NPL-99-0091, 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with

1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with

ML20207D640
Person / Time
Site:	Point Beach
Issue date:	12/31/1998
From:	Kaminskas V WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NPL-99-0091, NPL-99-91, NUDOCS 9903100033
Download: ML20207D640 (113)

Text

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Electnc POWER COMPANY Poht Beach Nuc' eor Plont. (920) 755-2321 l 6610 Nuclear Rd.. Two Rtvers. WI 54241 NPL 99-0091 -

l February 25,1999 ' i Document Control Desk U. S. NUCLEAR REGULATORY COMMISSION Mail Station PI-137 Washington, DC 20555 Ladies and Gentlemen:

d DOCKETS 50-266 AND 50-301 1998 ANNUAL RESULTS AND DATA REPORT f

- POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Enclosed is the 1998 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.

De report contains descriptions of facility changes, tests and experiments; personnel occupational exposures; steam generator in-service inspectiorm; and commitment change evaluations that occurred ,

during 1998. j r

i Sincerely,

' h/

, Vito A. Kaminskas f Manager, Regulatory Services & Licensing

~ tat. 1 Enclosure cc: NRC Regional Administrator, Region III Y' j NRC Resident Inspector 0 '

9903100033 981231 PDR ADOCU 05000266'. )

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i ANNUAL RESULTS AND i l

DATA REPORT l 1998 I

i POINT BEACH NUCLEAR PLANT '

UNITS 1 AND 2-m' i

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U. S. Nuclear Regulatory Commission Dockets Nos. 50-266 and 50-301 Facility Operating License Nos.

DPR-24 and DPR-27

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• - 1 This Annual Results and Data Report for 1998 is submitted in accordance with Point Beach Nuclear Plant, i Units 1and 2. Technical Specification 15.6.9.1.B and filed under Dockets 50-266 and 50-301 for Facility Operat Licenses DPR-24 and DPR-27, respectively. .

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TABLE OF CONTENTS PAGE

. I. - INTRODUCTION 3

11. HIGHLIGHTS 3

< 111..- A MENDMENTS TO FACILITY OPERATING LICENSES 4 IV. 10 CFR 50.59 AND 10 CFR 72.48 SAFETY EVALUATIONS 5 Procedure Changes 5 Modifications 28 Temporary Modifications 82 Miscellaneous Evaluations 87 V. COMMITMENTCHANGE EVALUATIONS 103 VI. JOB NUMBER OF PERSONNEL AND PERSON-REM BY WORK GROUP AND FUNCTION 104 Vll. STEAM GENERATOR INSERVICE INSPECTIONS . 105 Vill. REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES 112 Overpressure Protection During Normal Pressure and Temperature Operation 112 Overpressure Protection During Low Pressure and Temperature Operation 112 VIV. REACTOR COOLANT ACTIVITY ANALYSIS 112 l

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1. INTRODUCTION ne Point Beach Nuclear Plant, Units I and 2 utilize identical pressurized water reactors rated at 1518.5 Mwt each.

Each turbine-generator is capable of producing 497 MWe net (524 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. lilGilLIGitTS UNIT 1 liighlights for the period January 1,1998 through December 31,1998, included unit operation from January I through February 13,1998, when the unit entered its twenty-fourth refueling / maintenance outage. The unit was returned to service on June 30 and operated until S ptember 2,1998, when power was reduced to 50% power to repair a steam leak on the "A" main feedwater pump suction line. The unit was returned to full power on September 3 and remained at full power until September 28,1998, when power was reduced to 35% to repair the "A" main feedwater pump and governor. Unit I was retumed to service on October 3 and operated at full power until November 14,1998. A power reduction to 51% was made in order to repair the IP-28A steam generator feedwater pump. The unit was returned to full power operation on November 29,1998, and remained there through i December 31,1998.

Unit I operated at an average capacity factor of 60.5% (MDC net) and an electrical / thermal effidency of 34.8%.

Unit and reactor availability were 62.7% and 63.4%, respectively. Unit ? generated its 92 billior'h kilowatt hour on January 1,1998; its 93 billionth kilowatt hour on August II,1998; and its 94 billionth kilowatt hour on  :

October 31,1998.

UNIT 2  !

liighlights for the period January 1,1998 through December 31,1998 included the unit being off line until February 8,1998, following a November 1997 shutdown to accommodate testing of the reactor protection system logic pennissives. The unit operated at full power from February 8 until March 5,1998, at which time the unit was shut down during an evaluation of component cooling water system inoperability (see LER 301/98-002-00). The unit operated from March 29 through September 12,1998, when a reduction to 50% power was made to repair a  !

failed oil fitting on the "B" main feedwater pump. The unit retumed to full power operation on the same day and remained there until October 23,1908, when power was reduced to 22% to allow for disconnection of the 2X-02 {

l unit auxiliary transformer for troubleshooting. The unit was retumed to service on October 29,1998. Unit 2 then '

operated at reduced power until December 5,1998, to support end-of-life coastdown when it began its twenty-third refueling / maintenance outage.

Unit 2 operated at an average capacity factor of 73.5% (MDC net) and an electrical / thermal efficiency of 34%. Unit and reactor availability were 75.5% and 76%, respectively. Unit 2 generated its 91 billionth kilowatt hour on April 6,1998; its 92 billionth kilowatt hotr on July 10,1998; and its 93 billionth kilowatt hour on September 25,1998.

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l 111. AMENDMENTS TO FACILITY OPERATING LICENSES During 1998 there were four amendments issued by the U.S. Nuclear Regulatory Commission to facility Operating 7

License DPR-24 for Point Beach Nuclear Plant Unit 1, and five amendments issued to Facility Operating License DPR 27 for Point Beach Nuclear Plant Unit 2. The list also cont. ins an amendment approved in 1997 not included in last years report. He License amendments are listed by date ofissue and summarized below:

Amendment 181 to DPR-24; Amendment 185 to DPR-27, September 29,1997: The amendments revised the Technical Specifications (TS) to eliminate the one-time requirement for Unit 2 Type A testing since the testing has been completed. It also deleted the TS Bases relative to containment purge valve testing since the reference no -

longer applies.

Amendment 182 to DPR-24; Amendment 186 to DPR-27, March 17,1998: The amendments updated the TS radiation protection program references to comply with revised Title 10, Code of Federal Regulations, Part 20.

Amendment 183 to DPR-24; Amendment 187 to DPR-27, March 24,1998: The amendments changed the title of the individual meeting the llealth Physics Manager qualification requirements to Health Physicist and allowed the position to be a line supervisor.

Amendment 184 to DPR-24; Amendment 188 to DPR-27, July 13,1998: The amendments allowed for removal of the Radiological Effluent Technical Specifications (RETS) from TS to licensee-controlled documents. This is consistent with Generic Letter 89-01," Implementation of Programmatic Cantrols for Radiological Effluent Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program."

Amendment 185 to DPR-24; Amendment 189 to DPa-27, July 17,1998: The amendments revised the TS line item surveillance for the radiation monitoring system from a generic surveillance encompassing monitors to specifically address effluent monitors or those credited as functioning in our safety analysis. Requirements for other monitors were removed to licensee control.

Amendment 190 to DPR-27, July 21,1998: The mendment revised the implementation schedule for the boron concentration changes to address the planned com ersion of Unit 2 to an 18-month fuel cycle.

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IV.10 CFR 50.59 & 72.48 SAFETY EVALUATIONS PROCEDURE CHANGES The following procedure changes requir'mg safety evaluations (SEs) were implemented in 1998:

1. AOP-98 Units 1 & 2, Component Cooling System Malfunction, Revision 11. (Permanent)

The change allows for venting the component cooling water (CCW) system. The procedure change addresses CCW system operability concerns addressed by condition reports (CRs) 98-0564 and 98-0619. The procedure is performed during cold shutdown conditions.

Summary of Safety Evaluation: Damage to the CCW pumps could occur without proper venting of the pumps.

Restoration of CCW to the residual heat removal (RHR) heat exchanger is performed slowly to minimize thermal shock and potential water hammer. Restoration of cooling to other components is similar to establishing normal cooling flow. This change provides for quick restoration of the CCW system.

Radiological releases from venting CCW are monitored through the primary auxiliary building (PAB).

Possible release from the evolution is bounded by the current activity for a shutdown unit which is routinely vented via the containment purge system. He first venting beation is through the CCW surge tank vent. This has an automatic shut off for high radicactivity.

Since equipment is operated and its function maintained as defined in the current licensing basis (CLB), the margin of safety is not reduced. His change provides guidance to vent the CCW system following a break inside containment and is within the CLB. The change does not pose an unreviewed safety question (USQ) nor does it require a change to the Technical Specifications (TS). (SE 98-038)

2. AOP-9B, Units 1 & 2, Component Cooling System Malfunction, Revision 12. (Permanent)

The SE evaluates the radiological consequences of venting the CCW system. An administrative hold was placed on AOP-9B Revision 11 until completion of this evaluation.

l Summary of Safety Evaluation: Direction to recover the CCW system if a leak occurs coincident with a loss of coolant accident (LOCA)is provided in AOP-9B with reference from EOP-1.3," Transfer to Containment Sump Recirculation." Containment sump recirculation is established without CCW to the RHR heat exchanger.

This change restores CCW flow to the RiiR heat exchanger by slowly opening the inlet after venting and ,

starting the CCW pump. He inlet valve is then fully opened after 30 minutes to allow venting of the heat l exchanger up to system pressure, ne outlet valve is then opened to establish flow. This method limits thermal shock and possible water hammer.

l A CCW system rupture inside containment coincident with a LOCA resulting in core damage is not within the CLB for the CCW system. Therefore, a radiological release from venting the CCW system for the postulated accidents in the CLB is limited to the maximum activity level allowed in the reactor coolant system (RCS) coolant. The maximum possible activity and release is within the CLB and access to the CCW system is not afrected for restoration. The containment atmosphere that is contained in the closed system before shutting the coltainment isolation valves (CIVs) is released as a result of this venting procedure. A radiological release as a restJt of venting CCW is monitored through the PAB. The first venting loccion is through the CCW surge tank vent. This has an automatic shut feature for high radioactivity. Response to the vent valve shutting results  :

in further evaluation of the radiation monitoring system (RMS) alarms.

The change slowly restores cooling flow to the RHR heat exchanger and vents the CCW system following a break inside containment. This is within our CLB. The change does not pose a USQ nor does it require a change to the TS. (SE 98-047)

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3. DCS 3.1.7, Service Water Pump Operability, Revision 10. (Permanent)

DCS s.l.7 documents additional administrative restrictions concerning the permissible conditions for entry into the Limiting Conditions for Operation (LCOs) for the service water (SW) system a.* specified in TS 15.3.3.D.2.

nese restrictions limit the permissible LCOs for the situation of one or two SW pumps being out of service.

Specifically, TS 15.3.3.D.2.b, SW ring header flow path, may only be entered with one SW pump out of service for the purpose of splitting the pump bays at SW-2890 and SW-2891. His change also provides the conditions for entry into the LCOs and possible concurrent LCO entries in a table based on the number of pumps in service.

Summary of Safety Evaluation: DCS 3.1.7 restricts the allowed SW system operating configurations to those ~

more conservative than currently allowed by TS. The restrictions are based upon knowledge of the SW flow model, worst case SW flow model runs using the additional restrictions, and knowledge of containment fan '

cooler (CFC) performance testing results. The restrictions ensure the SW system performs its design basis functions to mitigate accidents or events as described in the CLB. Since the SW system performs its design basis functions and components operate within design limits as described in the CLB, the probability of a malfunction of equipment important to safety is not changed.

A TS interpretation was included concerning opening of the CFC outlet motor operated valves (MOVs). The interpretation states these valves must be operable when they are opened while in the TS LCO. This allows the MOVs to be shut if a safety injection (SI) signal is received on the opposite unit. Shutting the non-accident unit MOVs, SW-2907 and 2908, is specified in EOP-0," Reactor Trip or Safety injection," and is assumed to be completed in the SW flow calculations, if the valve was inoperable in the open position, it could not be shut.

This interpretation is consistent with the Basis for the TS that states that these valves are opened at various times to allow testing of the CFCs or adjustment of the system flow rates.

The margin of safety as defined in the Basis of the TS is not reduced by this change. The change adds restrictions to the existing SW operation that are more conservative than the TS requirements. These conservative restrictions ensure the SW system is capable of performing its design basis functions to mitigate an accident or event. The existing TS requirements are met. De change does not eliminate er conflict with existing requirements; it only adds restrictions on entry into the existing LCOs. De SW flow analysis is currently in progress to determine additional needed TS changes. When this analysis is complete, a TS change request will be submitted. (See License Amendments 174/178 dated July 9,1998 for this analysis.) The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-081)

4. DSP 4. Multi-Sealed Basket (MSB) Drain Down/ Saturation Time Limit Calculation, Revision 1. (Permanent)

The SE revision addresses the change in the MSB drain down formula.

10 CFR 72.48 Evaluation Summary: The change implements an interim, more conservative administrative limit. The limit is embodied in 2 new formula that is used until this condition for system use (CSU) is revised by the NRC. Sierra Nuclear Corporation (SNC), the certificate of compliance holder, submitted an interim position acceptable to the NRC, pending resolution of the issue via rulemaking. The revision parallels the interim position. The new administrative limit is based on in-situ temperature measurements of the MSB internal water and change:, in temperature over time, as bounded by conservative theoretical heatup rates, rather than an assumed initial temperature and theoretical heatup rate in the existing CSU.

The new formula is conservative with respect to the present CSU l.2.10 formula for calculating the drain down time limit. No new occupational exposure or environmental impact concerns are created by the new formula.

The change does not pose a USQ, significantly increase occupational exposure, create a significant unreviewed  ;

environmental impact, nor change the license conditions as contained in the Certificate of Compliance (C of C).

(SE 96-043-03)

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5. EM, Environmental Manual, Revision 13. (Permanent) l The revision clarifies the intent of air sample field tests, the addition of sampling sites, including a new figure showing those sites.

Summary of Safety Evaluation: De revision augments the current radiological environmental moni*oringjl program as previously written. He EM only applies to offsite environmental samples. It does not affect systems, structures, or components (SSCs) necessary for the safe operation of either unit or the independent spent fuel storage installation (ISFSI). Neither the possibility nor the consequences of a previously analyzed accident are increased. No new accident scenario is created. Equipment malfunctions and margins of safety

. are not affected. The change does not pose a USQ nor does it require a change to the TS. (SE 98-1S3) 6.

EOP-0, Units 1 & 2. Reactor Trip or Safety injection, Revision 25. (Permanent)

EOP-0, Units 1 & 2, Series Foldout Page, Revision 1. (Permanent)

EOP-I, Units I & 2 Loss of Reactor or Secondary Coolant, Revision 23. (Permanent)

The changes isolate CCW upon in<lication of a CCW rupture. The method ofisolation preserves the desired forced circulation cool down method if the rupture occurs coincident with other postulated accidents in the CLB. The design basis of the CCW system requires action to manually isolate a break in the system inside containment.

Summary of Safety Evaluation: A CCW system rupture inside containment coincident with a LOCA that results in core damage is .at within the CLB for the CCW system. Isolation of CCW based upon surge tank level and reactor coolant pump (RCP) cooling return flow ensures that a radiological release is minimized for this specific beyond design basis accident (DBA). There is no change to the radiological consequences for those accidents evaluated in the CLB as a result of the loss of RCP cooling water flow.

Two independent means are used to identify a CCW rupture inside containment. One is surge tank level and the other is coo og water flow from the RCPs. Loss of CCW to the RCP seals could result in a failure of the seals. Rupture of the CCW system inside containment during a steam line break (SLB) is not a postulated accident. Since a SLB does not alTect these indications, erroneous isolation of CCW to RCPs based upon th indications does not occur during a SLD. .

Current design and licensing bases require the ability to repair the CCW system after a LOCA. It is possible that isolation of a CCW system leak inside containment does not occur prior to air binding the CCW pumps.

The CCW pumps are accessible for repair and recovery of the system to assist with long-term decay heat removal (DHR).

The change to isolate CCW to the RCPs based upon surge tank level and RCP cooling water flow is within the

. CLB. The change does not pose a USQ nor does it require a change to the TS. (SE 9P-045) 7.

EOP-0, Reactor Trip or Safety injection, Revision 25. (Temporary)

The revision clarifies a step to locally shut non-essential SW MOVs that do not shut automatically or ma by use of the control room control switch. His action, if needed, is implied by existing wording that states

" Verify service water isolation valves shut." The step clarifies that the valves are to be locally shut (by man action in the field)if necessary.

Summary of Safety Evaluation: The SW design basis flow calculations require non-essential SW isolation valves (except the unaffected unit turbine hall MOV) to be shut prior to adjusting flow through the CCW heat exchangers to ensure adequate SW flow to required components during the sump recirculation phase of a DB If this action was not completed, SW flow may not be adequate throegh the CFCs after SW flow to the CCW beat ex. hangers is increased to the value specified by EOP-l.3, " Transfer to Containmen' Sump Recircul t

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There is sufficient time for the operators to shut the required valves prior to the PAB operator being directed to begin EOP 1.3 Attachment A,' Local Actions," for the sump recirculation cooling lineup. The ability of the operators to shut the required valves within the available time period was validated by a timed field walkdown during which operators actually shut the subject valves.

Analyzed SW flow is available to the required components during the sump recirculation phase of a DBA if actions within EOP L3 are taken. Hat procedure isolates the non-essential SW isolation valves prior to increasing SW flow to the CCW heat exchangers. He ability to achieve the necessary sump recirculation lineup prior to refueliing water storage tank (RWST) depletion is not affected by the local actions required to shut a train of non-essential SW MOVs if required because of a loss of an emergency power train during a DB A. 'she procedure does not pose a USQ nor does it require a change t i the TS. (SE 98-076) .

8. EOP-l.3, Units 1 & 2, Transfer to Containment Sump Recirculation, Revision 16. (Permanent)

Summary of Safety Evaluation: Isolation of the shell side of a RliR heat exchanger in the event of a loss of CCW prevents boiling from occurring in the RilR heat exchanger after establishing containment sump recirculation. A loss of CCW inventory results in a reduction in static pressure at the R11R heat exchanger so boiling could occur. Subsequent condensation in either the CCW supply or return line results in water hammer in the piping. Isolation of the heat exchanger shell side results in rising temperatures in the shell side. The installed relief valve on the shell side provides overpressure protection and is designed for this situation. He RiiR heat exchanger is designed beyond the temperature for the maximum containment sump temperature.

Isolation of the heat exchanger reduces the stresses applied by the water hammer to the CCW piping and improves survivability of the RilR heat exchanger and adjacent piping.

The peak containment accident pressure and the peak containment pressure are not affected by a LOCA to the RiiR heat exchanger. Therefore, the radiological release from containment is not increased. Loss of CCW coincident with a LOCA resulting in core damage is not within the CLB. The CCW system is available for recovery at a later time.

Since equipment is accessible, the local actions to establish containment sump recirculation are not time critical when there is no core damage. Locally isolating the shell side of the CCW heat exchanger does not affect the time required to establish containment sump recirculation and the risk oflosing core cooling is not increased.

The guidance to vent the CCW system following a break inside containment is within the CLB. The change does not pose a USQ nor does it require a change to the TS. (SE 98-046)

9. EOP-1.3, Units 1 & 2, Transfer to Containment Sump Recirculation, Revision 18. (Permanent) ne SW system configuration meets that assumed in the SW design basis flow calculations prior to aligning SW to the CCW heat exchangers to support sump recirculation cooling. Ifless than five SW pumps are running, or the SW system is in a ring header continuous flow path LCO (TS 15.3.3.D.2.b), the SW calculations fer the sump recirculation phase of a design basis LOCA assume the following three non-essential SW flow paths are isolated prior to adjusting SW to the CCW heat exchangers for sump recirculation cooling:

• 1&2SW-2880, turbine hall supply (for accident unit only)(auxiliary feedwater pump AFP room El. 8') .

. SW-2817, water treatment supply (AFP room El. 8')

. SW-LW-61 or SW-LW-62, radwaste systems supply return (PAB El. 26', near C-59 control panel)

Summary of Safety Evalurcion: Ifless than four SW pumps are running, an automatic non-essential SW isolation signal occurs and at least one train of the non-essential SW valves isolate. De other train of non-essential SW valves is shut by EOP-0," Reactor Trip or Safety injection."

The completion of these additional steps using control room switches to stan additional SW pumps and/or opening SW ring header MOVs is insignificant with respect to affecting the ability to achieve the necessary sump recirculation lineup prior to RWST depletion. The maximum tirne required to shut the non-essential SW Page 8 of 112

valves was determined by timed field walkdowns. Operators shut the valves by local manual action as directed by control room personnel. Based upon the reco ded times, the ability to complete the field actions to make the sump recirculation lineup prior to RWST depletion are affected.

The margin of safety as defined in the TS is not affected by this change. The ability to cool the core and maintain containment integrity is currently analyzed and does not change the analysis. The margins of safety for the peak cladding temperature and containment pressure are not afTected by this change.

The ability to achieve the necessary sump recirculation lineup prior to RWST depletion is e,ot affected by the manual actions required to verify the SW system is in the proper configuration to support lining up SW to the

' CCW heat exchangers for sump recirculation cooling. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-091)

10. EPIPs and ERO, New Emergency Response Organization and Emergency Plan Implementing Procedures.

(Pennanent)

The Emergency Response Organization (ERO), Emergency Plan (EP), Emergency Plan Implementing Procedures (EPIPs) and other ERO support documents were revised. Emergency response facility duties were redistributed, ERO personnel titles changed, response and activation times changed, etc. Procedures that support the new organization changed accordingly. Other documents that support the new ERO also changed.

Summary of Safety Evaluation: Smcc the ERO primary function is directing personnel response to an accident, there is a potential for affecting the radiologica consequences of an accident. The changes to the ERO are primarily administrative in nature and the original intent and function of the ERO has either remained the same or has been enhanced. Extensive classroom and practical training was conducted prior to implementation of the new ERO. lhe new ERO was also evaluated in mini-drills, table-top scenarios and utility drills prior to changeover. Emergerscy Plan changes were previously evaluated in accordance with 10 CFR 50.54q. The change does not pose a USQ nor does it require a chsage to the TS. [SE 98-172) i 1

11. IIPIP 3.52.2, Containment Forced Vent / Purge Permit Sampling, Revision 0. (New Procedure)

The procedure samples containment air from outside containment at the RE-2il/212 cubicle. It allows for temporary removal of test connection fittings from, and operation of valves RM-3200R and RM-3200V. I Samples of the containment atmosphere are taken with a sample pump that has flow rates less than P-707B forced vent pumps.

Summary of Safety Evaluation: Although RE-211/212 pressure and flow slightly changa during this sampling, the RE-211/212 alert setpoints (as used for RCS leakage detection) are not adversely affected. Because the sample pump used has flow rates lower than the P-707B forced vent pump, effects on RE-211/212 are negligible since pressure and flow changes encountered during HPIP 3.52.2," Containment Forced Vent / Purge Sample From RE-211/212 Cubicle," sampling are less significant to that encountered during similar modes of normal operation. A check valve provided at RM 3200V prevents the possibility of a RE-211/212 malfunction resulting from improper connection or operation of sampling equipment. Equipment is connected to non-seismic and non-safety related portions of piping outside the containment boundary. Equipment operation does not interfere with CIV operation.

The procedure does not increase the probability of equipment malfunction, nor does it create a new type of malfunction. Because there is no increase in eqdpment malfunction probability or new possibility of malfunction, accident probability (as a result of unchanged pre-LOCA leak detection capability) remains unchanged. RE-21 I/212 or manipulations made to RE-211/212 are not accident or event initiators, and could not create an accident or event of a different type than described in the CLB.

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I Radiological consequences of postulated accidents are not increased because sample valves are outside the containment isolation boundary. Other sampling is prohibited while the containment vent or purge is in effect.

