05" width and a length such that there is a 0.25" gap between the ends -[

prior to welding in place. The shield lid support ring measures 2.12" worse case over a length of 28". -

The proper ing elevation has been maintained throughout the fabrication to ensure proper height of the j shield lid anJ structural lid. i e The outside diameter of the shell assembly is out of specification. The drawings require the outside l diameter of the MSB shell to be 62.5"+0.0/-0.5". During final assembly, the actual shell outside diameter !

at two localized areas was 62.510" at 180* and 62.509" at 30* Other areas measured were within !

speci6 cation. 'A spider is used in the MSB top to control diameter during shipment. [

'i 10 CFR 72.48 Saf-tv Evaluation Summaw: The safety analysis report was reviewed to ensure that the structural, thermal and shielding analysis conservatively bound these deviations from the design. The design is f

conservative and the non-conformances are acceptable, j t

The deviations do rot pose an unreviewed safety question, significantly increase occupational exposure, create {

a significant unreviewed environmental impact, or change the license conditions as contained in the VSC-24 l Cerificate of Comp iance. (SER 96-034) !

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13. MR 92-120. (Comn.on), ISFSI. The evaluation describes .i deficiency identified during the fabrication of l MSB 3 shield lid assembly. During the final inspection of the MSB shield lid assembly, the side ring wall j thickness was measured to be 0.453" minimum in certain localized areas. The drawing requires the material to i he 0.50" thick stock. The tolerances for the material are +0h30/-0.010" The side ring in the shield lid ;

provides a boundary tto the RX-277, a grout-like material, ano does not proside structural support to the shield ?

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lid while in place.

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10 CFR 72.48 Safety Evaluation Summary: During movement of the shield lid, the shield lid side ring supports the lower portions of the shield lid assembly. There have been no changes to the welding requirements on the -

side ring and therefort, the side ring is capable of supporting the shield lid during transport. The shield lid side ring was not used in dose rate calculations and does not effect the overall shielding ability of the shield lid. The shield lid has no structural significance and is not considered in structural analysis. The accepted limits of the

. shielding calculations tre unaffected by the deviation. in addition, the dimensional, heat transfer, and structural characteristics of the M SB design are unaffected by the deviation.

' The safety analysis repo rt was reviewed to ensure that the structural, thermal and shielding analysis conservatis ely bound tha deviation from the design. The design is conservative and the deviation are acceptable.

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The change does not pose an unreviewed safety question, significantly increase occupational exposure, create a j significant unreviewed environmental impact, or change the license conditions as contained in the VSC-24 i Certificate of Compliance. (SER 96-035) l

14. MR 92-120. (Common), ISFSI. An improper ;ap size for the structural lid on MSB 3 was identified. The i resolution requires machining the lid to achieve an acceptable diameter. This resolution could cause the j structural lid recess of 0.5" to be slightly less than 0.5" The important parameter is that the gap (0.125-0.375")

does not exceed the width of the backing ring, which is also 0.5" !

10 CFR 72.48 Safety Evaluation Summarv: The probability of an accident is based on the probability of occurrence of the associated initiated event. The MSB structural lid change does not affect initiating events for

! , accidents described in VSC-24 safety evaluation report Chapter 11 because the final configuration is unchanged. Therefore, this change cannot increase the probability of occurrence of an accident previously evaluated in the ISFSI licensing basis documents.

The consequences of accidents are based on the type of accident and the function of systems, structures and components that affect the amount and rate of release of radiological material. The change maintains the functional capability of the MSB as required for leak tightness, shielding, heat transfer, and structural integrity.

When the MSB is sealed by welding the MSB structural lid to the MSB shell, surveillances are required by l VSC-24 Certificate of Compliance Sections 1.2.2 and 1.2.9. The surveillances provide assurance that the structural lid to MSB shell weld is satisfactory and provides leak tightness and structural capability. Therefore, this change does not affect the consequences of an accident previously evaluated in the ISFSI licensing basis doc ments, based on the fact that the critical safety parameters for the MSB have been preserved.

Accidents of a different type can only be created by different accident initiators or sequences. This change does not affect the functional capability of the MSB. Therefore, no new accident initiators or sequences have been !

l introduced and hence the possibility of an accident of a different type is not created.

i The change facilitates proper fit-up and welding of the MSB structural lid. Therefore, the time required to l perform this weld and the manner in which the structural lid is placed is not changed. Hence, occupational exposure is not affected by this change.

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The change does not pose an unreviewed safety question, significantly increase occupational exposure, create a significant unreviewed environmental impact, or change the license conditions as contained in the VSC-24 i

Certificate of Compliance. (SER 96-036)

15. MR 92-120,(Common), ISFSt. The time limit for draindown of the MSB following removal from the spent fuel pool is evaluated based on conservative values for the heat-up rate and an initial temperature of the spent fuel pool of 95'F. This methodology for calculating dra;ndown time remains ni place until additional detailed thermodynamic analysis can be completed; such as modeling the system, including the MSB and MTC, using

. the thermodynamic computer code - GOTHIC. For an MSB with 24 kW heat lead, a heat-up rate of 5.9'F/ hour, and a spent fuel pool temperature of 95*F; the 47-hour limit from the CSU l.2.10 would be reduced to 19.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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10 CFR 72.48 Safety Evaluation Summary: Establishing a draindown time limit which is based on more l conservative input assumptions than originally used in the analysis which established the 47-hour time limit in l the VSC-24 Certificate of Compliance, NRC Safety Evaluation Report, and VSC-24 safety analysis report. The l probability of occurrence and consequences of an accident, the probability of occurrence and consequences of a j malfunction, the possibility of creation of a new accident or malfunction, the margin of safety, the possibility of

! an increase in occupational exposure, and the possibility of a significant environmental impact are unchanged.

The use of a more conservative draindown time limit within our procedures provides added assurance that boiling will not occur in the MSB and provides a more conservative limit than allowed by the license condition contained in the VSC-24 Certificate of Compliance. Data show s heatup rates of 0.45 F/hr to 1.21 I /hr for casks inaded with total heat loads of 9.31 kW to 11.94 LW in addition. the data for the first cask loaded at l'q e r

PBNP shows a heat up rate of approximately 1.8'F/ hour for a heat load of 9.7 kW. The data relates to a heatup rate of <5.9'F/hr when linearly scaled to a full 24 kW heat load, allowed in the MSB. In addition, the draindown time limit is based on an initial spent fuel pool temperature of 95'F, which is a conservative assumption for PBNP under normal pool operating scenarios. Therefore, the use of a draindown time limit based on the 5.9*F/ hour heatup rate and higher than normal initial spent fuel pool temperature is conservative and preserves the margin of safety.

The license conditions as contained in the VSC-24 Certificate of Compliance are not changed. Section 1.2.10

" Conditions of System Use," states that the MSB must be drained within 47 hours5.439815e-4 days 0.0131 hours 7.771164e-5 weeks 1.78835e-5 months for the maximum heat load of 24 kW. Until additional analyses or evaluations are performed, the total heat load placed in an MSB will be such that the draindown time limit, assuming a 5.9'F/ hour heat up rate and a spent fuel pool temperature of 95'F, will be maintained greater than or equal to 47 hours5.439815e-4 days 0.0131 hours 7.771164e-5 weeks 1.78835e-5 months to ensure compliance with the VSC-24 Certificate of .

Compliance. This change does not pose an unreviewed safety question, significantly increase occupational exposure, create a significant unreviewed environmental impact, or change the license conditions as contained in the Certificate of Compliance. (SER 96-043) , ,

16. MR 93-025*4. (Common), Auxiliary Feedwater. MR 93-025*4 involves rewiring a portion of the Train A (P-38A) auxiliary feedwater pump (AFP), circuitry to satisfy train separation criteria internal to main control boards (MCB) C-01 and 2C-03. The work involves: Routing a total of eight new Train A wires through Train A conduits / wireways internal t0 the MCBs; declaring P-38A out of service to Unit 1; lifting old wire leads at its termination points; removing old wiring; landing new wires at the correct termination point; and performing tests on circuits rewired.

Summarv of Safety Evaluation: During installation P-38A and 2P-29 (Unit 2 turbine-driven AFP) are unable to perform their design functions. 2P-29 is unavailable because its steam supply motor-operated valve control switches are isolated shut and in pull out. During testing of 2LCAA and 2LCBA, its contacts change state which generates the open (auto-start) signal to steam supply MOVs 2MS-2019 and 2MS-2020. Since this work i plan is performed during U2R22 before Unit 2 is taken critical,2P-29 is not required to be available. The appropriate LCO is entered for P-38A when required by the work plan. !

It is important to note that this work plan does not change circuit functionally or add additional wiring, it only replaces existing wiring, that is not correctly routed internal to the MCBs, with new wiring using the correct MCB train separation requirements. This modification does not increase the probability or consequences of an accident or malfunction or create the possibility of an accident or malfunction of a different type previously evaluated in the FSAR. The margin of safety is not reduced. The installation does not involve an unreviewed safety question. (SER 96-098)

17. MR 93-025*58,(Common), Auxiliary Feed Water Pump. MR 93-025*58 rewires a portion of the Train B (P-388) auxiliary feedwater pump ( AFP). circuitry to satisfy train separation criteria internal to main control ;

board (MCB) C-01. This work insoh es routing a total of four new Train B wires through Train B conduits / wireways internal to the MCBs; declaring P-38B out of service to Unit I and Unit 2; lifting old wire leads at its termination points; removing old wiring; landing new/ existing wires at the correct termination point, and performing tests on circuits rewired. ;

Summary of Safetv Evaluation: During the time period when this work is installed, a LCO is declared taking P-38B out of service. Therefore, P-38B is not available to perform its design function in case of an accident.

The LCO called to remove P-388 from service controls the time frame in which P-38B may be taken out of service to Unit I and Unit 2.

This work does not change circuit functionality or add additional wires; it only replaces existing wires, that are not correctly routed internal to the MCBs, with new wires using the correct MCB train separation requirements.

This work does not increase the probability or consequences of an accident or malfunction or create a possibility of an accident or malfunction of a dif ferent ty pe presiously evaluated in the FSAR. The margin of j safety is not reduced The installation does not insohe an unresiewed safety question. LSliR 96-103)

Page 38 ,

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I 8. MR 93-025*A. (Common), Service Water. MR 93-025'A removes valve 2SW-2907 (HX-15A-D containment recirculation heat exchanger emergency fire control valve) from service by opening circuit breaker 2B52-3210F at motor control center (MCC) 2B-32 and then re-routes / replaces internal MCB C-01 wiring within the control circuits to provide adequate separation between redundant safety trains.

Summarv of Safetv Evaluation: SW-2907 is used to respond to accidents previously evaluated in the FSAR.

The valve safety function in the containment cooler service water discharge lines is to automatically open on a safeguards actuation signal to ensure adequate coolant flow through the containment accident fan coolers.

Evaluations in the FSAR show that the valve does not act as an initiator of these accidents. Therefore, the activity does not increase the probability of occurrence of a previously evaluated previously evaluated.

. The activities are performed with the control circuit for the 2SW-2907 valve de-energized. If additional service water return flow is required, redundant 2SW-2908 valve is not affected by this work and is available for use.

Either valve is capable of passing the full flow required for all four containment fan coolers. Therefore, the activities do not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.

The margin of safety for TS is not reduced since the work is performed when the 2SW-2907 valve is not required. The change does not involve an unreviewed safety question. (SER 96-100)

19. MR 93-025* A. (Common), RHR. MR 93-025* A replaces and reroutes main control board (MCB) wiring for residual heat removal pumps 2P-10A&B in train specific wireways to provide adequate separation between redundant safety-related wiring. RHR pump control circuit wiring in main control boards C-01 and 2C-03 is isolated by removing control power fuses.

Summarv of Safety Evaluation: The only applicable accident previously evaluated in FSAR Section 14.1.4-3 is the chemical and volume control boron dilution accident. The probability of this accident is not increased as the work directs that wire installations occur when the unit is defueled and the chemical and volume control system is not in use. There are r.o accidents previously evaluated that require RHR for mitigation while the unit is defueled. Therefore this work does not increase the probability of occurrence or the consequences of an accident previously evaluated in the FSAR.

The work only replaces / reroutes wiring in control circuits for RHR pumps 2P-10A&B. The configuration of these circuits are not changed. Wire replacements are performed with the RHR pump control circuits deenergized. Other precautions such as QC inspections are included to minimize the possibility of affecting safety-related circuitry in the main control boards. Afler the wire replacements are completed, testing verifies that circuits perform its inter;ded safety function and that wiring is installed correctly so as not to affect other safety-related equipment or change the operation of the RIIR pump control circuits. These provisions ensure ,

that there is no increase in the probability of occurrence of a malfunction of equipment nor an increase in the consequence of a malfunction of equipment important ta safety. Thus no new failure modes are added to create e the possibility of an accident of a different type than previously evaluated in the FSAR. The change does not involve an unreviewed safety question. (SER 96-104)

, , 20. MR 93-025* A,(Common), Containment Accident Recirculation Fans. MR 93-025*A installs nsw wireways in main control board (MCB) C-01 and in the transition between main control boards C-01 and C-02. It also removes containment accident recirculation fans 2W-1 AI and 2W lDI from service by removing the

! associated control circuit power fuses and then re-routing / replacing internal main control board C-01 wiring within these control circuits to provide adequate separation between redundant safety trains.

I Summary of Safety Evaluation: The containment accident recirculation fans are used to respond to accidents previously evaluated in the FSAR. The evaluations show the fans do not act as an initiator of these accidents.

Therefore, the activity does not increase the probability of occurrence of an accident previously evaluated.

1 i

IWe 39 s

T i MR 93-025*A is performed with Unit 2 in cold shutdown while the fons operations are not required. There are

! no accidents previously analyzed in the FSAR that require the containment accident recirculation fans with the unit in the cold shutdown condition. Therefore, the activities do not increase the consequences of an accident previously evaluated.

The activities are performed with the control circuits for the accident fans de-energized. If a reactor coolant pump is required to run while one or both of the fans are out of service, the other two fans are not affected by the work and therefore would be available for use. Therefore, the activities do not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR.

The margin of safety for the TS is not reduced since the work is performed when the fans are not required. The change does not involve an unreviewed safety question. (SER 96-106)

21. MR 93-025* A,(Unit 2),480 V Electrical Service. MR 93-025*A removes the 28-04 lockout relay 2-86/B04 from service by opening the associated control circuit power manually operr.ted breaker (MOB) and also .

removes the 2B-04 auxiliary lockout relay 2-86X/B04 from service by removing the associated control circuit power fuses. It then replaces internal main control board C-01 wiring within these control circuits to provide adequate separation between redundant safety trains.

Summary of Safetv Evaluation: The work on the 28-04 lockout relay and auxiliary lockout relay is performni in Section C-01 of the main control board. Section C-01 does not contain any wiring or components associated with RCCA movement, RCS integrity, normal feedwater flow, electrical distribution, fuel handling equipment, or any other circuitry capable of causing an accident of the type evaluated in the FSAR. Therefore, the activity does not increase the probability of occurrence of an accident previously evaluated in the FSAR.

The work and subsequent testing is performed when the lockout relay functions are not needed. No special equipment lineups are required and no LCOs are entered; therefore, there is no efTect on emergency diesel generator loading. There is no effect on shared systems between Unit I and Unit 2 (e.g., service water pumps powered from 2B-04). Therefore, this activity does not increase the consequences of an accident previously evaluated in the FSAR.

The only possible outcome of the work besides that desired would be to cause a short circuit, open circuit or ground in the control circuit being worked on, to land a wire in the wrong location, or to cause a short across the field wiring to the various trip circuits thus simulating the closing of the lockout relay contact in that trip circuit. With the except on of simulating the closing of the relay contact, the possibility of an accident of a different type than previously evaluated is not created since the post-maintenance testing detects an undesired outcome. Regarding simulating the closing of the relay contacts, only the circuits of non-vital loads would be affected. Therefore this activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.

The margin of safety defined in the Basis for TS is not reduced because the work and testing is performed when the 2B-04 lockout relay and auxiliary lockout relay are not required to function, because component cooling water pump 2-Pil A is required to be running belbre work beings. The work does not involve an unreviewed safety question. (SER 96-119) -

22. MR 93-025*A.(Unit 2), Residuallleat Removal. MR 93-025*A for 2SI-850A and 2SI-851B removes the Train A wire from the opposite train raceway and replaces it with a new wire routed in Train A raceway. The wires bundled together require those wires for the Train A valves be disconnected from the control circuit device (vah e position indicating light or control switch) and separated from the bundle. The Train A wires are rebundled, covered with a protective material which pmvides an effective barrier to the redundant Train B valve control circuit wire bundle, and reconnected to the control circuit device.

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l Summarv of Safety Evaluation: RilR valves 2SI-850A and 2SI-851 A and redundant Train B valves are l

. remotely-operated manual valves that receive no automatic signals to open or shut. They are placed into l

, service by operators following a large break loss of coolant accident to allow primary system and refueling ,

l water storage tank water that accumulates in containment sump B to be recirculated back to the core for long term decay heat removal. The FSAR accident evaluation does not show that the sump recirculation control !

circuits or the valves as initiators of the loss of coolant accident or other evaluated accident. The work does not i increase the probability of an accident previously evaluated. j The 2SI-850A and 2SI-851 A valve control circuits are not changed. Work controls inside main control board C-01 ensure the redundant Train B valve control circuits, other Unit 2 components and Unit I components are l not affected. No new equipment or failure modes are added to create the possibility of a different type of accident than previously evaluated in the FSAR. The only malfunctions that can be created are open or short circuits inside C-01 for 2SI-850A and 2SI-851 A or its redundant Train B valve control circuits. Electrical tests and operational checks are performed on the circuits during post-maintenance testing to verify the malfunctions are not introduced.

For the shutdown and defueled initial conditions of the work activities, decay heat removal capability is not required and containment integrity is not required. This assures optimal safety margin. The containment leakage boundary provided by the respective valve is not affected by the work. Valve position indication is verified following completion of the installation work. Therefore, the margin of safety as defined in the TS Bases is not reduced.

The work does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The work does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in TS is not reduced. (SER 96-122)

23. MR 93-04l' A and*B,(Common), HVAC. MR 93-041* A modifies the existing power supplies to control room HVAC supply and filter fans to provide EDG backed (safeguards) power sources from safeguards buses 1B-32, IB-42,2B-32, and 2B-42. The new power supplies are designed so that one filter fan and one supply fan is supplied from a single unit and the redundant fans are supplied from the other unit. There will also be one filter fan and one supply fan per t ain. In addition, loss of voltage relays and arming pushbuttons are installed to ensure control room HVAC fans do not automatically start afler a loss of offsite power and potentially overload the EDG in the injection phase.

MR 93-041 *B modifies the existing power supplies to this control circuit for the damper control to provide i EDG backed (safeguards) power sources from 120 Vac distribution panel 1Y-06 (normal supply) and 2Y-06 l (alternate supply). The switching from normal supply lY-06 to alternate supply 2Y-06 is an automatic function upon a loss of voltage to lY-06. In addition. indication lights are installed on the C-67 panel to identify that the power sources to the panel are available and which power source is feeding the panel.

Summarv of Safetv EvaluatiDn: Interim configurations allow the control room filter fans (W-14A&B) cable spreading room supply fans (W-13 AI&A2) and the control room supply fans (W-13Bl&B2) to remain operable during installation. Ilowever, at various times only one fan for each pair is available while its I redundant fan is being worked on.

Failure of the control room HVAC system is not an initiating event for an accident previously analyzed in the l FSAR. The new power sources supplied to the control room HVAC system allow operation of the system during a design basis event. Therefore, the new design for the control room HVAC systern does not increase j the probability of occurrence of an accident previously evaluated in the FSAR. l 1

I Page 41

The designed function of the control room IIVAC system has not been altered by these changes. The changes and additional actions ensure that the radiological habitability in the control room is maintained during a design basis event. Therefore, the changes do not create the possibility of an accident of a different type than previously evaluated in the FSAR. llaving control room filtration fans that are powered from EDG backed buses should increase the margin of safety of having the system operable. Therefore, the margin of safety for any TS is not reduced. The change does not involve an unreviewed safety question. (SER 96-004)

24. MR 94-012. (Common),4.16 kV. MR 94-012 upgrades the anchorage on station service transformers X-13 and X-14. X-13 and X-14 step down the 4160 V to 480 V lines and feed the B-03/B-04 safeguards buses.

Summarv of Safety Evaluat. 2 MR 94-012 is installed in three phases. P'iasc ! and 2 work is not tied to a specific time frame for comp!ction. Ilowever, Phase 3 was completed during UlR23. .

Precautionary measures protect the transformer tank and cooling coils from the welding arc during the Phase 3 work. Although the welding is performed a minimum distance of 4" from the transformer tank and cooling .

coils, an insulation material is installed to cover the local area of the trandormer tank and cooling coils where I the welding is occurring.

Work is performed sequentially (X-13 preceding X-14) with the refueling cavity flooded, or witn the opposite train of RHR (or DHR) in operation. This ensures redundant decay heat removal capability is available. In l addition, where applicable, the opposite train of spent fuel pool cooling will be running.

l There are no adverse affects on equipment important to safety required to mitigate the consequences of an accident or malfunction of equipment important to safety by upgrading the seismic anchorage on X-13 and X-14. The upgraded transformer seismic anchorage does not affect a previously analyzed accident consequences.

The new anchorage b a an accident initiator. Therefore, the activities do not create the possibility of an accident of a different type than previously evaluated in the FSAR. The upgraded anchorage does not create new failure modes of equipment. Therefore, the activity does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.

This modification does not affect TS parameters, therefore, there is no reduction in the margin of safety as defined is the Basis of TS. MR 94-012 does not involve an unreviewed safety question. (SER 96-009)

25. MR 94-051. (Unit 1), Safety injection. MR 88-097 installed the Unit I safety injection (SI) test line high point vent valve. The vent valve, ISI-V-23, is located at the ceiling in the spray pump cubicle. MR 94-051 extends the existing vent line from its present location to an area where the valve is readily accessible.