The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-009)

12. ICP 10.29, Seal Table Operations for Refueling, Revision 14. (Permanent)

The change allows cutting f'ive thimble tubes and pushing below the conduit opening and installing five seal table valve assemblies during Unit i Refueling 24 (UIR24). The tubes have fretting at the lower core plate and lower intemals. The change also removes the valve assembly following refueling and removes the worn thimble tubes. He change includes installation of the new thimble tubes and refurbishment of the high pressure seals. ,

Summary of Safety Evaluation: No accident scenarios or malfunctions as analyzed in the CLB are affected by the installation of the valve assemblies. No radiological consequences are increased by the use of the valve assemblies. The assemblies are plugged preventing leaks of refueling cavity water during core offload.

Refueling cavity level is not affected by installation of the valve assembly. He procedure change does not pose a USQ nor does it require a change to the TS. (SE 98-026)

Summary of Safety Evaluation: Installation of the new ferrule at the same location on the conduit as the original ferrule does not increase the probability of a LOCA. The conduit is prepared and refurbished following removal of the old ferrule. De procedure was successfully performed at other plants with no resultant leakage. The seal table is inspected during the leak test, which is performed prior to criticality

" following the refueling. This confinns the adequacy of the installation. He valve assembly installed on the thimble conduit has no other affect to the reactor coolant system and therefore can e-eate no possibility of an accident or event of a different type. The procedure change does et pose a USQ nor does it require a change to the TS. (SE 98-026-01)

13. IICP 11.482A&B, Safeguards Train A & B Test Switch Replacement, Revision O. (New Procedure)

The procedures reflect the modification that replaced the Unit I emergency safeguards feature (ESF) test switches and removed spare " press.to-test" indicator sockets. The new switches are equivalent to the old. The test requires deenergizing both trains of safeguards control power for Unit 1. Only one train of safeguards is deenergized at a time. The test is performed when Unit I is defueled, or in a cold or refueling shutdown with no refueling operations in progress.

Summary of Safety Evaluation: During replacement of the Train B ESF test switches, no restrictions or LCOs are required. This is based on no requirements for SW cooling of the Train B EDGs.

Prior to energizing safeguards, post-maintenance testing (PMT) verifies the circuitry and test switches.

Following energization of safeguards control power, voltage is checked across the actuation contacts of the containment isolation and containment ventilation isolation circuits to verify availability, and if required, operability for refueling. Finally, prior to leaving cold shutdown, procedures llCP 02.019,"ESF System Logic Shutdown Surveillance Test," and llCP 02.020," Post-Refueling Pre-Startup RPS and ESF Analog Surveillance Test," are performed to verify operability of the RPS/ESF analog and ESF logic systems. .

The ESF system is not required to be operable during these plant conditions. Removing each train from service with applicable isolations, entering applicable standby emergency power LCOs, establishing conditions for SW operability and capabilities for containment closure, and performing PMT, ensures that the activities do not increase the probability of occurrence or radiological consequences of a previously evaluated accident, event or malfunction. Use of qualified equipment and maintaining installation standards that do not affect the seismic qualification of the racks ensures that a different type of accident, event or malfunction is not introduced to the ESF system.

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Since the physical activities performed in the ESF racks maintain the system design, the safety margin of the ESF system is not reduced. Additionally, since the applicable standby emergency power LCOs are entered during the installation, and conditions are established for SW operability and capabilities for containment closure, the margin of safety for the affected TS are maintained. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-006)

14. IT 08A, Unit 1, Cold Start Testing of Turbine-Driven Auxiliary Feedwater Pump and Valve Test (Quarterly),

Revision 29. (Pcmanen_t) t IT 09A, Unit 2 Cold Start Testing of Turbine-Driven Auxiliary Feedwater Pump and Valve Test (Quarterly),

Revision 17. (Permanent)

The SE revision addressed comments made by the Offsite Review Committee. The SE revision contains furtherjustification of the new throttle valve settings and further evaluation discussions of steam generator tube rupture (SGTR), SLB and small break loss oicoolant accidents (SBLOCA).

Summary of Safety Evaluation: The auxiliary feedwater (AFW) flow requirement is based upon halting the increase in pressurizer liquid level before it reaches the level of the pressurizer power-operated relief valves '

(PORVs) and safety valves. The current Unit I licensing basis loss of normal feed (LONF) and loss of all AC (LOAC) safety analysis assumes that 60 seconds after the AFW actuation signal, the delivery of 100 gpm of feedwater to each of two steam generators (SGs)is sufficient to mitigate the accident. The current Unit 2 i

LONF/LOAC safety analysis (WCAP 14602) assumes that five minutes after the AFW actuation signal (low-low SG water level reactor trip) the delivery of 200 gpm of feedwater to a single SG is sufricient to mitigate the accident. This is consistent with the worst case of allowing the operators time to switch AFW suction to SW and align the system as necessary to meet the flow requirements.

De change does not increase the probability of an accident or event, create an accident or event of a different type, nor increase the rr.diological consequences of accidents or events previously evaluated in the CLB. He procedure does not pose a USQ nor does it require a change to the TS. (SE 97-208-02)

15. 0110, Boron Recycle, Revision 10. (Permanent)

The revision allows for either a batch process or a continuous bleed process of the liquid effluent in the chemical and volume control system (CVCS) holdup tanks (HUT) for boron recycle in accordance with FSAR Section 9.2.

Summary of Safety Evaluation: A continuous bleed process was added to minimize the restrictions placed on operations during the current boron recycle operation. The valve used to control liquid effluent changed from a 2" diaphragm valve (BS-1116) to a downstream 1" diaphragm valve (BS-1132B).

The change is not described'in TS and does not affect the operation of safety-related equipment. It does not introduce abnormal conditions in which the equipment must operate. The change does not hinder boron recycle operation and the continuous bleed process for bcron recycle is not affected by the operation of the boric acid evaporators. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-181)

16. 01 16A, Control and Processing of Contaminated Secondary inventory, Revision 0. (New Procedure)

The procedure controls the processing oflarge volumes of contaminated water generated as a result of a SGTR accident. The contaminated water is primarily contained in equipment in the main and reheat steam, condensate and feedwater, and steam generator blowdown (SGBD) systems. The procedure is to be used after the accident unit is in a stable condition, with RCS temperature <l20*F.

Summary of Safety Evaluation: The concentration of radioactive material in contaminated water is calculated to be greater than efiluent concentration limits of 10 CFR 20 and, therefore, the processed water is controlled to not exceed 10 CFR 20 limits nor violate TS requirements. Release paths by which the contaminated water may Page1I ofi12

l 5 l reach the environment have been identified and are controlled. Leakage to the environment may occur through j the circulating water system via condenser heat exchanger tube leaks. The procedure does not allow storage of contaminated water in the condenser hotwells at levels above 65" until after it is demonstrated that the condenser heat exchanger does not leak. Leakage of contaminated water from equipment located in the containment facade and the turbine building is collected in sump systems. These sump systems have the capability of pumping water to the efTluent sump and ultimately to the environment. He procedure ensures that the turbine building sump and the containment faqade sump systems are secured. He basement floor of the turbine building and the containment fa ade contain access points to the subsoil drainage system. To prevent uncontrolled release of the contamirated water to the environment, subsoil drainage system access points local to the area where temporary hoses are used are covered. The PAB basement does not contain access points to the subsoil drainage system and leakage to the environment from the building structure is '

minimal. The procedure does not reduce the margin of safety as defined in the Basis of TS, nor does it increase the radiological consequences of an accident or event previously evaluated in the CLB.

Both the PAB sump and the waste holdup tank (WHUT) have high level alarms which alert the operators to a hose failure. In the faqade, a hose rupture would release water to floor drains that drain to the facade sump.

De fa9ade sump has high level alarm indication in the control room. In the turbine building, a hose failure would drain to the turbine building sump which also has a high level alarm indicator. Therefore, the procedure does not increase the probability of an occurrence ofinternal flooding. The procedure does not affect equipment important to safety.

The contaminated water control in the main and reheat steam, condensate and feedwater, and SGBD systems creates new radiatioa areas. For example, calculated dose rates near the condenser hotwell are, depending on the hotwell level and the calculated concentrations in the water,3-5 mrem / hour. Above normal dose rate conditions will also exist near other secondary system equipment. The procedure contains provisions to ensure the appropriate radiological controls are implemented. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-092)

17. 01 151, HX-012C&D Component Cooling System Heat Exchanger Data Collection, Revision 0.

(New Procedure)

The procedure allows for data collection to ensure that HX-012C&D capacity is adequate to remove the required heat load at design limiting conditions. Highest heat loads for the CCW heat exchangers (HX) occur during unit shutdown when the RHR system is first placed on-line. During this pnase of plant operation and during post-LOCA plant operation, tne CCW heat exchangers help with DHR from the RHR system. One CCW heat exchanger must provide DHR for the safe shutdown of one unit. Data collected during this procedure is used to ensure that HX-012C&D can perform that function. Train B of the PAB battery and inverter room ventilation system is isolated. In addition, CCW heat exchanger HX-012C or HX-012D is isolated, depending on which HX is monitored. 01151 involves collecting data from non-intrusive ultrasonic test (UT) flowmeters externally-mounted RTDs. .

Summary of Safety Evaluation: The Unit 2 CCW heat exchanger not monitored is isolated during the procedure. Both the CCW beat exchangers and the PAB battery room ventilation system can be unisolated during procedure performance if needed. Heat removal from the CCW system is not reduced by passing all of the CCW flow for a unit through one HX, and higher flow rates within the HX do not cause adverse effects. l Only one train of the PAB battery and inverter room ventilation system is normally in operation at a time. One train is able to adequately cool the PAB battery and inverter room. The test instmmentation is non-intrusive and adds negligible loads to the piping. The data collection does not pose a USQ nor does it require a change to the TS. (SE 98-108)

Page 12 of i12

18. OP 1 B, Reactor Startup, Revision 31. (Permanent) i The revision rebaselines the plant process computer system (PPCS) inverse count rate ratio following withdrawal of the shutdown banks; adds a note to allow operators to adjust the source range audible count rate channel; adds a step to alent Chemistry of the alternate dilute mode; changes the temperature of startup to 547'F; adds steps to assist in determining the point of adding heat using the advanced digital reactivity computer; changes the optimum reactivity holdup to 60 pcm versus 50 pcm; and uses the alternate dilute mode l of the CVCS with CV-Il0C shut when diluting to initial criticality for a cycle. l 1

i Summary of Safety Evaluation: Use of the attemate dilute mode with CV-Il0C shut does not increase the probability of a previously analyzed accident. The method is not significantly different from the current mode except that dilution is continued until criticality is achieved and the concentrations in the pressurizer, volume l

control tank (VCT), and RCS remain relatively equal. The amount of positive reactivity from dilute water that '

could potentially remain in the system can be compensated by control rod motion without challenging the zero power insertion limit. The CVCS and valve CV-Il0C are designed to operate as described, therefore no ,

increase in the probability of malfunction of this equipment occurs. The radiological consequences of I previously analyzed accidents are not increased since the initial conditions and mitigating actions are not changed. No other accident or malfunction of a difTerent type is created. The margin of safety as defmed in TS i is not affected since the conditions for criticality are met and the zero power insertion limit is not challenged by the procedure. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-048)

19. OP 2A, Normal Power Operation, Revision 28. (Temporary) l l

The revision adds a statement to allow Unit 2 generation to remain at 100% power while bus section (BS) 2-3 breaker is out-of-service for maintenance. The temporary change incorporates the results of a grid l

stability study performed in 1992 in accordance with Calculation N-92-047, Revision 2 and its addendum. '

Summary of Safety Evaluation: This evaluation analyzed a loss of off-site power (LOOP). A LOOP is an j analyzed event and within the plant design bases. A LOOP is not an accident initiator. The change enhances '

the integrity of the grid power supply. This in tum maximizes the availability of the offsite power supply to PBNP. This ensures that plant process control and monitoring equipment operate within normal power supply parameters. The change does not reduce the margin of safety defined in the Basis of TS. The change does not pose a USQ nor does it require a change to the TS. (SE 98-059)

Summary of Safety Evaluation: There are no stability concerns associated with operation of Unit 2 at 100%

power with BS 2-3 breaker open. Additionally, reducing Unit 2 generation could adversely affect the offsite power supply to PBNP under some grid loading conditions. The change does not pose a USQ nor does it require a change to the TS. (SE 98-059-01)

20. OP 2A, Normal Power Operation, Revision 28. (Temporary)

The revision adds a section to allow power operation with an X-02 unit availability transformer mir.imally loaded. Another section added allows the A-01 and A-02 loads to be transferred from X-04 to X-02 once X-02 load restrictions are resolved. This permits continued opeilon of a unit with an X-02 load-related concern.

Summary of Safety Evaluation: The change allows actions to occur within OP 2A that are normally performed during a unit shut down per OP 3A. There are no time restrictions in OP 3hregarding operation of a unit with its non-safeguards buses supplied from its X-03 and X-04 transformer supply during power operation.

Supplying power to a unit's safeguards and non-safeguards buses from an X-03 and X-04 transformer supply is within the design bases for both transformers. Therefore, continuous operation with a unit's non-safeguards buses supplied from its X-03 and X-04 transformer supply is within the X-03 and X-04 transformer design basis.

Page 13 of112

a ei i

l Operation of the supply and tie breakers to the A-01 and A-02 buses is within the scope of normal plant I operations. ne.efore, the procedure does not pose a USQ nor does it require a change to the TS. (SE 98-133)

21. OP 9D, Dist harge of Gas Decay Tanks, Revision 16. (Permanent)

The r.: vision changes the normal position of WG-14A, T-20A D gas decay tank (GDT) vent throttle valve, from shut to locked shut. Additionally, the C-59 controls for WG-14 were changed from red locked shut to shut. WG-14A accommodates a padlock more easily than WG-14.

Summary of Safety Evaluation: A waste gas accident is defined in FSAR Section 14.2.3 as "an unexpected and uncontrolled release to the atmosphere of the radioactive xenon and krypton fission gases that are stored in the

• waste gas storage system." The FSAR also states that," failure of a gas decay tank or associatad piping could result in a release of this gaseous activity." Installing a padlock on the WG-14A valve handwheel is not a waste gas accident initiator. The waste gas system components are Seismic Class I, designed to the standards of ASME II' Class C ard USAS-B31.1. Installing a lightweight link chain to the valve handwheel with a padlock in a manner commonly used for other padlocked valves does not affect the structural integrity of these components nor does it increase the source term from a waste gas accident, affect the release rate or duration, nor create a new release mechanism or release path.

The changes are in compliance with FSAR Section 11.1.2. The procedure for a gaseous waste release described in the CLB is maintained. The margin of safety remains tha same. The procedure changes do not involve a USQ nor TS change. (SE 98-080)

22. OP 10E, CVCS B HUT, kevision 6. (Temporary)

A procedural temporary modification was installed and removed via OP 10E that provided nitrogen overpressure / makeup to T-8B CVCS HUT during sampling for discharge and through completion of the discharge. The "B" HUT is isolated from the waste gas header during the evolution.

Summary of Safety Evaluation: The "B" HUT is protected from overpressurization by tank relief valve BS-1266. The regulator is set to maintain tank pressure well below the relief valve setpoint of 15 psig. The fittings, valve regulator and tubing used for this temporary connection were rated for pressure at least as high as the nitrogen header relief valve setpoint of 150 psig. A check valve is installed downstream of the P-tubing at the connection to the "B" HUT to prevent back flow of radioactive gas or liquid into the nitrogen system. This also prevents leakage of radioactive gas or liquid to the PAB in the unlikely event of a fitting or tubing failure in the temporary lines. The "B" HUT must also be protected from being drawn into a vacuum, which could cause failure of the tank. The gas stripper feed pumps are used for discharges. The nitrogen supply is sized to provide adequate flow to makeup for the maximum allowed fiow rate in OP 10E. The waste gas header that normally supplies the gas makeup to the tank is controlled by a single regulator. He temporary gas makeup is supplied by a single regulator which is expected to be as reliable as the waste gas regulator. Based on this, the probability of collapsing the tank because of a negative pressure is not increased.

The equipment important to ssfety as related to a discharge of the "B" HUT includes the radiation monitors, associated alarms and automatic isolation function if activity increases above a predetermined level. The -

i installation and use of the nitrogen makeup capability rather than the waste gas header does not affect the ability of this equipment to perform its functions ofidentifying a discharge with higher than expected activity and isolating the discharge.

l 1

Page 14 of 112 l 1

TS 15.7.6 requires that prior to sampling for discharge, a batch release shall be isolated and mixed to ensure a representative sample occurs. The "B" HUT could not be isolated from the waste gas header as the waste gas header provides the makeup to the tank while the tank is being drained. De temporary modification allowed isolation of the waste gas header to the "B" HUT and provided makeup gas to the tank from the nitrogen header waich was supplied from the liquid nitrogen tank. It ensured that the tank lineup met the intent of the TS.

Therefore, it did not reduce the margin of safety as defined in the TS. The temporary nitrogen supply to the "B" HUT for discharges does not pose a USQ nor does it require a change to the H. (SE 98-007)

23. ORT 3 A, Unit 1, Safety injection Actuation with Loss of Engineered Safeguards AC (Train A), Revision 31.

(Permanent)

The change removes the Train B ESF testing from ORT 3A. The procedure initiates a Unit i Train A safety injection (SI) and 1 A-05/B-03 bus undervoltage (UV) with G-01 emergency diesel generator (EDG) aligned to ,

1 A-05/B-03. A subsequent Si and bus UV are initiated with G-02 aligned to I A-05/B-03. G-02 start is 1 inhibited during G-01 testing and G-01 start is inhibited during G-02 testing. His places both units in an s emergency power LCO per TS 15.3.7.B.I.h. Normal and standby emergency power are operable to Train B  !

equipment, and Train B ESF are operab!c as required by this TS.

The revision isolrtes 1B-03 from l A-05 to perform Unit 1480 V UV in Unit 2 SI sequencing circuit testing.

This isolation causes a sustained loss of normal an6 standby emergency power to bus IB-03. The change also incorporates Generic Letter 96-01 testing requirements.

Summary of Safety Evaluation: The test is performed with Unit 1 in cold or refueling shutdown with RCS temperature <140*F. Applicable LCOs for both units are entered. G-01 EDG, and subsequently G-02, are l aligned to bus 1 A-05 and a simultaneous Unit 1 Train A Si and bus I A-05 UV are initiated in accordance with TS 15.4.6.A.2. Automatically sequenced loads available on bus I A-05 are loaded onto G-01 and G-02 EDGs.

l The testing configuration and equipment operation is within the system and component requirements specified in the CLB. The capability of the equipment to perform under these loading scenarios was demonstrated by Point Beach Test Procedure (PBTP) 65, "G-01/G-02 Functional Test with Unit 2 Accident Loads and Unit I Cold Shutdown Loads." Cold shutdan requirements of Unit I are maintained by the train which is not being tested (Train A). Applicable LCOs are entered for Unit 2 operation. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-082) )

24. ORT 3B, Unit I, Safety injection Actuation with Loss of Engineered Safeguards AC (Train B), Revision 29.

(Permanent)

)

l The change removes the Train A ESF testing from ORT 3B. The procedure initiates a Unit 1 Train B Si and i

i A-06/B-04 bus UV with G-03 EDG aligned to I A-06/B-04. A subsequent Si and bus UV are initiated with '

G-04 aligned to 1 A-06/B-04. G-04 start is inhibited during G-03 testing and G-03 start is inhibited during G-04 testing. This places both units in an emergency power LCO per TS 15.3.7.B.I.h. Normal and standby emergency power are operable to Train A equipment and Train A ESF are operable as required oy this TS. The  !

change also incorporates Generic Letter 96-01 testing requirements.

Summary of Safety Evaluation: The test is performed with Unit I in cold or refueling shutdown with RCS temperature <l40*F. Applicable LCOs for both units are entered. G-03 EDG, and subsequently G-04, are aligned to b> s 1 A-06 and a simultaneous Unit 1 Train B Si and bus I A-06 UV are initiated in accordance with TS 15.4.6.A.2. Automatically sequenced loads available on bus I A-06 are loaded onto G-03 and G-04 EDGs.

Page 15 of 112

The testing conGguration and equipment operation equipment is within the system and component requirements specified in the CLB. The load carrying capabilities of G-01 and G-02 were demonstrated during the performance of PBTP 66, *G-03/G-04 Functional Test with Unit 2 Accident Loads and Unit 1 Cold Shutdown Loads." Cold shutdown requirements of Unit I are maintained by the train which is not tested (Train B).

Applicable LCOs are entered for Unit 2 operation. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-083)

25. ORT 3C, Unit 2, Auxiliary Feedwater System and AMSAC Actuation, Revision O. (New Procedure)

The testing is the same as that previously performed as part of ORT 3,3 A and 3B for Unit 2. ORT-3C demonstrates automatic operation of the AFW following anticipated transient without scram mitigation system actuation circuit (AMSAC) actuation; AMSAC actuation of turbine trip AMSAC reset block when AMSAC trip signal is present, and actuation of AMSAC trip computer alarm; automatic operation of A FW on SG low-low level; and automatic response of AFW because oflow AFP suction pressure.

The test is performed with Unit 2 in cold or refueling shutdown. Mctor starting duty limits are incorporated into the procedure.

Summary of Safety Evaluation: Steps are included in ORT 3C to coordinate with steps in ORT 3,3A and 3B to close the main feedwater pump breakers in the test position. The main feedwater pump breaker closing circuit interlocks for suction pressure, seal water and oil pressure are defeated for this evolution and restored upon test completion. A 7-day LCO entry (per TS 15.3.4.C.2) for P-38B electric-driven AFP is necessary for a portion of the test when the low pressure suction trip for P-38B is tested. He P-38A&B AFPs remain operable to Unit I throughout the other portions of the test. During a portion of the test, the discharge MOVs to Unit I for the motor-driven AFPs are opened and verified to shut on a Unit 2 Si actuation signal. In order to reduce the possibili ty of injecting relatively cold feedwater (-65-70* injected feedwater versus ~440*F main feedwater) into Unit I, a pre-job briefis performed to explain the expected system response and operator responsibilities during this portion of the test. A caution statement is included just prior to this portion of the test to verify that the valves to Unit I are shut and to take operator action to shut them if necessary to limit the discharge of relatively cold feedwater to the Unit I SGs. In the event that some relatively cold feedwater is injected into Unit 1, as is done during the normal quarterly inservice tests of the AFPs, there is no effect on Unit 1. During the normal quarterly inservice tests, a sustained injection of cold feedwater is required; therefore, a 2% power reduction is required to offset the addition of the cold feedwater.

The total amount of AFW that could be directed to the SGs during this portion of the test is minute in comparison to total feedwater. Therefore, there is little or no effect on the operation of Unit 1. Complete mixing of AFW and main feedwater is expected prior to the bulk fluid reaching SG components. If cold AFW was injected into the Unit 1 SGs, no adverse effects to components in the flow path or on the operation of Unit i occurs. Appropriate restrictions are incorporated in the procedure for motor starting duty times to ensure the AFPs are not adversely affected. The testing is routine and within the normal design limits of the equipment.

The shared safeguards equipment, in this case both P-38A&B AFPs, remain operable throughout the test (except once) to support accident mitigation on Unit 1. ne portion of ORT 3C Unit 2 that tests the low suction pressure trip for P-38B requires entry into a 7-day LCO per TS 15.3.4.C.2. No other safeguards equipment needed to mitigate an accident previously evaluated in the FSAR is affected. Unit 2 is in cold or refueling shutdown and AFW is not required. He margin of safety defined in the Basis of TS is not reduced. The testing does not pose a USQ ror does it require a change to the TS. (SE 98-069)

26. PBTP 077, Transient Response of G-02 Replacement Governor, Revision 0. (New Procedure)

The test loads G-02 EDG using bus 2.A05 loads to demonstrate satisfactory transient response of G-02 replacement govemor. A Unit 2 SI signal and 2A-05 bus UV is initiated with G-02 EDG aligned to bus 2A-05.

PBTP 077 simulates the maximum practicable accident basis step load.

Page 16 of 112 l

l

Summary of Safety Evaluation: PBTP 077 is performed with fuel in the reactor and R CS temperature < 140"F.

Applicable LCOs for Unit I are entered. G-02 is aligned to bus 2A-05 and a simultan :ous Unit 2 Train A Si and bus 2A-05 UV is initiated. Automatically sequenced loads available on bus 2A-0.5 are loaded onto G-02 in order to demonstrate G-02 EDG response to load demand after governor replacement.

CCW pump 1P-l1 A is not available for G-02 loading because of the inability to deenergize bus I A-05. The Si and bus UV trip of 2P 11 A CCW pump are inhibited to allow 2P-11 A to be used in place of IP-11 A for G-02 loading.

A Unit 2 SI signal is initiated by means of a manual SI pushbutton, which normally actuates both Trains A and

- B. The Train B inputs from the pushbuttons are inhibited to allow only a Train A Si signal to be initiated. .

l The test configuration and equipment operation associated with the performance of this test is within the system and component requirements specified in the CLB. The load carrying capabilities of G-02 were demonstrated '

during the performance of PBTP 65, "G-01/G-02 Functional Test with U2 Accident Loads and U1 Cold Shutdown Loads." Cold shutdown requirements of Unit 2 are maintained by the train not tested (Train B).

Applicable LCOs are entered for Unit I operation. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-001)

Summary of Safety Evaluation: One Si pump is disabled and Si blocked. The PORVs remain operable to comply with the low temperature operation protection (LTOP) requirements of TS 153.15. LTOP is not j

challenged during this testing since the Unit 2 Train A Si pump is operated on the test line. Unit 2 RCS is '

depressurized (<20 psig on 2PI-493) and double valve isolation is provided between the discharge of the Train

{

A SI pump and the RCS. The procedure does not pose a USQ nor does it require a change to the TS.

(SE 98-001-0_lj

27. PBTP 086, Mobture Carryover Testing, Unit 2, Revision 0. (New Procedure)

PBTP 086 collects data to determine the moisture carryover performance of the Unit 2 SGs after a radioactive tracer (Nauas sodium nitrate) is injected into the feedwater system at the suction header piping for P-28A or B main feedwater pumps using a temporary injection system. TM 98-008 installs the eq.iipment. Except for using manual SG level control, increasing SG level to $74% does not require changes in the operation, control, j

function, or operability of safety-related systems, structures, or components. The automatic feedwater isolation

{

on high SG level is operational during manual SG level control.

Summary of Safety Evaluation: PBTP 086 identifies the upper SG data point of 74% level as 4% below the high level feedwater isolation setpoint of 78%. 'Ihe 4% margin provides the operator adequate response time to avoid a high level feedwater isolation event. PBTP 086 provides detailed instructions for monitoring reactor ,

power using core AT in accordance with REI 1.0, " Power Level Date'mination and Guidelines," and REI 17.0, l

" Backup Methods for PPCS Monitoring Functions," prior to and .ng SG level changes. During SG level

- k changes while holding secondary plant load constant, reactor thet..al output (RTO) calculate by the computer is expected to be unreliable by 1.2 - 1.3%.  !

Although sodium is known to preferentially hide out in SG tubing crevices and for.n alkaline conditions, the SG tube material in Unit 2 is thermally treated Alloy 690 which is proven to hiwe improved resistance to initiation ofintergrannular attack (IGA) and stress corrosion cracking (SCC). Based on previous carryover tests, the predicted sodium concentration in the SGs during testing is sl4 ppb, which is kss than the allowable normal at power sodium concentration of s20 ppb listed in FSAR Table 10.2-1 and is also below Action Level I for sodium in NP 3.2.3,

• Secondary Water Chemistry Guidelines." In the event SG sodium concentration exceeds 20 ppb, PBTP 086 requires compliance with NP 3.2.3.

PBTP 086 requires 2RE-219 SG liquid discharge :nonitor be in " alarm off." RE-219 is placed in " alarm off" after Unit 2 SGBD has been shut down. RE-219 is one of three liquid effluent monitors identified in TS Table 15.7.3-1, associated with SGBD. With SGBD shut down, none of the monitors are required to be in service.

Page 17 of 112

Prior to returning SGBD to service, PBTP 086 requires verification that RE-219 is returned to service including verification ofits daily channel check requirements listed in TS 15.7.4-1 are satisfied. PBTP 086 does not affect the operability of other SSCs identified in the TS. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-070)

28. PBTP 089, Unit ISteam Generator Blowdown System, Revision O. (New Procedure)

PBTP 089 operates the Unit 1 SGBD system to allow data gathering during low power operation (<30% RTO) of Unit 1.

Summary of Safety Evaluation: PBTP 089 detennines when the condensate system (CS) return piping starts to '

excessively vibrate, the refore establishing operating guidelines that prevent the return piping vibration and precludes piping failures. PBTP 089 varies SGBD flow rate and condensate return temperature and collects data at each test point. De SGBD system is in the normal lineup for low power operation during PBTP 089 -

(through heat exchangers, bypass tank, pump, and filters to SW overboard). No other changes are made during the performance of PBTP 089 (e.g., reactivity changes, turbine generator output changes, etc.) thereby ensuring that Unit I operation is steady-state and plant operators are attuned to the test. SGBD flow rate is input into PPCS as 10 kib/hr throughout the duration of the test to conservatively account for changt.s in RTO because of SGBD flow rate changes. De test is performed during recovery from UIR24 prior to shifting condensate return from the condenser to the heater drain tank (IT-23) as controlled by OP-1C. " Low Power to Normal Power Operation." Acceptable operation of the SGBD system and CS return piping is based upon CS return pressure above saturation pressure of fluid. Operating with CS return pressure above the saturation pressure ensures that transported fluid flashes across the orifice plates in the return line rather than upstream of orifice plates (the cause of piping vibration). Data obtained is used to establish operating guidelines for both Unit I and Unit 2 SGBD systems.

PBTP 089 does not operate the SGBD system outside ofits design basis function. He safety-related components of the SGBD system are the CIVs for each train of SGBD. Valves IMS-5958 and IMS-5959 are the CIVs (Class IV per FSAR Chapter 5 Figures 5.2 50-1 and 5.2 51-1) for each SGBD line. PDTP 089 has no affect on these valves; therefore, the valves remain fully operational as defined by TS 15.3.6.A.I.B.2 throughout testing.

PBTP 089 does not adversely affect the SGBD system nor affect low power operation of Unit I during testing.

The procedure does not pose a USQ nor does it require a change to the TS. [SE 98-074)

29. PBTP 090, Transient Response Test of G01 Governor following Maintenance Adjustment, Revision 0.

(New Procedure)

The procedure perfonns a LOOP test of safeguards bus 1 A-05. The test verifies that G-01 EDG starts, attains required voltage and frequency, and energizes permanently connected loads within acceptable limits and time.

The test requires that Unit I be in cold sht'tdown, refueling shutdown or defueled. Unit 2 may be at-power based on the LCO provision identified in this evaluation.

Summary of Safety Evaluation: LCO entries are made for Train A SW, Train A AFW, and Train A standby - ,

emergency power. Bus I A-05 is deenergized (LOOP) and G-01 starts and supplies bus loads as designed.

IP-ISA SI pump and 1P-11 A CCW pump are operating prior to the LOOP. The pumps restart when the G-01 output breaker closes to I A-05. Train A SW pumps, P-32A and P-32B, automatically sequence on to G-01 as designed. Train A SW pump, P-32F, automatically starts and remains powered from offsite power via bus 2A-05. Sequenced loading of G-01 is complete within one minute of LOOP initiation. The loading is maintained for at least five additional minutes. Component run times are extended as required to comply with ,

published motor starting duties.

Page 18 of112 1

After supplying the I A-05 bus for at least six minutes, the single-load rejection test is performed by securing the IP-15A Si pump. This test demonstrates EDG capability to reject the single largest load while maintainin voltage and frequency requirements.

Appropriate TS LCO entries permit this test. Testing is similar to existing TS surveillance tests for the standby emergency power supplies. Testing and LCO entries are limited to Train A components that preclude a failure of the tested train from preventing accident / event mitigation using the redundant Train B equipment. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-077)

30. PBTP 091, Adjustment and Transient Load Response Test of G-02 Governor, Revision 0. (New Procedure) I The test requires that Unit 1 be in cold shutdown, refueling shutdown or defueled. LTOP requirements of TS 15.3.1.5 are satisfied by maintaining one SI pump disabled. De PORVs remain operable for the duration of

- this test as required by TS. Unit i RCS is not solid and the RCS is depressurized (<25 psig) so LTOP is not challenged. He Unit 1 Train A SI pump is operated on mini-recirculation during this test. Unit 2 may be {

j at-power based on LCO provisions. '

Summary of Safety Evaluation: PBTP 091 provides assurance of redundant circuit functionality in the G-02 EDG starting controls, adjusts its governor, then performs load stability and transient load response testing.

This is accomplished by simulating a LOOP to safeguards bus I A-05, verifying that G-02 starts, provides the required voltage and frequency and accepts load within specified limits and time. This test includes a single l largest load rejection and a full load rejection test of the generator. I The testing is similar to existing surveillance and maintenance tests for the standby emergency power supplies.

This test and required LCO entries are limited to Train A and do not affect the ability of the redundant Train B equipment to perform their design function. De procedure does not pose a USQ nor does it require a change toi the TS. (SE 98-084)

Summary of Safety Evaluation: The test is performed with Unit I fueled, RCS temperature >l40*F and <350*F and the RCS not solid. DHR is provided in accordance with TS 15.3.1.A.3.a(1). LTOP requirements of TS 15.3.15 are satisfied by maintaining one SI pump disabled and the PORVs operable for the duration of the test. The Unit 1 Train A SI pump is operated on recirculation during this ten. De procedure does not pose a USQ nor does it require a change to the TS. (SE 98-084-01)

31. PBTP 092, Unit 2, Turbine / Generator Startup Following LP Turbine Retrofit Outage, Revision 0.

(New Procedure)

PDTP 092 perfonns the initial turbine startup and identified PMT required following the Unit 2 low pressure turbine / generator retrofit modification. PBTP 092 is similar to PBTP 062; the Unit 1 LP turbine retrofit.

Summary of Safety Evaluation: During performance of PBTP 092, minimum turbine overspeed protection devices are maintained operable as required by FSAR Appendix T and the actual setting of the mechanical overspeed device is verified. Overspeed and torsional testing on Unit 2 requires the auxiliary governor and independent overspeed protection system (IOPS) to be out of service. The Unit 2 mechanical governor remains fully functional during overspeed and torsional testing. If required as a precautionary measure, IOPS may also be disabled on Unit L Unit I mechanical overspeed and auxiliary governor both remain fully functional and the Unit I crossover steam dump system remains operable. The procedure includes specific instructions for the restoration from the torsional testing and mechanical overspeed test. Following restoration from the torsional and mechanical overspeed testing, the procedure retums the plant back to 10-17% reactor power where plant control returns to OP IC,"L.ow Power Operation to Normal Power Operation," for performance of a normal escalation to 100% reactor power.

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PBTP 092 does not change setpoints nor controls that affect SSCs identified as important to safety. It does not '

impose new fire loading nor increase the probability of a fire to safeguards equipment areas. Torsional testing is performed with the unit offline and electrically isolated from the switchyard and station buses. Required generator and step up transformer protection is maintained functional during torsional testing. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-119)

32. REI 2.0, Power Range Detector Power Level Adjustment, Revision 12. (Permanent)

The procedure change requires setting the nuclear instrumentation system (NIS) power range high power trip setpoint to 85% if any power range channel was adjusted down and the power level is less than 70%. His puts the trip setpoint below the rod stop setpoint. His setpoint is then in effect until a secondary side power -

calorimetric is used above 70% power to adjust the power range channels.

Summary of Safety Evaluation: The revision prevents the malfunction of the reactor tripping function of the power range detectors that could otherwise occur under the conditions described in Westinghouse Technical Bulletin 92-14-RI. The bulletin discusses decalibration of NIS due to increased uncertainty in calculated reactor thermal output (RTO) at reduced power. The bulletin states that adjusting the NIS power range indication down uhen power level is below 70% could result in an NIS high power trip in excess of the limiting trip point assumed in safety analysis during a power increase transient with the trip set at its nominal setpoint.

Therefore, the power range high power trip setpoint needs to be reduced to an acceptable level prior to adjusting the NIS power range indication down when power level is below 70% to maintain accident analysis assumptions. The bulletin also states that at the beginning of a cycle, the NIS trip setpoint should be set at approximately 85% of rated power until a more accurate secondary side power calorimetric can be used for normalization purposes. Additional bulletin recommendations indicate that this "more accurate secondary side power calorimetric" is at greater than 70% power based on RTO. The values in the bulletin are based on a 2%

feedwater flow Ap uncertainty in channel span with a 120% flow span and two flow loops. A comparison of PBNP specific to the Westinghouse basis for the values show that the values in the bulletin are conservative for

. PBNP use. Also, the bases are applicable throughout the cycle. The NIS high power trip setpoint adjustment keeps the reactor tripping function of the power range detectors functioning as assumed in the safety analysis.

The procedure maintains the NIS power range high power trip function within the value assumed in the tecident analysis. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-151) 33, REI 10.1, Pre-Critical Alignment and Setup of the Advanced Digital Reactivity Computer, Revision 0.

(New Procedure)

REI 10.1 prepares the advanced digital reactivity computer for use. The computer uses the same signals as the previousi analog computer to determine reactivity. The signals are first processed through picoammeters before the signals are processed into reactivity. The new computer uses its own power supply to the power range NIS detectors. It also requires an input from the control rod drive system to monitor rod insertion. He procedure differs in that the power range channel used for the reactivity is placed in trip for the overtemperature and overtemperature rod stop and the fuses are removed from the instrument.

Sumrnary of Safety Evaluation: The installation does not increase the probability of previously analyzed .

-accidents or malfunctions. The test instrument used to monitor the " Rod In" light circuit is a high impedance ,

device that has no affect on the rod drive or renctor protection systems. The high voltage power supply to the power range detector does not affect the power range channel while it is out of service. The channel is checked prior to returning it to service to ensure that it is operable. Placing the power range channel removed from service in " trip" is a condition required by TS. Placing the channel in " trip" maintains the redundancy of the reactor protection system required by TS. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-094) i l

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34. 2RMP 9096, Unit 2, Reactor Vessel Head Removal and Installation, Revision 18. (Permanent)

The revision incorporates an optimized reactor vessel (RV) stud tensioning, detensioning and elongation I procedure. This reduces the required number of tensioning sequences from 34 to 20 and detensioning sequences from 32 to 24. It also expands the allowable stud elongation tolerance from an original tolerance of 0.037" to 0.041" to a new tolerance of 0.037" to 0.047" and provides requirements for determining

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acceptability of a lightly loaded stud. The determination is based on an average of five studs (a lightly loaded

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stud and two studs adjacent each side of a lightly loaded stud) meeting the minimum elongation requirement. l Summary of Safety Evaluation: Acceptance ofless stress on an individual stud is allowable based on still providing minimum Code-allowable reload on the particular area of the flange. In all cases, the requirements of ASME Section 111, Winter 1968 Edition are met. Stresses associated with the unequal preload have been analyzed and found to be within Code-allowable limits.

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l FSAR Section 9.4.2 is revised to change the current description of how the studs are tensioned and what '

elongation measuring devices are used. He change to the tensioning sequence provides two modified tensioning passes, in that half of the studs are tensioned to an intermediate level with the remainder of the studs tensioned to final tension. The intermediate tensioned studs are then tensioned to final tension in the second pass. The sequence changes provide a slightly different interim condition for the reactor vessel assembly in that different stud loading occurs while in process than that occurring during use of past procedure requirements.

These changes provide effectively the same result of either all studs detensioned in preparation for removing i the RV head or all studs tensioned to provide RV closure in preparation for returning the RV to service. l interim and final stud loading conditions were evaluated via Dominion Engineering, Inc., report R-4411-00-2,  !

" Reactor Vessel Bolting Evaluations Point Beach Unit 2," and are within Code-allowable stress limits and

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provide closure in accordance with ASME Section 111, Winter 1968 Edition. He alternate stud elongation '

measuring equipment can be maintenance and test equipment (M&TE) and are equivalent to using a micrometer for the degree of accuracy required.

No new affects on the RV or the plant, including the accident analyses, radiological hazards, or malfunction are probable. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-177)

35. RMP 9327, CC-745 Swing Check Valve inspection, Revision 0. (New Procedure) l l

The procedure controls the use of a freeze seal downstream of 2CC-745 to allow maintenance on the valve.

2CC-745 is a 6" check valve in the CCW return line from containment. It has internal damage that prevents it from shutting tightly.

Summary of Safety Evaluation: The activity is done in accordance with the guidelines of Battfle report,

" Development of Guidelines for the Use ofIce Plugs and Hydrostatic Testing." Failure of the ice plug is ,

considered extremely unlikely; however, if the ice plug should fail, it would fail gradually as the ice melts. The j plug should be restrained between the upstream and downstream elbows. AOP 98, " Component Cooling System Malfunction," is entered to ensure makeup to the system is accomplished. The RCS is filled, vented {

and cooled down to approximately 110'F. The SGs are available so natural circulation can be established, if '

, necessary. Containment closure is established. I i

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At the low temperatures required for the freeze seal, the physical properties of the metal are changed. As the 1 temperature decreases, the yield strength and ultimate strength increase and the fracture toughness decreases.

Since the temperature required for the freeze seal is approximately -320'F, the piping is well below the brittle-ductile transition. The piping is examined for defects prior to freezing. A failure of the piping is considered extremely unlikely. However, if the pipe fails at the freeze location, a contingency plan is implemented. An emergency piping repair isolates or contains the leak. De CCW system would be considered inoperable; therefore, the RHR system would be considered inoperable. If repairs can not be made, AOP 9B," Component Cooling System Malfunction," SEP 1, " Degraded RHR System Capability," and/or SEP 1.1, " Alternate Core Cooling," are entered as required. Natural circulation is established. Feed and bleed cooling could be established if required.

The limiting flooding analysis in PSA 6.0, Section 6.3.5.7 for the PAB is based on a maximum flow rate of 27,000 gpm from the SW system. He flow through the check valve body if the freeze plug failed could not approach 27,000 gpm. Herefore, flooding caused by a failure of the freeze plug is bounded by the existing flooding analysis. If the entire CCW system is drained to El. 19', the RHR pumps would not be flooded. The volume of the central RHR cubicle is over 27,000 gallons, which is greater than the entire volume of the CCW system. This does not include the volume of the sump. WL-4100 and WL-4101, the RHR pump room drains, are maintained shut so the pump rooms do not communicate with the central cubicle. The entire CCW system volume could be contained in the central RHR pump cubicle and sump. He procedure does not pose a USQ nor does it requi- change to the TS. (SE 98-037)

36. RMP 9362, Welding MSB Lids and Valve Cover Plates, Revision 1. (Permanent)

RMP 9362 coordinates the welding and nondestructive evaluation processes associated with the loading process of the VSC-24 dry cask system. The procedure runs parallel to operational activities outlined in RP 7 Part 6,

" Preparing a Multi-Assembly Scaled Basket (MSB) for Storage in a Ventilated Concrete Cask (VCC)."

10 CFR 72.48 Evaluation Summary; ne changes relate to the performance of the UT examination of the structural lid to-shield lid groove weld. As a result of the need to perform the UT examination, a change to CSU l.2.9, is required because of the weld cracking issue surrounding closure of the structural lid of the MSB.

UT examination of the completed structural lid-to-shell weld demonstrates its structural soundness.

ne revised CSU l.2.9 directs the UT examination to be done,". . in accordance with the criteria defined in the Guideline Requirements for the Time-of-Flight Diffraction Ultrasonic Examination of the VSC-24 Structural Lid to Shell Weld, VMSB-98-001, latest revision." ne CSU discussed the actions necessary if the flaws are determined to be unacceptable during the dye penetrant portion of the non-destructive examination (NDE).

Additionally, the CSU contains the flaw screening criteria (size) that is acceptable for temperatures > 30'F.

This is linked to the change made to CSU 12.13 via SE 98-101.

The change was determined acceptable without prior NRC approval because of the non-intrusive nature of the UT on the structural integrity, shielding and cooling capabilities of the VSC. The change does not pose a USQ, significantly increase occupational exposure, create a significant unreviewed environmental impact, nor change the license conditions contained in the Certificate of Compliance. (SE 98-099)

37. RMP 9367, Containment Paint and Coatings for Structural Steel Floors, Revision 0. (New Procedure)

The procedure controls application of Service Level I QA coatings on structural steel and steel surfaces inside containment. The procedure includes surface preparation, application of coatings and a listing of coating materials. Plant conditions required by the procedure are refueling or cold shutdown with the containment cleanup system shut down and the purge supply and exhaust system in service. ne charcoal filters in the purge system are removed or an evaluation is performed to determine the disposition of the filters prior to or afler the coating process.

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l Summary of Safety Evaluation: The coating activities do not change the function or operation of plant equipment nor create an interim or off-normal condition that could cause a malfunction of equipment important to safety No accident described in FSAR Chapter 14 is initiated by coatings nor does the activity introduce new initiators that could increase the occurrence of a previously evaluated accident or cause an accident of a different type. Since the function and operation of plant equipment important to safety is not affected by application of Service Level 1 QA coatings, an increase in the radiological consequences does not occur. Here is no reduction in the margin of safety as defined in the Basis of TS.

Compliance with this procedure ensures that only coatings approved for Service Level I are used on steel and steel structures in contsinment. No adverse conaitions occur while performing coating activities or as a result

• of the coatings after application. He procedure does not pose a USQ nor does it require a change to the TS.

(SE 98-063)

38. 2RMP 9370-5,2A-05 Deenergized Bus Test Procedure, Revision 0. (New Procedure)

The procedure tests the 2A-05 4160 V bus differential relay, bus lockout relay, EDG breaker interlocks, and individual breaker overcurrent trip and lockout relay operation. Relay and breaker trip testing is performed with the breakers in the test pc sition. Breakers are closed, then tripped from the bus differential relay, through  !

the bus lockout relay, and froin each breaker overcurrent device. Breakers, with the exception of the EDG 1 breakers and spare breakers, are tested simultaneously to trip from the bus lockout relay. The EDG breaker trip 1 from the lockout relay are test :d individually. Blocking of breaker close after a bus lockout is also tested in conjunction with the testing.

j Summary of Safety Evaluatiorg EDG breaker interlock testing requires the normal bus feed breaker and both  !

the normal and altemate EDG breakers to be racked into the de-energized bus. Temporaryjumpers are installed on the undervoltage niay contacts the block automatic closure of each EDG breaker into an energized bus. He jumpers allow break-:r operation to test interlocks between these three breakers. Thejumpert, are removed following testing.

The procedure governs testing such that breaker alignment and the installation of temporaryjumpers are directly controlled by the proc: dure. The testing is verification of previously installed functions that remain the same at the completion of the testing. The testing is done with the bus, breakers, protective relaying, undervoltage relaying, and EEGs out of service. Appropriate LCOs are entered for equipment placed out of service, The change does not pose a USQ nor does it require a change to the TS. (SE 98 186) 4

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39. RP 7 Part 5, Remove the Mult -Assembly Scaled Basket (MSB) and the MSB Transfer Cask (MTC) from the l Spent Fuel Pool, Revision 2. ! Permanent) l Summary of Safety Evaluatioit: The SE revision addresses CR 97-3709 that included administrative changes to this document. SE 95-074-02 contained missing words from the end of sentences and strikeouts from the o

original SE. The editorial cha ages do not pose a USQ nor does it require a change to the TS. (SE 95-074-02)

40. RP 7 Part 6, Preparing a Mult -Assembly Scaled Basket (MSB) for Storage in a Ventilated Concrete Cask (VCC), Revision 6. (Permant n_t) t The SE revision addresses Cli 97 2782. He condition report identified the original SE as not adequately supporting the procedure change to allow 15 psig to be developed inside of the MSB vice the original 10 psig.

His change enhances operatc r flexibil:ty in using nitrogen gas under certain operating scenarios. The revision is minor and supports the con:lusion that the change in allowable pressure inside the MSB does not require further evaluation.