Summarv of Safety Evaluations: During normal power operation, the Si high flow test line is isolated from the SI system by Si discharge cross-connect valves ISI-829A& B. These valves are tested per the Leakage Reduction and Preventive Maintenance Program. During the installation of this modification, the test line is isolated. When the high flow test line is lined up for testing, the operable Si train is isolated from the test by l either ISI-829A or ISI-8298, depending on the train being tested. In the event of a LOCA resulting in an SI -

! during a high flow test, the operators will terminate the test and isolate the test line. The installation of new isolation valve downstream of the existing high flow test line vent does not change these conditions. The new isolation valve and tubing are designed to meet or exceed the specifications of the existing system and therefore maintains the system integrity under the design conditions. Therefore, this modification does not increase the l consequences of an accident previously evaluated.

l The accidents evaluated in the FSAR that rely upon the Si system for post accident mitigation are a steam pipe l rupture, steam generator tube rupture and the RCS loss of coolant accident (LOCA). The Si high Dow test line is isolated from the Si system during normal power operations and remain isolated during the evaluated j accident conditions. The iubing and imlation salve used to extend the Si high now test line high point vent are 1

Page 42 1

isolated from the test line during the installation. This modification extends the existing high point vent isolation valve to an area that is more accessible. The new isolation valve and tubing are designed to meet or exceed the speciGcations of the existing system and therefore maintains the system integrity under the design j conditions. This modiGcation does not increase the probability of an accident previously evaluated or create !

the possibility of an accident of a different type. '

The SI system is required to be operable when the reactor is critical per TS 15.3.3. One Si train may be taken i out of service provided it is retutned to service within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the other trair. is operable. Neither Si train i is required to be taken out of service for the installation or testing of this modificat on. Men the high flow test !

line is lined up for testing in accordance with IT-520A, the operable Si train is isolated from the test line and the LCO requirements of TS 15.3.3.A.2 shall be met. Ti extension of the existing SI high flow test line vent does not change these requirements. Therefore, the installation of MR 94-051 does not reduce the margin of safety as defined in the Basis for TS. The extension of the Si test line high point vent in accordance with MR 94-051 das mt create an unreviewed safety question. (SER 96-047)

26. MR 94-066* A. (Unit 2), MR 95-059' A. (Unit ), Safety injection. The modifications upgrade the nitrogen supply piping to the Sl accumulators in Unit I and Unit 2. A soft face disk assembly is installed into the 2SI-823D containment isolation valve (C!V) check valve. The valve upgrade makes it more leaktight, thereby providing acceptable leakage per 10 CFR 50 Appendix J criteria. The work includes installation of a pressure relief valve from both Unit I and Unit 2 nitrogen supply piping. A pressure regulator is installed in conjunction with the relief valve to fulfill OSHA requirements for a compressed gas system. Included with the installation of the pressure regulator in Pipeway #3 is the replacement of a 1/2" tubing spoolpiece currently used during ORT 32. The spoolpiece is replaced by a tubing tee and a vent valve to allow for operator case in performing ORT 32.

Summary of Safety Evaluation: The modiGcations are installed during a refueling outage when conditions are allowed by TS. The installation window occurs while the piping is isolated from the Si accumulators and the nitrogen source. The installation does not affect the integrity of the nitregen supply piping to the Si accumulators. The check valve will still shut when required, allowing the pressure to be maintained within the accumulators when Si is required. The relief valves and the check valve do not have an adverse affect on I containment integrity. The relief valves protect the containment penetration piping from overpressurization during filling of the SI accumulators. The containment isolation function is performed by the check valve.

Containment integrity is not adve sely affected. The change puts the containment structure in a more safe condition than it was in prior to the modification. The piping is no longer subject to overpressurization from the nitrogen supply and the inside containment CIV check valve has a higher reliability due to the soft seat conGguration.

The modification enhances the ability to perform the function of containment isolation with the use of the soft seat check vah e disc. It also removes the possibility of overpressurization of containment penetration piping.

. Both of these items place the plant in a more safe condition than it was prior to this modification. Core cooling, containment isolation and reactivity control are equa' , or better than they were prior to this modification. The modification does not involve an unreviewed safety question. (SER 46-087)

! 27. MR 94-097. (Unit 2), Reactor Coolant. MR 94-097 removes existing valves and piping in the Unit 2 reactor l coolant system. The valves are 2RC-503,2RC-541,2RC-543,2RC-598 (MOV), and 2RC-599 (MOV). Tne valves are not required for safe operation of the plant.

l Summary of Safety Evaluation This modification is installed during a refueling outage when the reactor vessel l is defueled. The reactor coolant system (RCS)is vented to the atmosphere. The reactor coolant drain tank (RCDT) is vented, purged, and isolated. The breekers for the two motor-operated valves are danger tagged i

open_ The new fittings meet appropriate design requirements. The materials are compatible with the existing materials. New equipment meets Code and rating requirements.

l 1 9

4

The consequences of an accident previously evaluated in the FSAR are not increased by this modification. The ability to perform the functions of containment integrity, reactivity and core cooling is not degraded by MR 94-097.

Safety-related structures, systems and components are not adversely affected by this change. The probability of l occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. l The margin of safety as defined in the TS is not reduced. The change does not pose an unreviewed safety I question. (SER 96-062) i

28. MB.95.D22,(Unit 2), Safety injection. MR 95-029 provides a more reliable and accurate level indication system for the Si accumulators. The present system utilizes a capacitance probe to monitor the level in a 2" standpipe connected to the accumulator. There are two probes on each accumulator. The capacitance probes .

are replaced with differential pressure transmitters. MR 95-029 uses Rosemount Model 3051 Smart dp transmitters which are more accurate and reliable than the ones used in the past.

. l Summary of Safety Evaluation: Installation is done with Unit 2 in cold shutdown and with the accumulators isolated. For each instrumentation channel, this modification requires: Removal of the existing capacitance probe; installation of a differential pressure transmitter; rerouting of cable and conduit; reworking the instrumentation tubing; tapping into the 2" standpipe for the variable leg; and replacing the controller / power supply for the instrumentation loop.

The operation of the level indicators and alarms in the control room are not affected by this change. The accuracy and reliability of the difTerential pressure transmitters is substantially better than that provided by the old capacitance probes.

The level indication system for the Si accumulators is non-safety-related. The operability of the level indication system is not relied upon during accident conditions to verify accumulator injection. The QA-scope accumulator pressure channels are used for this purpose. The accumulator level indication and alarms are, however, used to verify that the accumulators are able to perform its safety-related function by allowing the i operators to maintain level within its TS limits. Redundancy and separation are maintained between the two level indicating channels for each of the accumulators.

The Si accumulators are passive components, relying only on the associated check valves to open in order to deliver flow to the core. This modification does not increase the possibility of a passive failure of the Si accumulators, or other component important to safety. This modification does not introduce an active failure ,

that would impact the operation of equipment important to safety.

The assumptions used in the 13 asis of TS associated with the Si accumulators remain valid. The Si accumulator level channels are used for indication and alarm purposes only and do not perform control er safety functions.

As such, the Si accumulator level channels are not addressed by TS.

MR 95-029 does not increase the probability or consequences of an accident or equipment important to safety.

No margin of safety described in the TS is reduced. This modification does not involve an unreviewed safety question. (SER 96-105) . ,

Summary of Safety Evaluation: Installation occurs win the unit in cold shutdown with the accumulators isolated. For each instrumentation channel this modification requires the following: 1) Remove existing capacitance probe along with the associated 2" standpipe; 2) install dp transmitters and reroute cable and conduit; 3) rework the instrumentation tubing, connecting the variable leg of the new transmitter to the root isolation valve near the bottom of the tank; and 4) replace the controller power supply for the instrumentation loop.

Page 44

The welding required for this installation is to socket weld an insert into the root isolation valve and to socket weld a tubing adapter to the insert to transfer to 3/8" tubing, in order to tie in the variable leg. This section of piping is within the boundary of ASME Section XI, therefore, it is within the Repair and Replacement Program.

Since welded connection is socket welded, radiographic testing is not required. Parts that make up the

! accumulator pressure boundary are stainless steel with appropriate temperature and pressure ratings for the pipe class, and meet QA requirements. The conclusions of SER 96-105 remain valid. (SER 96-105-01)

29. MR 95-058. (Unit 2), Steam Generator Replacement. As part of MR 95-058 and the Unit 2 steam generator i replacement, an equipment decontamination trailer is used for steam generator replacement decontamination activities. The trailers located within the protected area adjacent to the radiation controlled area (RCA) to allow !'

ease of transferring items to the trailer for decontamination. The trailer is equipped with a self-contained HEPA

, air filtration system.

Summary of Safety Evaluation: This activity has no association with accident initiation mechanisms. The assumptions of the FSAR design basis accident analyses are not affected in any way. Also, the activity has no )

effect on equipment important to safety previously evaluated in the FSAR. Therefore, the activity does not 1 increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

The activity has no association with the initiation of design basis accidents. An evaluation determined that for 4

, an advertant release of the most conservative expected amount of radiological materials within the trailer, the l resultant offsite dose consequences would be well within the limits of 10 CFR 50, Appendix I and 10 CFR 20.

The dose to the public at the site boundary control center from a liquid release (e.g., fire fighting runoff) would be insignificant as bounded by a previous dose calculation for the storage of radioactive materials in Warehouse !

No. 2. Further, the activity has no effect on accident mitigation equipment associated with design basis accident previously described in the FSAR. Therefore, the activity does not increase the consequences of an

)

accident or malfunction of equipment important to safety previously evaluated in the FSAR.

l I

l The activity has no impact to TS. No equipment in the Basis of TS is affected by use of the equipment.

l Therefore, the activity does not reduce the margin of safety as defined in the Basis of TS. The use of the j l equipment decontamination trailer does not involve an unreviewed safety question. (SER 96-081) )

Summarv of Safety Evaluation: This evaluation demonstrated that for inadvertent release of the most conservative expected amount of radiological materials within the trailer, the resultant offsite dose consequences would be well within the limits of 10 CFR 50, Appendix I and 10 CFR 20. (SER 96-081-01)

30. MR 95-058. (Unit 2), Steam Generator Replacement. As part of MR 95-058, tools and equipment for the i Unit 2 SGR are brought to the site and stored in structures located outside the protected area (e.g., Warehouse No. 4, ISFSI transporter storage buildmg). These may include low level radiologically contaminated items ,

I shipped here from other nuclear sites with unconditional release limit of 1000 dpm/100 cm2 for beta-gamma emitters, as opposed to PBNP conservative limit of 100 dpm/100 cm' If contamination is detected above PBNP limits, radiological controls are immediately established in accordance with procedures and 10 CFR 20 requirements.

l .

Summarv of Safety Evaluation: This activity has no association with the initiation of a design basis accident.

The consequences of inadvertent release of radiological materials stored within structures outside the protected area due to fire or other events was evaluated in Calculation SGRP-96-017. An evaluation demonstrates that for inadvertent release of the most conservative expected amount of radiological materials within these structures, the resultant offsite dose consequences would be well within the limits of 10 CFR 50, Appendix 1 and 10 CFR 100. The dose to the public at the site boundary control center from a liquid release (e.g., fire fighting runoff ) would be insignificant as bounded by the previous dose calculation for the storage of

( radiological materials in Warehouse No. 2. The actisity has no effect on the important to safety / accident mitigation equipment associated w ith design basis accidents previously described in the FSAR.

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l The activity has no impact on TS. No equipment in the Basis of TS is affected by storage of potentially contaminated tools and equipment. Therefore, the activity does not reduce the margin of safety as deGned in the Basis of TS. The temporary storage of radiologically contaminated tools and equipment does not involve an unreviewed safety question. (SER 46-063)

Summary of Safety Evaluation: The resultant offsite dose consequences ofinadvertent releases of most l conservative expected amount of radiological materials within these structures would be well within the limits of 10 CFR 50, Appendix I and 10 CFR 20. (SER 96-063-01)

Summary of Safety Evaluatan: Tools and equipment may be brought from Unit 2 containment for release to clean areas and stored in structures located outside the protected area (e.g., Warehouse No. 4, ISFSI Transporter .

Storage Building). The equipment and tools brought from Unit 2 containment is surveyed for removable -

l contamination and is less than 100 dpm/100 cm2 prior to removal from the containment access !

building / containment access covered enclosure RCA. (SER 96-063-02)

31. MR 95-058. (Unit 2), Steam Generator Replacement. This evaluation addresses changes to the setpoint document associated with reactor trip setpoints, runback setpoints, rod stop setpoints, steam generator level program, and rod insertion limit alarms, as required by the Unit 2 steam generator replacement final design and associated Technical Specification changes. It also covers selection of conservative OTAT and OPAT setpoints to be used during power ascension following steam generator replacement, and the temporary or permanent procedure changes required to implement temporary and permanent setpoint changes.

Summarv of Safety Evaluation: Reactor protection setpoints have been conservatively chosen to account for instrumentation rack error and ensure that protective actions occur in accordance with Technical Specification requirements. Uncertainties associated with not knowing the AT O's following steam generator replacement have been conservatively accounted for in the trip setpoints which depend upon an accurate AT O, and by choosing constants for the setpoint equations which are conservative to Technical SpeciGcation requirements. ,

The conservative setting allows for ascension to approximately 90% power. OTAT and OPAT runbacks and rod stops are returned to its original design difference from the OTAT and OPAT reactor trips, and rod insertion limit alarms are recalibrated based on a return to full reactor coolant system flow associated with new steam generators. Upon ascension into approximately 90% power, AT O's are determined, new setpoints calculated, and installed in the plant. At that point, setpoints will be in accordance with TS, and settings allow continuation to 100% power.

Conservative selection of setpoints ensures that protective actions occur as designed, and as assumed in the accident analyses. Rod insertion limit alarms, OPAT runback, and OTAT runback are restored to its original design. The activity does not increase the probability of occurrence or consequences of an accident previously evaluated in the FSAR. The activity does not increase the probability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated in the FSAR. The activity does not create ,

the possibility of an accident or a malfunction of equipment important to safety different than previously evaluated in the FSAR. The activity does not reduce the margin of safety as deGned in the Basis for Technical SpeciGcations, assuming the inclusion of changes based upon Technical SpeciGcation Change Requests 188 and 189. The changes do not involve an unreviewed safety question. (SER 96-114)

P.nze a

32. MR 95-058* A. (Unit 2), Steam Generator Replacement. Temporary Steam Generator Storage Facility. As part of the Unit 2 steam ger.erator replacement, the original steam generator (OSG) lower assemblies are placed in the temporary steam generator storage facility (TSGSF) for storage. The TSGSF is an existing site structure constructed in 1983 for the storage of the Unit I and Unit 2 OSGs. The structure is divided into two storage areas, the north bay and the south bay, intended in the original design to be used to separately house the Unit I and Unit 2 OSGs. The Unit i OSG lower assemblies are stored in the north bay. MR 95-058'A provides the structural and radiological reanalysis of the TSGSF for storage of both the Unit I and Unit 2 OSG lower assemblies in the TSGSF north bay. It also evaluates the implementation of temporary changes to the TSGSF j

, facade, removable wall sections, and fence required to allow placement of the Unit 2 OSG lower assemblies l into the TSGSF. MR 95-058*1 addresses the transportation of the Unit 2 OSG lower assemblies to the )

threshold of the TSGSF. '

Summary of Safety Evaluations: The structural and radiation dose analyses conclude that both the structural ,

and radiation dose effects of the revised OSG lower assembly storage configuration are within both design and regulatory requirements for the TSGSF. The changes to the removable wall sections and facade to allow movement of the Unit 2 OSG lower assemblies into the TSGSF are temporary in nature, and are returned to its as-found configuration or engineering approved alternate configuration following completion of positioning the l Unit I and Unit 2 OSG lower assemblies within the TSGSF. The TSGSF plays no role in the initiation of an accident described in the FSAR. MR 95-058*A does not adversely affect equipment failure mechanisms.

MR 95-058*A does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

The TSGSF is not required to perform a function which may influence the consequences of an accident evaluated in the FSAR. A radiation dose reanalysis verifies that dose levels associated with storage of the Unit I and Unit 2 OSG lower assemblies are within acceptable 10 CFR 20 limits. No design hsis accident assumptions are adversely affected. Therefore, the consequences of an accident or malfunction of equipment important to safety previously evaluated are not increased.

A structural reanalysis verified that loads associated with storage of the Unit 2 OSG lower assemblies are within the design limits of the TSGSF structure. No new accident initiators or new equipment failure mechanisms are created by the storage of the Unit 2 OSG lower assemt ies which are different from those previously evaluated.

Therefore the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR is not created.

TS do not require the operability of the TSGSF or other systems, structures. or components potentially affected by implementation of this modification. There are no adverse effects on safety limits, setpoints, or operating parameters as a result of MR 95-058*A. In addition, no fission product barriers which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel are degraded because of MR 95-058*A. It does not reduce the margin of safety defined in the Basis of TS. MR 95-058* A does not involve an unreviewed safety question. BER 96-050)

33. MR 95-058'B,(Unit 2), SGRP ElectricalInterferences. MR 95-058'B provides the design details for the temporary removal, interim measures to allow operation of specific equipment, and reinstallation of electrical

, interferences located within the steam generator A and B cubicles, the containment building equipment hatch, and the equipment chase area. Implementation measures are also specified to allow operation of certain equipment during the interim condition between removal and restoration of supporting interferences (e.g., electrical power / instrument air).

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Summarv of Safety Evaluation: MR 95-058'B is implemented only when the affected equipment is not required to perform an important to safety function which would be relied on to mitigate the consequences of a FSAR postulated accident. The cabling servicing the safety injection, residual heat removal, and containment

isolation functions is only affected while the Unit 2 reactor is defueled. Power is isolated to the valves during the time period when the cable tray and conduit are pushed out of the way; therefore, there is no potential for inadvertent actuation of an important to safety valve. The electrical interferences and the instrument air line for the fuel manipulator crane are temporarily removed and then returned to its as-found configuration following

! completion of steam generator replacement.

Components afTected by the design package are tested to verify proper operation after reinstallation and prior to returning the component to service. The implementation does not adversely affect other structures, systems, and components. Accident mitigation is unaffected since accident mitigation required equipment continues to -

meet applicable requirements. There are no new failure modes, such as common mode failures, that would place Unit 2 outside ofits design bases. MR 95-058'B does not adversely affect equipment failure initiation mechanisms. .

There are no changes to safety limits, setpoints or operating parameters as a result ofimplementation of this design package. In addition, no fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel, are degraded by implementation of MR 95-058'B. It does not reduce the margin of safety as defined in the Basis of TS. MR 95-058'B does not involve an unreviewed safety question.

(SER 95-058)

Summarv of Safetv Evaluation: Due to the important to safety function the Gai-tronics (public address system) serves during the procedural handling of nuclear fuel, Gai-tronics interference removal / restoration is administratively prohibited during refueling operations and refueling shutdown (with the reactor vessel head removed). Afler Gai-tronics interference removal, the unaffected portion of the containment Gai-tronics is verified to be operable. The conclusions of the original SER remain valid. (SER 96-058-01)

Summary of Safety Evaluation: Cable tray 2VE04 is removed, and the cable pushed back and secured while l the Unit 2 reactor is defueled, when the equipment served by the moved cabling is not required to be operable to perfonn function important to safety (T-34 A Si accumulator outlet valve, CC-761 A; SG blowdown containment isolation valve, CV-1299; and lights / receptacles for SG cubicle A. The conclusions of the original SER remain valid. (SER 96-058-02)

Summarv of Safety Evaluation: The cable trays removed are 2VEO4 and 2VEOS. (SER 96-058-03)

34. MR 95-058*C. (Unit 2), Steam Generator Replacement ilVAC Interferences. MR 95-0588C temporarily removes and reinstalls IIVAC interferences located in the Unit 2 containment equipment chase area, SG A & B cubicles, and the reactor vessel missile shield platform.

Summarv of Safetv Evaluation: MR 95-058'C temporarily removes, reinstalls, and functionally tests certain l portions ofIIVAC systems inside the Unit 2 containment with no permanent changes to the system design functions or performance. Implementation windows for removal of HVAC interferences are specified to ensure l that the affected systems and equipment are not required to perform important to safety functions during the -

period ofinoperability. The systems are not initiators for accidents previously evaluated in the FSAR.

l The disabling of the containment air recirculation cooling system by removal of a section of supply duct is l limited to Unit 2 reactor defueled condition when the system is not required to be operable. The duct section is !

j restored to its as-found configuration and functionally tested to establish operability, prior to commencement of refueling operations.

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During the implementation of MR 95-058'C, affected equipment is capable of performing its safety function as described in the Basis of TS for the conditions specified. There are no adverse effects on safety limits, setpoints or operating parameters as a result ofimplementation of MR 95-058'C. To address NUREG-0612 concerns, identified heavy loads are handled either during the Unit 2 reactor defueled condition, or during cold shutdown, refueling shutdown, or refueling operations using a safe load path. MR 95-058*C does not involve an unreviewed safety question. (SER 96-059)

35. MR 95-058* D. (Unit 2), Steam Generator Replacement. Structural interferences in Unit 2 containment equipment chase area are reconfigured to provide clearances necessary for the removal of the original steam generators (OSGs) and the installation of the replacement steam generators (RSGs). Removed structural interferences are returned to their as-found configuration or an engineered approved configuration.

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Summarv of Safetv Evaluations: MR 95-058'D affects the following equipment chase area structural interferences: Platform at El 25'-10 3/4"(top of steel); portions of platforms at El 29'-4 3/4" / 51'-10 3/4"(top of steel); portions of reinformed concrete slabs at El 46' / 66'; and equipment hatch monorail support steel.

MR 95-058'D does not affect the ability of the affected structures, systems, and components to perform its intended important to safety functions. The equipment chase structural interferences are restored to its as-found configuration or an acceptable engineered configuration following completion of steam generator handling activities and prior to plant conditions that require its operability. No important to safety equipment is adversely impacted by MR 95-058'D. Therefore, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR are not required.

MR 95-058'D does not adversely affect equipment failure mechanisms. Temporary inoperability of the equipment hatch monorail due to design package implementation only results during the reactor defueled condition when containment closure is not required. The monorail capability to close the equipment hatch is restored and the equipment hatch reinstalled prior to the commencement of refueling operations. No important to safety equipment is adversely impacted. Therefore, the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the FSAR is not created.

TS do not require the operability of the affected components during implementation windows. Inspections following reinstallation of the equipment chase structural interferences ensure that the affected equipment is returned to an operable condition following steam generator handling activities. There are no changes to safety I limits, setpoints, or operating parameters as a result of the modification, in addition, no fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel, are degraded by the changes. MR 95-058'D does not reduce the margin of safety defined in the Basis for TS. MR 95-058'D does not involve an unreviewed safety question. (SER 96-051)

Summary of Safety Evaluation: Any wood used for work on platform at El 25'-10 3/4" complies with NP 1.9.9," Transient Combustible Contro' "

1 Additional structural interference removed for the Unit 2 SGRP include: Reactor vessel head storage stand i stanchions / ring at El 215 The northernmost reactor vessel and the reactor vessel head storage stand ring are relocated temporarily south ofits current location. The conclusions of the original SER remain valid.

(SER 05-051-01)

36. MR 95-058'E. (Unit 2), Steam Generator Replacement. MR 95-058'E temporarily removes and reinstalls structural steel platforms, handrails, personnel tie-off rails, removable wall sections and hydraulic tu' 'ng to the SG upper lateral snubbers.

Summarv of Safety Evaluation: The design basis of the SG upper lateral snubbers is to prevent unrestrained I motion under dynamic loads that occur during an earthquake or severe transient to maintain primary and I secondary side integrity of the SG and attached piping. In order to remove the secondary manway' platforms at El 81'6"(SG Cubicle A) and El 81%"(SG Cubicle 11) during cold shutdown refueling shutdow n, refueling Pge40

l l

operations, or plant defueled condition. Upper lateral support (ULS) snubber tubing which service local l hydraulic oil reservoir level gauges for SG A and B is temporarily disconnected and capped.