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10 CFR 72.48 Evaluation Summary: ne design limit pressure for the MSB is 41.6 psig. The 41.6 psig limit is based on Calculation 95-0074 that demonstrates this pressure level is acceptable based on ASME Section lit, Service Level A criteria. Service Level A represents the most conservative of four service levels described in Section !!! from an allowable stress perspective, and represents what would be considered normal operating stresses for the MSB. Establishment of this design pretsure was evaluated via SE 95-118. Although an accident prnsure of 34.6 psig is mentioned in licensing documents because of internal pressurization of a loaded MSB because of fission gas release, the Basis merely states that this pressure level falls within the I accident (Service Level D) stress allowables for the MSB. In fact, this pressure level not only falls within Service Level D, it also falls well within Service Levels A, B, and C. He procedure does not pose a USQ nor does it require a change to the TS. (SE 97106-01)

41. RP 7 Part 8, Transfer and Placement of the Ventilated Storage Cask (VSC) to the Independent Spent Fuel I

Storage Installation (ISFSI), Revision 6. (Permanent) ne revision incorporates lessons learned from loading and unloading dry runs. The revision also includes changes related to the update of Certificate of Compliance 72 1007, CSU l.2.13, " Minimum Temperature for Moving the MSB."

10 CFR 72.48 Evaluation Summay: The CSU, as currently written, limits movement of the MSB (loaded MSB implied) inside of the VCC to ambient temperatures of 0*F or above. As stated in the current basis and objective statements, the temperature limitation on movement of the MSB is intended to minir .ize the potential for brittle fracture. The licensing basis for the VSC-24 system currently requires all pressure e oundary materials comprising the MSBs to demonstrate a Charpy impact test absorbed energy of 15 fb ib at -50*F. His level of absorbed energy provides assurance that the pressure boundary materials will behave in a ductile manner at the minimum movement temperature (0*F) with a margin of 50*F.

A change to the current CSU l.2.13 is required because of weld cracking issues surrounding the closure welding of the MSB. One of the elements of closure is to perform a UT examination of the completed structural lid-to-shell weld to demonstrate its structural soundness. As the allowable flaw size in the structural lid weld is a function of the weld temperature, an evaluation temperature of 30*F was selected as this weld temperature offered an optimal compromise between flaw evaluation capability and ability to handle the MSBs under the expected range of ambient conditions.

The revised CSU l.2.13 requires t'.4t " Movement of the loaded MSB while inside the VCC will only be allowed at ambient temperatures of 0*F or above, coincident with a closure weld temperature of 30'F or above." His is accomplished by the action, " Confirm before removing the loaded MSB, while inside the VCC, that the ambient temperature is at 30*F or above." This verifies that both the general shell material is at 0*F or above and that the closure weld is at 30'F or above, as the MSB always has an internal heat load of some magnitude during its licensed service life. ne revised CSU is conservative or equivalent to the original CSU.

No USQ, significant increases in occupational exposure, nor significant unreviewed environmental impact exist. He change does, however, constitute a change in the license conditions contained in the C of C. This change is conservative relative to the original CSU basis and is endorsed by the NRC via letter dated July 17, 1998. (SE 98-101)

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42. RP-8, Unloading the Multi-Assembly Scaled Basket (MSB), Revision 6. (Permanent)

RP 8 Part 4, Placing the MSB Transfer Cask (MTC) into the Spent Fuel Pool, Revision 4. (Permanent) 10 CFR 72.48 Evaluation Summary: The SE revision addresses CR 97-3969 that cites an erroneous value in the time to clean up zine in the spent fuel pool (SFP) following cask load operations while using demineralizer U-6. The value is changed from 50.7 days to 73.9 days. The increase in zine concentrations in the SFP have adverse effects on U-6 demineralizer resin. This requires that the resin be changed out earlier, it takes 73.9 days to clean up the SFP. These clean up durations are based on conservative zinc and boron reaction rates. The procedure does not pose a USQ nor does it require a change to the TS. (SE 96-115-02)

43. SLP 1 and SLP 2, Units i & 2, items Lifted by Containment Polar Crane, Revision 9. (Permanent)

The revision incorporates changes to the safe load paths (SLPs) for the upper and lower RV internals. The SLPs of the upper and lower RV internals changed from single lines to a SLP area to allow more flexibility for the operators and riggers to move the load around interferences.

Summary of Safety Evaluation: Based on the PBNP submittal of January 1982 and review of the submittal by the NRC, the SLPs were developed to avoid movement of the loads over safety-related equipment (including the RV). The SLPs, however, did not contain dimensions or specific locations allowing the operator to avoid interferences. The new SLP for the RV upper intemals uses the same criteria for development of the SLP as the original path. The new SLP area avoids safety-related equipment and minimizes movement over the RV.

There is no increase in the probability for a load drop on irradiated fuel or redundant safety-related equipment.

Therefore, the activity does not increase the probability of occurrence of an accident or event, the probability of occurrence of a malfunction of equipment or the radiological consequences of an accident, event or malfunction of equipment important to safety. It does not create the possibility of an accident or event of a different type nor creates the possibility of a malfunction of equipment important to safety of a different type previously evaluated in the CLB. The change also does not reduce the margin of safety defined in the Basis of TS. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-182)

44. SLP 3, Turbine Building Crane, Revision 7. (Permanent)

The revision clarifles a limitation within SLP 3 to prevent lifting loads over the Operations of0cc if both swing safety-related battery D-305 and charger D-09 are in service to a safety-related bus if both are in service, they must be aligned to ensure that the potential for dropping a heavy load over the Operations office only affects a single train of safe shutdown equipment.

Summary of Safety Evaluation: The D-305 safety-related swing battery is capable of connecting to any one of normal 125 V de distribution buses D-01, D-02, D-03, or D-04 if the battery nonnally associated with the bus is out for maintenance or testing. The D-09 swing battery charger is only capable of being aligned to D-01 and D-02. The swing battery and swing battery charger are interlocked so the battery and charger can only be

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connected to one 125 V de distribution bus at a time; however, it is physically possible to have the D-305 swing battery and D-09 swing battery charger aligned to supply 125 V de to different trains of safe shutdown equipment. Clarification of the allowable alignments of the D-305 swing battery and D-09 swing battery charger ensures that the potential drop of a heavy load over the Operations office only affects a single train of safe shutdown equipment, thereby maintaining the requirements of NUREG-0612. The actions required in the event the safety-related swing battery and/or swing battery charger failure while aligned to a safety-related 125 V de bus are provided in TS 15.3.7.B.I. '

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The turbine building main crane SLP criteria originally approved by the NRC is conservative in that it takes no credit for the actual physical separation between safe shutdown equipment nor the physical dimensions of the heavy loads. The low pressure (LP) turbine hoods and LP2 rotors we the longest loads expected to be lifted over the SLP. The physical separation between the Train A EDGs and the safety-related swing battery and battery charger prevena a simultaneous loss of both Train A EDGs, the safety-related swing battery, and the safety-related swing battery chaiger.

While addition of the safety-related swing battery and swing distribution bus add more safety-related equipment to the SLP area, the probability of dropping a load is not changed, and the criteria of NUREG-0612 and NRC SER are not violated. Herefore, the procedure does not pose a USQ nor does it require a change to the TS.

(SE 98-179) ,

45. SMP 1193, ISW-2854 Replacement, Revision 0. (New Procedure)

SMP 1193 directs and controls the replacement of ISW 2854, including use of a fabricated flange assembly and an inflatable plug to provide downstream isolation. Replacement is in accordance with spare parts equivalency evaluation document (SPEED)98-059. ISW 2854 is a 4", air-operated valve on the 6" SW return line from the turbine tube oil coolers it controls the temperature of the lube oil to the main turbine bearings.

The replacement valve has a post-guided trim design that climinates the tight tolerances found in the old valve.

The new valve is adjusted to provide flow characte?istics similar to the old valve.

Summary of Safety Evaluation: A significant amount of water is spilled during this work when the flanges are spread to install the flange assembly. He water is contained within the tube oil dike area. The south doors to the vital switchgear area are placarded to prevent SW spray from reaching vital switchgear. This is a precautionary step since significant spray is not expected. His work is done with SW overboard aligned to Unit I circulating water to eliminate the possibility of Unit 2 condenser waterbox level problems. He plug is restrained on a lanyard to prevent the possibility oflosing it in the piping. Losing the plug is extremely unlikely since the system pressure at iSW-2854 is positive (~3 psig). It is certain that the flow rate is significantly less than the 600 gpm that normally flows through this line when the tube oil coolers are in service. Therefore, the SW flow model analysis is not affected.

If the Dange asembly or the inflatable plug failed, it is possible for SW returning from components in the PAB and containment to flow through the return header to the Unit I turbine building and onto the floor. If one of the llXs in these areas fails, it is possible for the SW to be contaminated. ISW-2854 is located adjacent to the turbine lube oil coolers inside the dike area. If the flange or plug failed, the SW would be collected in the dike area and would eventually be routed to the retention pond. If this occurs, an unmonitored release path is not created because the retention pond is monitored before being released to Lake Michigan. If SW return was contaminated,it would be identified by alarms on IRE-229 and/or 2RE-229. The water in the dike could be sampled before discharging it to the retention pond. As a contingency, SMP 1193 requires that a variety of isolation equipment be staged in the work area, meluding a second 6" inflatable plug, and a 4" 150 lb blind a

flange. If the flange assembly or inflatable plug fails, one or more of these centingencies are used to isolate the leak. The flooding analysis in PSA Section 6.3.5.7 identifies the Unit I turbine building as Flood Zone 3.

Flood Zone 3 is equipped with a 44"x44" flood damper specifically designed to channel water from the Unit I turbine building to the yard. The flooding analysis for Flood Zone ', considers a maximum leakage rate of .

314,000 gpm (runout capacity of two circulating water pumps). This is based on a complete failure of a circulating water line. The potential flooding caused by a failure of flange assembly or the innatable plug is well within the bounds of the existing flooding analysis, if the flange assembly and/or the inflatable plug failed, there would be an insignificant effect on the SW system pressures and flows. The SW return header is I ..m mally at a vacuum state. Leakage past a failed flange / plug might reduce that pressurr Hightly. If this I

occurs, overall system Dow increases slightly. The increase is distributed through every flow path. 'lhe effect on any single flowpath is insignificant. Overall system operation is not significantly affected.

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r SMP 1193 does not increase the proba'sility of an accident, event or equipment malfunction it does not increase radiological consequences of an accident, event or equipment utalfunction. It does not create a new type of accident, event .4 equipment malfunct.on. It does not reduce TS basis margin of safety. SMP 1193 does not involve a USQ nor require a TS change. (SE 98-090)

46. STPT 21.1, Sheet 1," Unit 1 Main Generator," Revision 1. (Permanent)

The Unit I voltage regulator minimum excitation limiter (MEL) setting change allows operation at higher megawatt (MW) output to the grid and retains the aoility to control generator output voltage. The main )

generator KLF (loss of excitation) protective relay setting change ensures that sufficient margin exists between

, its settings and the MEL settings of the Unit i voltage regulator.  ;

1 Summary of Safety Evaluation: The MEL does not allow an excitation reduction because of the additional MW l

' loading on the generator. MW load was reduced to regain voltage control. He recent LP turbine rotor upgrade improved cycle efficiency by approximately 3% so " full power" is $35 to 540 MWe band.

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The MEL is adjusted in accordance with Westinghouse standard methods using Westinghouse equipment and qualified Westinghouse personnel. The KLF relay is adjusted and calibration checked using Westinghouse Information Leaflet 41-748E. Adjustments and calibration checks are performed by a qualified relay technician using appropriately caabrated test equipment. J I

The turbine generator is rated for 580 MVA. The turbine generator continues to operate within its ratings.

Changing the MEL and KLF relay settings allows the generator to operate slightly closer to the steady-state stability limit of the turbine generator. He steady-state stability limit is approximately 650 MW at unity power factor. It is not possible for the turbine generator to approach this power level. Due to SG, governor and control valve efficiencies, the upper power level limit is approximately 580 MW. As generator excitation is reduced, power factor becomes more lagging. As excitation is reduced, the MEL eventually functions to not  !

allow further reductions in field excitation. The probabilities of an operator lowering excitation or the voltage i regulator failing in the lower direction are not affected. MEL and KLF settir.gs are coordinated so the MEL functions to limit reductions in field excitation prior to KLF relay actuation.

Therefore, the probability of occurrence of an accident, event or malfunction previously evaluated in the CLB is not increased. The MEL and KLF relay setting changes do not affect their probability of malfunctioning and do not affect structures, systems or components necessary to limit radiological consequences of analyzed accidents and events. The radiological consequences of an accident, event or malfunction of equipment important to safety previously evaluated Li the CLB are not increased. De accidents or events evaluated for MEL and KI. F malfunctions are the loss of electrical load, loss of AC power and turbine generator overspeed.

Since the MEL and KLF relay settings only involve turbine generator loading limitations, no other events are considered to be ensdible for malfunctions of the MEL or KLF relay. The procedure does not pose a USQ nor does it require a change to the TS. (SE 98-086) 9

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J MODIFICATIONS .l The following modifications were implemented in 1998: i MR 88-178+A, (Unit 2), Containment Purge Supply and Exhaust (VNPSE). i 1.

MR 88-178*A provides a more reliable control system while supplementing the existing instmment air (IA) l accumulator supply with a safety-related nitrogen backup system. ,

Summary of Safety Evaluation: ne new system meets separation and seismic criteria for each containment I penetratior.. De nitrogen supply is connected to the existing IA supply with a pressure regulating valve that

• i opens when the downstream lA system pressure drops below the regulator setting. De nitrogen supply [

exceeds the current IA accumulator volume and provides features for nitrogen bottle replacement without loss -[

of pneumatic supply. The solenoid exhaust port is sized to provide an adequate vent path for the actuator to '

ensure valve closure within the 60-second FSAR limit. The revised control system relies upon a normally shut ,

l solenoid valve to control the opening and shutting of the valve disk as well as controlling the pneumatic power ' i to a slaved switching valve. De switching valve operates to control the pressurization of the T-ring seal. De  !

I new design allows the T-ring to deflate prior to the valve opening and ir.flates after the valve has fully shut.

The purge supply and exhaust valves, are locked shut during normal unit operation. Herefore, the valves cannot cause or contribute to accident scenarios during operation. De valves are opened at the beginning of t cach refueling outage to purge containment for acccas and to provide heating and ventilation for the containment. When the valves are open, automatic or meaual containment isolation signals shut both the  !

- internal and extemol CIVs. During refueling operations, these valves shut to isolate containment, thereby mitigating the consequences of a refueling accident in containment. The final medification configuration j improves the opening and shutting features of the external valves.

. The modification is installed during cold shutdown, with no fuel handling activities' performed in containment. 'l The affected CIVs are blocked open to provide containment heating and ventilation. De external CIVs are [

mechanically blocked open because the valve components fail shut when y.. Or is isolated. Testing provides for stroking each valve shut and then open to verify proper operation of al w solenoid valve and slaved [

switching valve for each CIV. Functional testing, including verification of de integrity of the system, is . i performed to verify proper operation of the installed pneumatic supply system. De modification does not pose j l

USQ nor does it require a change to the TS. (SE 98 130)-

' 2. MR 91-134 and MR 91 135,(Unit I and 2), Main Steam.

. This SE supersedes SEs94-034, SE 95-101, and SE 96-013, As a result of design deficiencies and maintenance difficulties, MRs91-134 and 91-135 were initiated to repbce atmospheric steam dump valves (MS-2015 and L

MS-2016) on both Units 1 and 2. ~

Summary of Safety Evaluation: The replacement valves have a larger flow coefficient, and are capable of more i flow than the original installation. FSAR Section 10.1.2 is updated to include a description, including .{

minimum component quirements. Along with the new valve installations, %" test connections are installed. .

l Upon completion ofinstallation, PMT is performed on the new pressure retaining components, in accordance l with ASME B31.1. De valves are stroke tested. j i t f

I

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a Accident and equipment malfunction probability (e.g. MSLB, va' **iled open/ uncontrolled plant cool down) are not changed. The replacement components and installation pu .ses are in accordance with Section XI requirements to ensure the final configuration is superior or equivalent to original construction codes and standards. The new valves are designed to operate with the stem in the horizontal position. As a result, stem sticking problems (caused by actuator deadweight bending) no longer exist. He trim is less likely to wear or bind as a result of the balanced plug design.

Accident consequences are not changed as a result of this modification. Valve work is performed during cold or refueling shutdown. Radiol gical consequences of the SGTR accident analysis are estimated by mass-energy balances and required core decay heat removal for the duration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> event. liigher

• instantaneous flow rates do not change SGTR accident consequences as calculated by mass-energy balances.

Additionally, radiological consequences from a SGTR or MSLB (valve stuck open) are bounded by existing MSLB analysis, assuming flow from a larger pipe section or stuck open passive safety relief valve.

ne modifications do not change valve availability or performance as described in the Basis of the TS. The modifications do not pose a USQ nor do they require a change to the TS. (SE 98-175) ,

1

3. MR 92 120, (Common), ISFSI.

The SE evaluates a change to the MSB sleeve thickness from 0.20" wall thickness to 0.1875." He calculation used to determine criticality,"VSC-24 Criticality Safety Analysis (Palisades ISFSI)," assumed the MSB sleeve material is 0.200" thick and was not reevaluated for the use of 3/16" sleeve material Credit for burnup of fuel is not taken and because the loading of bumed fuel assernblies in unborated water is not an accident that was part of the designer safety analysis report (SAR) submittal to the NRC. He change from 0.200" to 3/16" material removed mawrial from the intemal basket area, thus allowing additional moderator to take the place of the material. His removal of material has the potential to increase Keffin the MSB.

13 CFR 72.48 Evaluation Summary: The change involves using a thinner material for the storage sleeves for

)

the fuel assemblies in the MSB than analyzed in the ISFSI SAR. He effects were previously analyzed in '

SE 95-079 and SE 95-079-01, but a recent reexamination of only the criticality aspects of this change (looking l at the criticality issue during the loading process when the loaded VSC is flooded with water) and exammmg j the change while assuming unborated water vice borated water is used, prompted this reevaluation in  !

accordance with 10 CFR 72.48. I h examination concluded that the effect of using unborated water in the criticality examination of the change creates a higher Keff., as a measure of the reactivity existing in the MSB during the dynamic loading configuration than previously calculated. The examination also looks at the approach to the licensing basis-stipulated margin to criticality of keeping Keff <0.95, and concludes for the VSCs/MSBs affected by the change, in consideration of the characteristics of the fuel loaded or planned to be loaded, that the margin of i safety represented by a Keff, of <0.95 being maintained. He modification does not pose a USQ nor does it l require a change to the TS. (SE 98-140) '

Page 29 of 112

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4. MR 92 120, (Common), ISFSt.

The SE revision addresses changes made to dry cask loading and unloading procedures. Unique lifting sequences occur during the loading and unloading processes when the loaded MSB is to be lowered or raised between the ventilated concrete cask (VCC) and the MSB transfer cask (MTC). nese lifting sequenas are unique because they involve scenarios during the VSC-24 loading and unloading process where a loaded MSB is being transferred into or out of the MTC by a crane that is given an input where the MTC doors are open.

The significance of this is that if the MSB was erroneously raised once it was fully positioned within the MTC (with the MTC doors open), the MSB could lift the MTC ofTof the VCC via the MTC lid, thereby causing a potential radiological exposure.

10 CFR 72.48 Evaluation Summary: The SE revision evaluates the potential occupational exposure risks associated with possible inadvenent !ifts during the loading and unloading processes. Additional potential for synificant increases in occupational exposure is afforded by the possibility ofinadvenent lifts during the -

loading and unloading processes. To provide reasonable assurance that such an increase does not occur, procaiural controls are implemented that involve use of critical lift speed for operation of the crane during the: e lifts; stationing of an operator at the crane main d;sconnect with radio communication to an individual closely observing the lift, and radiation protection (RP) hold points that include warnings about the hazard and a requirement that non-essential personnel be evacuated from the area. However unlikely, the potential for an inadvertent lift / movement cannot be eliminated. The measures implemented provide reasonable assurance that a significant increase in occupational exposure does not occur. The change does not pose a USQ nor does it require a change to the TS (SE 94-041-05)

Summarv of Safety Evaluation: Ti e SE revision incorporates the addition of RP 8, Part 3 to the scope of procedures potentially affected by the inadvertent lift scenario. It also incorporates changes in draindown time limit methodology as referenced in SE 96-043-03. He modification does not pose a USQ nor does it require a change to the TS.

10 CFR 72.48 Evaluation Summary: He SE revision adds RP 8, Pan 3 to the scope ofloading and unloading procedures potentially affected by inadvenent lift scenarios. The potential for inadvenent lifts during loading and unloading evolur5ns was evaluated. The evaluation still applies and remains valid to this SE redsion.

This revision also addresses a change in draindown time limit determination by reference to upda'.:d SE 96-043-03 that provides complete discussion and evaluation of this change in CSU l.2.10. The modification does not pose a USQ nor does it require a change to the TS. (SE 94-041-06)

5. MR 92-120, (Common), ISFSt.

The SE revision represents communications among Sierra Nuclear Corporation (SNC), the three General i

Licensees for the VSC44 Dry Cask Storage System, and the NRC, regarding the issue of C of C, CSU l.2.10,

" Time Limit for Draining the MSB." The methodology established is conservative with respect to the existing and proposed CSU l.2.10, and is consistent with the methodology already approved by the NRC for the Palisades Nuclear Plant.

10 CFR 72.48 Evaluation Summary: he change implements a more conservative, interim administrative limit. .

embodied in a new formula, that is used until CSU l.2.10 is revised by the NRC. SNC submitted a proposed '

revision to this CSU after consultation with the three General Licensees for the VSC-24 Dry Cask Storage System. The new administrative limit is based on in-situ temperature measurements of the MSB internal water and changes in temperature overtime, as bounded by conservative theoretical heatup rates, rather than an assumed initial temperature and theoretical heatup rate in the past CSU.

I ne new formula is conservative with respect to the past CSU l.2.10 formula for calculating the drain down time limit, and no new occupational exposure or environmental impact concerns are instituted by the new formula. The change does not pose a USQ nor does it require a change to the TS. (SE 96-043-02) 1 Page 30 of 112 l

6. MR 92-120, (Common), ISFSt.

Engineering Change Report (ECR) 98-0121 allows removal of material from the yoke hooks and beams, as required, in the form of 2.6" outside diameter cores that leave a 3" diameter hole where removed. He holes are specified in locations that provide for minimal affects on the strength of the yoke. Calculation 95-0M2, Revision 3, shows that up to one core from each of the yoke hooks and up to two cores from one of the yoke beams may be removed without affecting yoke compliance with the factor of safety requirements of NUREG-0612 and ANSI N14.6 as they relate to the design of special lifting devices.

Summarv of Safety Evaluation: The MTC and its associated lifting yoke are special lifting devices as defined

• by ANSI N14.6. ANSI N14.6 applies to the design, fabrication, testing, maintenance, and quality assurance of these devices which are scoped by the standard to lift containers of radioactive materials weighing 10,000 lbs.

or more. The ASTM A514 materials for the yoke hooks (4" thickness; heat C8086) and beams (2" thickness, heat C7648) were not impact tested during original fabrication to show a notch toughness of 15 ft-!b at 0*F as required by ANSI N14.6 Section 4.2.6. He use of safe 7 factors of 6 to yield and 10 to ultimate, with an applied dynamic factor of 10%, serves to render the yke single failure proof from a NUREG-0612 standpoint.

Removal of this relatively small amount of material (the cores represent less than 1/2 of 1% of the weight of the yoke) has no affect on the ability of the yoke to perform its intended function and ultimately serves to provide the yoke with full compliance to the requirements of ANSI N 14.6. Additionally, the yoke is load tested to 300% of rated capacity upon completion of the change in accordance with ANSI N14.6 Section 6.3.2 that requires acceptance testing be repeated upon completion of" .. replacement or removal cf significant quantities of metal.. " The modification does not pose a USQ nor does it require a change to the TS.

10 CFR 72.48 Evaluation Summary: The change allows the yoke to safely perform its intended function relative to the VSC-24 system. De change does not affect the ability of the yoke to perform its function of carrying a heavy load (MTC/MSB) into, out of or around the SFP. ne yokes compliance with the requirements of NUREG-0612 and ANSI N14.6 render it as single-failure proof. nis compliance is not compromised by this change. No USQ, significant increases in occupational exposure, significant unreviewed environmental impact, nor changes in the license conditions of the C of C are created. (SE 98-085)

7. MR 92-130 ar.d 92-131,(Unit I and 2), Fuel Manipulator.

MR 92-130 replaces the Z-16 fuel manipulator control system. The new system is more reliable, casier to maintain and includes several operational enhancements including variable frequency drive, programmable logic controller (T1545 PLC), an upgrade to the mast rotation device, interlocks, separate jog control switches, a new weight monitoring system with digital readout, replacement of the bridge position indication system, a digital readout of the mast venical position, and a manual override device.

Summary of Safety Evaluation: Installation is accomplished when the unit is shut down. Fuel handling accidents inside containment are described in FSAR Section 14.2.1 with funher analysis contained in a WE letter to the NRC dated March 9,1977, regarding a fuel handling accident in containment, and NRC SER dated June 20,1979. Dese accidents ar prevented through procedural adherence, supervisory oversight and safety interlocks designed into the fuel handling equipment. Fuel manipulator malfunctions and accidents that may result from them are bounded by the analysis in NRC SER dated June 20,1979, that conservatively assumes the fuel pins in two fuel assemblies are ruptured as a result of a dropped fuel assembly. Therefore, no new accident or malfunction is created by the change. The manipulator is not credited for mitigating a fuel handling accident and the amount of fuel damage that would occur from a dropped assembly is not afrected by this modification.

Therefore, the consequences of an accident, event, or malfunction of equipment is not increased by the change.

The seismic capabilities of the manipulator are maintained. The modifications do not pose a USQ nor does it require a change to the TS. (SEs98-004 and 98-154)

Page 31 of 112

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pmmary of the Safety Evaluation: The SE revision addresses installation while Unit 1 is shut down after the l core is refueled. Foreign material exclusion (FME) precautions are administered during installation and testing of the modification. (SE 98-004-01)

Summary of Safety Evaluation: He SE revision changes plant conditions. Insta!!ation begins during UIR24 with disconnection and/or removal of the components intended to be upgraded / replaced. If time permits, termination and/or installation of the components intended to be upgraded are completed during UlR24; otherwise, installation portion takes place in UIR25. He motor control center and control console are deenergized and the wires are disconnected for the motor control center. (SE 98-004-02)

8. MR 93-025'B,(Unit 1), IC-04 Main Control Board - CVCS. .

The SE evaluates the acceptability of rerouting and application of Siltemp fire retardant material, as appropriate, to control wires for the Unit I chemical and volume control system (CVCS). Additionally, some W-2 control switches are replaced. The installation reroutes Train A and/or B wires; installs Siltemp fire retardant material on Train A wires, and replaces the W-2 control switches. The new W-2 switches are like-for-like replacements. Installation occurs during UlR24 or when Unit I is in cold shutdown.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It assumes that train separation is appropriately implemented. He work plans identify post-maintenance testing required to assure each component is retumed to service with the ability to perform its intended safety functions. Installation of Siltemp may include disconnecting wires and sleeving, or wrapping without disconnecting selected wires.

The plant condition and work plan controls during installation and the post-installation testing verifying operability assure that the probability of occurrence of an accident, event or malfunction of equipment important to safety is not increased. For the same reasons, the radiological consequences of an accident, event or malfunction of equipment important to safety and is not increased. The margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Herefore, no possibility of an accident, event or malfunction of a different type is not created. The modification does not involve a USQ nor require a TS change. (SE 98-017)

9. MR 93-025'B,(Unit 1), C-01 Main Control Board - Instrument Air.

The SE evaluates the acceptability of the rerouting control wires ofinstrument air ystem header UIC inlet control valve IIA-3047 control switch. The installation consists of rerouting Train A wires associated with IIA 3047. Installation occurs during UlR24 or when Unit 1 is in cold shutdown. Dit 2 may be in any operating mode. ,

Summary of Safety Evaluation: The final configuration does not change components nor their functiors. It assures that the required train scparation has been appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions.

The plant condition and work plan controls during installation and the post-installation testing verifying operability assure that the probability of occurrence of an accident, event or malfunction of equipment important to safety is not increased. For the same reasons, the radiological consequt nces of an accident, event or malfunction of equipment important to safety are not increased. The margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event, or malfunction of a different type is not created. The modification does not involve a USQ nor require a TS change. (SE 98-018) l Page 32 of 112 l

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10. MR 93-025'B,(Unit 1), C-01 Main Control Board-Service Water.  ;

' The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant materials to control wires of the service water system. Additionally, the W 2 control rwitches are replaced on P-32A, P-328 and l

P 32F SW pumps. l t

Summary of Safety Evaluation: Installation of Siltemp may include disconnecting wires and sleeving, or wrapping without disconnecting wires. The final configuration does not change the components nor their '

design functions. It assures train separation is appropriately implemented. De work plans identify j post-maintenance testing required to assure that each component is returned to service with the ability to  !

. perform its intended safety functions.  !

i

. The plant condition and work plan consols during installation and the post-installation testing verifying i operability assure that the probability of occurrence of an accident, event or malfunction of equipme it -  !

important to safety is not increased For the same reasons, the radiological consequences of an accident, event j or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS i

is not reduced. He wont does not introduce new failure mechanisms. Herefore, the possibility of an accident, l

- event or malfunction of a different type is not created. De modification does not pose a USQ nor does it  ;

require a change to the TS. (SE 98-019)

I1, MR 93-025'B,(Unit 1),1C-03 Main Control Board -Component Cooling.

The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant material to control ,

wires of the Unit I CCW system. Additionally, the W-2 control switch is replaced on the P-11 A&B CCW j pumps. Installation of the Siltemp material may include disconnecting wires and sleeving, or wrapping without  !

disconnecting wires. t Summary of Safety Evaluation: Installation occurs during UIR24 or when Unit I is in a cold shutdown condition.

i The plant conditions and work plan controls verify component operability and ensure that the probability of f occurrence of an accident, event or malfunction of an equipment impoitant to safety are not increas.:d. For these same reasons, the radiological consequences of an accident, event or malfunction of equipment important

- to safety is not increased and the margin of safety as defined in the TS is not reduced. The work does not

' introduce new failure mechanisms. Therefore, the possibility of an accident,' event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-022) j

12. : MR 93-025'B,(Unit 1), C-02 Main Control Board-4160 V ac.

The SE evaluates the acceptab'ility of the rerouting and application of Siltemp fire retardant material to control

+

wires of various devices en the Unit 14160 V ac system. <

l Summary of Safety Evaluation: The final configuration does not change components nor their functions. I' asssures that train separation is appropriately implemented. De work plans identify the post-maintenance I testing required to assure that each component is returned to service with the ability to perform its intended safety function.

The plant condition and work plan controls verify operability and assure that the probability of occurrence of an accident, event or malfunction of equipment important to safety is not increased. For the same reasons, the radiological consequences of an accident, event, or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new

. failure mechanisms. Therefore, the possibility of an accident, event, or malfunction of a different type is not created.; ne modification does not pose a USQ nor does it require a change to the TS. (SE 98-024) l Page 33 of112 i

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,,. . - - - , - nr -- -~ ~ ~ ^

' ' ' ~ '

13. MR 93-025'B,(Unit 1), IC-03 Main Control Board-Condensate and Feedwater.

His SE evaluates the acceptability of the rerouting, rewi'ing and application of Siltemp fire retardant material to control wires of various devices of the Unit I condensate and feedwater systems. He instalkion consists of rerouting Train A and/or B wires and rewiring and installation of Sikemp fire retardant material on Train A wires. Siltemp matetial may be applied eithen by disconnecting and sleeving or by wrapping it around control wires without disconnecting them. Installation occurs during UlR24 or when Unit I is in cold or hot shutdown conditions. Unit 2 may be in any operating mode. i Summary of Safety Evaluation: ne final configuration does not change components nor their functions while ensuring that train separation is appropriately implemented. He work plans identify the post-maintenance ~

testing required to assure that each component is returned to service with the ability to perform its intended safety functions. He radiological consequences of an accident, event or malfunction of equipment important to  ;

I safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-036)

14. MR 93-025*B,(Unit 1), C-01 Main Control Board - Radiation Monitoring.

7 The SE evaluates the acceptability of application of Siltemp fire retardant material or color coding to control wires of the Unit I radiation monitoring system. ne installation consists ofinstalling Siltemp fire retardant material or color coding tape on Train A wires. Sittemp material may be applied either by disconnecting and s!ceving or by wrapping it around control wires without disconnecting them. Installation occurs during UIR24  !

or with Unit i in a cold shutdown condition, except during fuel motion. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It assures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each componect is returned to service with the ability to perform its intended safety functions. He plant ccnditions and work plan controls during installation and post-installation testing verifying component operability, ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. He radiological consequences of an accident, event or malfunction of equipment important to safety are not increased. The margin of safety as defined in the TS is not reduced. He work does not introduce new failure mechanisms. Therefore, no possibility of an accident, event or malfunction of a different type is created. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-043)

15. MR 93-025'B,(Unit I), C-01, IC-03 and IC-04 Main Control Boards - Reactor Protection.

The SE evaluates the acceptability of rerouting control wires with new designation numbers as appropriate, relocating terminations to risers of the correct train, removing the old wires and replacing them with new wires if necessary, and application of Siltemp fire retardant material or color coding to control wires in the Unit 1 reactor protection system. He installation consists of rerouting Train A and/or B wires with new designation numbers as appropriate, relocating terminations to risers of the correct train and installing Siltemp fire retardant material or color coding ; ape on Train A wires. Siltemp may be applied either by disconnecting and sleeving or -

by wrapping it around control wires without disconnecting them. Installation occurs during UIR24 or with '

Unit 1 in cold shutdown. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: ne final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component h returned to service with the ability to perform its intended safety functions. He plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an eccident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in Page 34 of112 l

^he TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an I

accident, event or malfunction of a difTerent type is not created.

l l

The modification does not pose a USQ nor does it require a change to the TS. (SE 98-044)

16. MR 93-025*B,(Unit 1), IC-04 Main Control Board - Nuclear instrumentation.

1 The SE evaluates the acceptability of the rerouting, color coding and application of Siltemp fire retardant l material to control wires in the Unit I nuclear instrumentation system. He installation consists of rerouting Train A and/or B wires, color coding and installation of Siltemp fire retardant material on Train A wires. l

• Sittemp may be applied either by disconnecting and sleeving or by wrapping Siltemp around control wires  ;

without disconnecting them. Work plans require the associated circuits to be deenergized prior to installations. I Summary of Safety Evaluation: The final configuration does not change components nor their functions. It 1 ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is retumed to service with the ability to perform its intended  !

safety functions. The plant conditions and work plan controls during ir stallation and post-installation testmg  ;

verifying component operability ensure that the probability of an occurrence of an accident or event or '

malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Herefore, the possibility of an '

accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-049)

I

17. MR 93-025'B, (Unit 1), C-01 Main Control Board - Containment Cooling, ne SE evaluates the acceptability of the rerouting, color coding and application of Siltemp fire retardant material to control wires of the Unit I containment cooling system.

Summary of Safety Evaluatior}: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. ne work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-050)

I8. MR 93-025'B,(Unit 1), C-01 Main Control Board - EDG Fuel Oil.

He SE evaluates the acceptability of the rerouting, color coding and application of Siltemp fire retardant material to control wires for the G-01/G-02 EDGs fuel oil system.

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Page 35 of 112

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Summary of Safety Evaluation: The final configuration does not change componcats nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-051)

~

19. MR 93-025'B,(Unit I), IC-04 Main Control Board - Containment Purge Supply and Exhaust.

The SE evaluates the acceptability of the rerouting and replacing existing cables with new cables, color coding and application of Siltemp fire retardant material to control wires for the Unit I containment purge supply and exhaust system.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment imponant to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a difTerent type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-055)

20. MR 93-025*B,(Unit 1), C-01 Main Control Board - Waste Liquid.

This SE evaluates the acceptability of the rerouting, color coding and appucation of Sil!emp fire retardant n.aterial to control wires of the Unit I waste liquid system.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is retumed to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. 'Ihe modification does not pose a USQ nor does it require a change to the TS. (SE 98-056) i l

Page 36 of I12 l l

21. MR 93-025'B and E,(Unit 1), C-01 and C-02 Main Control Boards-480 V ac.

The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant material to control wires of the Unit 1480 V ac system. The installation consists of modifications for Train A and/or B wires and installation of Siltemp fire retardant material on applicable Train A wires. Sillemp may be applied either by disconnecting and sleeving or by wrapping it around control wires without disconnecting them. Installation occurs during UlR24 or with Unit I in cold shutdown. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance

~ testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-02_lj

22. MR 93-025'B, and E, (Unit 1), C-01, IC-03 and IC-04 Main Control Boards - Reactor Coolant.

The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant material to control wires for the Unit I reactor coolant system. The installation consists of modifications for Train A and/or B wires and installation of Siltemp fire retardant material on applicable Train A wires. Siltemp may be applied either by disconnecting and sleeving or by wrapping it around control wires without disconnecting them.

Installation occurs during U1R24 or with Unit 1 in cold shutdown. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance  ;

testing required to assure that each component is retumed to service with the ability to perform its intended  !

safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. De radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new faiLre mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-023)

23. MR 93-025'B, and E,(Unit 1), C-01, C-02, IC-03 and IC 04 Main Control Boards - 125 V dc.

I The SE evaluates the acceptability of rerouting, relocating terminations to risers of the correct train, monthly operating breaker (MOB) replacement, and application of Siltemp fire retardant material to control wires for the Unit 1 125 vde system. The installation consists ofrerouting Train A and/or B wires, relocating terminations to .

risers of the correct train, MOB replacement, and installing of Siltemp fire retardant material on Train A wires.

Siltemp may be applied either by disconnecting and sleeving or wrapping it around control wires without {

disconnecting them. Installation occurs during UIR24 or with Unit 1 in cold shutdown. Unit 2 may be in any l operating mode. '

3 l

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended j

safety functions. He plant conditions and work plan controls during installation and post-installation testing '

verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, Page 37 of112 i a

e-event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-027)

24. MR 93-025'B, and E,(Unit I), IC-03 Main Control Board - RHR.

The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant material to control wires for the Unit I residual heat removal system. He installation consists of rerouting Train A and/or B wires, installing Siltemp fire retardant material on Train A wires and replacement of the W-2 control switches.

Installation occurs during UIR24 or with Unit I in cold shutdown. .

Summary of Safety Evaluation: ne final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. He work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installatisn testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. De radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. De work does not introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-030)

25. MR 93-025*B, E, and F,(Unit I), IC-03 and C-01 Main Control Boards -Main Steam.

The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant material to control wires for the Unit I main steam system. The installation consists of rerouting Train A and/or B wires and installing Siltemp fire retardant material on Train A wires. Installation occurs during UIR24 or with Unit 1 in cold shutdown. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: ne plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident, event or malfunction of equipment important to safety is not increased. He radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the  :

margin of safety as defined in the TS is not reduced. De work does not introduce new failure mechanisms.  !

Derefore, the possibility of an accident, event or malfunction of a different type is not created. De modification does not pose a USQ nor does it require a change to the TS. (SE 98-020)

26. MR 93-025'B, E, and F,(Unit 1), IC-03 and IC-04 Main Control Boards - 120 V ac.

The SE evaluates the acceptability of rerouting of control wires with new designation numbers as appropriate, i relocating terminations to risers of the correct train, MOB replacement, and application of Siltemp fire retardant material or color coding to control wires of the Unit i 120 V ac system.

l Summary of Safety Evaluation: The final configuration does not change components nor their functions. It I I

ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created, ne modification does not pose a USQ nor does it require a change to the TS. (SE 98-034)

Page 38 of i12

m-

27. MR 93-025*B and *F,(Unit I), IC-03 and C-01 Main Control Boards - Auxiliary Feedwater.

The SE evaluates the acceptability of rerouting Train A and/or B wires and installing of Siltemp fire retardant material to contrel wires in the auxiliary feedwater system. The installation consists of rerouting Train A and/or B wires and the installing Siltemp fire retardant material on Train A wires. lastallation occurs during UIR24 or with Unit 1 in cold shutd,wn. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: The final configuration does not change components nor their functions, it ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perfomt its intended safety functions. ne plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. De radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Herefore, the possibility of an ccident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-028)

28. p!R 93-025*B and 'F,(Unit 1) C-01 Main Control Board - SI.

The SE evaluates the acceptability of the rerouting and application of Siltemp fire retardant material, as appropriate, to control wires for the safety injection system. ne installation consists of rerouting Train A and/or B wires, installation of Siltemp fire retardant material on Train A wires and replacement of the W-2 control switches. Installation occurs during UIR24 or with Unit I in cold shutdown. Unit 2 may be in any operating mode.

Summary of Safety Evaluation: ne final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. He work plans identify the post-maintenance testing required to assure thEt each component is returned to service with the ability to perform its intended safety functions. The plant conditions and vork plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. He radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-029)

Summary of Safety Evaluation: The SE revision adds 1-PB/MSI-1,2 to the scope of this design package. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-029-01) 4

29. MR 93-025'B and F,(Unit 1), C-01, IC-03, and IC-04 Main Control Boards - ESF.

The SE evaluates the acceptability of the reing, color coding and application of Siltemp fire retardant material to control wires for the Unit i engineered afety features system. He installation consists of rerouting Train A and/or B wires, color coding and installing Siltemp fire retardant material on Train A wires. Siltemp may be applied either by disconnecting and sleeving or by wrapping it around control wires without disconnecting them Installation occurs during UlR24 or with Unit I in cold or hot shutdown. Unit 2 may be in

' any operating mode.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each compor'ent is returned to service with the ability to perfonn its intended i

safety functions. The plant conditions and work plan controls during installation and post-installation testing Page 39 of 112 i

i

d I

verifying component operability ensure that the probability of an occurrence of an accident er event or malfunction of equipment important to safety is not increased The radiological consequence.s of an accidem l event or malfunction of equipment important to safety are not increased and the margin of safety as defined in  !

the TS is not reduced. The work does not introduce new failure mechanisms. Herefore, the possibility of an  !

accident, event or malfunction of a different type is not created. ne modification does not pose a USQ nor , j does it require a change to the TS. (SE 98-035) i

30. MR 93-025*C, { Unit 2),2C-04 Main Control Board - Containment Purge Supply and Exhaust. r The SE evaluates the acceptability of the modification to control wires of various control switches and relays in i the Unit 2 containment purge supply and exhaust system. Installation consists or rerouting and replacing Train i A and/or B wires, physicallyseparating Train B wires from Train A wims by resupporting them, installing .

color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires. .l6 Summary of Safety Evaluation: De final configuration does not change components nor their functions. It -

ensures that train separation is appropriately implemented. He work plans identify the post-maintenance .

testing required to assure that each component is returned to service with the ability to perfonn its intended - .

safety functions. He plant conditions and work plan controls during installation and post-installation testing  ;

verifying component operability ensure that the probability of an occurrence of an accident or event or  ;

malfunction of equipment important to safety is not increased. De radiological consequences of an accident,  !

event or malfunction of equipment important to safety are not increased and the margin of scfety as defined in .

the TS is not reduced. He work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. De modification does not pose a USQ nor does it require a change to the TS. (SE 98-135)

31. MR 93-025*C,(Unit 2),2C-03 Main Control Board- Auxiliary Feedwater.

ne SE evaluates the acceptability of modification to control wires for control switches and relays in the Unit 2  ;

?

auxiliary feedwater system, installation consists or rerouting and replacing Train A and/or B wires, physically separating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, j and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It  ;

ensures that train separation is appropriately implemented. De work plans identify the post-maintenance ['

testing required to assure that each component is returned to service with the ability to perform its intended safety functions. De plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or l malfunction of equipment impoitant to safety is not increased De radiological consequences of an accident,  ;

I event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. De work does not introduce new failure mechanisms. Herefoir, the possibility of an j accident, event or malfunction of a different type is not created. He modification does not pose a USQ nor  :

does it require a change to the TS. (SE 98-144) l

32. MR 93-025'C,(Unit 2),2C-03 and C-01 Main Control Boards - Main Steam. -

l t

The SE evaluates the acceptability of the modification to control wires of control switches, pushbuttons, indicating lights and relays in the Unit 2 main steam system. Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resuppoiting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires. .

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. De work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. He plant conditions and work plan controls during installation and post-installation testing Page 40 of112

verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. He work does not introduce new failure mechanisms. Therefore, the possibility of an l

accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-145)

Summary of Safety Evaluation: ne SE revision addresses changes to applicable work orders (WOs) performed in accordance with this SE. The SE attachment describing the WOs deleted WOs 9811769, 9811770,9811772,9811773,9811778,9811779,9811780,9811781, and transferred WOs 9811800,9811849,

- and 9811868 to other MCB wire separation modification design packages. The change does not pose a USQ nor does it require a change to the TS. (SE 98-145-01)

33. MR 93-025'C,(Unit 2),2C-03 Main Control Board - ESF.

I 1

he SE evaluates the acceptability of the modification to control wires for control switches, pushbuttons and  !

relays for the Unit 2 engineered safeguards feature system. Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance  !

testing required to assure that each component is returned to service with the ability to perform its intended safety functions. De plant conditions and work plan controls during installation and post-installation testing i verifying component operability ensure that the probability of an occurrena of an accident or event or malfunction of equipment important to safety is not increased. De radiciogical consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. TFe work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-147)

Summary of Safety Evaluation: n- SE r'evision addresses changes with WOs performed in accordance with this SE. The SE attachment describing the WOs deleted WOs 9811764,9811766,9811848,9812006, and added WOs 9811793,9811820,9811868, and 9817328 to this modification design package.

1 l

The change does not pose a USQ nor does it require a change to the TS. (SE 98-147-01)

34. MR 93-025'C,(Unit 2),2C-03 Main Control Board - RHR.

The SE evaluates the acceptability of the modification to control wires control switches and relays in Unit 2 residual heat removal system. Installation consists or rerouting and replacing Train A and/or B wires, physically separe. ting Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train 8 wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. h ensures that train separation is appropriately implemented. The work plans identify the post-maintenance

)

testing required to assure that each component is returned to service with the ability to perform its intended

{'

safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. De radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS (SE 98-148)

Page 41 of i12

l Sunimary of Safety Evaluation: The SE revision addresses changes to applicable WOs performed in accordance with this SE. The SE attachment describing the WOs deleted WOs 9811782 and 9811789 since the WOs were evaluated under a separate evaluation. He change does not pose' a USQ nor does it require a change to the TS. (SE 98148-01)

35. MR 93-025*C, (Unit 2), C-01 Main Control Board - Waste Liquid.

De SE evaluates the acceptability of the modification to control wires for control switches and reisys in the Unit 2-related waste liquid system installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B w ' ires from Train A wires by resupporting them, installing color-coded tape on the ~

wires, and installing Siltemp fire retardant material on Train B wires.