Braces are added to El 81'6" SG Cubicle A platform to prevent adverse efTects to the attached snubber tubing and SG water level instrumentation tubing during an earthquake. MR 95-058* E maintains the seismic integrity of partially disassembled SG cubicle platforms during plant fueled conditions by restraining the :ffected platforms in accordance with good rigging practices.

Since the affected systems and equipment are capable of performing their original important to safety functions and MR 95-058'E does not change accident intention mechanisms. Therefore, the assumptions of the design basis accident analyses are not adversely impacted. Affected components continue to satisfy design requirements. There is no adverse impact on potential failure mode mechanisms. Therefore, the probability of -

an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

Affected components are capable of performing its important to safety functions as described in the Basis of the -

TS. The affected equipment is only inoperable when not required to perform a safety ftmetion. The TS do not require the operability of the affected components during the implementation windows. Components are returned to an operable condition and are fully available as required by TS. SG ULS snubber tubing affected by MR 95-058'E are leak checked following reinstallation to assure operability. There are no changes to safety limits, serpoints or operating parameters as a result of these changes. In addition, no fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by these changes. Therefore, the margin of safety as defined in the Basis of TS is not reduced. MR 95-058'E does not involve an unreviewed safety question. (SER 96-080)

37. MR 95-058*F. (Unit 2), Steam Generator Replacement Reactor Cavity Decking. MR 95-058'F fabricates and installs a temporary structural steel deck over the reactor cavity at El 66*. This temporary reactor cavity decking provides additional work space during the performance of the Unit 2 steam generator replacement, and provides a foreign material exclusion (FME) barrier to protect the reactor cavity liner and components residing inside the cavity from damage during the activity.

Summary of Safety Evaluation: Calculation SGRP-95-003 demonstrates that no adverse effects result to the '

reactor cavity liner and other affected containment structures as a result of MR 95-058'F. The reactor cavity decking does not serve an important to safety function. Reactor cavity decking installation, use during steam generator replacement, and removal is limited to the reactor defueled condition. During the defueled condition, there are no structures, systems, or components affected by the reactor cavity decking witch are required to be operable to perform an important to safety function.

MR 95-058'F is implemented when the Unit 2 reactor is defueled. Dur ng this condition, there is no important to safety equipment affected by MR 95-058'F implementation that is required to be operable to mitigate the consequences of malfunction of equipment important to safety previously evaluated in the FSAR. Therefore, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR are not increased.

Mk 95-058'F ensures that affected components are capable of performing its safety limction as described in -

the Basis of TS for the conditiens specified. The design requirements of the reactor cavity and liner, as well as I

other affected containment structures are maintained. After disassembly and removal of the reactor cavity decking, a survey of the reactor cavity liner and rails to verify no damage and verification of reactor cavity cleanliness is performed to ensure that plant structures and equipment are returned to its as-found configuration.

There are no changes to safe'y limits, setpoints or operating parameters as a result of implementation of MR 95-058'F. In addition, no fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel, are degraded. . The margin of safety as def'med in the Basis of TS is not reduced. MR 95-058*F does not involve an unreview ed safety question. (SER 96-064)

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38. MR 954)58*G. (Unit 2), Steam Generator Replacement Building Analysis and Shoring. MR 95-058*G installs -

and removes temporary shoring required to support the containment structural members on which the hatch transfer system is mounted. It also installs and removes temporary jib cranes in the vicinity of the equipment chase area.

Summarv of Safety Evaluation: Installation of the temporary shoring in the facade is performed during Unit 2 cold shutdown (following removal of the equipment hatch missile shield), refueling shutdown, refueling operations, or defueled condition. Upon completion of the heavy component hatch transfers and prior to the replacement of the equipment hatch missile shield, the temporary shoring in the facade is removed. Facade shoring areas are veriGed by walkdowns as directed to be restored to its as-found configuration prior to completion of project.

Use of the temporary jib cranes, which is limited to the Unit 2 defueled condition does not require shoring or bracing and has no adverse effects on plant systems, structures, and components. Evaluations demonstrate that the temporary jib cranes may be installed and semoved during cold shutdown, refueling shutdown, refueling

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operations, or when the Unit 2 is defueled with no adverse effects, including seismic effects, on plant systems, structures, and components. During defueled conditions, installed jib cranes are physically secured to prevent inadvertent operation.

Affected equipment is capable of performing its safety function as described in the Basis of TS for the conditions specined. Calculations demonstrate that the containmert structural integrity is not compromised.

Other than the containment structures, no plant systems or components are affected as a result of MR 95-058*G. There are no adverse effects on safety limits, seismic requirements, setpoints or operating parameters as a result of MR 95-058*G. In addition, no Ossion product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded by MR 95-058*G.

MR 95-058*G does not reduce the margin of safety as denned in the Basis of TS. MR 95-058*G does not involve an unreviewed safety question. (SER 96-078)

Summarv of Safety Evaluation: Temporary shoring members may be brought into the containment during cold shutdown, refueling shutdown, refueling operations, or in the Unit 2 defueled condition. Temporary shoring brought into the containment during Unit 2 fueled conditions shall be placed in a seismically safe condition (e.g., seismic event would result in no damage to important to safety systems, structures, and components).

Installation of temporary shoring inside the containment and use of shoring to impart supporting forces to containment structures are limited to the Unit 2 defueled condition.

If required for support of the hatch transfer system, the temporary bracing in the facade may be installed during Unit 2 cold shutdown (following removal of the equipment hatch missile shield), refueling shutdown, refueling operations, or Unit 2 defueled condition. Conclusions of original SER remain valid. (SER 96-078-01)

39. MR 95-058*H,(Unit 2), Steam Generator Replacement Temporary SG Supports and SG Gap Requirements.

MR 95-058*l1 installs and removes the temporary supports for the OSGs. It provides for temporary removal and reinstallation of the upper lateral support (ULS) ring as well as the ULS snubbers for each SG in order to provide adequate clearance for the removal of the OSGs.

Summary of Safety Evaluation: The important to safety function of the ULS in providing seismic support to the reactor coolant system (RCS) is maintained during cold shutdown, refueling shutdown, and refueling operations. Except for the original SG lower assembly lateral restraints, portions of the SG temporary restraint system may be installed during cold shutdown, refueling shutdown, or refueling operations, provided that the SG support columns are allowed to move independent of one another. The additional weight is insigniGcant and does not adversely affect the seismic integrity of the RCS required during defueled plant conditions.

MR 95-058*H ensures that the affected equipment is not required to perform functions important to safety during the time of inoperability. The SG lateral supports are returned to its as-found conGyuration within the acceptable cold and hot gap tolerances following completion of the project. Since the affected equipment is Pre 51

l l

! capable of performing its original important to safety functions, and the design package does not change

(- accident initiation / equipment failure mechanisms, the assumptions of the design basis accident analyses and equipment failure are not adversely affected, j Affected equipment is capable of performing its safety function as described in the Basis of TS for the I

conditions specified. The TS do not require the operability of the affected equipment during implementation windows, and the equipment is returned to an operable condition and fully available as required by TS. There are no adverse effects on safety limits, setpoints or operating parameters as a result of MR 95-058*H. In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded by implementation of MR 95-058*11. MR 95-058*H does not reduce the margin of safety as defined in the Basis of TS. (SER 96-W2) t Summarv of Safety Evaluation: The OSG lower assembly is temporarily laterally restrained above the upper lateral supports at approximately El 67'. Similar temporary lateral restraint of the RSG lower assemblies may be applied during RSG lower assembly installation. Except for the OSG/RSG lower assembly lateral restraints, .

portions of the SG temporary restraint system may be installed during cold shutdown, refueling shutdown or ,

refueling operations, provided that the SG support columns are allowed to move independently of one another. l Conclusions of original SER remain valid. (SER 96-067-01i !

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40. MR 95-058'J,(Unit 2), Steam Generator Replacement Offload and Transpo1. MR 95-058*1 uses rigging and handling equipment to transport the OSG and RSG steam drums and lower assemblies during the Unit 2 SGRP.

Summarv of Safety Evaluation. Accident initiation mechanisms are not changed, and the assumptions of the design basis accident analyses are not adversely affected by MR 95-058'I. It does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

In the unlikely event of a drop of a Unit 2 OSG lower assembly, radiological consequences for offsite dose to the general public have been postulated and evaluated in Calculation SGRP-95-005. It concludes that a Unit 2 tower assembly drop as its closest approach to the site boundary control center, which results in the rupturing of penetration welds that allow release of radioactive material from within the component, resultant offsite dose to the general public would be well within 10 CFR 50, Appendix ! and 10 CFR 100 limits. This accident offsite dose is also less than the radiation dose calculated for the similar accident of a rupture of a gas decay tank, as described in FSAR Section 14.23.

Affected equipment is capable of performing its safety function as described in the Basis of TS. There are no adverse effects on safety limits, setpoints or operating parameters as a result of MR 95-058'I. In addition, no fission product barriers, which include tr,c conminment, reactor coolant pressure boundary, fuel cladding, and j fuel, are degraded by the changes. MR 95-058'l does not reduce the margin of safety as defined in the Basis of )

1S. MR 95-05Pi does not involve an umeviewed safety question. (SER 96-065) '

Summary of Safetv Evaluatinn: Calculation SGRP-95-005 concludes that a Unit 2 lower assembly drop at its closest approach to the Site Boundary Control Center, which results in the rupturing of penetration welds that allows release of radioactive material from within the component, resultant offsite dose to the general public would be well within 10 CFR 50, Appendix I and 10 CFR 20 limits. The conclusions of the original SER -

remain valid. (SER 96-065-01)

41. MR 95-058*L (Unit 2), Steam Generator Replacement Temporary Lifting Hoist. MR 95-058'J installs, operates, and removes polar crane temporary support components and a temporary lifting hoist for lifting and maneuvering the OSG and RSG steam drums and lower assemblies.

Summarv of Safetv Evaluation: The licensing bases consider lifting of the reactor vessel head above the reactor vessel to be a core alteration, due to the possibility of an inadvertent drop which could disturb the fuel assemblies. Therefore, in making this lift of the RV head, the polar crane performs an important to safety function. MR 95-058'J activities which would impair the function of the polar crane and esisting trolley to Page 52

perform lifting of the reactor vessel head are limited to the time period after replacing the RV head on the vessel following reactor defueling and prior to commencement of reactor re u' eling. Restoration measures I ensure that the polar crane is capable of performing its important to safety frction oflifting the RV head with I fuel in the core, with no increase in the probability of occurrence of an accide W p eviously evaluated in the FSAR.

MR 95-058'J results in no adver se effects on plant systems and components which could result in increasing l the probability of occurrence of t?e malfunction of equipment important to safety previously evaluated in the l FSAR. There are no adverse effects 'n equipment failure initiation mechanisms. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR are not i increased. i MR 95-058'J is implemented with Unit 2 plant conditions configured such that important to safety equipment required to mitigate the consequences of an FSAR postulated accident is not required to be operable. ;

Restoration of plant equipment is assured that the affected equipment is capable of performing its original

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- important to safety functions, and thus accident mitigation analyses are not adversely affected. The l consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.

In the unlikely event of an OSG low er assembly drop during temporary lining hoist operation, damage to its lower assembly must be assumed along with radiological consequences resulting from the release of radioactive '

material from within the lower assembly itself and secondary radiological contamination sources during the l reactor defueled condition (e.g., ruptured RCS piping). The resultant offsite dose to the general public from an l OSG lower assembly drop at its closest approach to the site boundary control center has been evaluated to be j well within 10 CFR 50, Appendix I and 10 CFR 100 limits. The evaluated accident dose is also less than the l radiation dose calculated for the similar accident of a rupture of a gas decay tank, as described in FSAR j Section 14.2.3. By comparison, the offsite dose consequences of a drop by the temporary lifting hoist withm the containment would be less than the drop described, because of the effects of the containment and associated ventilation systems in confining the radiation while the OSG lower assembly is still within the containment, and the larger distance from the original SG radiation source to the site boundary control center. For dose comparison purposes, the possible secondary radiological contamination sources within the containment during i the reactor defueled condition are considered to be negligible when compared to the rupture of the OSG lower assembly at its closest approach to the site boundary control center.

Affected equipment is capable of performing its safety function as described in the Basis of TS. The TS do not 4

require the operability of the affected equipment during implementation windows, and the equipment is returned to an operable condition and fully available as required by TS. There is no adverse effects on safety limits, se points, or operating parameters as a results of MR 95 058'J. MR 95-058*J does not reduce the margin of safety as defined in the Basis for TS. MR 95-058'J does not involve an unreviewed safety question.

LSfiB 96-066)

Summary of Safety Evaluation: MR 95-058'J also relocates the RV missile shield to the south end of the reactor cavity with the existing polar crane trolley w hile Unit 2 is defueled. Relocation of the missile shield is required so that it does not interfere with TLil installation and OSG/RSG component rigging and handling.

Once positioned at the south end of the reactor cavity, the missile shield wheels /end trucks may be disassembled as required to enable the polar crane to lift the missile shield and place it temporarily on blocking located adjacent to the fuel manipulator crane rails. Following missile shield replacement onto its normal rail configuration, the missile shield is functionally tested to ensure proper operation. Missile shield temporary storage areas are inspected to verify no damage to the fuel manipulator crane rails and reactor cavity liner plate.

Pape 53

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For dose comparison purposes, if the worst case secondary radiological contaminathn sources within the containment during the reactor defueled condition are considered to be 4 times the gaantity released for the rupture of an OSG lower assembly at its closest approach to the site boundary control center (as evaluated in Calculation SGRP-95-005), the resultant offsite dose to the general public would be well within 10 CFR 50, ;

Appendix I and 10 CFR 20 limits. The conclusions of the original SER remain validfSER 96-066-01) l Summarv of Safety Evaluation: During the defueled condition when the missile shield and OSG/RSG components are handled, residual heat removal, service water, component cooling water, and waste gas decay systems are isolated from the Unit 2 containment at a location exterior to its respective containment penetrations. Also, the fuel transfer tube is isolated during the same conditions by means ofisolating the fuel transfer tube gate valve outside the containment and closing the two fuel transfer canal doors within the spent ;

fuel pool. The conclusions of the original SER remain valid. (SER 96-066-02) . !

42. MR 95-058*K. (Unit 2), Steam Generator Replacement. MR 95-058*K modifies the Unit 2 polar crane for use in the replacement of the Unit 2 steam generators. It makes permanent modifications to the polar crane necessary fro use in rigging the OSG/RSG steam drums and lower assemblies.

Summary of Safety Evaluation: MR 95 058*K involves the permanent installation of horizontal stiffeners to the bridge girder webs and installation of push-pull tracks to the upper flange of each bridge girder. Addition of the horizontal stiffeners on the polar crane bridge girders is required for the rigging of the OSG/RSG steam drums and lower assemblies and does not affect the capability of the polar crane to make lifts associated with normal plant operations and maintenance.

MR 95-058*K is performed such that Unit 2 requirements for seismic, containment integrity, containment closure, foreign material exclusion, and polar crane availability are met. The welds, as inspected and accepted, do not degrade the structural integrity of the polar crane. Once installation of horizontal stiffeners which involves temporary removal of the trolley but as commenced, this work must be fully completed and trolley operation tested prior to use of the trolley for lifling. MR 95-058*K is coordinated with the temporary removal of the trolley seismic restraint (interference removed / reinstalled in MR 95-058'J) to ensure that this restraint remains operable while Unit 2 is refueled.

Accident initiation mechanisms are not changed, and the assumptions of the design basis accident analyses are not adversely affected. MR 95-058*K does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Affected equipment is capable of performing its safety function as described in the Basis of TS for the conditions specified. Lifting and handling of OSG/RSG components and test weights in conjunction with centerpost, horizontal stiffener / push-pull track installation within the evaluated travel limits of the temporary lifling hoistlexisting trolley, do not exceed the allowable design stress / loads stipulated in CM A A Specification No. 70, AISC Manual of Steel Construction, and plant design criteria. There are no adverse effects on safety limits, setpoints or operating parameters as a result of MR 95-058*K. In addition, no fission product barriers, which include the containment (containment integrity / containment closure functions), reactor coolant pressure l boundary, fuel cladding, and fuel are degraded by MR 95-058*K. MR 95-058*K does not reduce the margin of safety as defined in the Basis of TS. MR 95-058* K does not involve an unreviewed safety question. -

(SER 96-082)

43. MR 95-058*L,(Unit 2), SGRP Transfer System. MR 95-058*L installs and removes the hatch transfer system l (llTS) that provides the mechanical means for transferring the steam generator components through the Unit 2 equipment hatch. The llTS is constructed of concrete foundations, steel shapes and plates over which rolling devices and stabilizing saddles transfer the original and replacement SG steam drums and lower assemblies.

The force to accomplish movement of these components along the llTS is provided by hydraulic push-pull units on the llTS structure.

Page 54

Summary of Safety Evaluation: Calculations SGRP-95-001, SGRP-95-007, and SGRP 96-010 demonstrate that no adverse effects are experienced by plant systems or components. The calculations address HTS design, i locations of buried utilities, and load effects on the Unit 2 containment facade retaining wall. The integrated effects of the HTS loadings on the containment structure, along with the load effects of other SGRP items and l support equipment, are evaluated in MR 95-058*G," Building Analysis and Shoring." ;

To prevent malfunction of the 345 kV transmission lines connected to the Unit 2 main transformers, administrative precautions maintain a minimal personnel and equipment approach distance of 12'6" from the i lines. Fur *.her administrative precautions limit the use of construction and disassembly equipment (e.g., cranes) I during adverse environmental conditions (e.g., fog, high winds). l

)

. MR 95-058'L does not increase the probability of an accident or malfunction of equipment important to safety l previously evaluated in the FSAR.

~

MR 95-058'L is implemented with the Unit 2 plant conditions configured such that the effected equipment F l not required to mitigate the consequences of FSAR postulated accidents. Accident mitigation is unaffected j since accident mitigation required equipment continues to meet applicable requirements. In the unlikely event ;

of damage to the 345 kV lines, the consequences of such damage is the same and is bounded by the j consequences analyzed for a loss-of-offsite-power event in the FSAR.

i During the reactor defueled condition, the consequences of a drop of a heavy load, such as an OSG lower assembly, is limited to damage of components required to be operable to maintain nuclear fuel integrity. With the Unit 2 reactor defueled condition, there is no important to safety functions performed by equipment in the equipment hatch area. However, in the unlikely event of an OSG lower assembly drop in the equipment hatch area during HTS operation, damage to its lower assembly must be assumed along with radiological consequences resulting from the release of radioactive material from within the lower assembly itself. In comparison to the effects of a drop of an OSG lower assembly at its closest point to the Site Boundary Control ;

Center outside of containment during its lower assembly transport, as performed in Calculation SGRP-95-005, 1 the offsite dose consequences of a drop at the HTS would be les's because of the effects of the containment and the larger distance from its radiation source to the site boundary control center. The resultant offsite dose is also less than the radiation dose calculated for the similar accident of a rupture of a gas decay tank, as described in FSAR Section 14.2.3," Accidental Release-Waste Gas."

There are no adverse effects on safety limits, setpoints or operating parameters as a result ofimplementation of MR 95-058*L ltems mo,ed in/out of the containment during refueling operations and/or refueling shutdown are limited to those items which do not affect containment closure of the hatch third door. As a result, adverse l

effects to the fuel are precluded during HTS and supporting equipment installation, operation and removal.

Fission product barriers, v hich include the containment, reactor coolant pressure boundary, fuel cladding, and j fuel, are not degraded by these changes. MR 95-058'l does not reduce the margin of safety defined in the l TS Basis. MR 95-058'L " Hatch Transfer System," does not insolve an unreviewed safety question.

(SER 96-049)

I i 44. MR 95-058'M,(Unit 2), Steam Generator Replacement Temporary Piping S/Rs. MR 95-058* M temporarily alters existing piping support / restraints and the fabrication / installation of temporary piping supports and

l. restraints.

The design basis for the temporary piping supports and restraints is to provide pipe stability during cutting, beveling, fitup, and welding operations. MR 95-058'M is implemented when the affected systems and l equipment are not required to perform an important to safety function. The changes are temporary, and affected systems and equipment are returned to its as-found configuration within acceptable tolerances.

Inspections performed verify correct restoration and no permanent effects to the piping. The probability of an accident or malfunction of equipment important to safety previously esaluated in the FS AR is not increased.

Page 55

For the interim condition when the temporary supporting of the RCS hot and cold leg piping is installed, the Unit 2 reactor is defueled. For the interim conditions when the main steam and main feedwLter piping temporary supports are installed, the Unit 2 steam generators are not required to remove RCS decay heat.

MR 95-058'M does not change accident initiation or failure mechanisms in any way. The possibility of an accident or a malfunction ofimpi rtant to safety equipment of a different type than previously evaluated in the FSAR is not created.

- Affected equipment is capable of performing its safety functions as described in the Basis to the TS. TS do not require the operability of the affected equipment during the implementation windows, and the equipment is returned to an operable condition as required by the TS. There are no adverse effects on safety limits, setpoints, or operating parameters as a result ofimplementation of MR 95-058*M. In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded .

by the implementation of this design package. MR 95-058*M does not reduce the margin of safety defined in the Basis for TS. MR 95-058*M does not involve an unreviewed safety question. (SER 96-060) l

45. MR 95-058'M and MR 95-058*0. (Common), Steam Generator Replacement. The modificaions provide for i the reconGguration of main feedwater and auxiliary feedwater lines necessary to achieve fitup of the Unit 2 RSGs. It provides for the removal of approximately 2" of the vertical sections of SG A and B main feedwater and auxiliary feedwater lines to achieve fitup to the RSGs. Cuts are made in the run of pipe between the steam generators and the El 66' elevation slab in containment. Affected main feedwater and auxiliary feedwater pipe suppons are adjusted to maintain the existing piping displacement requirements. The adjustments include Shimming, grinding, relocation of bumper stops, and rod adjustment / replacement.

Summary of Safety Evaluation: Important to safety components which mitigate the consequences of design basis accidents are taken out of service only when the plant is configured in such a manner that the component is not required to be operational. Components required to mitigate the consequences of an accident evaluated in the FSAR will be fully available to ful0ll its importance to safety functions when required. Equipment used for accident mitigatien are not affected in any way that would adversely innuence accident assumptions, timing of accident scenarios or accident source terms. FME measures are employed which preclude foreign material in the as-restored main feedwater and auxiliary feedwater piping affected by implementation. The activities have no adverse effects which would result in increasing the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.

Affected components are capable of performing their safety functie is described in the Basis of TS.

Modifications of the main feedwater and auxiliary feedwater pipinb and suppons are engineered-equivalent to the original construction Code and design requirements, and the margin of safety designed / constructed into the pressure boundary is not reduced.

There are no changes to safety limit setpoints or operating parameters as a result of the changes to main feedwater and auxiliary feedwater piping and supports. In a.ddition, no Ossion product barriers w hich include

~

containment, the reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by the changes.

Therefore, the margin of safety as defined in the Basis of TS has not been reduced. The work does not involve an unreviewed safety question. (SER 96-129)

46. MR 95-058'E,(Unit 2), Steam Generator Replacement RCS Elbow Replacement. If Stup problems with an RSG and its RCS cold leg cibow occurs, MR 058'N will be utilized to provide the design and implementation instructions to allow replacement of the affected RCS cold leg elbow (s) to achieve acceptable alignment of the RCS piping to the replacement SG nozzles. MR 95-058'N is prepared as a contingency plan.

The RCS piping cold leg elbow (s) w ill be replaced only if necessary to obtain acceptable fitup of the RCS piping to the replacement nozzles.

page 56

Summarv of Safety Evaluatiom The licensing design basis for the RCS cold leg elbow (s) is to maintain the integrity of the reactor coolant system (RCS) pressure boundary / fission product release barrier, function to remove RCS decay heat, and minimize the likelihood of a high energy line break (e.g., loss of coolant accident). The replacement RCS cold leg elbow (s) is equivalent to the existing elbow (s) in both form and function, as reconciled to the original design requirements and construction code. The replacement cibow(s) provides the same pressure boundary function, material, and piping geometry as the existing elbow (s) such that RCS ftmetion and operability are not affected. The design equivalence of the replacement RCS cold leg elbow (s), along with measures taken to ensure proper installation, and inspection and testing ofinstallation welds ensure that the applicable FSAR safety analysis assumptions are maintained.

There are no changes to safety limits, setpoints, or operating parameters as a result of the changes. In addition,

, no fission product barriers, which include containment, the reestablished reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by the implementation of MR 95-058'N. The margin of safety as defined in the basis of TS is not reduced. MR 95-058*N does not involve an unreviewed safety question.

(SER 96 0(2J.)

Summarv of Safety Evaluation: The replacement elbow (s) provides the same pressure boundary function, material, and piping geometry as the existing elbow (s) such that RCS function and operability are not affected.

The design parameters (e.g., size, wall thickness, material) for the replacement cold leg elbows are unchanged from the original design. 'Iherefore, the seismic response of the RCS piping / loop supports is not affected. The conclusions of the original SER remain valid. (SER 96-061-01)

47. MR 95-058*O (Unit 2), Steam Generator Replacement. MR 95-058*O physically replaces the OSGs with RSGs. It cuts and rea*taches affected piping, including the reactor coolant system (RCS) hot leg and cold leg piping, main steam (MS) piping, main feedwater (FW) piping, blowdown piping (removal only), wet layup piping (flanged installation only), MS vent piping (removal only), SG shell drain piping, and SG channel head vent piping.

Summarv of Safety Evaluation: The licensing design basis for the affected piping systems is to maintain the integrity of the RCS pressure boundary / fission product release barrier, maintain the integrity of the secondary side pressure boundary containment barrier, function to remove RCS decay heat, and minimize the likelihood of a high energy line break (e.g., loss of coolant accident, steam line break or feedwater line break). Installation activities (e.g , pipe cutting, preparation and welding), piping modification design, and installation weld testing / inspection are performed in accordance with ASME Code requirements, as reconciled to original plant construction code, USAS B31.1-1967, including ASA B31 Code Cases N-7 and N-10 for the RCS piping and fittings.

New material and/or modifications of the affected piping systems are engineered-equivalent to the original construction code and design requirements to be equivalent in pipe routing and support, therefore pipe failure mechanisms are not adversely affected. Use of ASME Code-qualified w elding processes and inspections and

. testing of the welds in accordance with Code requirements ensure the integrity of the primary and secondary side pressure boundary installation welds. Piping and equipment reliability is verified so that the probability of a malfunction of equipment important to safety previously evaluated in the FSAR are not increased.

Important to safety components which mitigate the consequences of design basis accidents are taken out of service only when the plant is configured in such a manner that the component is not required to be operational.

Following completion of MR 95-058*0, components required to mitigate the consequences of accidents evaluated in the FSAR are fully available to fulfill its important to safety functions when required.

Affected components are capable of performing its safety function as described in the Basis of TS for conditions spe:ified.

P.ge 57

There are no changes to safety limits, setpoints, or operating parameters as a result of these changes. In addition, no fission product barriers, which include containment, the reestablished reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by MR 95458*0. The margin of safety as defined in the Basis of TS are not reduced. MR 95-058'O does not involve an unreviewed safety question. (SER 96470)

48. MR 95-058*P. (Unit 2), Steam Generator Replacement. MR 95-058'P temporarily removes and reinstalls the SG wide range and narrow range water level instrumentation piping necessary for the water level instrumentation configuration of the RSGs. Its includes the piping, tubing, ralves, piping connections and its associated supports between the SGs and the water level transmitters. l Summarv of Safety Evaluation: Implementation activities above the SG cubicle walls in the vicinity of the reactor cavity are limited during refueling shutdown (when the reactor vessel head is removed) and refueling ,

operations in order to preclude malfunction of equipment important to safety from foreign material entering the !

reactor cavity. MR 95-058'P does not adversely affect equipment failure mechanisms. MR 95-058'P does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in ,

the FSAR.

MR 95-058*P does not affect the ability of affected equipment to perform its important to safety functions. SG water level instrumentation piping / tubing is designed to be equivalent in material used, installation design and seismic qualification. There are no adverse effects to accident initiation mechanisms. The changes do not adversely affect other structures, systems, or components. MR 95-058'P does not affect accident initiation mechanisms. Therefore, the possibility of an accident or a malfunction of equipment important to safety of a difTerent type than previously evaluated in the FSAR is not created.

The SG water level piping / tubing is capable of performing its safety functions as described in the Basis of TS.

The TS do not require the operability of the affected components during plant conditions specified for MR 95-058*T. Equipment is returned to an operable condition as required by TS. No fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by implementation of MR 95-058'P. MR 95-058'P does not reduce the margin of safety as defined in the Basis of TS MR 95-058'P does not involve an unreviewed safety question. (SER 96-079)

49. MR 95-053L*Q,(Unit 2), Steam Generator Replacement. MR 95-05G*Q replaces the insulation of the Unit 2 SGs and the SG piping connections.

Summary of Safety Evaluation: The Nukon insulation system does not adversely affect the containment air recirculation cooling system design function of maintaining the average containment bulk atmospheric temperature s120*F, which is the initial containment temperature used in FSAR accident analyses for containment pressure and temperature. Also, the Nukon insulation system is constructed of chemically inert material so as to not produce hydrogen gas during LOCA conditions or adversely affect post-LOCA reactor coolant system (RCS) water chemistry. Following SG replacement, the replacement insulation is capable of perfonning its important to safety functions equivalent to that of the original SG insulation, and thus the -

accident mitigation analyses are not adversely affected.

Common mode failures that disable both trains of safety equipment are not created by MR 95-058*Q. .

Accident mitigation is unaffected since accident mitigation required equipment continue to meet applicable requirements. The consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR are not increased. )

MR 95-058'Q does not add equipment failure mechanisms or new accident initiators. Design analyses demonstrated that Nukon insulation debris generated by a LOCA does not adversely afTect the flow of water through the ECCS sump screens, which could cause a common mode failure of the ECCS. Therefore, the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated in the l'SAR is not created.

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.. - - - - . = - .. .-. ~ - ~ .

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The Nukon insulation system does not adversely affect the ECCS, the stainless steel pressure boundaries or that .!

initial containment temperature used in FSAR accident analyses. During a worst case pipe break LOCA, the j resultant Nukon insulation debris causes an insignificant head Icss across the vertical ECCS sump screens i which is not detrimental to the effectiveness of the ECCS.

The Nukon insulation system, including blankets, stainless steeljacketing, and support system remains intact

' during postulated earthquake conditions to prevent accident mitigating systems from adversely being affected ,

by falling insulation debris. Installation of the replacement insulation does not affect the ability of equipment to perform its safety function as described in the Basis of TS. There are no adverse effects on safety limits, i setpoints, or operating parameters as a result of MR 95-058'Q. In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding and fuel, are degraded by these

, changes. MR 95-058*Q does not reduce the margin of safety as defined in the Basis of TS. MR 95-058*Q does not involve an unreviewed safety question. (SER 96-075) i A

50. MR 95-058'R. (Unit 2), Steam Generator Replacement Upper Cubicle Work Platform. MR 95-058*R installs and removes temporary work platforms that provide support for scaffolding, additional laydown, and

workspace for activities around the SG during steam generator replacement.

Summarv of Safety Evaluation: The temporary platforms have no adverse impact on SG cubicle structures resulting from temporary platform installation and support of design loads. Procedural controls coordinate platform use so platform design bases are maintained.

Platform installation is limited to cold shutdown (and when the SGs are not required to be operable for decay heat removal), refueling shutdown, refueling operations, or the reactor defueled condition. Associated interferences to platform installation are temporarily removed and reinstalled to its as-found configuration following platform removal. The installation locations for the temporary SG upper cubicle work platforms are proximate to important to safety components (e.g., SG water level instrumentation, upper lateral supports

. snubber hydraulic tubing, and ESF system). However, during plant conditions specified for platform j installation, use, and removal / restoration activities, these components are not initiators for accidents previously ;

evaluated in the FSAR.

Affected components are capable of performing its safety function as described in the Basis of TS for conditions specified. Full access to the SG upper lateral support ring hydraulic snubbers are maintained, so that i the capability to satisfy snubber operability requirements is retained. TS do not require the operability of components affected by MR 95-058*R implementation during the plant conditions specified. Components are returned to an operable condition and fully available as required by TS. There are no changes to safety limits, setpoints, or operating parameters as a result of MR 95-058*R. In addition, no fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by MR 95-058* R. The margin of safety as del'med in the Basis of TS is not reduced. MR 95-058'R does not j involve an unreviewed safety question. (SER 96-0M) >

Summary of Safety Evaluation: MR 95-058*R also installs a temporary jib crane on each OSG steam drum to provide the rigging capability needed to assemble the upper cubicle work platforms and to rig other cubicle loads associated with removal of the OSG steam drum / lower assembly. i l

OSG steam drum temporary jib crane and upper cubicle work platform installation and removal including interference removal and reinstallation, is performed during cold shutdown (and when the SGs are not required to be operable for decay heat removal), refueling shutdown, refueling operations or with the reactor defueled.

Prior to removing the OSG steam drums from the containment the OSG steam drum temporary jib cranes are removed from its respective steam drum. The temporary jib crancs are only used in conjunction with the OSG steam drums which serve as the crane mounting structure,(e.g., it is not used' mounted on the RSG steam drums). The conclusions of the original SER remain valid. (SER 96-068-01)

51. MR 95-058's. (Unit 2), Steam Generator Replacement. MR 95-058'S replaces the Umt 2 Model 44 SGs with Model A47 SGs.

Summarv of Safety Evaluation: The modification does not result in a condition where the material and construction standards, which were applicable prior to change, are altered. System integrity is maintained.

Design changes are incorporated to the SGs to enhance the overall reliability and maintainability of the SGs.

1 The RSGs are not initiators of accidents previously evaluated in the FSAR, with the exception of an SG tube rupture. The pressure boundary integrity is the same as originally licensed, and provides enhanced resistance to corrosion which reduces the potential for tube degradation. In addition, other design features reduce the potential for the tube degradation or damage. Thus, the probability of a tube rupture is decreased. The existing separation of the control and protection fanctions related to the RSGs such as SG level control and low-low , ,

level reactor trip setpoints are not adversely impacted and appropriate separation is maintained. Therefore, the probability of an accident previously evaluated in the FSAR is not increased by the modification.

The systems and components evaluation performed included evaluation of systems, structures, or components at uprated power level, over a full power Tavg range of 557-i at either 2000 or 2250 psia RCS pressure, in addition to the evaluation with the replacement SGs. While the evaluation identified concerns for SSC related I to the uprated power level and the Tave range, no adverse impacts on systems, structures, or components were identified due to the RSGs.

The SG tube rupture analysis was re-evaluated with the RSG parameters and the results v ere that the radiological releases remained well below acceptance criteria of"a small fraction of" 10 CFR 100 values.

The overtemperature, pressurizer and SG low-low level setpoints have the potential to affect the consequences of a previously evaluated accident. The TS setpoints are based on the setpoints used in the Westinghouse safety analysis. The safety analysis demonstrates that the consequences of previously evaluated accidents are not increased.

A structural and thermal-hydraulic evaluation of the RSGs was performed. The safety analysis clearly demonstrates that the SG can function as required to mitigate the consequences of previously evaluated accidents, and in the case of an SGTR is designed such that the consequences remain within the acceptance limits. The RSGs do not change, degrade or preclude the prevention of mitigation or the radiological consequences of accidents described in the FSAR.

The RSGs have no effect on the availability, operability, or performence of the safety-related systems and components described in TS. The replacement SGs do require a change to the design features section of the TS under the description of the RCS bu' does not impact the limiting conditions of operation, the inspections or surveillances required by TS. A TS change is necessary to increase the total RCS volume for Unit 2 to incorporate ihc !arger primary side solume of the Model A47 SGs. The RSGs do not involve an unreviewed safety question. f SER 96-084) -

Summarv ofSitfgty Evaluation: This revision clarifies the previously evaluated scenarios of high energy line break (llELB) outside containment, Appendix R, anticipated transient without scram (ATWS) and station ,

blackout are bounded by the design of the RSGs and the revised accident analysis.

The station blackout evaluation has been reviewed. Calculation 89-019 was revised to serify that the TS required volume in the condensate storage tanks is adequate to ensure the ability to mairtain the unit in a hot shutdown condition for at least one hour concurrent with a loss of power as defined in the Basis for TS 15.3.4.A3. T hus the margin of safety as defined in the Basis for this TS is not affected by the RSGs. The conclusions of the original SER remain valid. (SER 96-084-01) l page 60

-~ - - - - , .

Summarv of Safety Evaluation: This revision clarifies that the effects of the slightly greater weight of the RSGs and the different center of gravity have been included in the analysis and evaluations performed to demonstrate safe operation. It also discusses the effects of the design change regarding the feedring in the RSGs. In the RSGs, the feed ring is elevated from its location in the OSGs. The results of the small break loss of coolant analysis were revised to reflect an additional assessment on the peak cladding temperature.

Weight, center of gravity and thermal expansion of the RSGs are included in the evaluation and analyses ;

performed. This included RCS loading, seismic loading, RSG loading and nozzle loading evaluations.

Calculations support these evaluations and demonstrate applicable results.

i llowever, there is an additional issue related to the design of the elevated feed ring when continuous water level is not maintained to the RSGs. During this condition, due to the elevated feed ring, the probability that the feed ,

ring could become uncovered is increased. To address this concern, the level program is changed to a constant l level program. This raises the operating water level during hot shutdown and at low power operation to ensure :

the feed ring remains covered. ]

in addition, procedural controls are in place to address conditions below 0% power. Procedures direct the j operators to not allow the feedring to be uncovered for more than 15 minutes without feed flow. This ensures i water remains in the feed piping within the RSG. This takes into account the losses of water due to boil off and ,

main feed check valve leakage. Another control restricts feed flow to a maximum of 100 gpm whenever the i feed ring is uncovered. This is a low enough flow to ensure a steam void could get pushed out of the feedwater piping prior to causing' water hammer damage to the piping. The conclusions of the original SER remain valid.

(SER 96-084-02)

52. MR 95-058'U. (Unit 2), Steam Generator Replacement, Main Steam Vent Valve Relocation. MR 95-058*U removes the vent valves 2MS-211/212 piping and vent lines located in a 1-1/2" branch line immediately downstream of the steam generator main steam outlet nozzles. This allows the steam generators to be rigged in ,

and out of the steam generator cubicles. The modification also reinstalls the vent piping and relocates the vent valves downstream of its present location to improve its operation.

Summarv of Safety Evaluation: The consequences of a steam line rupture are determined by the results of i analyses based on the initial plant conditions and operation and failure of equipment. The relocation of the main steam vent valves does not change assumptions or parameters used to perform this analysis. The valve relocations do not subject essential plant equipment in the area to the effects of a high energy pipe break. The modification relocates the existing valves or an approved equivalent downstream ofits existing location to permit safe plant access for plant personnel. The consequences of an accident or malfunction of equipment important to safety is not increased.

'I he mam steam vent valve relocation design and installation meets or exceeds the existing system design l requirements. Implementation of the modification does not affect the ability of the main steam system to perform its decay heat removal function nor does it affect the margin of safety defined in the TS Basis.

MR 95-058*U does not create an unreviewed safety question. (SER 96-053) !

53. MR 95-058*W. (Unit 2), Steam Generator Replacement. MR 95-058*W provides the piping and littings necessary to make the transition from the 2-1/2" bottom blowdown connection to the existing 2" blowdown piping for the Unit 2 steam generator replacement. It also adds a new 2-1/2" drain valve edjacent to each SG blowdown connection. The valve provides a connection to the secondary system which can be used to facilitate draining the SGs or place it in wet lay-up.

Summary of Safety Evaluation: The valves provide a connection to the secondary system which can be used to facilitate draining the SGs or to place them in wet lay-up. The new valves meet existing system design requirements. The valves are a doubic-disk gate valve w ith a positive seal backseat. The vahe shall be provided with an equalizing port to assure that fluid cannot be trapped in the bonnet u hile the valve is shut.

Seat leakage and hydrostatic testing of the vah e shall be performed by the manufacturer. A failure of the page 61 i

j

1 modified piping / valves would have similar affects to the secondary system as a loss of normal feedwater accident. The plant response to these conditions is evaluated in the existing analysis for a loss of normal feedwater accident. The plant response to a failure of the modiGed blowdown piping is bounded by this existing analysis.

The modiGed piping system meets or exceeds the original design and construction requirements.

MR 95-058* W does not effect the assumptions of accident analysis or accident initiating mechanisms.

Therefore this modification does not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

MR 95-058*W is performed when the SGs are no longer required for decay heat removal. During this time the affected portions of the SG blowdown system is not required to perform important safety ftmetions as described ,

4 in the Basis of TS. The modified piping design and installation ensures that the system functions as originally )

designed and meets the existing design requirements. The new valves are of the same type and material as the existing system valves and a blind flange is provided downstream of the valves to assure system integrity. The margin of safety designed / constructed into the secondary side pressure boundary is not reduced.

There are no changes to safety limits, setpoints, or operating parameters as a result of the implementation of MR 95-058*W. No fission product barriers are degraded. Therefore the margin of safety as defined in the Basis of TS is not reduced. MR 95-058*W does not involve an unreviewed safety question. (SER 96-094)

54. MR 95-058'X. (Unit 2), Steam Generator Replacement, Temporary Lead Shielding. MR 95-058'X installs temporary lead shielding on piping / components (e.g., reactor coolant system (RCS), residual heat removal (RHR) and structun:s within the Unit 2 containment.)

Summarv of Safety Evaluation: Evaluations demonstrate no adverse impact on important to safety structures, systems and components to maintain require system operability during temporary shielding installed conditions.

The amount of shielding weighs so that piping allowable stresses are not exceeded considering static and dynamic loads. Temporary shielding installation and removal is limited to cold shutdown, refueling shutdown, refueling operations, or the reactor defueled.

No permanent effects to components are experienced as a result of MR 95-058*X. Provisions ensure no lead contamination to stainless steel piping occurs. MR 95-058'X during specified plant conditions does not adversely effect initiators for the accidents previously evaluated in the FSAR. Also, no equipment failure mechanisms are adversely affected. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

MR 95-058'X does not interfere with the operation of equipment. As a result of engineering and i implementation measures, adserse effects to important to safety equipment is precluded. No new accident l initiators are created by this design package that are different from those previously evaluated. No new failure 1 modes different from previously evaluated are created by this design package. Therefore, the possibility of an -

accident or a malfunction of equipment irnportant to safety of a different type than previously evaluated in the FSAR is not created.

Components are capable of performing its safety function as described in the Basis of TS. No piping stress allowable is exceeded by the installation of the temporary shielding. There are no changes to safety limits, setpoints, or operating parameters as a result of MR 95-058'X. In addition, no fission product barriers, which include containment, reactor coolant pressure boundary, fuel cladding, and the fuel are degraded by MR 95-058'X. Therefore, the margin of safety as defined in the Basis of TS is not reduced. MR 95-058'X does not involve an unreviewed safety question. (SER 96-00)

Pge (Q

55. MR 95-059. (Unit 1), Safety injection. MR 95-059 upgrades the nitrogen supply piping to the safety injection l (SI) accumulators. It installs a testable containment isolation valve (ISI-834D) associated with Penetration P-14c. A check valve and test connection / drain valve (1S1-1427) is installed inside containment to allow Type C local leak rate testing conducted to meet 10 CFR 50 Appendix J criteria.

Summarv of Safety Evaluation: The new valves, fittings and c piping meet appropriate design requirements. !

The materials used for installation are compatible with the existing system materials. New equipment meets the ;

required Code and rating rcquirements as well as containment isolation valve requirements. l 1

nis installation is not an accident initiator. A malfunction of this valve in this system will not cause an l accident. The consequences of an accident previously evaluated in the FSAR is not increased by the modification. The new check valve and test connection does not affect the integrity of the nitrogen supply to the Si accumulators. It does not affect containment integrity. The function of containment isolation is performed by this new containment isolation valve.

Containment integrity and core cooling is not reduced by this modification. The consequences of a malfunction are not altered by the modification. The nitrogen supply line to the Si accumulators is properly isolated. Unit 1 is in cold shutdown during installation.

Safety-related structures, systems, and components are not adversely affected by this work. The probability of occurrence or the consequences of an accident or malfunction of equipment imponant to safety is not increased.

De margin of safety as defined in the TS is not reduced. The modification does not pose an unreviewed safety question. (SER 96-010)

56. MR 96-022. (Common),480 V. MR 96-022 resupplies the de control power for the 1B-04 and 2B-03 buses.

Summarv of Safety Evaluation: The de control power to the safeguards buses is used to respond to accidents 1 previously evaluated in the FSAR. 'l he evaluations do not show the de control power to the 480 V switchgear i buses act as an initiator of these accidents. Therefore, MR 96-022 does not increase the probability of occurrence of an accident previously evaluated in the FSAR.

To prevent spurious operation of equipment during an Appendix R fire scenario, the de control power to the 480 V safeguards buses is de-energized. An abnormal operating procedure used to respond to such scenarios is revised to provide appropriate direction to de-energize the new de control power supplies to the IB-04 and 2B-03 buses. l The new de control power to the iB-04 and 2B-03 switchgear replaces the existing de control power to these l buses. The existing alternate de control power is spared out and the existing normal de control power becomes I alternate feed with the new feed acting as the normal. Since the existing alternate feeds are from the opposite train as the switchgear itself. accidents that could arise from utdizing the alternate feed from the opposite train

, . is eliminated. The new feeds perform the same function as the existing normal feeds under the same scenarios.