1 Summary of Safety Evaluation: ne final configuration does not change components nor their functions. It 1 ensures that train separation is appropriately implemented. De work plans identify the post-maintenance 1

testing required to assure that each component is returned to service with the ability to perform its intended  :

safety functions. De plant conditions and work plan controls during installation and post-installation testing  ;

i verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. De radiologica! consequences of an accident, j event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced, ne work does not introduce new failure mechanisms. Therefore, the possibility of an  ;

accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-152)

36. MR 93-025*C,(Unit 2),2C-03 Main Control Board -Containment Spray. t Summary of Safety Evaluation: De SE evaluates the acceptability of modification to control wires for control l switches and relays in the Unit 2 containment spray system. Installation consists or rerouting and replacing l Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires. l I

Summary of Safety Evaluation: De final configuration does not' change components nor their functions. It ensures that train separation is appropriately implemented, ne work plans identify the post. maintenance testing required to assure that each component is returned to service with the ability to perfonn its intended safety functions. ne plant conditions and work plan controls during installation and post-installation testing  ;

verifying component operability ensure that the probability of an occurrence of an accident or event or ,

malfunction of equipment important to safety is not increased. The radiological consequences of an accident, l event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. De work does not introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. ne modification does not pose a USQ nor does it require a change to the TS. (SE 98-153)  ;

37. MR 93-025*C,(Common), C-01 Main Control Board -Instrument Air.

. i The SE evaluates the acceptability of the modification to control wires for pushbuttons in the Unit 2-related instrument air system. Installation consists or rerouting ar.d replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the  !

wires, and installing Siltemp fire retard nt material on Train B wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended i safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or l Page 42 of 112

malfunction of equipment important to safety is not increased. The radiologica! consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. De modification does not pose a USQ nor does it require a change to the TS. (SE 98-155) 1

38. MR 93-025*C,(Unit 2),2C-03 Main Control Board- Component Cooling.  ;

- The SE evaluates the acceptability of the modification to control wires for control switches and relay in the Unit

'2 CCW system. Installation consists or rerouting and replacing Train A and/or B wires physicallyseparating

• Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing ,

Siltemp fire retardant material on Train B wires. ,

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. De work plans identify the post-maintenance  ;

testing required to assure that each component is returned to service with the ability to perform its intended i

safety functions. The plant conditions and work plan controls during installation and post-installation testing

)

verifying component operability ensure tinat the probability of an occurrence of an accident or event or i

malfunction of equipment important to safety is not increased. The radiological consequences of an accident, j event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an i accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-156)

Summary of Safety Evaluation: The SE revision addresses changes with applicable WOs performed in accordance with this SE. The SE attachment describing the WOs added WO 9811754 to this tnodification design packap. The change does not pose a USQ nor does it require a change to the TS. (SE 98-156-01)

39. pjR 93-025*C,(Unit 2), C-01 Main Control Board -Containment Cooling.

The SE evaluates the acceptability of the modification to control wires for control switches and relays in the Unit 2-related containment cooling system. Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: ne final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment imponant to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it requve a change to the TS.' (SE 98-160)

40. MR 93 025'C,(Unit 2), C-02 Main Control Board- Reactor Protection.

The SE evaluates the acceptability of the modification to control wires for pushbuttons and relays in the Unit 2 related reactor protection system. Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupponing them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Traia B wires.

1 Page 43 of112

l l

l 1

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. He work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. %e radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in  :

the TS is not reduced. The work does not introduce new failure mechanians. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-16_1)

41. MR 93-025*C,(Unit 2), C-02 Main Control Board - EDGs.

ne SE evaluates the acceptability of the modification to control wires for control switches, synchronizing switches, pushbuttons, wattmeters, voltage regulator, ammeter, lights, varmeter, hertzmeter and relays in the -

Unit 2-related emergency diesel generators. Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. He work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in  ;

the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a difTerent type is not created. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-162)

Summary of Safety Evaluation: The SE revision addresses changes made to the SE attachment that describes the WOs for this design package. SE Attachment A canceled WO 9812157 and added WO 9818155. The change does not pose a USQ nor does it require a change to the TS. (SE 98-162-01)

Summary of Safety Evaluation: The SE revision addresses changes made to the SE attachment that describes the WOs for this design package. SE Attachment A was revised to remove the condition from WO 9818155 that G-04 EDG be taken out of service during this work. WO 9811784,9811813,9812150, and 9812151 plant conditions were revised to remove the requirement to take G-03 EDG out of service during this work. The change does not pose a USQ nor does it require a change to the TS. (SE 98-162-02)

42. MR 93-02PC,(Unit 2), C-01 Main Control Board- Radiation Monitoring. -

The SE evaluates the accepability of the modification to control wires for control switches and relays in the Unit 2-related radiation monitoring system. Installation consists or rerouting and replacing Train A and/or B . l wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires. ,

I 1

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separatior i appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended l safety functions. The plant conditions and work plan controls during installation and post-installation testing i verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in Page 44 of 112

the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a difTerent type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-152)

43. MR 93-025'C,(Unit 2), C-02 Main Control Board -4160 V ac.

The SE evaluates the acceptability of the tuodification to control wires for control switches and relays in the Unit 2-related 4160 V ac system Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires. I Summary of Safety Evaluation: De final configuration does not change components nor their functions, it ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is retumed to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. He radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. He work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-165)

Summary of Safety Evaluation: ne SE revision addresses changes made to the SE attachment that describes the WOs for this design package. SE Attachment A was revised to clarify the required plant conditions for WOs 9811739,9812122 and 9812123. Buses 2A-05 and 2A-06 are isolated from their affected circuit during i its work. Therefore, the prior requirement specified in these WOs to remove the buses from service was  !

deleted. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-165-01)  ;

l

44. MR 93-025'C, (Unit 2),C-01 Main Control Board - Reactor Coolant.

The SE evaluates the acceptability of the modification to control wires for control switches, breaker and j selector switch operator in the Unit 2-related reactor coolant system Installation consists or rerouting and  !

replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: ne final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance l

testing required to assure that each component is returned to service with the ability to perform its intended l safety functions. The plant conditions and work plan controls during installation and post-installation testing I verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident,  ;

event or malfunction of equipment important to safety are not increased and the margin of safety as defined in i the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. This modification does not pose a USQ, nor does it require a change to the TS. (SE 98-169) 45, MR 93-025'C and G,(Unit 2) 2C-04 and C-01 Main Control Boards - SI.

The SE evaluates the acceptability of the modification to control wires for control switches and relays of the Unit 2 safety injection system. Instaliation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Page 45 of 112

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. He work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. He radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in l the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an i accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. [SE 98-146)

Summary of Safety Evaluation: The SE revision addresses changes made to the SE attachment that describes  !

the WOs for this design package. SE Attachment A was revised to clarify the WOs performed by this SE. Nine WOs were deleted from this design package. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-146-01)

Summary of Safety Evaluation: The SE revision updates SE Attachment A to clarify the required plant conditions for three WOs specified in this design package. WOs 9811819 and 9811830 were updated to remove the reference to isolating the refueling water storage tank (RWST) and a statement was added that valve SI-878A is shut with its power removed during these WOs. WO 9812090 was clarified by removing reference to Unit 2 RHR being isolated and not required and by adding a statement that valve SI-871 A is shut with its power removed if RHR is in operation during WO 9812090. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-146-02)

46. MR 93-025'C and G,(Unit 2),2C-04 Main Control Board - CVCS.

The SE evaluates the acceptability of the modification described to control wires for control switches and relays in the Unit 2 chemical and volume control system. Installation consists or rerouting and replacing Train A and/or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. This modification does not pose a USQ nor ~

does it require a change to the TS. (SE 98-149)

Summary of Safety Evaluation: The SE revision updates SE Attachment A by adding WO 9811805 to this work package. WO 9811812 also added another device to this design package. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-149-01)

47. MR 93-025'C and G,(Unit 2), C-01 and C-02 Main Control Boards - 480 V ac.

The SE evaluates the acceptability of the modification described to control wires for control switches and relays in the Unit 2-related 480 V ac system. Installation consists or rerouting and replacing Train A and'or B wires, physicallyseparating Train B wires from Train A wires by resupporting them, installing color-coded tape on the wires, and installing Sittemp fire retardant material on Train B wires. I l

Page 46 of 112

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Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. He work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post installation testing  :

verifying component operability ensure that the probability of an occurrence of an accident or event or  ;

malfunction of equipment important to safety is not increased. De radiological consequences of an accident,

)

event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor i does it require a change to the TS. (SE 98-164)

Summary of Safety Evaluation: The SE revision updates SE Attachment A to clarify the required plant condition for WO 9812108. A statement was added to note that B-03 is isolated during this work. De modification does not pose a USQ nor does it require a change to the TS. (SE 98-164-01)

48. MR 93-025'F,(Unit I), IC-03 Main Control Board - Feedwater, j

- The SE evaluates the acceptability of the application of Siltemp fire retardant material by sleeving or wrapping l to instrumentation wires in the Unit I feedwater system with electronic instrumentation.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It

)

ensures that train separation is appropriately implemented. The work plans identify the post-maintenance '

testing required to assure that each component is returned to service with the abihty to perform its intended i safety functions. The plant conditions and work plan controls during installation and post-installation testing  ;

verifying component operability ensure that the probability of an occurrence of an accident or event or l

malfunctio of equipment important to safety is not increased. He radiological consequences of an accident, l event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Therefore, the possibility of an j accident, event or malfunction of a different type is not created. De modification does not pose a USQ, nor  !

does it require a change to the TS. (SE 98-042) '

49. MR 93-025*G,(Unit 2),2C-04 and 2C-03 Main Control Boards- 125 V dc.

I The SE evaluates the acceptability of the modification to control wires for control switches and relays in the Unit 2-related 125 V de system. De installation process consists of bundling rerouting and replacing Train A and/or B wires, MOB replacement, physically separating Train B wires from Train A wires by resupporting them, installing correct color-coded tape on the wires, and installing Siltemp fire retardant material on Train B wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. 'The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment importr.:n is safety is not increased. De radiological consequences of an accident, event or mr.lfunction of equipment ira tant to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does .ot introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-158) i l

1 Page 47 of112  ;

50. MR 93-025 *G, (Unit 2),2C-04 Main Control Board - Nuclear Instrumentation.

The SE evaluates the acceptability of the modification to control wires for control switches and relays in the l Unit 2 nuclear instrumentation system. Installation consists or rerouting and replacing Train A and/or B wires, physically separating Train B wires from Train A wires by resupporting them, installing color-coded tape on I the wires, and installing Siltemp fire retardant material on Train B wires. l i

Summary of Safety Evaluation: ne final configuration does not change components nor their functions, it ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing

• verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. De radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. He work does not introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-159)

51. MR 93-025'G and H,(Unit 2),2C-04,2C-03, and C-01 Main Control Boards - 120 V ac.

This SE evaluates the acceptability of the modification to control wires for breakers and receptacles in the Unit 2120 V ac system. Installation consists of rerouting and replacing installed wires with new terminal designation numbers, physically separating Train B wires from Train A wires by resupporting them, relocating terminations of risers of the correct train / channel, MOB replacement, installing Siltemp fire retardant material, by installing correct color-coded tape on the Train B or Channel wires.

Summary of Safety Evaluation: The final configuration does not change components nor their functions. It ensures that train separation is appropriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. The work does not introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-12_6)

52. MR 93-025'H,(Unit 2),2C-04,2C-03 and C-01 Main Control Boards- Regulatory Guide 1.97 Devices.

The SE evaluates the acceptability of the modification to control wires for control switches and relays in the ~

Unit 2 and common Regulatory Guide 1.97 devices. Installation consists of rerouting and replacing wires, physically separating different channel wires by resupporting them, installing color-coded tape on wires, relabeling and reluging wires if necessary, and an application of Siltemp fire retardant material.

. 1 l

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Summary of Safety Evaluation: The final configuration does not change components nor their functions. It 1

ensures that train separation is appsopriately implemented. The work plans identify the post-maintenance testing required to assure that each component is returned to service with the ability to perform its intended safety functions. The plant conditions and work plan controls during installation and post-installation testing verifying component operability ensure that the probability of an occurrence of an accident or event or malfunction of equipment important to safety is not increased. The radiological consequences of an accident, event or malfunction of equipment important to safety are not increased and the margin of safety as defined in the TS is not reduced. De work does not introduce new failure mechanisms. Herefore, the possibility of an accident, event or malfunction of a different type is not created. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-127)

53. MR 95-004/004*A,(Unit 2), Low Pressure Turbine.

MR 95-004 replaces the LP turbine rotors with new rotors having improved turbine shaft materials and forged monoblock design with integral forged disc wheels.

Summary of Safety Evaluation: MR 95-004 does not affect the operation of the turbine control and protection circuits nor does it affect the operation, operability or function of systems, structures or components identified as important to safety. In addition, with the exception of some potential residual fixed contamination on internal turbine surfaces from previous SG tube leakage, MR 95-004 does not add or relocate radioactive material, does not affect systems or components that contain radioactive materials, or does not affect radiological barriers. During installation, there are no restrictions imposed on Unit I nor does the installation t require degradation of fire barriers or fire suppression equipment. Installation is performed when steam is isolated from the turbine. Following installation, the turbine generator is fully functional. Therefore, there are no interim conditions associated with MR 95-004. The modification does not pose a USQ nor does it require a change to the TS. (SER 98-114)

54. MR 95-005 and MR 95-006,(Unit I and 2), B-03 and B-04 Anchorage Upgrade.

The modifications upgrade the anchorage of the B-03 and B-04 cabinets located in the cable spreading room.

The anchorage upgrade provides a positive connection between the cabinets and the floor slab which is 4 accomplished by installing structural steel angles along the north and south lengths of the pad using Hilti bolts; welding the embedded channel to the angle, and welding the fra ne of each cabinet to a channel embedded in its '

{

respective equipment pad. The configuration satisfies the SQUG criteria.

Summary of Safety Evaluation: Installation implemented on one cabinet at a time. Essentialloads such as SW pumps, instrument air, SFP cooling pumps, CCW pumps, and RHR pumps are swapped to the bus opposite that i being worked on. Excessive dust is not created to avoid inadvertent activation of controls in the cable i

spreading room. During weld installation, the smoke detection system may be taken out of service so smok.e from the welding process does not inadvertently actuate the Halon suppression system. The internal cubicle components are protected with fireproof blankets during welding and the breakers associated with the cubicles

{

are opened during welding to ensure personnel safety. The work does not affect the function of the components within the cubicles Compensatory measures, such as a fire watch, are taken to ensure that no conditions are l

created that are outside of TS requirements. In addition, the welding machine is located outside of the cable {'

rpreading room to mitigate potential electric and magnetic field (EMF) interference. These precautions ensure that interim conditions associated with the modification are not created that could cause, increase the probability of, or increase the radiological consequences of an accident or event.

i Page 49 of112

The installed angles do not directly interact with plant equipment and are intended to provide seismic support for the cabinets. De upgraded anchorage for the cabinet frame to the embedded channel do not interact with other plant equipment. He effect of the seismic upgrade is to increase overall plant safety by ensuring that the B-03 and B-04 cabinets remain intact during and after a seismic event. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-012) l

55. MR 95-026,(Common), Reactor Coolant System (RCS).

The modification replaces the RCS Loop A pad B flow elbow tap isolation root valves (t1C-501 A-C,-502,

-505,-506 A-C) and its deformed downstream tubing. The original 1500 lb class 3/4" globe Rockwell Model A3624J valves are replaced with Anchor / Darling 1878 lb class 3/4" globe vahes. The valves are rated

• for the RCS design temperature and pressure. ,

Summary of Safety Evaluation: Lead shields are installed on the old 3/4" flow elbow tap valves to reduce dose in the area. De shields protect workers from radiological hot spots associated with the valves.

FSAR Table 4.2 1 defines the material requirements for components associated with the RCS. The new valves meet these material requirements and are designed and fabricated to ASME Section til Subsection NB for Class I components. NDE requirements for cast valves include radiography and liquid penetrant examinations. ,

The valves were hydrostatic alley pressure tested to demonstrate their ability to maintain pressure. Accessible surfaces of the new valves were liquid penetrant examined; however, the valves were not radiographed. This is acceptable because the physical properties of the material are used to establish the allowable stresses which i

affect the design basis. The methodology for verifying physical properties (NDE inspection) has no impact on design. The new valves weigh 4 lbs more than the original valves. This increase in weight was evaluated as acceptable by Calculation W-SMT-98-007. He consequences of a failure of components associated with this modification are boended by the small break LOCA discussed in FSAR Section 14.3. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-033)

56. MR 95-041'A,(Unit 1), Safety injection.  ;

The SE revision addresses removal of the lower rings of packing (below the lantern ring) and replacement with a spacer to assist with pressure locking concems for valves ISI-857A&B.

Summary of Safety Evaluation: The modification corrects pressure locking concerns. If operation of the '

I SI-857A&B valve was difficult and the vent lines opened for bonnet pressure relief to get in the sump recirculation configuration, EOP-1.3," Transfer to Containment Sump Recirculation,"is entered. That procedere directs the operator to shut both vent isolations to provide two valve isolation. EOP-l.3 prevents potential human errors that allow for release of radiation to the environment. Therefore, this activity does not ,

increase the radiological consequences of an accident, event or malfunction of equipment important to safety previously evaluated in the CLB. The modification does not pose a USQ nor does it require a change to the TS. *

(SE 97-196-01) l

57. MR 95-041*B,(Unit 1), Reri.fual Heat Removal (RHR).

The modification caps the leakofflines to valves IRH-700, IRH-701 and IRH-720 when the valves are repacked in accordance with MI 32.2," Valve Packing." The maintenance instreetion did not discuss capping the leakoff line after the valve is repacked. The capping prevents liquid (at system pressure) from traveling out the leakoff connection and flowing to the reactor cociant drain tank (RCDT). The line is also capped on the side nearest the RCDT to prevent gases or liquids from traveling back through the tubing and escaping into containment atmosphere.

1 I

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Summary of Safety Evt luation: Repacking the valves and a change in the packing arrangement does not affect the operability of the vs.lves. Valve operability is not affected by the modification. The adequacy of the new packing arrangement ( .g., having only 5 rings of packing vice 10) was evaluated as acceptable. Work is performed with the unit defueled.

Installation of a tubing cap on the leakoffline does not increase the probability of an accident or an event as described in the CLB. The opportunity for leakage is decreased by use of the cap vice the tubing. ne tubing used is 3/8." If a brea ( occurred, the FSAR states that the charging system can provide charging makeup for a 3/8" hole to allow an orderly shut down without activation of emergency core cooling. Since a break of this line would not activat e the emergency core cooling systems, the leakage is not considered a small break LOCA.

The installation of a f abing cap on the leakofTconnection for 1Ril-700, IRil 701 and IRil 720 cannot initiate an accident of a different type. The installation has no effect on the ability of 1Ril-700, IRil-701 or 1Ril-720 to perform their required safety functions of staying shut when required, and opening when the unit needs to go.

on to RiiR. The che nge does not affect operability of IRil-700, IRil-701 and IRil-720 valves. The modification does not pose a USQ nor does it requires a change to the TS. (SE 97-090)

58. MR 95-041*C,(Unit 1), R11R.

Modifications are iemporarily installed on motor-operated valves 1Ril-701 and 1 Ril-720, as well as on the piping between IF11700 and 701. The modifications remain in place until the Unit I reactor is defueled and a -

permanent fix can be installed. The interim changes install bonnet relief valves on the gland leakofflines and on the vent line between IRll 700 and IRIl-701. The valves are repacked to allow bonnet pressure to activate i the reliefs as nece ssary. The setpoints for the relief valves are above the 2485 psi pressurizer safety valve setpoint, but are adequate to protect the piping and valves from overpressurization. De modifications are installed during cold shutdown with fuci in the core.

Summary of Safety Evaluation: The changes eliminate potential damage to valves and piping that could i

increase the probability of a malfunction. The tubing is installed to the correct design Code requirements and i rated to RCS temperature and pressure. He leakoff and vent lines are seismically adequate. The relief valves are tested per ASME Section XI requirements. If a tubingjoint fails or a relief valve fails open, a maximum 3/8" diameter 1:ak is created. He leakage is not considered a small break LOCA as evaluated in the CLB and is within the normal makeup capacity of the unit. The ability to conduct a normal plant shutdown is maintained.

The 1Ril-701 and 1Ril-720 valves are designed with backseats to allow for maintenance. This feature is utilized to allow for packing adjustments without having to isolate the RHR system. The backseats of the valves are expected to be in good condition. There is no history of backseating these valves as a general practice and recent work on 2Ril-700,2Ril-701 and 720 showed the backseats were in good condition. In ,

addition, the valves are worked individually to ensure the ability to isolate the Ri1R system from the RCS is '

maintained. The modification does not pose a USQ nor does it require a change to the TS. (SE 97-146-01)

59. MR 95-042,(Unit 2), Safety injection.

He SE revision addresses CR 974683 that documents the work on Ril-700 was not performed as originally stated.

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Summary of Safety Evaluation: The 2SI-857A&B valves are casier to operate following installation.

Therefore, the unit can be placed in the sump recirculation phase in a more timely fashion should an accident l occur. In the normal configuration, two valve isolation is maintained because 2SI-857A&B are normally shut. l EOP-1.3," Transfer to Containment Sump Recirculation," directs the operator to shut both vent valves to  !

provide two-valve isolation. EOP-1.3 prevents potential human errors that could allow for release of radiation to the environment. The modification does not pose a USQ nor does it require a change to the TS.

(SE 97-002-01)

60. MR 95-042*B, (Unit 2), RHR.

The modification installs a venting capability on valve bonnets 2RH-704 A&B that allows a manual pressure '

relief of the valve bonnet before attempting valve opening. The new routings, as well as their interfaces, are designed, fabricated and tested to the same standards and criteria as the original systems.

Summary of Safety Evaluation: The RHR vent valves are typically shut. The shut vent valves render the parent valve unchanged regarding its sealing capability in both direction. When the valve is required to be open, the operator first opens its vent valve located in an area normally accessible with varying radiation doses.

The overall operator radiation exposure is reduced by the modification. Since the parent valves remain shut and need not be manipulated during accident conditions, they are not provided with reach rods.

The modification is installed in Unit 2 Pipeway #3 with the unit in cold shutdown and the reactor defueled. The valves cannot be isolated to allow simultaneous work and RHR system availability during other plant conditions. Since the work is performed on systems that handle reactor coolant, precautions that are associated with such actions, including exclusion of foreign materials from the system, are strictly enforced. A pressure test and an initial service leak test are performed as acceptance of the modification. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-117)

61. MR 95-044,(Unit 2), Refueling Cavity.

The modification removes the refueling cavity drain barrel, the associated piping to the barrel, including the 4" cross fitting and drop leg, the flapper valve, and the reducing diffuser in the cavity drain. The drain barrel and reducing diffuser are removed because of their effectiveness in reducing radiation dose in the area. The barrel has turned into a high dose contributor. The flapper valve and holding mechanism are removed to facilitate the installation of the debris strainer.

Summary of Safety Evaluation: The modification upgrades a postion of the refueling cavity drain line, including the first isolation valve, to safety-related status and Seismit 3 wass I piping. The drain pipe can then be relied upon to pass / retain water from, or in, the lower refueling cat ity. The modification ensures that the containment spray /RCS fluid that enters the refueling cavity during a large break LOCA drains and is made available to the containment sump. This leaves a minimal holdup volume in the cavity.

The FSAR is revised to the drain barrel discussion and the changes to the cavity holdup volume. The cavity drain system ensures that the inventory of water that is directed to the lower refueling cavity during a large break LOCA is available or returned to the containment sump. This modification meets commitments made in -

TS Change Request 204 regarding Unit 2 sump levels. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-166)

62. MR 95-048,(Unit I),4160 V. ,

l MR 95-048 adds a time delay relay circuit to the Unit 14160 V loss of voltage relay circuit. Time delay relays are added to the 4160 V loss of voltage relays input to the EDG output breaker closing circuit for 4160 V breakers I A52-60,66,80 and 86. These new relays ensure that a time delay occurs between when the 4160 V loss of voltage relays pick up, and when the EDG output circuit breaker closes.

Page 52 of 112

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Summary of Safetv Evaluation: The EDGs supply emergency power to safeguards loads by reenergizing the respective 4160 V bus when a loss of vcitage from the normal supply occurs. A time delay is installed in the  !

EDG output breaker close circuits for i A52-60,66, 80 and 86, that is in effect during a loss of voltage on either 4160 V bus I A-05 or i A 06. However, the time delay is short enough not to have an effect on the closing of the EDG output breaker, except when the EDG is in a standby (running) condition and the loss of voltage occurs from the normal supply. During this condition, the permissive to close the EDG breaker is a function of 1

the new time delay scheme. This time delay is greater than the time between when the 4160 V loss of voltage  !

. relays pick up and when the 480 V loss of voltage relays pickup. However, the time is less than the accident analysis assumed time to start the EDGs (10 seconds). His enhances the availability of emergency power to '

power safeguard loads during an accident or event.