Therefore, the possibility of an accident of a different type than previously evaluated is not created.

Components are adequately sized to prevent its failure under design conditions. Additionally, it has been

( verified the additional load of the de system during testing does not prevent the de systems from performing its l safety function. Therefore, the activity does not create the possibility of a malfunction of equipment important j to safety of a different type than previously evaluated in the FSAR.

MR 96-022 does not reduce the margin of safety as defined in the Basis of TS. It does not involve an unreviewed safety question, f SER 96-.072) t!. I l

57. MR 96-026. (Unit 2), Pipe Supports. MR 96-026 upgrades 20 pipe supports in the Unit 2 feedwater, auxiliary feedwater, main steam, and safety injection systems, in addition, one pipe suppon is removed and another installed in the SI system. These changes, additions and deletions meet our commitments to NRC Bulletin 79-14 and ensure that no piping or other system components are stressed in excess of Code allowables.

Summary of Safety Evaluatim: Operability analyses for power operations (all piping systems are considered operable) and Code compliance analyses were performed. Additionally, its design accounts for concerns such as the piping systems deadwet ght load, hydroweight (especially for main steam piping), maximum thennal expansion, OBE and SSE seismic loads, and maximum dynamic deflections. The work is performed when the affected piping system is not required to be operable or when the work does not affect operability of the piping system.

The design changes enhance the affected piping systems ability to withstand loading conditions and are acceptable under normal, emergency, faulted, and operability conditions. It is performed in a manner such that the affected systems capacity and operability is not degraded, nor is the systems function impaired. There is no , ,

effect on piping system pressure boundary or the integrity of Unit 2 containment pressure boundary from these modifications. The modiHcations do not involve an unreviewed safety question nor does it require a change to i TS. (SER 96-121)

58. MR 96-048. (Common), Main Transformer. MR 96-048 repowers X-27 NSB main transformer and its )

associated loads from H-01 breaker H52-14 to H-01 breaker H52-16. The load side cables on H52-14 are >

removed, pulled back to manhole 16; rerouted to H52-16 and connected to the load side of breaker H52-16.

The method of cable termination in the H52-16 cubicle is equivalent to that used in the existing H52-14 cubicle.

Summarv of Safety Evaluation: Repowering X-27 from H52-14 to H52-16 does not alter systems, structures or components described in the FSAR. Although the 13.8 kV system is described in the FSAR,its function is unchanged. H52-14 and H52-16 are capable ofinterrupting fault conditions applied to the 13.8 kV system.

Relaying is in place to isolate bus H-01 from fault due to 1152-14, H52-16, X-27 or cabling connected to the systems. The protective relaying installed is consistent with the existing protective relaying on the system and is considered adequate to assure isolation of fault condition. Since no formal calculation existed on the acceptability of the protective relay settings on breaker 1152-14, Calculation 96-0197 documents the acceptability of these settings. (SER 96-09Q

59. MR 96-048. (Common),13.8 kV. MR 96-043 powers the new switchyard auxiliary transformer from breaker H52-14 on 13.8 kV bus H-01.

Summarv of Safety Evaluation: The protective relaying on H52-14 is consistent with the existing protective relaying system and is considered adequate to ass ure isolation of fault conditions. The H52-14 breaker is capable ofinterrupting and isolating fault supplied from the 13.8 kV system.

The new transformer adds a maximum of 21 amps to H-01. Therefore, the maximum current on 11-01 with one -

of the two bus tie breakers to 11-02 or H-03 closed a1d assuming transfor.ners are loaded to capacity is 1986 amps. This is less than the 2000 amp rating of the bus. Therefore the addition of this load does not increase the probability of a loss of offsite power. .

The 13.8 kV system is not required for FSAR accidents. However, it is used for Station Blackout and Appendix R. In the event of a loss of offsite power, the 13.8 kV system can be supplied from G-05 combustion turbine. The worst case loading for station blackout or Appendix R is approximately 2542 kW. The additional load added by this modification does not affect the ability of G-05 to supply safe shutdown loads for either station blackout or Appendix R. Therefore, this modi 0 cation does not increase the probability or consequences of an accident previously evaluated in the FSAR. This activity does not increase the probability of a loss of offsite ptmer, station blackout, or Appendix R. Protective devices isolate the new load in the event of a fault or m erload 1 his is consistent with the design of other H-O l loads.

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The modification does not affect the ability of the 13.8 kV buses to supply offsite power to plant equipment nor !

does it affect the ability of G-05 to supply power to the 13.8 kV system. In addition,it does not reduce the i availability of offsite power during installation or testing. The modi 5 cation does not reduce the margin of

' safety defined in the Basis for Technical Specification or Station Blackout / Appendix R commitments.

MR 96-048 does not involve an unreviewed safety question. (SER 96-113)

60. MR 96-049 and MR 96-050. (Unit 1), Safety injection. Present Si test line pressure gauges 1&2PI-913 A&B ]

have a poor history of retaining calibration accuracy. They are normally isolated and cut in only during the l quarterly Si pump Dow and pressure test. Often the gauges are over-ranged while being cut in, shich adds to l the LCO time while a replacement gauge is being procured. The modTications remove Si test line pressure !

gauges 1&2PI-913 A&B which return the pressure transmitter tubing to its original configuration. It installs a dropping resistor in the cunent loop of each SI pump discharge pressure transmitter (l&2-PT-922 & 923) and a Moore digital panel meter to monitor the voltage drop across the resistor.

Summary of Safety Evaluatim: During the installation, if the pumps are required to run, increased operator monitoring is directed and SI pump flow and current draw are still available as the primary means of cMermining pump performance during accidents. The modifications do not affect the accuracy or mechanical integrity of the pressure transmitters or indicators which would be ured for post-accident monitoring of Si pump performance.

Since '.he instrumentation involved has no accident initiation mechanism, and has no control or safety protective functian, the changes do not increase the probability of occurrence or consequences of an accident or increase the probability of occurrence or consequemes of a malfunction of equipment important to safety previously evaluated in the FSAR.

l The modifications do not create the possibility of a malfunction of equipment important to safety of a different !

type than previously evaluated in the FSAR, but in fact reduces the number of Httings and tubing associated j with the pressure transmitter. This reduces the possibility of failures which would affect equipment. Proper j post-maintenance testing ensures acceptable mechanical and indication performance. This change also does not i effect the margin of safety defined in the Basis for TS. The modifications improve the instrumentation used in the Inservice Test program and dees not involve an unreviewed safety question. (SER 96-1.l'[)

61. MR 96-055'fL (Common), Main Control Board. MR 96-055'B relocates the 1&2 C-20 annunciator horn from I the inside of the rear panel to the outside of the front panel of 1&2C-20 for better audibility of the horn. The l annunciator horn and annunciator panels l&2C-20A&C are temporarily inoperable during disconnection and I subsequent reconnection in the new circuit.

Summary of Safety Evaluation: Train separation is maintained in the new wire routing. The new horn installation is seismic. The horn and new wire runs are installed first and then disconnections and new terminations done last to minimize the amount of time the annunciator horn and annunciators are inoperable. A test of the new horn circuit is performed after installation.

The modification does not change the alarm (annunciator) function or descriptions as stated in the FSAR. The main control board integrity is not compromised by the mounting of the new horn. The change does not insolve an unreviewed safety question. (SER 96-088)

62. MR 96 063. (Unit 2),345 kV. MR 96-063 replaces the Unit 2 main generator output circuit breaker,2F52-142.

The breaker provides the primary connection between the Unit 2 main power transformer and the 345 kV electrical grid. The existing GE Type ATB breaker utilizes an air-blast operating mechanism and has demonstrated substantial air seal leakage during periods of extremely cold weather. The replacement ABB Type PM 362KV SF6 breaker is expected to provide improved cold weather performance, as well as better spare parts asailability and better overall reliability than the existing breaker over the remainder of plant life.

Page M

Summarv of Safety Evaluation: Although breaker 2F42-142 is not classified as safety-related, it is capable of initiating several FSAR accidents. The breaker can therefore be considered a system, structure, or component important to safety. The replacement of the breaker via MR 96-063 does not increase the probability of failure of the breaker or its associated circuits. Therefore, neither the configuration changes nor the interim conditions associated with the modification increases the probability of malfunction of system, structure, or component imponant to safety.

Replacement of breaker 2F52-142 does not introduce new failure mechanisms for circuits or equipment. SF6 is already utilized as an insulating medium. Therefore, neither the con 6gurations nor the interim conditions associated with the modification creates the possibility of new types of accidents or equipment malfunctions not previously evaluated in the FSAR.

TS Section 15.3.7.A.I defines the requirements for offsite power availability to sa6 guards buses, as well as 345 kV line requirements for power generation from one on both units. The configuration changes associated with replacement of breaker 2F42-142 does not increase the probability of failure of the breaker of the breaker or its associated circuits, and therefore does not increase the probability of a loss of offsite power or 345 kV ;

lines. The installation and testing is performed with Unit 2 in refueling shutdown and 345 kV bus Section 4 deenergized and grounded. Under these conditions, the TS requi e that only two 345 kV lines be available. A minimum of two lines will be available throughout the installation The interim conditions associated with the installation do not reduce TS margins of safety. This modification does not involve an unreviewed safety question. (SER 96-112)

63. MR 96-068*B. (Unit 2), Containment. MR 96-068'B cuts and caps supply and return lines in containment heating steam lines. It affects the system piping passing through containment penetrations #52 and #53 and has no impact on the containment penetrations.

Summary cf Safety Evaluation: The design safety function of containment was considered since the modi 6 cation involves the pressure boundary of containment. The evaluation determined that the modification does not affect the design safety function of containment since the function of several exicting containment isolation valves is determined by permanently welded pipe caps inside containment. The cansequences of failure of the modification were determincd to be unchanged from the existing configuratio t The modification is designed with both seismic Class 1 and Seismic 2/1 considerations and therefore has no a dverse effects on safety-related equipment. The modification has been determined to have no effect on the '.nitiation of accidents described in the FSAR, since containment is an accident mitigaer. The margin of safety has slightly improved due to the improved reliability of the piping configuration by the modification. The welded caps are more reliable against leakage than valves. Therefore, the modification does not involve .m unreviewed safety question. (SER 96-133) 6-L MR 96-072,(Unit 2). Engineered Safeguards Feature. MR 96-072 replaces Unit 2 EOF test switches and removes spare " press-to-test" indicator sockets. The new switches are equivalent to the xisting switches. The work requires deenergizing both traira of safeguards control power for Unit 2. Only one trab of safeguards is -

deenergized at a time. The work is performed when Unit 2 is defueled or in cold or refueling shtadown with no refueling operation in progress.

Summary of Safetv Evahiation: With the specified plant conditions, the only TS safeguards functior. required is service water pump automatic start. During replacement of the Train A test switches, all three se vice water timers are out-of-service. During this period, automatic service water cooling cannot be assured to the EDG aligned to the Unit 2 Train A bus. Therefore, an LCO is entered on both units, prior to starting Train A replacement, on the standby emergency power supply to 2A-05. Alignment of the other Train A EDG does not correct the deficiency because the timers have no power. An LCO is not required for Train B because Train B EDGs are not dependent upon service water for cooling. Additional requirements are placed on G-01 EDG operability during Train A replacement because of cross unit electrical alignments required in DCS 3. L I 7 for outside air temperature and fuel oil supply. During Train A replacement alignment of service water pumps is specified to ensure service water system operability and cooling of potentially connected EDGs. If performed Page 66

l i

with fuel in the core, administrative restrictions are in place to ensure containment isolation and containment ;

ventilation isolation can be achieved within the shutdown condition time requirements. This activity does not l increase the probability or consequences of a malfunction of equipment important to safety and does not reduce the margin of safety specined in the basis of TS. Therefore, replacement of the Unit 2 ESF test switches and removal of spare " press-to-test" indicator sockets does not involve an unreviewed safety question.

(SER 96-126) i 1

65. MR 96-073. (Unit 2), Residual 11 eat Removal. MR 96-073 removes three supports and replaces Ove other )

existing pipe supports with enhanced support designs in Unit 2 residual heat removal / letdown and cross-tie piping (RH/CV); replaces an existing insulation saddle on an RiiR support with plate steel and stainless steel shim stock; removes a snubber from the auxiliary feedwater system, and adds one pipe support to the component cooling water system. These changes maintain compliance with design basis requirements for pipe stress and support load capacity and ensures that no piping supports, structures, or in-line components are stressed in excess of Code allowables.

i 1 Summarv of Safetv Evaluation: This work is completed prior to relying on the RHR system for core reload.

The Code compliance analyses demonstrate that once the modi 6 cations are completed, piping will be within design basis Code allowables for pipe stress and support loads. The material for modincation of AC-60lR-R40 meets the compatibility requirements for materials inside containment. )

l The design changes enhance the affected piping systems ability to withstand loading conditions, and are l acceptable under normal, emergency, faulted and operability conditions. It is performed in a manner such that i the affected system capacity and operability are not degraded, nor is the system functions impaired. There is no )

effect on piping system pressure boundary (with the exception of surface grinding on the RHR piping that is tested to ensure the pressure boundary is unaffected), or on the integrity of Unit 2 containment pressure boundary from any of the modifications. The changes do not involve an unreviewed safety question.

(SER 96-134) i I

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TEMPORARY MODIFICATIONS The following temporary modifications were implemented in 1996:

1. IM 96-001. Containment. TM 96-001 installed ACR Smart Readers in containment to monitor ambient temperatures of ennronmentally-qualified (EQ) equipment. Four-channel ACR Smart Readers temperature

loggers and one eight-channel ACR Smart Reader temperature logger with up to seven remote temperature probes are used to log and store temperatures for one refueling cycle. The data loggers are located outside the missile shield. Data loggers are attached to equipment supports with Terzel Ty-Raps. Remote temperature -

probes are attached to equipment support or conduits with Tefzel Ty-Raps.

Summarv of Safetv Evaluation: The readers are independently powered and completely telf-contained. The only identified potential impact on safety-related equipment is a potential to partially block the sump screen in the event of a loss of coolant accident (LOCA). LOCA may include a maximum temperature 280*F, exposure to chemical spray and radiation doses up to 1.5 x 10' rads.

The ACR ' Smart Readers are contained in Noryl(polyphenylene oxide) cases. The melting temperature is

>450*F. Polyphenylene oxide is resistant to nonoxidizing acids (e.g.,20% sulfuric acid and boric acid),

aqueous salt solutions (Nacl), aqueous alkali solutions (NaOH), and water.

EPRI NP-2129 states: "The material (polyphenylene oxide) maintains slightly greater than original strength to approximately 10' rads." The 1-year integrated gamma dose level inside containment per FSAR Figure 14.3.4-15 is 1.5 x 10' rads. Therefore, the temperature logger is expected to maintain its structural integrity during and following a LOCA.

The data logger contains no aluminum. Therefore it does not contribute to the generation of hydrogen as a result of the sodium hydroxide spray.

The Ty-Raps will remain in place during a LOCA. Therefore, the Ty-Raps do not contribute to sump clogging or affect an accident in another way. The change does not involve an unreviewed question. (SER 96-01 n

2. TM-96-002, Containment SGRP Temporary Power. TM 96-002 provides temporary electrical power to SGRP loads within Unit 2 containment during steam generator replacement. It installs, uses and removes a temporary electrical distribution system which connects to the electrical power feeders for the 480 vac backup pressurizer ,

heater groups 2-Tl A and 2-Tlf3 (fed from non-safety power panels PP-15 and PP-16), and one 4.16 kV reactor coolant pump (RCP) motor feeder.

Summary of Safety Evaluation: The establishment of single RilR capability is a prerequisite to RCP feeder '

availability. Tag-out and energization of pressurizer heater circuits require defeat of the pressurizer level interlock. QA/QC centrols, w hich include conductor lif t/and verification ensure that removed material and disconnected terminations are returned to its as-found configuration and the affected components function properly. Since pressurizer heater and RCP motor feeder circuits do not initiate accidents during the conditions specified, and TM 96-002 design does not change accident initiation mechanisms, the assumptions of the FSAR design basis accident analyses are not affected. No design basis accident or 10 CFR 50, Appendix R assumptions are changed.

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

I Equipment described in the Basis of TS is not affected by TM 96-002. There are no adverse effects on safety limits, setpoints, or operating parameters as a result of TM 96-002. In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded by this temporary modification. Therefore, TM 96-002 does not reduce the margin of safety as defined in the Basis of TS. TM 96-002 does not involve an unreviewed safety question. (SER 96-076) l

3. TM 96 003. Steam Generator Replacement, Containment Access Building (CAB). As part of performing the replacement of the Unit 2 steam generators (MR 95-058), a temporary CAB, consisting of mobile office )

structures (trailers) are installed outside of containment in the area between the Unit 2 containment facade and q Warehouse No. 3. The CAB is used to provide office space, lunchroom and change facilities for construction !

and radiological protection personnel working on the steam generator replacement project. TM 96-003 j provides the design details for the placement of the CAB trailerr and for its removal after completion of the "

project. Upon removal of the CAB, the area will be restored to its original, as-found condition or to an acceptable condition.

Summarv of Saferv Evaluation: As demonstrated in Calculation SGRP-95-001 and TM 96-003, there are no important to safety utilities buried in the vicinity of the CAB. An evaluation demonstrated that for inadvertent ;

release of the most conservative expected amount of radiological materials within the CAB, the resultant offsite dose consequences would be well within the limits of 10 CFR 50, Appendix 1 and 10 CFR 100. The CAB fire sprinkler system is designed in accordance with the NFPA requirements. Installation and removal of the CAB 1 fire sprinkler system tie-in to the Warehouse No. 3 sprinkler system is implemented so that disabling of .

Warehouse No. 3 fire wet suppression capability is minimized. For the interim period during which the Warehouse No. 3 fire sprinkler system is out-of-service, hot work activities (e.g., cutting, grinding, welding) are administratively prohibited. Since the temporary modification does not adversely affect equipment failure mechanisms, it does not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Equipment including the safe shutdown area fire protection systems are capable of performing its safety function, as described in the Basis of TS. There are no adverse effects on safety limits, setpoints, or operating -

parameters as a result ofimplementation of this temporary modification. In addition, no fission product ,

barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded i by the changes. TM 96-003 does not reduce the margin of safety defined in the Basis of TS. TM 96-003 does not involve an unreviewed safety question. (SER 96-040)

Summarv of Saferv Evaluation: The attachment point for the wet pipe sprinkler system is downstream of the alarm valve for the Warehouse No. 3 sprinkler system, so that the alarm in the control room is activated when either the Warehouse No. 3 sprinkler system or CAB sprinkler system is in operation. The conclusions of the original SER remain valid. (SER 96 040-0,_1.)

Summary of Safety Evaluation: plant Gai-tronics is prmided to the CAB for emergency notification purposes.

. The conclusions of the original SER remain valid. (SER 96-040-02) ;

Summarv of Safety Evaluation: The CAB complies with NFPA and plant fire protection requirements, including the use of a dedicated wet sprinkler system for fire suppression, and the provision of Class 2-A:C fire extinguishers for each CAB trailer. This evaluation demonstrates that for inadvertent release of the most conservative expected amount of radiological materials within the CAB, the resultant offsite dose consequences is well within the limits of 10 CFR 50 Appendix ! and 10 CFR 20. The conclusions of the original SER remain valid. (SER 96 040-03)

Summarv of Safety Evaluation: TM 96-003 will use a more conservative minimum personnel approach distance of 15' and the minimum equipment approach distance of 20' to the 345 kV lines required by OSHA.

The approach distances to the 345 kV lines are maintained during all phases of CAB installation, use and removal.

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A closed cycle, self-contained decontamination sink is provided in the CAB. The decontamination sink has an onboard, limited water supply capacity (approximately 25 gallons) with a drain tank capable of receiving the supply. The decontamination sink water supply is located in the heated (occupied) area of the CAB. Freeze protection is not required. The conclusions of the original SER remain valid. (SER 96-040-04)

4. TM 96-004. Containment Access Covered Enclosure (CACE). A temporary CACE structure is built outside of the containment in the area immediately adjacent to the equipment hatch facade door in support of the Unit 2 SG replacement The CACE is used to shelter personnel, equipment and SG movement into containment. The structure provides additional laydown space and is an extension of the radiation controlled area (RCA).

Summarv of Safety Evaluation: In order to not adversely affect the 345 kV transmission lines connected to the Unit 2 main transformers in the vicinity of the CACE, and to also protect personnel and equipment from the ,

hazards these energized lines present, precautions are implemented in accordance with Wisconsin Electric Safety Manual, Section 5, and the National Electrical Safety Code,1993 Edition (ANSI C2-1993). TM 96-004 uses the more conservative minimum personnel and equipment approach distance to the 345 kV lines of 12'6" ,

required by the National Electrical Safety Code. This minimum personnel and equipment approach distance to the 345 kV lines is maintained during all phases of CACE installatien, use, and removal following completion of the steam generator replacement. Actual placement distance between the CACE structure and these 345 kV lines is no closer than 19'1" Precautions are taken to ground the CACE structure, as well as temporary structures and equipment used in CACE installation and removal.

Evaluations determined that iraportant to safety structures, systems, and components are not adversely affected.

This temporary modification does not adversely affect any equipment failure mechanisms, and no new accident initiators nor new equipment failure mechanisms are created. Therefore, the implementation of TM 96-004 does not create the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.

Equipment will be capable of performing their safety function as described in the Basis to the TS. There are no adverse effects on any safety limit, setpoints, or operating parameters as a result ofimplementation of this temporary modification. In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel, are degraded by these changes. Therefore, TM 96-004 does not reduce the margin of safety defined in the Basis for TS. TM 96-004 does not involve an unreviewed safety question. (SER 96-041)

Summarv of Safety Evaluation: The consequences ofinadvertent release of radiological materials present within the CACE due to fire or other event has been evaluated via Calculation SGRP-96-016. The conclusions of the original SER remain valid. (SER 96-041-01)

Summaiv of Safety Evaluatin: TM 96-004 uses a more conservatise minimum personnel approach distance of 15' and the minimum equipment approach distance of 20' to the 345 kV lines required by the OSif A. The approach distances to the 345 kV lines are maintained during all phases of CACE installation, use and removal .

following completion of steam generator replacement. The evaluation demonstrates that for inadvertent release of the most conservative expected amount of radiological materials within the CACE, the resultant offsite dose consequences is well within the limits of 10 CFR 50, Appendix ! and 10 CFR 20. The conclusions of the ,

original SER remain valid. (SER 96-041-02)

5. TM 96-005. CAD /CACE Temporary Power. TM 96-005 installs cabling, protective electrical metallic tubing (EMT) and disconnect switches to provide temporary electrical power to the containment access building (CAB) and the containment access covered enclosure (CACE)

Summary of Safety Evaluation: A calculation shows that load centers B-08/B-09 have additional capacity to provide the necessary power to the CAB and CACE without adversely affecting existing loads on B-08/B-09.

Spare breakers 54C and 5KD, along with the fused disconnect su itches, ensures thn no faults are propagated to imponant to safety equipment The cables uhich tie into power equipment meet hre retardancy standards Page 70

idatified in Tables 4.1.5-1,4.1.5-2, and 4.1.5-3 of tl.c Fire Protection Evaluation Report, and the temporary wiring is in accordance with the National Electrical Code (NFPA 70-1993). The cables are megger-tested prior to circuit energization and the circuits tested for proper voltage levels and proper phase sequence before powering loads. In conjunction with CAD /CACE removal, the temporary cables are disconnected and removed, and load centers B-08/B-09 are restored to their as-found configuration. The implementation of TM 96-005 does not adversely affect equipment failure or accident initiation mechanisms, and no design basis accident or 10 CFR 50, Appendix R assumptions are changed. Therefore, the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

Equipment is capable of performing its safety function as described in the Basis to the ~1 S. There are no adverse effects on safety limits, setpoints, or operating parameters as a result ofimplementation of TM 96-005.

In addition, no fission product barriers, which include the containment, reactor coolant pressure boundary, fuel cladding, and fuel, are degraded by the temporary modification. Therefore, TM 96-005 does not reduce the margin of safety defined in the Basis for TS. TM 96-005 does not involve an unreviewed safety question.

(SElit f42)

6. TM 96-011, Containment Penetration P-56. TM 96-01I routes communications cabling through containment penett. vion 56. It involves substituting a temporary electrical penetration assembly for the inside flange and consisting of a metal flange with sealed electrical fire-retardant cabling that terminates at electrical quick disconnects on both sides of the assembly.

Summarv of Safety Evaluation: 'Ihe additional loadings placed on the penetration flanges, including seismic loadings, were evaluated as acceptable. No accident initiation mechanisms are affected; and therefore, the assumptions of the design basis accident analyses are not adversely affected.

Accider.. witigation is not affected since accident mitigation required equipment continues to meet applicable requirements. No accident mitigating equipment is adversely affected by this temporary modification. The assumptions of the design basis accident analyses are not affected. Therefore, TM 96-011 does not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Implementation of TM 96-011 does not adversely affect the ability of plant systems, structures, or components to perfonn its important to safety functions, nor does it affect accident initiation mechanisms. Containment Penetration 56 maintains the accident mitigation function of Unit 2 containment during implementation windows. No important to safety structures, systems or components are adversely affected and no new equipment failure mechanisms are created by TM 96-011. No new accident initiators are created by this temporary modification. Therefore, implementation of TM 96-011 does not create the possibility of an accident or malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.

Affected equipment is capable of performing its safety function as described in the Basis of TS for the

, conditions specified. Following penetration restoration and prior to Unit 2 exiting cold shutdown, TS-required leak rate testing be performed via ORT 58 to verify the as-left penetration is capable of performing its safety-related containment integrity function. There are no adverse effects on safety limits setpoints or operating parameters as a result ofimplementation of TM 96-01). In addition, no fission product barriers, whicn include the containment, reactor coolant pressure boundary, fuel cladding, and fuel are degraded by TM 96-011. Therefore, TM 96-011 does not reduce the margin of safety as defined in the Basis of TS.

TM 96-011 does not involve an unreviewed safety question. (SER 96-095)

Summary of Safety EvaluatieD: The accident mitigation function of Unit 2 containment is maintained by having Containment Penetration 'No. 56 performed during the reactor coolant system (RCS) fueled condition with the temporary electrical penetration assembly installed and verified acceptable (pressure tested at 6 psig, leakage s 20 slpm). the outside containment blank flange is bolted in place or is capable of being bolted in place in the time required by Cl -l E. " Containment Closure Checklist.'

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Also temporary electrical assembly installation and removal sequences are designed to ensure that a containment closure capability is provided on at least one side of penetration 56 during Unit 2 shutdown fueled conditions. The penetration remains in a sealed condition during Unit 2 shutdown fueled conditions, either with the test rig, the temporary penetration assembly, or one of the existing blind flanges.

When Unit 2 is defueled, the penetration assembly may be removed and the cabling directly passed through penetration 56. Approved temporary fire barrier material is installed in the penetration for this interim configuration. The conclusions of the originct SER remain valid. (SER 96-095-01)

7. TM 96-015, Service Water. TM 96-015 removes valve SW-660, a service water throttle valve and replaces it with two blind flanges. After installation, flow will be through SW-661 instead of both or either SW-660 and SW-661. TM 96-015 is removed after SW-660 is repaired and reinstalled. .

I Summarv of Safety Evaluation: SW-660 is an ASME Class III Section 3 valve. Therefore, the blind end flanges are procured as ASME Section 3 to maintain the system design. A seismic review per ,

Calculation N-92-092 ensures the adequacy of the piping with valve SW-660 removed.

TF - n argin of safety for the SFP cooling is not defined in the TS, but is stated in Section 9.3 as being able to have one train of SFP cooling available. SER 96-090 addresses the isolation of SFP cooling during removal and installation of SW-660. With SW-660 removed, SW-661 is able to pass 1250 gpm of service water to adequately cool one SFP heat exchanger. Therefore, the margin of safety is not reduced.

During removal of SW-660, SFP cooling is isolated. The' concern is to ensure boiling does not occur. Repair of SW-660 is anticipated to be completed prior to U2R22, but it is acceptable to leave SW-660 out of the piping system during the refueling outage since SW-661 is capable of passing sufficient service water flow to remove the heat load of one SFP heat exchanger. No dry cask loading or fuel handling is performed during removal l

and reinstallation of SW-660 since service water cooling is isolated via 01-115,"SFP Service Water Cooling for Maintenance ' The change does not involve an unreviewed safety question. (SER 96-092)

Summarvpf Safety Evaluation: Procedure 01-115 states that a contingency plan is developed for work performed. The contingency for the repair of SW-660 is the installation of blind flanges, which is the same as the work perfomed. The conclusions of the original SER remain valid. (SER 96-092-01)

8. TM 96-017 and 96-018. Spent Fuel Pool Coo:ing. TMs96-017 and 96-018 repair outlet motor-operated valves of the spent fuel pool (SFP) heat exchangers. These valve automatically shut after a safety injection (SI) actuation signal when less than four service water (SW) pumps start to isolate non-essential service water flow.

Repair involves isolation of service water to and from both SFP heat exchangers; removal of one of the valves; installation of a blank in the valve's place; normalizing service water to the remaining SFP heat exchanger; repair of the sahe; isolation of service water to and from both SFP heat exchangers; removing the blank flange; reinstalling the valve; and repeating these steps for the other vahe. Isolation and normalization of the SFP cooling system is controlled via 01-115,"SFP Service Water Cooling Isolation for Maintenance." -

Summary of Safety Evaluation: The blank flange is designed to retain the piping design pressure and meet the requirements of ASME Section 111 Class 3. The temporary piping configuration is seismically acceptable when the blank is installed in place of the valve. The blank flange weighs less than the removed valve assembly.

Also, the blank flange is installed in the same manner as the valve, between the flanged ends of the piping. The piping remains rigidly connected. In addition, the inlet service water isolation valves to both heat exchangers are leak tight. Therefore, a flooding concern during valve removal and reinstallation is alleviated.

The design capacity of one train of SFP cooling was verified by Calculation 96-0087. The results indicate that one SFP heat exchanger can maintain the SFP within its temperature limits with a service water temperature of 75'F and one-third of the core off-loaded. Since the work is done at a time when the decay heat load of the S1 P is less than that analped in this calculation operating nith one SFP heat exchanger is acceptable during the activity .

Page 72

The valve repairs are done prior to the Unit 2 refueling outage (U2R22) and when no dry cask loading is performed. Recent SFP heat-up tests conducted with both SFP heat exchangers out of service indicated a temperature rise of 1.5'F/hr. The estimated time for removal or reinstallation of each SW-2930 valve is less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . At 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , though, the temperature rise would be 12*F. The normal operating temperature of the SFP is less that 100*F/hr. The temperature limit of the spent fuel pool is 120'F which allows 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months before boiling in the SFP according to OP-8A and AOP-8F. OP-8A is the SFP cooling water system operating procedure, and AOP-8F is the loss of SFP cooling operating procedure. Twenty-eight hours is a conservative value based on the current SFP heatup rate. Because of the limited time both SFP heat exchangers are required to be out of service to remove or reinstall each SW-2930 valve, boiling in the SFP is not of concern. The change does not involve an unreviewed safety question. (SER 96-091)

Summary of Safety Evaluation: 01-115 requires a contingency plan for the work performed. The contingency plan for the repair of valves SW-2930A&B is the installation of a blank Range in place of the removed valve.

This contingency is contained in the work plans for WOs 9609261 and 9608520. The conclusions of the original SER remain valid. (SER 96-091-01)

Summarv of Safety Evaluation: This revision removes the reference to completing the repair of both SW-2930A&B valves prior to U2R22. Although SW-2930A valve was repaired and returned to service, the SW-2930B valve was not repaired prior to the start of U2R22. The restriction on the repair of these valves is that both must be repaired prior to the start of fuel motion not prior to the start of the outage. This revision does not change the purpose, conclusions or evaluations contained in SER 96-091-01. The conclusions of the original SER remain valid. (SER 96-091-02)

9. TM 96-019. Unit 2 Containment. TM 96-019 installs an ACR Sman Reader in Unit 2 containment to monitor ambient at FT-474, FT-475 and PT-498. One 8-channel ACR Smart Reader temperature logger with up to 7 remote temperature probes are used to log and store temperatures for one refueling cycle. The data logger is located on El 66' of Unit 2 containment El 66'.

Summarv of Safety Evaluation: This equipment does not affect containment sump operation in the event of a loss of coolant accident. The data logger contains no aluminum. Therefore, it does not contribute to the generation of hydrogen as a result of the sodium hydroxide spray.

Since the data logger case is resistant to fluids to which it may be exposed to during an accident, it will maintain its structural integrity following a LOCA.

The Ty-Raps installed will remain in place, even through a loss of coolant accident. Therefore, the Ty-Raps do not contribute to sump clogging or affect an accident in any other ways. The change does not involve an unreviewed safety question. (SER 96-IDD O

a! ( ,

SPARE PARTS EOUIVALENCY EVALUATION DOCUMENTS (SPEEDS)

The following SPEED was implemented in 1996:

SPEED 96-050. SPEED 96-050 replaces existing safety-related 4.15 kV Dil breakers with vacuum breakers of the same ratings and qualifications.

Summarv of Safety Evaluation: The new breakers are lest subject to wear and are more easily maintained. The failure rate of the new breakers is expected to be as low as or lower than that of the existing breakers. ,,

Therefore, the probability of the loss of power to safety-related equipment supplied by these breakers is not increased. A single failure of any of the 4.16 kV breakers cannot result in the loss of all offsite power. Testing and qualification on the breakers ensures that no common failure modes exist. Besides the loss of offsite power, there are no accidents evaluated in the FSAR which could be initiated by these breakers. Therefore, the probability of occurrence of an accident previously evaluated in the FSAR is not increased. The ability of the breakers to perform its functions during accident conditions is not reduced. Therefore, the new breakers do not increase the consequences of an accident previously evaluated in the FSAR.

The primary failure mechanism is essentially unchanged for the existing and replacement breakers, namely the inability to interrupt a current are during contact separation. The failure in vacuum applications is generally localized to the contacts within the bottle compared to a more catastrophic failure for air circuit breakers.

Therefore, a failure of a new breaker will most likely result in less severe consequences than the failure of an existing breaker. Other breaker failure modes including failure to open or close when required would result in identical consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.

Since the ratings and qualifications of the new breaker are equal to or better than the existing breakers, the margin of safety as defined in the 13 asis for TS is not reduced. This change does not involve an unreviewed safety question. (SER 96-!?])

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MISCELLANEOUS EVALUATIONS The following miscellaneous evaluations requiring safety evaluations were implemented in 1996:

1. Carbohydrazide as a Feed train Wet Layuo Chemical Treatment. During wet layup, carbohydrazide (CHZ) is added to the secondary plant feed train system to provide increased protection of the feed train tube side protective magnetite layer and to scavenge dissolved oxygen.

Summary of Safety Evaluation: Catalyzed hydrazine was used for feed train wet layup treatment chemical to scavenge oxygen and help control pH. Catalyzed hydrazine was chosen since it reacts faster with dissolved oxygen than uncatalyzed hydrazine at room temperature, it is added to the chemical addition pots at candensate pumps IP-25A or 2P-25B just prior to breaking vacuum as the unit enter its respective refueling shutdown. The wet layup chemicals mixed with the feedtrain bulkwater as it flows through the feedwater heaters and back to the condenser. When the wet layup chemical reaches the appropriate concentration as measured at the No. 5 feedwater heater outlet sample point, the feedtrain is considered to be in wet layup and the condensate pumps and feedwater heaters are then isolated. CHZ is added to the feed train for wet layup the same way.

CHZ can be measured directly using spectrophotometric methods. However, the time of analysis is rather long.

To reduce the analysis time, hydrazine is added with the CHZ for feedtrain wet layup. Instead of analyzing for CHZ, Chemistry analyzes for hydrazine since the hydrazine analysis is relatively quick and simple. CHZ will not interfere with the analytical method for hydrazine. To confirm this, known concentrations of CHZ and hydrazine were added together and analyzed. Results confirmed no chemical interference existed between CHZ and hydrazine. Therefore, with hydrazine and CHZ added to known concentrations in the feedtrain for wet layup, verification of wet layup CHZ concentration is accomplished by analyzing for hydrazine.

The probability and consequences ef a previously evaluated accident, malfunction of equipment, or possibly an accident of a different type is not increased since CHZ serves as a corrosion inhibitor by passivating metal surfaces and scavenging oxygen. In addition, equipment important to safety is not affected since CHZ reduces corrosion. The results show that using CHZ as a feed train wet layup additive does not involve an unreviewed safety question nor a change to Technical Specifications. (SER 96-111)

2. CR 96-247. Possible 1.oose Part in Unit 1 or Unit 2 Reactor Coolant System (RCS) or RCS Connected Fluid

. Systems in the Snent Fuel Pool. or the Independent Spent Fuel Storage installation (IMS CR 96-247 documents foreign materials that were dropped in the refueling cavity as a result of the mini-submarine impacting the core barrel stand in the lower cavity. parts were retrieved from the lower cavity including several small misce!!aneous parts that were not from the mini-submarine. Upon subsequent inspection of the mini-submarine, all pans were accounted for with the exception of a small stainless screw, a No. 4 screw approximately 3/8" in length. This screw is one of four screws located underneath the mini-submarine shroud.

It was the only one that was found detached from the sub. The screws that held the shroud on were sheared off when the shroud detached from the submarine. There was no indication that this screw was sheared off.

Summary of Safety Evaluation: Evaluations have been performed in the past regarding different debris or parts found or potentially left in the RCS, chemical volume and control system, residual heat removal or emergency core cooling systems. Items addressed include split pin remnants, nonle dam flange ring inserts, miscellaneous parts unaccounted for during UlRl7. and debris found in the I CCS and refueling uater storage tank (RWST) dming U2RIR The evaluations found no unres icued safety questions associated uith these metallic loose Page 75

parts potentially being present in these systems. With the exception of the cotter pin evaluation, the debris evaluated was generally larger in size than this screw. The evaluations and conclusions remain valid for this debris. One area clarined in this safety evaluation is the potential for the screw to End its way through the debris Glter bottom nozzles (DFBNs) of the fuel assemblies. The potential for this to be in the SFP or the ISFSI loading equipment is also addressed.

Areas previously evaluated for the effects of debris include: fuel assemblies; bottom nozzle plate; rod cluster control assembly; reactor internals and vessel; reactor coolant pumps; steam generators; residual heat removal; valves, pumps, and heat exchangers; letdown isolation valves and orifices; control rod drive mechanisms; pressurizer; safety valves; RTD bypass lines; safety injection pumps and valves; and containment spray pumps and valves.

Based upon previous evaluations for transport through the affected systems and the effects of the debris, and the uncertainty of the actual location of the screw, it is not considered probable that this debris will cause any damage to a fuel rod.

Past safety evaluations discuss the effects of potential debris on equipment or systems required to mitigate the consequences of an accident previously evaluated in the FSAR. The conclusions reached are that the probability of equipment malfunction or system degradation due to debris is not increased, and therefore the consequences of a previously evaluated accident are not increased.

Ber wse of the size of the screw and the uncertainty ofits location, there is a remote chance that it may flow tbr ,h a DFBN and up into a fuel assembly where, due to the clearances, it likely would get lodged between a n

o strap and a fuel assembly. Ifit did get lodged between a fuel pin and a grid strap, it could grate against, and possibly create a pinhole in the cladding (unlikely also due to the small size of the screw). Should this occur, the activity of the RCS would increase. Procedures and TS limits are in place that address actions to be

, taken if the activity in the RCS should increase. This is only one screw and one fuel rod would be affected.

Our TS limits would not be exceeded should this occur.

The consequences analysis for many of the accidents or malfunctions of equipment evaluated in the FSAR include an assumption of up to 1% failed fuel as an initial condition. A substantial number of fuel rods would have to be defective for this number to be approached. TS limits are set well below this number and if the activity in the RCS were to exceed the TS limit and could not be corrected, the unit would be placed in the cold shutdown condition. Thus, if debris were able to cause damage to a fuel rod, this would be detected and appropriate actions taken to ensure the consequences of an accident or malfunction of equipment previously evaluated would not be increased.

No new initiators or failure modes of equipment are cr,eated and the possibility of an accident or malfunction of equipment of a different ty pe is not created. The possibility of this screw being in any of these locations does not insolve an unreviewed safety question. (SER 06-024) 3.

DCS 3.1.7. Service Water System. Administrative restrictions ensure the availability of 3 service water pumps under limiting design basis scenarios. This is an interim measure which exceeds the restrictions already imposed by TS 15.3.3.D. That TS similar to other criginal TS, were predicated on a simplined single failure analysis that provides different results than those performed to meet current general design criteria. Until the Basis of TS 15.3.3.D is reconciled with current single failure criteria, the administrative restriction is a conservative measure to ensure that adequate SW pressure and How is available for the design basis accident.

Pace %

Summarv of Safety Evaluation: TS 15.3.3.D states that 4 SW pumps must be operable, and that one of the four can be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . CR 96-026 describes a condition that the SW temperature at the outlet of the containment fan cooler (CFC) could reach saturation. This condition was discovered during analysis of the SW system for resolution of CR 95-331, which identified a problem with achieving adequate SW flow through the CFCs. The operability of the CFCs and SW system is based on the following: (1) Administrative restrictions in DCS 3.1.7 that provide assurance that at least 3 SW pumps would operate for design basis accident mitigation; (2) results of testing show that CFC fouling is present (which mitigates boiling); and (3) judgment that if saturation is reached, the SW system continues to perform its function, as required.

The probabilities of accidents previously evaluated are based on the probability ofinitiating events for these accidents. Initiating events for accidents previously evaluated: Control rod withdrawal and drop; CVCS malfunction (boron dilution); startup of an inactive reactor coolant loop; reduction in feedwater enthalpy; excessive load increase; losses of reactor coolant flow; loss of external electrical load; loss of normal feedwater; loss of all ac power to the auxiliaries; turbine overspeed; fuel handling accidents; accidental releases of waste liquid or gas; steam generator tube rupture; steam pipe rupture; control rod ejection; and primary coolant system ruptures. The SW system and the CFCs do not cause or affect the probability of initiating events.

Therefore, maintaining 6 SW pumps operable or limiting unit operation based on less than 6 SW pumps being operable does not change the probability of occurrence of accidents previously evaluated in the FSAR.

Subsequent analyses have shown that 3 SW pumps may be needed for accident mitigation. This effectively has reverted the situation to inconsistency between TS and the design of the system. To compensate for this situation, an administrative restriction provides additional assurance that at least 3 SW pumps are operational for accident mitigation.

The consequences of malfunctions of equipment important to safety are based on the accident that has occurred and the system that has failed. In this case, an administrative restriction is used to provide additional assurance that sufficient equipment is operational for accident mitigation. Therefore, the consequences of SW system equipment malfunctions are not changed. Additionally, the failure of one train of equipment has been considered in determining the consequences of accidents previously evaluated in the FSAR. This is not being changed and hence the consequences of malfunctions of equipment important to safety have not been increased.

Although the TS do not specifically delineate a margin of safety associated with SW system operability, some statements are made in the Basis of TS 15.3.3. Based on the original TS Basis for SW pump operability and if the margin of safety for SW pump operability is considered to be the difference between the number of pumps that exist and the number that are required to be operable the margin of safety is unchanged. Additionally, the margin of safety under the administrative restriction could be considered to have increased. Using the definition of SW pump operability margin of safety being the difference between the minimum number of pumps required to be operable without tetriction and the number required for accident mitigation, the margin of safety under the administrative restriction actually increases by one (6-3=3. compared to 4-2=2). As stated previously, the SW system is operable. This ensures that other margins of safety, which are based on systems,

, structures and components that rely on SW to be operable, are preserved. Therefore, no margin of safety as defined in the TS Basis has been reduced.

in recognition that the SW system was originally analyzed for the use of 3 pumps, the condition described in this evaluation, in which subsequent analyses have shown that 3 SW pumps are needed for accident mitigation again, basically re-establishes this situation. Therefore, it was concluded that this condition does not pose an unreviewed safety questior.. (SER 96-056) 4 FSAR Section 9.6.2, Service Water System. The FSAR is being revised to acknowledge that of certain automatic isolation valves located in the service water (SW) supply lines between essential accident loads and the Lake Michigan. The FSAR suggested that such auto natic valves may be unacceptable due to the potential to block an essential flow path. However, automatic SW valves for emergency diesel heat exchangers were pros ided in the original design, and automatic SW sah es for ausiliary feedw ater pump ( AFP) coolers u ere prouded in 1972.

Pye 77 l

Symmary of Safety Evaluation: Since the time ofits issuance, FSAR Section 9.6.2 has stated: "To preclude the possibility of an inadvertent blockage of cooling water flow, there is no automatically operated valves in the service water system between the lake and the components which require cooling following a design basis accident." Emergency containment cooler isolation valves are the only exceptions noted in the FSAR.

As opposed to a passive service water supply, the existence of these isolation valves may pose some additional potential for a flow blockage that could impede the capability of accident-mitigating equipment. However, such a blockage would constitute a " single active failure" which is acceptable in the plant licensing basis. The existence of these valves does not adversely affect the capability of the ESF capability to mitigate accidents.

Furthermore, the existence of these valves improves the reliability of the associated SF equipment by preventing condensation on idle equipment.

The valve fail-open design ensures that a single active failure that prevents the valve from opening on demand would only affect one train of safety-related equipment. These components are safety-related active components that are subject to the same single failure criterion as other components in a safety-related train of equipment. In this respect, the failure of an isolation valve to open may cause the host component to trip, which is no more limiting than a failure of the host component to start. In either case, the single nctive failure would have the same worst-case consequences as a ! ass ofone equipment train. Train independence ensures the capability of the redundant train to perform the safety functions which ensure that the accident consequences are no worse than those described in the FSAR.