Unit 1 is in cold or refueling shutdown or defueled during installation and testing of this modification. During installation and testing, the 4160 V undervoltage circuitry and respective EDG breaker control circuits (for each i

' bus 1 A-05 and 1 A-06) are removed from service (one train at a time). Applicable TS LCOs for loss of emergency power are entered. One train of power is worked and tested until restoration prior to performing i

. work on the opposite train. A loss of emergency power to an entire AC power train is assumed in the accident 3

analysis (when it is the most limiting accident condition). No components are disabled beyond those currently ,

analyzed in the accident analyses and no equipment is operated beyond the acceptance criteria described in the l

TS. Therefore, installation and testing of this modification does not increase the probability of an accident or l

event, create an accident or event of a different type, increase the radiological consequences of an accident or event, increase the probability of a malfunction of equipment impodant to safety, create the possibility of a I malfunction of equipment important to safety of a different type, nor decrease the margin of safety defined in 'I the TS Basis. MR 95-048 does not pose a USQ nor does it require a change to the TS. (SE 98-057)

Summary of Safety Evaluation: The SE revision deletes the reference to PB-546, " Specification for Electrical Installation," because that document is used when outside contractors are hired to perform work. In this case,  !

PBNP Maintenance is performing installation activities. Therefore, work is covered under PBNP maintenance  ;

and administrative procedures. The change does not pose a USQ nor does it require a change to the TS.

(SE 98 057-0I) ['

Sum nary of Safety Evaluation: De SE revision changes the frequency for periodic testing of the newly installed relays from monthly to each refueling interval (not to exceed 18 months). It also removes reference to I

- IRMP 9071 1 and 2 that perform the monthly testing of the I A-05 and I A-06 undervoltage relays required by TS. This change is consistent with PBNP philosophy for calibration and testing of Agastat ETR relays that perform a safety related function. Based on industry experience with Agastat ETR relays, calibration and ,

testing of the relays at each refueling interval satisfies the testing requirements for ETRs used in a safety-related J application. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-057-02)  ;

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63. MR 95-049,(Unit 2),4160 V.

MR 95-049 installs a time delay relay in the 4160 V loss of voltage relay circuit input to the EDG output breaker closing circuits for 4160 V breakers 2A52-67,73,87 and 93. His ensures that 480 V load shedding occurs prior to an EDG (in running standby condition) picking up a 4160 V bus during a loss of voltage

- condition. Calculation N-94-130 shows that during such a condition, when an EDG is in the standby (running) ,

mode and an undervoltage condition occurs, the EDG could potentially load on to the respective 4160 V bus i prior to the 480 V loss of voltage relays stripping the 480 V loads if the 480 V and 4160 V loss of voltage relays were set at their extreme limits as specified in TS Table 15.3.5-1. This could cause the EDG to load unstripped 480 V buses, resulting in a possible EDG overload. As noted in WE to NRC letter dated April 27, ,

1995, and in an NRC SER dated December 27,1995, this condition is inconsistent with the design and ~

licensing basis requirements for the setting of these relays. The time delay relay circuitry added by MR 95-049 corrects this potential condition. The modification installs new relays with redundancy within the new time

. delay circuit in the 4160 V cubicles, installs test pushbuttons and test points on the 4160 V cubicles, and -

performs internal wiring changes within the 4160 V cubicles on 2A-05 and 2A-06 to facilitate the new tim '

- delay relay scheme and changes in the EDG closing control circuits.

- Summary of Safety Evaluation: EDGs supply emergency power to safeguard loads by reenergizing the . ,

respective 4160 V bus when a loss of voltage from the normal supply occurs and the 480 V load shedding i occurs via the 480 V loss of voltage relays. The time delay is affected during a loss of voltage on either 4160 V  :

bus 2A-05 or 2A-06 by installation of a time delay in the EDG output breaker close circuits for 2A52-67,73,87 'l and 93. However, the time delay is short enough that no effect on the closing of the EDG output breaker occurs l except when the EDG is in a standby (running) condition and the loss of voltage occurs from the normal supply During this condition, the permissive to close the EDG breaker becomes a function of the new time  :

delay scheme. The time delay is greater than the time elapsed between dropout of the 4160 V loss of voltage l' relays and dropout of the 480 V loss of voltage relays; but less than the 10 second EDG start time assumed in

. the accident analysis. His enhances the availability of emergency power to safeguards loads during an accident or event. Unit 2 is in cold shutdown, refueling shutdown or defueled during installation and testing of ,

this modification. De 4160 V undervoltage protection circuitry and respective EDG breaker control circuits ,

(for each bus 2A-05 and 2 A-06) are removed from service (one train at a time) during this work. A loss of emergency power to an entire AC power train is assumed in the accident analysis when it is the most limiting ,

accident condition. Derefoie, during installation and testing of this modification, no components are disabled beyond those currently analyzed in the accident analyses. No equipment is operated beyond the acceptance

- criteria described in the TS. ne modification does not pose a USQ nor does it require a change to the TS.

(SE 98125) i

64. MR 95-059' A,(Unit 1), Safety injection.

i The modification installs a pressure relief valve (SI-834E) in both Unit I nitrogen supply piping lines. A pressure regulator (ISI-688) is installed in conjunction with the relief valve to fulfill OSHA requirements for a [

compressed gas system. He spoolpiece is replaced with a tubing tee and vent valve (ISI-686) to allow for I operator ease during test performance.

i Summary of Safety Evaluation: De modification occurs during cold shutdown conditions, while the piping is -

1 isolated from the SI accumulators and the nitrogen source. He new check valve disc does not affect the ir tegrity of the nitrogen supply piping to the Si accumulators. The check valve function to shut to allow pressure to be maintained within the accumulators when Si is required does not change. The valves do not have'an adverse affect on containment integrity. De relief valves protect the containment penetration piping from overpressurization during filling of the SI accumulators. The modification does not pose a USQ nor does  !

it require a change to the TS. -(SE 96-087-01) l l

Page 54 of 112

65. MR 4007, { Unit I),480 V.

MR 96-007 installs a test switch assenbly in the IB-03 and IB-04 480 V safeguards buses. Each assembly consists of ten individual knife-edge test switches. Test switches are wired into the closing circuits of seven 480 V breakers: I B52-04B, normal supply to IB-01; 1852-05B, normal supply to IB-02; IB52-16B, norn.al supply to IB-03; IB52-17B, normal supply to 1B-04; IB52-16C, IB-03/IB-04 tie breaker; IB52-15C, 18-03/1B-01 tie breaker; and IB52-18C,1B-04/1B-02 tie breaker.

Summary of Safety Evaluation: The modification allows defeating the dead bus transfer scheme employed for the 480 V distribution system. The Unit 14160/480 V safeguards buses are tbd together to perform test switch installation for breakers 1BS2-16B and IB52-17B. Calculation 98-0021 verifies the ability of this tiered configuration to carry the Unit I safeguards loads required to operate to mitigate a DBA in Unit 2. He work plan restricts loading on transformers IX-13 and IX-14 during the cross-tie in order to meet the requirements of the calculation.

Operatien of the test switches imposes no functional changes upon operation of plant equipment. De test switches perform the same function that is fulfilled by the temporaryjumpers used to defeat the breaker den,d bus interlocks. The interim configurations that the Unit 1480 V distribution system is placed in to support test switch installation are approved and administratively controlled via TSs 15.3.~l.B.I.d,15.3.7.B.I.f, 15.3.7.B.I.g, and 15.3.3.D.2.a.

The test switches do not affect accidents evaluated in the accident analyses and do not affect the safe shutdown of the units. The activity does not increase the radiological consequences of an accident, event or malfunction of equipment important to safety previously evaluated in the CLB.

The modification only affects the supply and tie breakers for buses 18-01, IB-02,1B-03, and IB-04 and does not create the possibility of an accident or event of a different type than previously evaluated in the CLB.

Momentary paralleling of power to the buses is provided to ensure no loss ofloads and to allow breakers to be removed and returned to service. The modification does not pose a USQ nor does it require a change to the TS.

(SE 98-054)

I Summary of Safety Evaluation: The SE revision addresses CR 98-1889. ne changes are administrative and '

do not change the conclusions of SE 98-054. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-054-01)

66. MR 96-008,(Unit 2),480 V.

The modification changes the seven circuit breakers that comprise the dead-bus transfer scheme for the Unit 2 480 V distribution system to allow the scheme to be defeated.

Summary of Safety Evaluation: Safeguards buses 28-03 and 28-04 are deenergized to allow installation of the test switch assemblies and are cross-connected to perform PMT of breakers 2B52-25B and 2B52-40B.

l Approved procedures gover'n these evolutions. IWP 96-008-01 and IWP 96-008-02 require 8] nit 2 to be  !

defueled to perform the test switch assembly modifications. Tne modification installs local close/ trip pushbutton assemblies in the control circuits for three non-vital 480 V circuit breakers in buses 2B-01 and 2B-02. To allow the pushbutton installations, these buses are deenergized via other approved procedures. The ,

unit is required to be in a hot shutdown condition, or a more conservative mode, to perform the pushbutton j assembly modifications. The modification does not pose a USQ nor does it require a change to the TS.

(SE 98-167) l' Fqe 55 of112

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' 67. MR 96-014*A and 96-014*B,(Unit I and 2), Main Steam.  ;

The modification replaces the main steam isolation solenoid valves. De old valves are no longer f manufactured, are considered obsolete, and rebuild kits are no longer available. The modification replaces the j valves with the manufacturer's recommended replacement valves.  ;

Summary of Safety Evaluation: The modification replaces intemal piping / valve assemblies in panels RK-33/34  :

that provide control air for the main steam isoir ion valves (MSIVs). The isolation globe valves are replaced  !

with ball valves. De old pressure gauges and pressure switches are reused. l

?

There is a conservative change in loading on the 125 V de power system because of the new solenoid valves.  !

ne new valves are either identical to or better in mechanical attributes to the old valves. The components l installed meet or exceed system design requirements. De replacement r,olenoid valves are tested to  :

demonstrate operability before or after a seismic event. Testing ensures that the control air system is functional ; i and meets TS requirements for valve closure times. ne majority of equipment is passive in nature and is suitable for the requirements of the system and the environment it is installed in. De modification installs  !

- compatible or identical equipment. Therefore, there is no increase to the probability of failure rates, no changes [

to accident analyses, or impacts to safety-related components or structures. [

i This modification does not increase the probability of occurrence of an accident, event, or malfunction of l

. equipment important to safety, nor does it reduce the margin of s fety. No new release paths of radioactive j material are created that may affect the ssfety and the health of the public. ne modification does not pose a p US / nor does it require a change to the TS. (SEs98-041 and 98-179) l f

68. MR 96-034,(Unit 1), Steam Generator.

The modification relocates the Unit 1 SG level tap instrumentation. The instrumentation used increases the  ;

span of the narrow range protection system instrumentation by lowering the lower pressure tap associated with  !

the level transmitter variable leg sensing line. In addition, the SG water level corresponding to the protection j system low and low low level setpoints is lowered.

(

Summary of Safety Evaluation: The relocation of the six lower narrow range level taps involves machining the SG and rerouting the instrumentation tubing for the lower and upper narrow range sensing lines as well as one }

of the upper wide range sensing lines. Evaluations and calculations demonstrate that the new level tap  ;

f configuration and instrumentation tubing / supports are designed in accordance with appropriate codes. Stresses in the level tap assembly, the SG shell, and tubing was confirmed to be bounded by ASME Code allowables.

Therefore, the modification represents a condition equivalent to the original design basis of the plant. De ft location of the associated shell penetrations does not exceed the ASME Code minimum value regarding l proximity of pressure vessel penetrations. j

- The span of the three narrow range water level transmitters is increased from 143" to 206." His is f accomplished by decreasing the lower fluid connections for these transmitters. The increased span results in i less instrument tubing used, thus reducing the probability of a tubing leak affecting a safety system. The use of j one valve at the SG instead of two decreases the probability of a valve packing leak. The increased span allows .  !

' setting some reactor trip signals to a lower water level. His allows for operator case to address operational l

- transients and avoids reactor trips thet challenge safety systems. De increased span decreases the probability i of a reactor trip when at a low SG water level. If a trip occurs, there is less water in the SG at the time of the j trip. His lower water inventory at the time of a trip was a consideration used in the accident analyses, ne j analyses yield:d acceptable results so the consequences of an accident are not increased. Because of the  ;

increased span, trip and control signals that are based on narrow range water level are recalibrated. The  ;

suitability of the prior installed narrow range water level transmitters for the increased span was checked and determined to be satisfactory, both for operation and the situation where the reference leg was filled and the SG 1 water level was below the lower tap. The low-low water level setpoint of 25% of narrow range span and the TS low-low water level limit of 220% of narrow range span was selected by analytically aim; instrumentation Page 56 of112

uncertainties and processing measurement effects and margin to the analytical value of 10% used in the events analyses. The coping analysis associated with Station Blackout was revised to reflect the lower water inventories of the SGs at the new low-low water level setpoint. He coping analysis indicates that there is an adequate inventory of water in the SGs and conde:ncte storage tank for Station Blackout. He Appendix R evaluation is not afTected. The high energy line break evaluation in FSAR Appendix E is not altered by the new taps, tap locations or setpoints. The assumptions forjet impingement and environmental qualification bound the operation with tN new tap locations and setpoints. The modification does not pose a USQ nor does it require a change to the TS. (SE 97-204)

Summary of Safety Evaluation: The SE revision removes the hydrostatic test requirement for the sensing lines.

Hydrostatic testing is not Code required and is replaced with an inservice leak test. The change does not pose a USQ nor does it require a change to the TS. (SE 97-204-01)

Summary of Safety Evaluation: He SE revision adds a temperature range for the inservice leak test to be performed. The range of t10*F provides flexibility during performance of the test, ne test pressure associated with the new temperature range is above normal operating pressure for the SG secondary side and j

meets the requirements of ASME Section XI. The change does not pose a USQ nor does it require a change to the TS. (SE 97-204-02) l

69. MR 96-047'B,(Unit 2), RCS.

MR 96-047'B replaces power-operated relief valves (PORVs) RC-515 and RC-516. He old 3" Velan flex wedge gate valves are replaced with 3" BW/IP parallel disc gate valves. Only the valve bodies and internals are l

replaced. The old Limitorque SMB 00 motor operators are reused.

Summary of Safety Evaluation: Malfunction of the PORV block valves is not considered in the FSAR. Should the block valves fail shut, the pressurizer safety valves are designed to accommodate the maximum surge rate resulting from a complete loss ofload without a reactor trip or other control. The block valves isolate failed PORVs. The replacement valves reuse the motor operators and have a similar stroke time to the original I valves. The new valves meet the original design requirements and correct design deficiencies of the original valves. The new valves are not susceptible to pressure locking, thermal binding or stem embrittlement issues.

EPRI has developed a gate valve perfonnance prediction program that recommends rounding the disc edges and hard facing disc guides to ensure proper operation under design conditions. De BW/IP parallel disc design '

has these design characteristics as applicable and a much lower unseating load than the original Velan valves.

These features ensure that the valves perform under design conditions. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-121)

70. MR 96-064* A and 96-064*B, (Unit 1), Service Water.

The modification upgrades the capability of pipe supports and adds pipe supports in the SW system to accommodate higher loading and piping therm:;l movements due to higher temperatures in the CFC return lines, and hydraulic loads resulting from water hammer in the CFC return lines. The higher temperatures in the CFC return lines have been postulated based on analyzed CFC performance with lower than design fouling factors. The water hammer loads have been postulated based on the LOOP and LOCA scenarios identified in Westinghouse Nuclear Safety Advisory Letter (NSAL) 96-03 and NRC Generic Letter 96-06.

MR 96-064*A adds three new pipe supports and modifies nine existing pipe supports in the Unit I containment portion of the CFC return lines. The modification maintains 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.

Page 57 of112

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1 t

I MR 96-064*B adds two new pipe supports, modifies 21 existing pipe suppons and removes nine existing pipe supports for the Unit i outside containment portion of the CFC return lines. The modification maintains compliance with design basis requirements for pipe stress and suppon load capacity, and ensures that no piping, supports, structures or in-line components are stressed in excess of Code allowables.

Summary of Safety Evaluation: Modifications to the SW suppon are performed with Unit 1 in a cold shutdown, refueling shutdown or defueled condition. The installation does not degrade the SW system pressure boundary. The changes do not degrade the suppon system nor involve welding to the inservice pressure boundary.

The water hammer evaluations performed used a transient two-phase flow analysis of the CFCs to determine the internal pipe forces on each CFC train. The methodology used benchmarking vice experimental test data.

The calculation is a prediction of the maximum pipe pressure and the time dependent forces at various locations along the pipe circuit for both the loss of power and the LOCA coincident with loss of power transients. These results bound all other accident conditions. The thermal analysis was performed at 286*F to bound the -

maximum calculated pest-accident containment temperature per FSAR Section 6.3.1. The analyses demonstrate that once the modifications are completed, piping is within design basis Code allowables for components, pipe stress and support loads as defined within the CLB.

The design changes enhance the affected piping systems' ability to withstand loading conditions and are acceptable under normal, emergency and faulted conditions. System capacity and operability are not degraded, nor are the system functions impaired. There is no effect on piping system pressure boundaries or on the integrity of the Unit I contr.inment pressure boundary. The Unit 1 CFCs do not require removal from service for this work, and are maintained available as veded to support shutdown safety. The SW system remains fully functional and capable of responding to a Unit 2 accident, if required, during installation. The modifications do not pose a USQ nor do they require a change to the TS. (SEs97-188,97 191)

The SE revision addresses CR 98-0819 and ensures the SE documents the scope of work performed via MR 96-064' A. Grinding occurred during this work, which was not addressed in the previous SE.

Summary of Safety Evaluation: Pipe saddles are removed from existing suppon during the modification.

Welds are removed by grinding. Pipe wall thickness is verified after weld removal. A VT and PT/MT of the affected areas are performed to verify the integrity of the pressure boundary. The changes do not degrade the support system or involve welding to the inservice pressure boundary. Once the modifications for the supports and addition of the new supports are implemented, pipe stress, suppon load and component Code allowables are satisfied under design basis conditions. The change does not pose a USQ nor does it require a change to the TS. (SE 97-188-Oli The SE revision adds information related to the removal of existing pipe saddles. The work occurs during UIR24 with Unit I in a cold shutdown, refueling shutdown or defueled condition.

Summary of Safety Evaluation: Grinding is not performed if the minimum wall thickness is less than allowable. A VT and PT/MT of the affected areas are performed to verify the integrity of the pressure boundary. Once modifications for the 21 suppons, addition of the two new suppons, and removal of nine -

existing suppons are implemented, pipe stress and suppon load Code allowables are satisfied under design basis conditions. The modification does not pose a USQ nor does it require a change to the TS.

(SE 97-191-0J l

Page 58 of112 l

71, MR 96-064*C,(Unit 2), Service Water.

MR 96-064*C modifies 23 existing pipe supports in the Unit 2 containment portion of the CFC return lines.

Dese modifications are required to maintain compliance with design basis requirements for pipe stress and suppon load capacity, and ensures that no piping, supports, structures, or in-line components are stressed in excess of Code allowables.

Summary of Safety Fvaluation: The design changes enhance the affected piping system ability to withstand loading conditions, ao) are acceptable under normal, emergency and faulted conditions. The system capacity and operability is not degraded, nor is the system functions impaired. There is no effect on piping system pressure boundary or on the integrity of the Unit 2 containment pressure boundary because of the modifications. The Unit 2 CFCs are not removed from service during this work. Therefore, the CFCs are maintained available as needed. The SW system remains fully functional and capable of responding to a Unit I accident, if required. Changes to SW system suppons outside containment are documented in MR 96-064*D.

MR 9-064*C and *D together resolve concems with the SW piping described by NRC Generic Letter 96-06.

The modification does not pose a USQ nor does it require a change to the TS. (SE 98-141)

72. MR 96-064*D,(Unit 2), Service Water.

MR 96-064*D modifies 14 existing pipe supports and removes 2 existing pipe supports from the Unit 2 CFC return lines outside containment.

Summary of Safety Evaluation: The design changes enhance the affected piping systems' ability to withstand loading conditions, and are acceptable under normal, emergency, and faulted conditions. Bey are performed in a manner such that the affected system capacity and operability is not degraded, nor are the system functions impaired. Dere is no effect on piping system pressure boundaries or on the integrity of the Unit 2 containment msure boundaries. The Unit 2 CFCs are not removed from service during this work. He CFCs are aintained available as needed to support shutdown safety issues, ne SW system remains fully functional and capable of responding to a Unit I accident, if required, during the installation. Changes to the SW system supports inside containment are documented in MR 96-064*C. MR 96-064*C and *D together resolve the concerns with the SW piping raised NRC Generic Letter 96-06 for Unit 2. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-142)

73. MR 96-068*A,(Unit 1), Containment Heating Steam.

MR 96-068* A isolates the heating steam system from Unit I containment, and eliminates Unit I heating steam containment isolation valves (CIVs). Pressure boundaries currently provided by CIVs llV-808,809,818,632, and 633 and isolation valve test connections (IVTCs) are replaced with welded caps inside containment, and capped valves outside containment are replaced with test connections.

SummaryofSafety Evaluation: FSAR Section 5.2," Containment Isolation System," changes are required. The CIVs are replaced with welded boundaries; therefore, portions referring to CPP-52 and 53 are deleted. The final welded cap configuration is missile protected, not subject to significant thermal stresses, without resilient sealing materials, and not subject to flow accelerated corrosion. Most importantly, the modified configuration no longer performs an active " containment isolation system" function like other systems described in the FSAR section. " Closure of mechanical barriers" is no longer applicable.

He final design consists of closed and passive components, and does not initiate an accident or event under postulated post-LOCA or DBE conditions. Containment closure requirements are met during installation because the CIVs are removed while ensuring a single closed barrier is maintained. Replacing valves with a welded cap inside containment decreases the probability of occurrence of a containment isolation system malfunction as a welded configuration has a lower probability ofleaking than do CIV seals with resilient materials. The final design eliminates active barriers, thereby reducing the possibility of human error.

Page 59 of112

Radiological consequences of a fuel handling accident are not increased because containment closure is maintained during the containment isolation system modificttion. Piping is leak tested prior to unit startup, to ensure containment integrity for TS and 10 CFR 50 Appendix J compliance. l The modifications do not decrease the margins of safety as defined in the Basis of TS. He CIV replacements with welded pipe caps decrease post-IACA total containment leakage (below 13 ) as described in the TS and 10 CFR 50 Appendix J. The modification does not pose a USQ nor does it require a change to the TS.

(SE 97-169)

74. hjR 96-076*A,(Unit 1), Main Feedwater. I

~

The modification replaces the trim in the main feedwater regulating bypass control valves (ICS-00480 and  !

ICS-00481) while the unit is shut down for U IR24. Final acceptance testing is performed during unit startup.

The modification increases the flow capacity of the bypass valves. He increased flow rate allows the bypass valves to be used for longer periods during startup or low-load operations. Therefore, the main feedwater regulating valves are not required to control flow at extremely low flow rates. The trim replacement enhance.s 1 performance of the bypass system.

Summary of Safety Evaluation: The main feedwater bypass control valve provides adequate flow control at  ;

low power where a small change in the stroke travel of the large main control valve would result in a large percentage change in flow control. He valve isolates feedwater flow upon receipt of a main Si signal.

Feedwt.cr isolation is required to limit additional flow to the SG during SLB accident as described in FSAR Section 14.2.5. The valve spring is sized to ensure valve closure capability.

The new operating rang,e for the valve is achieved by revising valve trim and increasing spring size so it performs the same functions and has the same design as the previous valves. The increased trim does not add new active components to the system. He original valve body with new trim of compatible material and upgraded spring maintains the pressure boundary and isolation functions. Even with the increased capacity, failure of the bypass valve to shut is bounded by the excessive load increase incident. He modification does not affect the valve control circuits or control logic. He modification does not aher the design basis of the main feedwater system. The modification does not pose a USQ nor does it require a change to the TS.

(SE 97-200)

75. MR 96-080,(Common), Cable Spreading Room (CSR).

The modification enhances support anchorage in the CSR and adds, replaces or removes Unistruts in the area.

In ;nost cases the new Unistrut is bolted into place; however, in some cases, the Unistrut is welded to existing embedded plates or other types of elements such as threaded rod, or Hilti concrete expansion anchors are installed.