There are only two si;nificant effects of an inadvertent blockage of service water flow to these essential l

components. Most significant y, the loss of cooling could cause the associated component to fail. Secondly, the trapped service water could heat up and expand if the isolation valve inadverteritly shut during operation of the associated component. The effect of heating the trapped service water is inconsequential since the fluid could expand into the open service water lines and not increase the line pressure nor challenge the integrity of the service water system. Therefore, there are no significant effects of this activity wiJch would create the possibility of a new type of accident.

Since the change has no adverse effects on the plant's capability to mitigate design basis accidents and it fully complies with the General Design Criteria, it is not an unreviewed safety question. Since this component description is not part of the TS and does not affect the margin of safety, there is no need for prior NRC approval to make the FSAR changes. The FSAR changes are necessary to more accurately represent the actual plant configuration and the licensing basis for the service water cooling supply to essential components.

L R 96-003)

5. FSAR Section 10.2.2, Steam Dump System.

ESAR Table 10.3-1. Steam and Power Consersion System Single Failure Analysis.

This FSAR revision more clearly describes the condenser steam dump system design basis and its requirements, -

especially in consideration recent startup flow capacity testing results. Two of the eight condenser steam dump valves (1MS-2050 and 1MS-2055) were tested and delivered 5.8% and 6.2% of steam generator capacity, respectively.

Summary of Safety Evaluation: The upper bound of the system is limited by the accident as described in FSAR 14.2.5,'" Rupture of a Steam Pipe." The failure of a condenser valve (s) to shut falls within the definition of this accident. This accident remains the worst cese scenario for uncontrolled steam flow. Its analysis results and conclusion remains unchanged. The rupture of a steam line is far more limiting than the failure (valve fail .

open) of one condenser steam dump valve, whether its capacity be 5% or 6.2% of total steam flow. This is supported by the accident analysis assumptions which include a steam pipe break equivalent to the steam release through one steam generator safety vah e with outside pow er available, w hich is 25% of tiow from one steam generator Furthermore, the consequences are mitigated by the fast shutting isolation valve and check vahe that each sicam line possesses.

Page 78

The lower bound of the system is limited by the accident as described in FSAR 14.2.4," Steam Generator Tube Rupture." Under the worst case scenario where a steam generator tube rupture occurs coincident with loss of available offsite power, the steam dump to the condenser is isolated when condenser vacuum is lost.

The connguration and slight increase in How capacity of the replacement condenser steam dump valves and the revision to FSAR Section 10.2.2," Design Features - Steam Dump System," and FSAR Table 10.3-1," Steam

~

and Power Conversion System Single Failure Analysis," do not pose an unrev;ewed safety question.

(SER 96-071)

6. FSAR Section 11.2.3. Radiation Monitoring System. FSAR inconsistencies exist in Chapter Ii Tables. FSAR Table 11.2-78 states that a RDA-SS scintillation counter is installed in the component cooling water liquid monitor. The "S" at the end of the detector model number signines that the detector has a stainless steel housing. An RDA-5 A detector with an aluminum housing is installed in the Unit 2 component cooling water liquid monitor.

This change also includes removing background compensation for SPING beta particulate, low-range noble gas, and medium-range noble gas channels. The counting ranges and efHciencies for SPING channels are also corrected.

Summary of Safety Evaluation: The RMS has no possible accident initiatian mechanism. The RMS provides information that assists the operator in diagnosing accident conditions. The FSAR changes do not degrade the ability of the RMS in diagnosing accidents. The functions of the RMS are not affected by these FSAR changes.

The changes do not impact the ability of the RMS to provide radiological information to the operator or the ability of RMS components to perform its control function. No new types of failure modes are created as a result of the changes.

The radiation monitors described in FSAR Section i 1.2.3 are not necessary for performance of a safety-related function. The reliability of the RMS was not adversely impacted by making the PPCS the primary operator interface. This is an enhancement to the system because of the close proximity of the PPCS to the control boards. The other FSAR changes concern RMS technical information and calibration data that does not impact the ability of the RMS to perform its functions.

The radiation monitors described m FSAR Section i1.2.3 are not necessary for perfonnance of a safety-related function. The reliability of the RMS was not adversely impacted by making the PPCS the primary operator interface. This is an enhancement to the system because of the close proximity of the PPCS to the control boards. The other FSAR changes concern RMS technical information and calibration data that does not impact the ability of the RMS to perform its functions.

RMS surveillance requirements are specified by TS Table 15.4.1-1, item 36. None of the 1 S operability or

, surveillance requirements are altered by the FSAR changes identined.

The FSAR revisions correct errors and improve the readability of RMS information. The revision to FSAR Section 11.2.3 does not involve an unreviewed safety question. (SER 97-074)

7. FSAR Section 14.3.4: TS 15.3.3 and 15.5.2. Containment Integrity Evaluation with Reduced Fan Cooler Capacity. Reduced post-LOCA containment heat removal on the containment integrity analysis as described in FSAR Section 14.3.4 was evaluated. From a long-term LOCA/ containment integrity perspective, an evaluation model was developed and baselined against Case 1 (the "All Available Energy Case" from FSAR Figure 14.3.4.6). The baseline results of the model matched FSAR Figure 14.3.4-6 to within =l%. The model adopted the FSAR assumption of an EDG failure thtt results in only two containment accident fan coolers (CFCs) and one containment spray pump available for containment cooling starting at 60 seconds and running continuously for the accident duration. The model was then rerun assuming reduced containment heat remmal performance. equisalent to 1.5 Cl C and one containment spray pump running for the I;rst hour; followed by Page 79

containment spray flow termination at switchover to the sump recirculation phase and containment heat removal thereafter by only 1.5 equivalent CFCs (e.g., no containment spray after one hour).

Summarv of Safety Evaluation: Reduced CFC heat removal capacity and containment spray termination after one hour influence the containment pressure and temperature profiles resulting from the accident, but do not influence on the probability of accident initiation. The reliability of the equipment assumed to function in the safety analysis is not affected. Therefore, the p;obability of the occurrence of the accident is not affected by the change.

l The containment peak pressure for a LOCA as in the FSAR analysis (FSAR Figure 14.3.4-6) is not affected by the change. In addition, the analyses for offsite dose (FSAR Section 14.3.5) and control room dose are also not affected by the change. Both dose analyses assume a containment leak rate (at the containment design pressure -

of 60 psig) of 0.4 wt Wday for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of a LOCA, followed by a leak rate of 0.2 wt Wday for the next 29 days. The corresponding assumptions for containment pressure to cause these leak rates (60 psig for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then 15 psig for 29 days) are higher than the pressure profile resuiting from this change, such that the .

conservative leakage assumptions in the dose analyses are unaffected by this change. Therefore, there is no increased dose consequences as a result of the change, and the change does not increase the consequences of an accident previously evaluated in the FSAR.

The change does not affect the ability of environmentally qualified equipment to perform its function following a LOCA. The capability to take a post-LOCA containment air sample was reviewed for impact from the revised containment pressure profile resulting from this change. Based on the realistic net containment cooling etTect occurring one hour after en inc-ident, it can be concluded that the capability to take a post-LOCA containment air sample is not affected by this change. ne capability of the containment purge valve T-ring seals to maintain its sealing function under the revised containment pressure profile was reviewed and determined to be unaffected over the analyzed duration of the transient. Therefore, the reliability of the equipment assumed to be operable in the FSAR analysis is not affected, the ability of equipment to perform the necessary mitigating functions is not impaired, and there is no increase in the consequences of a malfunction of equipment important to safety as a result of this change.

The Basis of TS 15.3.3 indicates that two CFCs and one CS pump provides sufficient cooling to reduce containment pressure following a LOCA. The change still provides sufficient cooling with no reduction in the margin of safety. Although the change affects the rate of containment depressurization, the change does not increase the peak containment pressure and does not affect the overall capability to gradually reduce containment pressure after a LOCA. TS 15.5.2 (Containment), states that the four CFCs have a total heat removal capability of 55.600 Bru/sec following a LOCA. This is only a capability statement, and does not form the basis for the margin of safety. Therefore, there is no reduction in the margin of safety as defined in the TS.

The change does not involve an unreviewed safety question. (SER 96-055)

8. G-02 EDG Operabihty SERs 93-025-25 and 95-073-01 state that any time emergency power is not available to I A-05, G-02 EDG shall be declared inoperable. This is based on the requirements for independence and -

redundancy contained in GDC-39 and recognition of the situation that two of the three service water (SW) pumps in Train A are powered from l A-05 (the G-02 EDG is cooled by SW). A circumstance has arisen in which emergency power to I A-05 was removed from service to perform necessary maintenance on G-01 EDG. -

At the time of entry into this LCO, the G-02/l A-05 dependency was not recognized. Later that day the dependency was recognized and G-02 (emergency power to 2A-05) was declared inoperable changing both units from being in a 7-day LCO per TS 15.3.7.B.l.f to both units being in a 7-day LCO per TS 15.3.7.B.I.h.

During the subsequent verification of required redundancy for Unit 2, it was discovered that the Unit 2, Train B, component cooling water pump. 2p-l 1B. was inoperable. This was in accordance with page 80

TS requirements for required redundant engineered safety features being operable (TS 15.3.7.B.1.F and TS 15.3.0.D), because the component cooling water system redundancy with one pump out of service is still available to both units based on three pumps being operable and the availability of manual cross-connect of the component cooling water system, if necessary for accident mitigation. This is in accordance with TS 15.3.3.C.1 that allows one component cooling water pump to be inoperable indefinitely without limitation for a two-unit operation.

Summarv of Safety Evaluation: 7his evaluation specifically addressed equipment used to mitigate accidents, the TS requirements for maintaining TS operability and the TS limiting conditions for operation of the systems and equipment. The emergency power system and component cooling water system operability requirements do not affect factors that determine the probability of accidents. Therefore, the situation described does not change the probability of occurrence of an accident previously evaluated in the FSAR.

The consequences of the accidents previously evaluated in the FSAR are determined by the results of analyses that are based on initial conditions of the plant, the type of accident, transient response of the plant, and the operation and failure of equipment and systems. TS LCOs are established to allow temporary relaxation of single failure criterion, consistent with overall reliability considerations, to allow time periods during which corrective action may be taken to restore the system to full operability.

Recognition

hat G-02 is operable for emergency power to 2A-05/B-03 actually decreases the probability of a malfunction of Tram A equipment, compared to the situation of considering G-02 inoperable. Relaxation of Train B emergency power single failure allows G-02 to be considered operable based on operability of sufDcient service water pumps to cool G-02. Therefore, normal and emergency power are operable for I P-l l B and 2P-11 A, and normal power is operable for 1P-11 A. Under the condition with emergency power not operable to I A-05 and B-03,1&2P-1l A were considered operable from normal power and the redundant pump, IP-1 IB was considered operable from normal and emergency power. Both situations meet the requirements of the TS.

The consequences of malfunction of equipment are determined by the results of accident analyses. By entry into an LCO, in this case TS 15.3.7.B.I.f, relaxation of single failure is invoked. Relaxation of single failure requirement establishes that the most limiting malfunctions of equipment important to safety as described in the FSAR are still met. The recognition that G-02 is operable for emergency power to 2A-05/B-03 does not change this situation. Therefore, the consequences of malfunction of equipment important to safety previously evaluated in the FSAR are not changed.

New or different kinds of accidents can only be created by new or different accident initiators or sequences.

New and different types of accidents (different from those originally analyzed) were evaluated and incorporated into the licensing basis. Examples of different accidents incorporated into the' licensing basis include anticipated transients without scram and station blackout. The recognition that G-02 is operable for emergency power to 2A-05/B-03 does not create new or different accident initiators or sequences. The relaxation of single

- failure criterion and the limitation on the time that emergency power can be inoperable, provides the justi6 cation for not creating any new or different accident sequences (that would be caused by the malfunction of equipment beyond what is currently assumed for the operability of emergency power).

Adherence to TS operability requirements and LCOs provides the assurance that adequate equipment important to safety is available to mitigate accidents as described in the FSAR. The use of Train B service water pumps to maintain operability of the Train A G-02 EDG was not previously assumed, but based on the relaxation of single failure criterion as allowed by TS 15.3.7.B.l.f, the operability of G-02 as emergency power to 2A-05/B-03 can be demonstrated. Therefore, this situation does not create the possibility of malfunction of equipment important to safety of different type than previously evaluated in the FSAR.

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1 I

l ne margins of safety are based on the design and operation of the reactor and containment and the safety systems that provide their protection. Adherence to the TS operability requirements and 1 COs provides assurance that appropriate equipment are operable to mitigate accidents and safely operate PBNP. Therefore, the condition does not reduce the margin of safety, nor does it involve an unreviewed safety question.

(SER 96-006)

9. Radioactive Materials Storage Area in the South. Bay of the Steam Generator Storage Facility. Radioactive materials stored in Warehou e 2 for long-term storage were transferred to the Steam Generator Storage Facility (SGSF) south bay for long-term storage. He new location for radioactive material storage is a secured area located outside the radiation controlled area (RCA) as defined in FSAR Section i 1.2. The radioactive materiel to be stored at this location consists largely of tools and spare parts which receive periodic use and contains very low level fixed and loose contamination. The materials are stored in an area that is physically separate -

from the area used to store low level radioactive waste. No liquid or gaseous materials are stored here.

Summarv of Safety Evaluation: SERs 92-011 and 93-005 addressed the long-term storage of the materials .

within Warehouse 2. This SER addressed the storage of those same materials in an area established within the south bay of the SGSF. The hazard evaluations performed in SERs 92-011 and 93-005 for the storage of these materials is directly applicable to the new storage area and the constraints established in those SERs on the storage of radioactive materials are also directly applicable to the new storage area. The long-term storage of radioactive materials in the SGSF south bay does not pose an unreviewed safety question as long as the constraints established in SERs 92-011 and 93-005, and this safety evaluation are adhered to. These constraints were modified to reflect the new storage location:

Materials with loose contamination shall be stored in strong, tight packaging. Strong, tight packaging is defined as packaging that will not leak radioactive material during normal storage conditions.

Materials with loose contan'.ination shall be stored in packaging that does not support combustion.

Packages containing loose contamination shall be opened in the RCA.

e llealth Physics shall impose controls to limit access to the storage area to those persons authorized for entry by plant supervisors and Health Physics personnel.

The area where the radioactive materials are to be stored shall be surveyed, classified, and conspicuosly posted with a sign (s) bearing the radiation caution symbol and the words: CAUTION (or DANGER)-

Radioactive Materials.

Radioactive materials shall be stored off the ground. This may be accomplished through the use of pallets, shelving, or integral features of the packaging.

No radioactive waste is to be stored in the radioactive materials storage area. -

The storage area shall be secured so as to restrict unauthorized removal of the (licensed) materials.

The total activity of the materials stored in the radioactive materials storage are and the waste stored in the low-level radioactive waste storage area shall not exceed 3260 curies. The site radwaste group shall administratively control the storage of packages in the two areas of the SGSF south bay to meet this requirement.

No more than 100 packages shall be stored in the storage area. (SER 96-03'F)

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10. Release of Dedicated Operator for P-38A Motor-Driven Auxiliary Feedwater Pumo Discharge Valve AF-4012 To Control Discharce Flow. This evaluates releasing the dedicated operator described in SER 96-023 who controls P-38A motor-driven AF pump discharge valve AF-4012 to control discharge total flow to 200 gpm or less, when this load has been automatically sequenced on the emergency power supply. Since the designation of a dedicated operator to declare P-38A operable, the following actions have been taken:

The no-load frequency and droop has been changed on emergency diesel generators G-01 and G-02.

. Standard EOP cautions for warning that motor-driven AFP breaker may trip at Dows of >320 gpm has been changed to Dows of >200 gpm. The background document description for this caution reflects this new infonnation.

~

Operators received training on the EOP changes and the expectations for actions to control P-38A AFP discharge flow similar to the instructions provided to the dedicated operator. Training was completed for each operator prior to taking a watch without the dedicated operator being assigned.

Summary of Safety Evaluations: Operator actions in the emergency operating procedure set are required to mitigate the consequences of an accident as stated in FSAR Section 12.4.1. The auxiliary feedwater system starts automatically and delivers up to 800 gpm by design, but only 200 gpm is required by accident analysis.

Operator action is required to limit the flow and prevent excessive cooldown and steam generator filling for the following events: Steam generator tube rupture; reactor trip; small break LOCA; and a faulted steam generator.

The auxiliary feedwater system requires skill of the craft for a variety of postulated events to limit consequences. Redundant AFPs continue to more than required flow in some situations but are required for other postulated accidents. Other specific operator actions required for the operation of this system is switchover to service water supply after condensate storage tank depletion. Operator actions do not negate the fact that the plant design is based on assuming credible initiating accidents and that the protective and engineering safeguards systems are provided to limit the consequences of unlikely events. The EOP changes do not affect operator action but rath.r warns of potential consequences of actions to deviate from design flow and procedure guidance because of plant conditions beyond design basis. A dedicated operator is not required to perform these actions since the actions are already specified in the emergency operating procedure set.

The dedicated operator for P-3SA auxiliary feedwater pump was put in place to ensure operability of P-38A following assumption of Unit 1 Train A safeguards bus loads by an emergency diesel generator operating above 60 Hz frequency. P-38A was found to trip after approximately 6 minutes of operation because of high current condition caused by the EDG speed / frequency. The EDG governor was adjusted to reduce the no-load frequency and speed droop. That governor adjustment significantly reduced the maximum frequency supplied to the bus and when combined with the expected EDG loading reduces the likelihood of P-38A AFP motor breaker trip. Evalcmn of the emergency operating procedure set, the time required to reach the cautions for the auxiliary feedwater flow, in actual experience and in evaluated simulator scenarios, provide assurance that the operators are able to adjust AFW to 200 gpm prior to the expected time / current condition (minimum 250

- seconds) which could cause a P-38A AFP motor breaker trip. Additionally, operator action specified in emergency operating procedures ensure the operator takes action to prevent the motor breaker from tripping should the pump initially be operating greater than design Dow of 200 gpm. (SER 96-028) i1. Unit I Cycle 24 Core Reload, The safety evaluation covers mechanical design, nuclear design, thermal-hydraulic design, power capability, FSAR accidents and 'lJ changes that apply to the Unit 1 Cycle 24 (UlC24) reactor core.

The UIC24 core contains 121 upgraded optimized fuel assemblies (OFA). Sixteen assemblies with integral fuel burnable absorber (IFBA) and 12 peripheral power suppression assemblies (PPSA) are also included in the core. Feed assemblies have axial blankets.

Pape 83

Summarv of Safety Evaluation: The applicable limiting values for UIC24 have been compared to and found to be consistent with those used in previous safety evaluations. It is concluded that the UIC24 design does not cause safety limits to be exceeded, provided that the following conditions are met:

E e UIC23 burnup is bounded by 11,300 and 12,300 MWD /MTU. Actual U!C23 burnup was 11,939.2 MWD /MTU.

. UIC24 burnup is limited to the end-of-full-power-capability (EOFPC), which is defined as the burnup of fuel when all control rods are fully withdrawn, and s 10 ppm of boric acid at the UIC24 rated power condition of 1518.5 Mwt, plus 1500 MWD /MTU power coastdown operation.

e There is adherence to the plant operating limitations given in the TS. .

Administrative controls are in place to ensure the boron concentration in the Unit I refueling water storage tank is at least 2l00 ppm.

Operation of the UIC24 core does not involve an increase in the probability or consequences of accidents previously considered; does not involve a decrease in safety margin, and does not involve a significant hazard consideration. Therefore, provided that stanup physics testing dets not result in any discrepancies with the analysis assumptions, operation of UIC24 is in accordance with TS and does not involve an unreviewed safety question.

The non-LOCA transient evaluations and analyses show that applicable safety criteria, as presented in the FSAR are satisfied. Differential rod worths, shutdown margin, ejected rod worths and trip reactivity remain within analysis limits. Core peaking factors for the dropped RCCA event, the rod ejection incident, and the steam line break incident were reviewed and found to be acceptable and within the bounds of the analysis limits.

The design basis that the core remain subcritical on soluble boron alone in long-term cooling following a large break LOCA is satisfied for UIC24 operation based on a minimum refueling water storage tank (RWST) boron concentration of 2100 ppm. The boron concentration in the containment sump fo!!owing a LOCA is evaluated for every reload core. The evaluation for UIC23 showed that the core could return to a critical condition following a LOCA with a boron concentration of 2000 ppm in the RWST. A boron concentration of at least 2100 ppm prevents a post-LOCA return to criticality for UIC23 with additional margin for future cycles. An administrative control was implemented for UIC23 by changing the minimum RWST boron concentration from 2000 to 2 l00 ppm. The administrative control remains in effect for UIC24.

Enrichment limits for the new fuel vault and the spent fuel pool are satisfied. TS 15.5.4 limits new and spent fuel in fuel storage areas to 46.8 grams of U-235 per axial centimeter of OFA fuel which corresponds to 4.73 wt%. the highest enrichment fuel being loaded in UIC24 is 3.8 wt%. Use of the UIC24 core does not involve an unreviewed safety question. (SER 96-018) -

12. Unit 2 Cycle 23 Reload. This evaluates the Unit 2 Cycle 23 (U2C23) core loading pattern. The core contains 121 upgraded optimized fuel assemblies (OFA). Four feed assemblies contain integral fuel burnable absorber (IFBA) and 12 peripheral power suppression assemblies (PPSA). All 20 feed assemblies have solid pellet axial blankets. This evaluation is applicable from the time the final core leading pattern is established during U2R22 until the core geometry is changed in the next refueling outage.

Summary of Sa[cly Evaluation: There are two significant changes being made to Unit 2 for Cycle 23. Steam generators are replaced with Model A47 steam generators. Although the generators are functionally equivalent to the original Model 44 steam generators, many analyses and evaluations have been performed to support the replacement steam generators (RSGs). Key parameters that changed inc!ude the primary and secondary system volumes. metal mass of the Model A47, Integral flow limiting devices and narrow range lev'el span.

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The Cycle 23 reload core design meets applicable design criteria and assures that pertinent licensing basis acceptance criteria are met. it has been determined that the OFA fuel reload design and safety analysis limits remain applicable and that these limits are supported by the applicable TS for Cycle 23 through Amendment 172. The demonstrated adherence to applicable standards and acceptance criteria precludes new challenges to components and systems that could: a) Increase the probability of previously evaluated accidents; b) adversely affect the ability of the existing components and systems to mitigate the consequences of accidents; and/or c) adversely affect the integrity of the fuel rod cladding as a fission product barrier.

Adherence to applicable standards and criteria ensures that the fission product barriers maintain design margin of safety. The cladding integrity is maintained and the structural integrity of the fuel rods, fuel assembly and core is act affected. Predictions presented in the FSAR are not sensitive to the fuel rod cladding material or other mechanical changes that do not alter the metallurgical composition of the core.

The Region 25 fuel assemblies satisfy the same design bases as that were used for fuel assemblies in the other fuel regions. There are no mechanical changes to the Region 25 fuel assemblies. The Region 25 fuel assemblies do not change performance requirements of any system or component such that any design criteria will be exceeded nor will they cause the core to operate in excess of pertinent design basis operating limits.

The Region 25 fuel assemblies have no impact on fuel rod performance, dimensional stability, chemical, physical, or mechanical properties. No new modes of operation or limiting single failures have been created for systems, components, or pieces of equipment. No new performance requirements are imposed on systems or components such that design criteria is exceeded. He assemblies do not create the possibility of a malfunction ,

of equipment important to safety of a different type than previously evaluated in the FSAR.

There is no reduction in the margiu of safety as defined in the Basis of TS. He use of these fuel assemblies takes into consideration normal core operating conditions within the allowable TS values. For each cycle reload core, these assemblies are evaluated using approved reload design :nethods and fuel rod design models and methods. This includes consideration of the core physics analysis peaking factors and core average linear heat rate effects.

Excessive stress or high fatigue may increase the probability of equipment malfunction. Systems and components analysis to support the RSG and Tavg range is documented in WCAP-14602. Results confirm continued compliance with industry Codes and standards, regulatory requirements, and applicable performance and design basis requirements. Component fatigue usage factors and stresses are below design limits.

Consequences of a malfunction in the electrical systems, such as a station blackout, may increase due to changes in steam generator inventory. Calculations determine the condensate storage tanks (CST) have sufGcient volume to maintain decay heat removal capability using the steam generators for at least one hour during a station blackout. Results show that the CST minimum volume of 13,000 gallons is more than adequate for station blackout concerns.

The probability or consequences of accidents and malfunctions previously evalcated in the FSAR are not

. increased. No new accidents or malfunctions are created. The margin of safety is not reduced by the Unit 2 Cycle 23 core reload. Therefore, the Unit 2 Cycle 23 reload core design does not involve an unreviewed safety analysis. (SER 96-130) 1.1. Unit 2 Original Steam Generator Steam Drum Dismantjement in the ISFS! Transnorter Storage. The Unit 2 OSG steam drums are transported to the ISFSI transporter storage building for dismantlement. Oxyacetylene and plasma are torches or other equivalent equipment are used to cut the steam drums into manageable pieces.

The cutting is done inside a tent enclosure which is kept at a slight negative pressure by exhausting air through a high efHciency particulate activity (HEPA) Ulter. The llEPA exhaust is monitored by an alarming air monitor capable of detecting particulate, halogen, and noble gas activity.

Summary of Safety Es aluating The steam drums contain low-lesel Oxed and loose contamination. The steam drums are sune)ed for radioactise contamination using standard suney practices. Contaminated pieces are transported to an offsite s endor for further processing and disposal.

Pne85

This activity is calculated to produce a maximum dose to the liver of the hypothetical child at the Site Boundary Control Center of 4.6E-06 mrem. This is well within the 10 CFR 50, Appendix I limit of 15 mrem. The activity does not pose an unreviewed safety question. (SER 96-083)

Summary of Safety Evaluation: Radiologically contaminated material is contained inside a tent enclosure which is kept at a slight negative pressure by exhausting air through a high efficiency particulate activity (HEPA) filter. The conclusions of the original SER remain valid. (SER 96-083-01)

14. Work Orders 9602478. 9602491. 9602495. 9602497. and 9602498 The Elgar inverters that supply the white and yellow instrument buses contain a circuit that causes an inverter thutdown if de input voltage drops to a preset value. This circuit is designed to prevent excessive discharge of the & supply battery. He red and blue .

instrument bus inverters, manufactured by Solidstate Controls, Inc., do not have this feature. This circuit was disabled in each inverter to eliminate the maintenance costs associated with calibration. It also eliminates a potential failure mode of the inverters.

Summary of Safety Evaluation: The equipment specification for the Elgar inverters does not require a low voltage shutdown circuit. The existing setpoint for Elgar inverter shutdown is 100 Vdc. The inverters are required to be operable for de voltages between 105 and 140 Vdc. The batteries supplying the inverters are sized to maintain voltage above 105 Vdc during the worst case emergency discharge. The low de shutdown circuit requires periodic calibration to ensure that the setpoint has not drifled higher than 105 Vdc. Calibration of this circuit requires a significant amount of time because the circuit card must be removed from the mverter and bench calibrated. The low de shutdown feature would prevent degraded inverter operation epon low de input voltage, which could result in misoperation of connected loads (misleading indications, etc.). However, this was not the intended design function of this feature, and other failure modes of the inverter which result in low ac output voltage are not prevented. He issue ofinadequate instrument bus voltage and resulting equipment misoperation is being addressed through operator training.

This change dees not affect the ability of the inverters to supply the instrument bus load; with the inverter design limits. The probability of an inverter failure while operating within design limi's is not increased as a result of this change. The batteries supplying the inverters would continue to perform their design function, with no increased probability of failure. Derefore, the probability of any accident which could result from the loss of an instrument bus is not increased.

The ability of the batteries and instrument bus inverters to supply its associated loads, within its design limits, during postulated accident conditions is not affected by this change. The consequences oflow de voltage would be degraded operation or complete shutdown of the associated inverters. This would be considered a single failure, which has been previously analyzed. The effect upon plant systems wotic Se the same as for other possible ins erter failures. Therefore, the consequences of an accident previously evaluated in the FSAR is not increased.

The probability of an inverter or battery failure during operation within the design limits are not increased.

Operation of batteries or inverters outside its design limits is not assumed or analyzed in the FSAR. The loss of an inverter or battery has been previously analyzed. No accidents of a ditferent type than previously evaluated are possible.

The inverters are equipped with circuits, such as low logic voltage protection, which are designed to prevent inverter damage if dc voltage drops below design limits. Operator action is relied upon to operate the batteries within its design limits. Other battery loads do not have features to prevent excessive battery discharge.

Therefore, removal of this feature from the Elgar inverters does not increase the consequence of a malfunction of equipment important to safety.

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Operation of the mverters within the design limits for de input voltage is not affected. DC voltage below the design limits of the inverters could result in degraded operation or complete shutdown of the associated inverters. These ailure modes are already possible with the current configuration. Other battery loads, including red and Llue instrument bus inverters, do not have features to prevent excessive battery discharge.

Therefore, removal of this feature from the Elgar inverters does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the FSAR.

The degree of redundancy and design limits of the inverters and batteries are not affected by this change.

Therefore, this change does not reduce the margin of safety defined in the Basis for any TS. This change does not involve an unreviewed safety question. (SER 96ff2.1)

15. WO 9605848. WO 9605848 repairs Unit 1 A main feedwater pump low suction pressure trip / interlock pressure switch PS-2197.

Summary of Safety Evaluation: The low suction pressure main feed pump trip is described on FSAR Page 10.2-10. The trip and interlock are also shown on FSAR Figure 10.2-2 Sheet 2. The low suction pressure starting interlock, the low suction pressure alarm, and the low suction pressure feed pump trip protect the main feed pump from damage due to operation with inadequate suction pressure.

The activity has no effect en the probability of occurrence of an accident or consequences of an accident previously evaluated in the FSAR. He possible effects are confined to malfunction of or damage to the 1P-28A main feedwater pump. Loss of feedwater accident is analyzed in FSAR Section 14.1.10, which assumes a complete loss of feedwater. His analysis is bounding for the loss of one main feedwater pump.

Should the main feedwater pump be damaged due to operation with low suction pressure, it is not expected that such damage would result in damage to adjacent equipment. The main feedwater pump is located in a non-seismic, non-safety-related area, in the turbine hall on El 8'. No safety-related components are in the immediate area. The main feedwater regulator valves are located on turbine hall El 26' and are protected from potential damage by the El 26' Coor.

The activity could only affect the main feedwater pumps, loss of which has already been analyzed in the complete loss of feed section of the accident analysis (FSAR Section 14.1.10). Should the activity be performed incorrectly, a likely result is a Unit I trip, with unavailability of IP-28A. A reactor trip is a safe condition for the plant, which has already been analyzed. Preliminary steps of the work plan ensures that the trip and alarm are properly defeated. These steps are based on the AMSAC test procedure which is performed every refueling outage. The activity does not affect equipment important to safety, and does not result in an accident of a different type than analyzed in the FSAR. The activity does not involve an unreviewed safety question nor does it affect the margin of safety (SER 96-048)

16. WO 9606192 and WO 9606103. Installation of Containment Fan Cooler. WOs 9606192/9606193 installs combination vacuum-pressure gauges downstream of the containment fan cooler (CFC) outlet motor-operated valves. The gauges were installed downstream of 3/4" drain valves iSW-167 and 2SW-276. There is a cap downstream of ISW-167 and an open ended welded fitting downstream of 2SW-276. These gauges provide differential pressure measurements in the service water piping for use in setting containment fan cooler How rates.

Summary of Safety Evaluation: The addition of pressure gauges to the service water return piping does not increase the probability of a previously analyzed accident since the service water system is primarily considered an accident mitigating system it does not impact the ability of the service water system to supply cooling to equipment since the gauges are connected to the eturn piping. Leakage from the gauges does not r

reduce Cow from other components. The gauges do not degrade the pressure boundary integrity of the service water piping since the gauges withstand pressuies and vacuums well in excess of the service water return line pressures and sacuums that could be experienced under a normal or emergency conditions The seismic integrity of the service water piping is not affected by the addition of the relatisely light ueight gauges to the 3/4" piping When the gauges are not in use, its isolation vahes me normally shut.

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Flooding or spray down of equipment in the primary auxiliary building (PAB) could possibly initiate an accident or cause a malfunction of equipment. However, flooding or spray down because of the failure of these gauges is not considered credible because ofits location, small size and since leakage could be isolated by the upstream isolation valves.

The margin of safety as defined in the TS is not affected by the addition of these gauges since the operation and structural integrity of the service water system is not impacted by the addition of the gauges. This chang: does not involve an unreviewed safety question. (SER 96-054)

17. WO 9606835. Plant Process Computer System. WO 9606835 attaches an input point to the plant process computer system (PPCS) to ensure that an electrical actuation of the Model 20-ET circuit is detected as part of .

the sequence of event recording during a trip transient. This additional PPCS point provides positive indication of the Model 20-ET circuit energizing.

Summarv of Safety Evaluation: The 20 ET and 20 AST circuits are not safety-related equipment. While the two circuits duplicate the function of each other, it is not routinely tested (with the opposite circuit disabled) during unit operation. However, since the trip circuits do act as backups to each other, deenergizing one circuit does not disable the protective features. The EH reservoir level lackout is meant to protect the EH pumps from a loss of EH fluid by tripping the EH pumps in the event of a Low-Low EH reservoir level. The alarm portion of this function is still available with the 20 ET circuit deenergized.

The redundancy of the 20-ET and 20-AST circuitry ensures the turbine trip function occurs if required during the installation of the monitoring point. When the installation is complete, the administrative controls (double veritication) ensures that the 20 ET circuit is returned to a functional status. Overspeed trip protection is not effected since it is provided by the mechanical overspeed trip device as well as by the independent overspeed system and ensuring that no other tests are occurring are specified as prerequisites to the installation.

WO 9606835 does not involve an unreviewed safety question. (SER 96-069)

18. WO 9607130. 9607131. 9607132. 9607133. 9607134. 9607135. 9607136. 9607137. The work orders remove the mechanical interlocks between the main breakers ofinstrument bus panels 2Y-01,2Y-02,2Y-03,2Y-04, 2Y-101,2Y-102, 2Y-103,2Y-104.

Summary of Safety Evaluation: Removal of the mechanical interlocks allows paralleling ofinverters during instrument bus shifting. This eliminates the momentary power interruptions and resulting alarms.

This change affects the method of shifting bus power supplies. Failures during shifting could result in the loss of pow er to instrument bus loads. The failure of an instrument bus power supply is not an initiator of accidents previously evaluated in the FSAR. The reliability and availability of the instrument buses during and following an accident is not decreased as the result of this change. Therefore, the reliability of accident mitigating components supplied from the instrument buses is not decreased. Therefore, the probability of the occurrence -

or consequences of these accidents is not increased.

The reliability of the instrument buses is not decreased as a result of this change. Separation between instrument bus trains is not affected. In fact, the reduction in nuisance alarms and spurious actions improves operator cognizance of actual changes in plant status. The qualification of the instrument bus panels is maintained. Therefore, this change does not reduce the margin of safety defined in the Basis for TS. The change does not involve an unreviewed safety question. (SER 96-118)

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19. WOs 9608380 and 9608381. D-03 and D-04 Breakers. The work orders implement changes to the magnetic trip settings for the following de breakers: D72-03-02, feed to D-03 from swing battery D-305; D72-03-03, feed to D-03 from station battery D-105; D72-03-07, power to D-31 de distribution panel; D72-04-02, feed to D-04 from swing battery D-305; D72-04-03, feed to D-04 from station battery D-106; D72-04-07, power to D-41 de distribution panel.

Summarv of Safety Evaluation: The de power from D-03 and D-04 buses is used to respond to accidents previously evaluated in the FSAR. The evaluations show the de power from these buses do not act as an initiator of these accidents. Therefore, WOs 9608380 and 9608381 do not increase the probability of occurrence of an accident previously evaluated.

The actual work only involves turning a dial on the front of each of the breakers. The breaker manufacturer confirmed that turning the dial to increase the breaker magnetic trip setting increases the trip setpoint and indicates no additional testing is required to confirm this. Positive actuation / manipulation is inherently built into the mechanical configuration of the breaker magnetic trip components where a dial is turned to vary the air gap distance in an electro-magnet which is used to give the breaker magnetic trip. A coordination study shows that the increase of the magnetic trip setting does not achieve full coordination, it only improves it. Bus, if the magnetic trip setting were to be increased to a level greater than that expected from the manufacturers trip curves, the coordination would be improved beyond what is expected. Therefore, while additional breaker trip testing is not performed to verify the new settings, the probability of malfunction of these breakers due to the new trip settings is not increased.

Besides the safety-related equipment that would be disabled if the D-31 or D-41 breaker was opened (for which LCOs are entered upon breaker trip), the protective circuitry for the buses which receive its control power from the panel (including 13.8 kV fast bus transfer scheme) would be disabled. There are no LCOs associated with the loss of this bus protective circuitry. The actions of this work plan do not increase the probability of occurrence of malfunction of equipment that would require this protective circuitry. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR is not increased.

De consequences of a malfunction are not increased since the DC system will not be disabled due to the work.

The battery feed breakers will have their magnetic trip settings changed while the breakers are open and therefore the battery will be OOS. The D-31 and D-41 feed breakers will have their magnetic trip settings changed with the breakers closed. While these breakers are not expected to open due to the work, if they do open, all equipment dependent upon the DC power fed from these busses will be declared OOS and all associated LCOs will be entered. Additionstly, all Appendix R compensatory measures will be established for the required equipment being OOS. By entering the applicable LCOs, temporary relaxation of single failure criterion is provided consistent with overall reliability considerations. For example, under a hypothetical partial loss of Offsite Power, which could utilize the 13.8 kV fast bus transfer system in response if available, sufficient on-site power sources uould still be available to mitigate any malfunction of equipment important to

, safety. Therefore, the consequences of a malfunction of equipment important to safety is not increased.

In the case of the D-31 and D-41 feeder breakers, the work is performed with the breakers closed. The only possible consequences of this would be to open the breaker. Since only one system is worked on a time and each system has a redundant system, the possibility of an accident of a different type than previously evaluated in the FSAR is not created.

Since the activities do not affect equipment of the redundant systems for the systems worked on, the margin of safety as defined in TS is not reduced. The change does not involve an unreviewed safety question.

(SER 96-09).)

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I V. NUMBER OF PERSONNEL AND PERSON-REM BY WORK GROUP AND JOB FUNCTION- 1996 Number of Work Function and Total Dose. rem Personnel Total job Group Greater rem for Station Employees Than Job Group 100 mrem Reactor Operations & Routine Special Waste Surveillance Maintenance inspections Maintenance Processing Refueling Operations 43 12.860 9.730 0.900 0.250 1.980 Maintenance 44 25.100 16.400 0.030 1.100 7.570 Chemistrv & Ilealth Physics 24 7.870 l

7.310 0.560 Instrumentation & Control l 12 3.460 2.220 0.040 1.200 Administration & Engineering.

Regulatory Services 0 2.400 1.200 1.200 j

Utility Employees 47 17.976 0.870 13.930 2.010 1.166 i Contractor Workers & Others 483 206.108 0.860 9.380 194.408 1.460 GRAND TOl'ALS 653 275.774* 19.970 32.550 13.560 196.674 2.270 10.750

176.834 rem due to Unit 2 steam generator replacement.

l 1648 individuals were monitored exempt from the provisions of 10 CFR 20.

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VI. STEAM GENERATOR INSERVICE INSPECTIONS STEAM GENERATOR EDDY CURRENT TESTING The following abbreviations are used throughout this report section.

LISTOF ABBIEVIATIONS MBM Manufacturing Burnish Mark

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. MMB Multiple Manufacturing Burnish Mark DNG Ding TSli Tubesheet Hot Leg FBil Flow Bame liot Leg FBC Flow Baffle Cold Leg xil # x Tubesupport Plate 110t Leg xC # x Tubesupport Plate Cold Leg AVx Anti-Vibration Bar # x UNIT 1 No eddy current testing was performed on Unit I during 1996.

UNIT 2 Inspection Plan: Unit 2 steam generators were replaced during U2R22. Baseline eddy current testing was performed May 25,1996 through June 3,1996. Rows 3 and above were inspected full length using a bobbin coil (3394 tubes in each SG). Rows I and 2 were inspected over the straight length using a bobbin coil and a Plus Point was utilized for the U-bend (105 tubes in each SG). Additionally, at least one tube from each heat was inspected using a Plus Point (34 tubes in A SG,24 Tubes in B SG). These inspections were performed at each structure (top of tubesheet, bame plate, and each of the 7 tube support plates) intersection +/- 2" Finally, manufacturing burnish marks were inspected using a Plus Point (42 tubes in A SG,16 tubes in B SG).

Inspection Results: The following table lists the number of tubes found with indications:

A SG B SG 110t Leg Cold Leg 110t Leg Cold Leg DNT 5 0 1 0 h1BN1 33 23 13 15 MMU l 0 0 0 Page 91

"A" Steam Generator IndicatiDn3 Row Column Indication Location inch Mark 2 5 DNG FBli 12.78 2 5 DNG FBli 14.04 2 5 DNG FBli 15.61 2 5 DNG FBli 17.52 2 5 DNG FBil 18.34 7 90 DNG 3 11 17.85 2 5 MBM FBH 18.1 -

28 21 MDM 6C 35.91 28 21 MBM 6C 35.71 64 21 MBM 5C 28.65

64 21 MBM SC 28.48 39 22 MBM AV2 4.49 15 28 MBM 4 11 27.6 15 28 MBM 4 11 27.96 23 28 MBM IC 7.02 23 28 MBM 1C 6.92 39 28 MBM 2 11 21.47 71 30 MBM 4C 32.04 71 30 MBM 4C 31.67 38 31 MBM 111 3.71 38 31 MGM 111 3.86 61 34 MBM 2C 23.03 61 34 MBM 2C 22.65 12 35 MBM 511 5.92 38 35 MBM AVI 9.43 38 37 MBM AVI i1.65 75 40 MBM FBC 11.22 75 40 MBM FBC 11.24 31 48 MBM 511 23.57 31 48 MBM 511 31.9 31 48 MBM 511 38.74 31 48 MBM 5 11 41.82 -

31 48 MBM 511 35.3 31 48 MBM 511 24.81 31 48 MBM 511 33.58 31 48 MBM 511 41.82 27 50 MBM 3C 6.89 27 50 MBM 3C 6.85 17 52 MBM TSil 17.09 17 52 MBM 411 40.7 17 52 MBM 6l1 11.21 17 52 MBM 6C 36.34 17 52 MBM 6C 28 66 l' age 92

Row Column Indication Location inch Mark 17 52 MBM 6C

~

35.91 17 52 MBM TSH 17.79 17 52 MBM 4H 40.65 17 52 MBM 611 11.76 4 59 MBM 211 3.63 79 70 MBM IC 42.74 41 72 MBM AVI 10.12 75 72 MBM 2 11 25.04 75 72 MBM 2H 25.04 44 83 MBM 111 39.67 44 83 MBM 511 35.48 44 83 MBM 111 39.67 44 83 MBM 511 35.48 45 86 MBM 6C 28.68 7 90 MBM 311 17.49 18 93 MBM 5C 12.07 i 102 MBM 1C 6.13 1 102 MBM 1C 32.35 1 102 MBM FBC 17.95 1 56 MMB TSli 9.76 "II" Steam Generator Indicalimu Row Column Indication Location inch Mark 55 66 DNG TSli 2.65 66 21 MBM 5H 13.34 66 21 MBM 511 13.37 26 27 MBM 2C 33.49 26 27 MBM 2C 33.02 61 36 MBM 3C 33.99 61 36 MBM 3C 33.86 19 38 MBM 2C 42.56 19 38 MBM 2C 42.59 84 51 MBM 3H 4.86 14 53 MBM 4C 13.65

. 60 55 MBM 4C 10.52 60 55 MBM 4C 10.59 9 62 MBM 411 9 9 62 MBM 4H 8.92 45 62 MBM AV5 13.34 64 65 MBM FBC 13.51 21 66 MBM 6C 2.9 21 66 MBM 6C 2.99 55 66 MilM 'lSil 2.43 l

l' age 93

Row Column Indication Location inch Mark 55 66 MBM TSIi 2.53 53 68 MBM 6C 5.9 53 68 MBM 6C 6.13 1 70 MBM TSil 24.77 1 70 MBM TSH 24.65 80 71 MBM TSli 2.6 80 71 MBM TSH 2.61 14 91 MBM 511 36.76 14 91 MBM 511 36.64 F.epaired or Plugged Tubes: No repairs or plugging was required.

9 b

+

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Page 94

VII. REACTOR COOLANT SYSTEM RELIEF VALVE CIIALLENGES Overpressure Protection During Normal Pressure and Temperature Operation There were no challenges to the Unit 1 or Unit 2 reactor coolant system power-operated relief valves or safety valves at normal operating pressure and temperature in 1996.

. Overpressure Protection During Low Pressure and Temperature Operation The Unit I reactor coolant system power-operated relief valves were challenged on March 31,1996 during low pressure and temperature operation. The valves functioned properly and remain operable. The Unit 2 valves were not challenged in 1996.

VIII. REACTOR COOLANT ACTIVITY ANALYSIS There were no indications during operation of Unit 1 or Unit 2 in 1996 where reactor coolant activity exceeded that allowed by Technical Specifications.

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Page 95

L-97-006, 1996 Annual Results & Data Rept for Pbnp,Units 1 & 2

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L-97-006, 1996 Annual Results & Data Rept for Pbnp,Units 1 & 2

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