Summary of Safety Evaluation: No cable is rerouted during the modification nor are cable tray penetration fire barriers affected by the work. In addition, netting or tool lanyards or other means (such as barriers) are employed to maintain separation between tools or other components and safety-related components to prevent accidental damage or bumping of safety-related equipment. He integrity of the cable tray support system is -

maintained during the installation. Access to individual controls and cabinets are maintained at all times. In addition, precautions are taken when welding is performed to prevent activation of the fire suppression system.

Fire watches are provided during periods of welding. The welding machine is kept outside the CSR to minimize possible EMF interferences with sensitive equipment.

Page 60 of 112

The modification ensures that the cable tray supports in the CSR are qualified to the SQUG criteria. The modification enhances plant configuration and increases overall plant safety. The funciion of the CSR cable tray supports is not changed by the modification. The modification does not pose a USQ nor does it require a change to the TS. (SE 97-212)

76. MR 96-081* A,(Unit 1),120 V ac Electrical.

MR 96-081* A changes the methodology for transferring normal and alternate supplies for the safety-related 120 V ac instrument buses. Tt.e installation of make-before-break manual transfer switches a!!ows transfer of instrument bus power supplies without interruption.

Summary of Safety Evaluation: The worst case failure mode associated with this modification is the loss of power to an instrument bus. The failure of an instrument bus power supply is not an initiator of acc! dents previously evaluated in the FSAR. The reliability and availability of the instrument buses and accident mitigating components supplied from the instrument buses during and following previously analyzed accident are not decreased as a result of the change. Recovery actions for timely restoration of power to affected buses are not complicated as a result of the new configuration.

The previous methodology used was highly dependent upon operator skill to shift bus supplies without momentary power interruptions. With the new method of transferring buses, a loss of power could result from transfer switch or inverter failures during bus shifting. Operator action resulting in momentary power interruptions is less likely in this scenario. Paralleling ofinverters for approximately 10 milliseconds during switching ofinstrumen' bus supplies does not adversely affect the associated inverters or instrument bus loads.

The use ofindicating lights and a voltmeter to ensure that both instrument bus supplies are energized and in syz prior to shifting further reduces the possibility of power supply transients. He reliability of the transfer switches is considered to be as good as or better than the breakers. The new manual transfer switch enclosures and devices are seismically qualified. The new voltmeters and indicating lights are separately fused to isolate these non-safety-related components from the safety-related buses in the event of a component failure.

l Addition of a manual transfer switch to the system does not introduce new failcre modes for the inverters or instrument bus loads and does not affect other plant systems or components. Failure of a transfer switch during shifting has the same result (loss ofinstrument bus power) as a breaker failure. Loss of power to an instrument bus was previously analyzed in the FSAR.

1 ne reliability of the instrument buses is not decreased by the change. He reduction in nuisance alarms and spurious actuation helps improve operator cognizance of actual changes in plant status. Separation between  !

instrument bus trains is not affected. The qualification of the instrument bus panels are maintained. The I modification does not pose a USQ nor does it require a change to TS. (SE 97-190)  !

77. MR 96-081*B,(Unit 2),120 V ac. l MR 96-081 *B changes the methodology of transferring normal and alternative supplies for the safety-related 120 V ac instrument buses. In the past, the two main breakers in each instrument bus panel were used for bus

. transfer.

Summary of Safety Evaluation: The modification affects the method of shifting instrument bus power supplies.

The worst case failure mode associated with this change is the loss of power to an instrument bus. He 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 and accident mitigating components supplied from the l instrument buses during and following previously analysed accidents are not decreased. Recovery actions for timely restoration of power to affected buses are not complicated as a result of the new configuration.

Page 61 of 112

Indication lights and a voltmeter ensure that both instrument bus supplies are energized and in sync prior to  ;

shifting. This reduces the possibility of power supply transients. The reliability of the transfer switches is i considered to be as good as or better than the breakers. The manual transfer switch enclosures and devices are seismically qualified. The voltmeter, indication lights and pushbutton are fused to isolate these components from the safety-related buses.

Adding a manual transfer switch to the system does not introduce a new failure mode for the inverters or instrument bus loads and does not affect other plant systems or components. Failure of a transfer switch during shifting has the same afTect as a breaker failure causing loss ofinstrument bus power. His type ofloss of power to an instrument bus is analyzed in the FSAR. The reliability of the instrument buses is not decreased as a result of this change. The reduction in nuisance alarms and spurious actuation improves reliability of actual '

plant status. Separation between instrument bus trains is not affected. Qualification of the instrument bus panels is maintained. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-122)

78. MR 97-002,(Unit I and 2), Boric Acid Recirculation.

MR 97-002 replaces 1&2P-116, boric acid recirculation pumps. The replacement pump has a magnetic coupling and no seal. The pump change resolves seal leakage problems by eliminating the seal that was installed on the prior pump. Three new pipe supports are installed to minimize the flange loading on the pump.

Summary of Safety Evaluation: The new pump casing, impeller and shaft are 316 SS, which is appropriate for boric acid service. Since the new motor has a higher urrent rating (2.9A vice 2.3A), the overload relay heaters are resized and replaced in accordance with MI 18.6," Instructions for lleater Element Selection for Westinghouse Motor Starters." The additional bus loading of 0.6A was evaluted as acceptable. The flange-to-flange distance on the new pump is shorter than the old pump, and the flanges are positioned slightly lower. This requires a minor adjustment in the piping configuration to accommodate the new dimensions. De piping is in accordance with the original B31.1 t equirements. De operation and control of the new pump are not changed. ,

The flooding that could occur with a potential structural failure of the boric acid recirculation pump is limited by the size of the pump and the 2" piping. The open design of the PAD can accommodate significant flooding.

Potential flooding that could occur is bounded by existing flooding analyses associated with failure of a much larger system, such as SW.

The boric acid storage tanks (BAST) are not contaminated. Therefore, potential failure of the pump pressure boundary has no radiological consequences. Postulated offsite and onsite dose rates are not increased.

TS 15.3.2 addresses the CVCS system and the required boration flow paths. The Basis for TS 15.3.2 discusses the RWST as well as the BASTS as sources of borated water. De BASTS and the flow path to the charging pumps are operable when the boric acid recirculation pump is isolated. Since the RWST is the safety-related source of boric acid, replacing the boric acid recirculation pump can not affect the ability to mitigate the consequences of an accident or malfunction. The modification does not pose a USQ nor does it require a change to the TS. (SE 97-063)

79. MR 97-009' A/* B,(Units I and 2), Containment Penetimions. t The modification welds the P-12b and P-30a containmea penetrations. The prior isolation valves, drain valves I I

and piping are cut offinside containment at approximately 6" from the point of piping entry into the inside penetration cap. A pipe cap is socket welded into place on the open pipe end inside containment. Leak testing j and VT are performed to confirm the integrity of the inside pipe cap weld. De remaining open pipe end is j threaded and a threaded pipe cap is installed. l Page 62 of 112

Summary of Safety Evaluation: The modification is installed during shutdown conditions. One barrier remains intact on each penetration throughout the installation. He penetrations being modified function to maintain containment integrity. The replacement of valved CIVs with welded / threaded caps is a more conservative i design (less likely to leak) than the existing configuration. The change has no negative afTect on rnalyzed events because the design is more conservative. The modification does not introduce unanalyzed conditions.

After installation and leak testing, the welded inside pipe caps are considered part of the containment liner and no future local leak rate test (LLRT) of the f ormer penetrations are required.

l Applicable design and operation criteria remain valid. The modification eliminates personnel dose associated

{

with performance of LLRTs on penetrations P-12b and P-30a. The work also reduces the potential for

~ containment leakage at these locations. The modification does not pose a USQ nor does it require a change to  !

the TS. (SE 97158)

)

I

80. MR 97-010*A, and MR 97-010'B,(Units I and 2), CVCS. I The modification replaces CV-286, CV 287, CV-288, CV-289, CV-290, and CV-291, charging pump discharge valves. Engineering Work Request (EWR)94-260 identified that the charging pump discharge ,

valves require excessive force to operate and were causing personnel injuries. The old discharge isolation I valves were Velan Model 13920 globe valves installed with the flow over the seat (with the exceptions of 2CV-287,2CV-288,2CV-291 that have the flow under the seat). The old valves had a rising rotating stem with bronze bushings. The replacement valve is a double disk gate valve with needle bearings. ne bearings are ,

located external to the piping system. His modification also changes FSAR Figure 9.2-1.

l 1

Summary of Safety Evaluation: The work is perfonned during refueling shutdown when the charging pumps i are not required to be operable. The affected charging pump is isolated and tagged out. De two discharge I valves are cut out and removed, thus providing space for the new valves to be installed. The new valver are  !

siightly larger, heavier and have a larger Cv then the original. A seismic evaluation verified that the system i meets the system seismic requirements because of the increase in valve weight. The increase in Cv slightly l reduces the pressure losses in the CVCS piping system, thus providing slightly lower back pressure on the discharge pulsation dampers and charging pumps. The pressure losses through the fully open existing 3" globe

{

valve at 60.5 psi are less than 1 psi. Thus, the reduction in pressure drop seen with the new gate valves is not I significant. The replacement valves meet the design requirements of the CVCS system as identified in the FSAR. )

l TS 15.3.2 states that the reactor shall not be tal:en critical unless the following CVCS system conditions are met:

1) a minimum of two charging pumps for that reactor shall be operable; and,2) the system piping and valves shall be operable to the extent of establishing two flow paths from the boric acid tank (s) and/or the j refueling water storage tank to the reactor coolant systerti. During the interim period when one of the six '

charging pump discharge valves is removed, all three charging pumps for that unit are considered inoperable.

The modification is performed when the respective unit is shut down, containment integrity is not required and coordinated with either a core offload or when the CVCS boration path is not required. During this period, ,

CVCS containment closure is met by the inside CIV (one check valve per penetratim inside containment, three in total). A new valve is installed, tested and the charging pumps and piping system are declared operable. As

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j part of the CVCS closed loop outside of containment, the new valves are leak checked as defined by the leakage reduction and preventive maintenance program. The modification does not pose a USQ ncr does it require a change to the TS. (SE 97-175)

81. MR 97-014*A,(Unit 2),125 vdc.

The modification replaces existing safety-related 125 vdc panelboard D-302 and installs new non-safety-related panelboard 2D-202. New panel D-202 includes an isolMion device for the cable between D-302 and swing distribution panel D-30. Thus resolves an Appendix R issue described in CR 97-1965.

Page 63 of 112

i Summary of Safety Evaluation: Replacement of panel D-302 provides a means ofisolating cable between I D-302 and D-301. This reduces the consequences of a fault in this cable. There are no accidents or events  !

l previously evaluated in the CLB associated with this equipment. New panel 2D-202 provides a means of isolating the cable between D-302 and D-301. This reduces the consequences of a fault in this cable. There are no accidents or events previously evaluated in the CLB associated with this equipment. Mounting of the new panel, including conduit rework, is seismically qualified. Installation of two indication lights on 2C-20 does not affect the seitmic qualification of 2C-20. This is based on the small size of the mounting holes and the small weight increase. The circuit supplying the new indication lights has adequate capacity for the very small load increase resulting from the lights. Panel 2C 202 is non-safety-related and does not affect equipment important to safety. New conduit for 2D-202 is installed in non-seismic areas of the plant. Panel 2D-202 is mounted on a seismic wall but does not affect the seismic capabilities of the wall because rebar is not cut and

• the relatively small weight change.

Replacement of panet D-302 and installation of panel 2D-202 does not affect the ',*ty functions of the 125 V de system. The only failure modes associated with the new DC panels are open or shoit circuits. These failure modes exist with the existing panels. New panel D-302 contains only manually operated disconnect switches with no automatic functions. The modification ooes not reduce the independence of the main de buses. There are no credible accidents or events that could be caused by replacement of panel D-302 or installation of panel 2D-202. Four safety-related de liuses remain in service during this installation. There are no TS requirements regarding the availability of the swing safety-related battery to supply the de buses. The modi 0 cation does not affect the degree ofindependence of safety-related de system trains. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-132)

82. MR 97-014'B,(Unit 2),125 vdc.

The modiGcation replaces existing 125 vde circuit breaker panelboards D-03 and D31 with new fused switch panelboards. During replacement of panel D-03,inveners IDY-03 and DY-0C are supplied from swing safety-related panel D-302, which are supplied from battery D-305 and charger D-109. This allows the Unit I and 2 white instrument buses to remain energized during replacement of panel D-03.

Summary of Safety Evaluat' ion: Several loads supplied from the 125 vde system are deenergized during this work. Accidents associated with these loads include loss of offsite power and Unit 2 turbine overspeed. Since Unit 2 is shut down during the time that a Unit 2 independent overspeed protection channel is out of service, turbine overspeed is not possible. During shifting of de control power supplies to 13.8 kV, and 480 V switchgear, de control power to the affected buses are deenergized for a short period. This, however, does not increase the probability of a fault or other event resulting in the loss of offsite ac power.

The panel replacement corrects problems with age-related degradation of molded-case circuit breakers. The new panels contain fused switches instead of molded-case circuit breakers. The fused switches improve de system selective coordination and increase fault clearing capability. Mounting of the new panels, including conduit rework is seismically qualified. Te~mporary supply to the inveners is safety-related and seismically qualified. Selective coordination between the temporary inverter supply fuses in D-302 and the fuses in upstream Panel D-301 is improved. Rerouting of cables to 2D-202 and D-18 does not violate train separation criteria. The only failure modes associated with the new de panels are open or short circuits. This modification l does not introduce new failure modes nor reduce the independence of the main de buses, including interim conditions. Accidents or events associated with the loss of a single de train have been previously evaluated.

The mo#fication does not pose a USQ nor does it require a change to the TS. (SE 98-168)

83. MR 97,0,48,(Unit 1), Steam Generator.

1 MR 97-048 upgrades the duct and old suppons of I W-87A&B channelhead blower fans, provides designs and calculations for new vertical and lateral supports for the intake ducts to place the "A" compartment duct back into the "A" compartment, and repairs a portion of the "A" duct that is adjacent to the IW 87A filter box; additional suppons are installed in each compartment. The new supports are prefabricated outside of Page 64 of H2

containment and are mounted to the compartment walls with anchor bolts. The modification includes installation of new plate steel and anchor bolts.

Summary of Safety Evaluation: De modification calculations show that the final configuration of the duct is adequate for seismic 2/1 considerations, with respect to high energy line breaks (HELB), and with respect to environmental effects because of a LOCA. Dynamic effects from double guillotine failure of the RCS need not be considered since PBNP is a leak-before-break plant with respect to the primary loops. The installation is completed pnor to the leaving cold shutdown conditions following UIR24.

The modification reduces the probability of an occurrence of the malfunction of equipment imponant to safety

• or an accident because the installation of the suppons prevents the Asetwork from falling on equipment, such as the SG level indicators, that are important to safety during and aficr a seismic event, LOCA or HELB. In addition, the modification prevents the compounding of cont.eque nces because of an equipment malfunction by ensuring that the ductwork does not affect equipment imponant tc safety during or after a seismic event, a LOCA or a llELB. The radiological consequences of the accidents mad events described in the FSAR are not increased by the modification since the entire installation is completed within containment and does not affect radiological barriers.

We ducts are made of galvanized metal; however, FSAR Section 5.6.2.1 explicitly states that protective coatings such as zine do not significantly add to post-LOCA hydrogen buildup and are acceptable for use in containment. He load impaned to the structure by the new support is negligible given the capacity of the structure. Additionally, the installation is performed during cold shutdown, which funber reduces the possibility of the work causing a radiological release. He addition of the support does not change the function of the ductwork nor other components important to safety. The Basis of TS relatui to the SG companments references the operability of one SG when core temperature is >350' F, the functioning of the effluent monitoring instrumentation, and a working SG tube inspection program. Since the work is not related to these systems or tests, the margin of safety defined in TS is % affected. The modification does not pose a USQ nor does it require a change to the TS. (SE 97 21l)

84. MR 97-072* A, and MR 97-d72*B,(Unit I and 2). RHR.

MR 97-072 replaces the existing 2" stainlass steel globe valve with a 2" stainless steel gate valve. De modification is performed during refueling shutdown with the core defueled when RHR is not required to be operable.

Summary of Safety Evaluation: The valve isolates the cross connect line between the RIIR system and the CVCS. The line was initially installed to allow for additional letdown flow to occur when the existing line did not provide adequate flow because oflow RCS pressure. He valve is normally fully shut. Replacing the old globe valve with a gate valve should reduce ECCS leakage through the valve during leak rate tests. This aids in

' maintaining ECCS leak rates within the TS allowable limits. There are no changes to safety limits, setpoints or operating parameters as a result of the installation.

The new valve has Tristelle EB 5183 hardfacing material and is cobalt-free. The valve is procured and tested to e

meet Seismic Class I criteria and QA requirements. Here is no change in the function or method of operation of RH-715C nor the RHR system. The pressure drop across the full open gate valve is less than the drop across the old globe valve. The minor difference in flow characteristics does not adversely affect RHR system operation. The replacement valve meets pressure and temperature requirements for its application. The modification does not pose a USQ nor does it require a change to the TS. (SE 97-180)

Page 65 of 112 k

t Summary of Safety Evaluation: The new valve has Tristelle EB 5183 hardfacing material and is cobalt-free.

De valve is procured and tested to meet Seismic Class I criteria and QA requirements. Here is no change in the function or method of operation of RH-715C nor of the RHR system. De pressure drop across the full open gate valve is less than the drop across the old globe valve. De minor difference in flow characteristics does not adversely affect RHR system operation. De replacement vrilve meets pressure and temperature requirements for its application. He modification does not pose a USQ nor does it require a change to the TS. ,

(SE 97-180-01) l I

85. MR 97-085,(Unit 2), St. i MR 97-085 converts manual valves 2SI-857A(B), RHR heat exchanger to SI pump suction, into MOVs with ,  !

appropriate interlocks. De modification also converts the 2SI-89A&B Si recirculation line retum from fail  ;

shut to fail open air-operated valves. {

l Summary of Safety Evaluation: Failure of the systems and components a&cted by this modification are not i precursors to accidents or events previously evaluated. Train separation between the new 2SI-857A&B valves (i and revised 2SI-897A&B valves are maintained. Loss of one train was previously evaluated. The use of imerlocks on a train specific basis to enforce system operating requirements does not increase the probability of i occurrence of a malfunction of equipment important to safety. The occurrence of an Appendix R fire, which l could cause spurious valve ope ations coincident with a DBA is no: within our licensing basis. Potential {

malfunctions created as a resub of adding MOVs such as pressure Jocking or hot smart shorts, were previously l evaluated. . The modith::'on does not pose a USQ nor does it require a change to the "Ili. (SE 98-171) {

i 86, MR 97-086,(Unit 1), Steam Generators.

The modification cuts a new opening in the wrapper seal plate and installs a new wrapper plug. De wrapper f plug prevents cross flow from the downcomer to the tute bundle that could affect thermal performance of the l steam generator. There is no effect on the primary-to-secondary pressure boundary of the RCS.

Summary of Safety Evaluation: The modification has no effect on the pressure retaining shell of the SGs. -l There is no adverse effect on the S(i tubes either during installation or during subsequent plant operation. A i mechanical stop on the cutting tool prevents contact with the tubes and the piece of metal removed during cutting is retained inside the tool. Foreign material exclusion practices and a 'horough cleanliness inspection  ;

ensure that no debris is lea inside the SGs. Analysis demonstrates that the plug remains in place during normal  ;

and accide... :ondkions. Analysis also demonstrates that the seismic loads on the plug result in extremely small j and negligible stresses. The materials for the new plug are compatible with the chemistry environrnent of the i SGs.' The nr olug is designed to prevent cross flow from the downcomer to d;c tube bundle, so there is no reduction in thermal or hydraulic performance of the SGs. He modification does not pose a USQ nor does it l i

require a change to the TS. (SE 98-039)

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87. MR 97-087 and 97-088,(Unit I and 2), Reactor Protection System (RPS).

The modifications add a new test pushbutton in RPS Train A and B logic test cabinets to the P-10 permissive "

logic circuit, In addition, eight new banana test jacks are installed in each of the unit's RPS Train A and B logic test cabinets associated with matrix logic relays for the P-9 permissive, and the P-10 permissive logic .

circuits. The modification provides testing facilities for full testieg of the P-9 and P-10 perminive logic circuits and the associated nuclear instrumentation system trip block logic. Addition of the new test pushbunons and test jacks allows the P-9 and P-10 permissive logic circuits to be completely tested to meet the testing expectations of NRC Generic Letter 96-01, " Testing of Safety-Related Logic Circuits." The new test pushbuttons and testjacks are not required to actively function during or following a design basis event; however, the components are an integral component of the RPS logic circuits and therefore, are treated as QA scope equipment.

Page 66 of112 L . __

Summary of Safety Evaluation: The RPS is used for accident mitigation and is not an initiator of accidents previously evalua'ed in the 1 SAR. De addition of new test pushbuttons and testjacks in the RPS test cabinets has no effect on events or conditions, including radiological, which initiate or are the result of accidents, malfunctions or events previously evaluated in the CLB. He modification does not affect structural or mechanical pressure boundaries and no other systems or components are physically impacted by the modification. The design, location and operation of the new test pushbuttons and testjacks are similar to other existing RPS test pushbuttons and test jacks. New testing components are located inside the locked RPS test cabinets to ensure access is restricted and controlled. Postulated failure modes for the new components are encompassed or similar to the failure modes for other existing RPS pushbuttons and test jacks located inside the same test cabinets. The risk of failure / malfunction is significantly reduced using equipment that has been

- proven to be effective and reliable. Operating procedures OP 1C and OP-3 A require the operator to verify the successful operation of the P-9 and P-10 permissives during unit startup and shutdown. He modification does not affect the intended design or function of the RPS. Therefore, the installation does not affect the ability of the RPS to perfonn its safety-related function of tripping the reactor.

TS 15.4.1 Basis describes the tetting requirements for the RPS logic channels. Instrumentation testing frequencies are specified in TS Table 15.4.1 to maintain the status of the equipment and systems to assure safe operation of the plant. Testing is an integral part of maintaining the reliability of each RPS scheme. Adding the new pushbuttons and test jacks increases the reliability and margin of safety by meeting the testing expectations of GL 96-01. The modification does not pose a USQ nor does it require a change to the TS.

(SEs98-013 and 98-124)

88. MR 97-091,(Common),13.8 kV Electrical Feed.

MR 97-091 installs new switch fuse unit, H-08 and 13.8 kV/480 V auxiliary transformer X-65 to facilitate loads required for the new Nuclear Engineering Services (NES) building.

New switch fuse unit H-08 consists of two separate sets of disconnect switches and fuses to allow isolation of and protection for new transformer X-65. De new subswitch fuse unit can accommodate a future transformer.

This new unit is supplied from spare breaker H52-2313.8 kV bus H-02. X-65 is rated for 500 kVA and steps down the 13.8'kV voltage to a utilization voltage of 480 V. The new transformer feeds a new 480 V panel at the NES buik'ing, that accommodates loads for the building.

Summary of Safety Evaluation: The loss of all ac to the station auxiliaries usumes the loss of the 13.8 kV bus.

The affect of this change does not affect the inputs nor the conclusions of the analysis of that event. The 13.8 kV system is required to supply ac power to the station auxiliaries for mitigation of Station Blackout and Appendix R events. The ability of the 13.9 kV system to perform as described in these analyses is not affected by MR 97-091.

Breaker H52-23 is used to supply new switch fuse unit H-08 and transfonner X-65. To assure that the a

probability of accidents or events is not increased, the breaker is equipped with an overcurrent device to protect the cable and isolate the upstream equipment from faults. This protection philosophy is consistent with other breakers and their associated overcurrent devices used on the 13.8 kV system.

e The new circuitry, cabling and breaker H52-23 are tested and found acceptable prior to energization of new transformer X-65. Isotation requirements are in accordance with the danger tagging procedure.

The loss of all ac to the station auxiliaries, Station Blackout, and 10 CFR 50 Appendix R scenarios are bounding events or accidents that cover the changes and associated malfunctions evaluated. The change does not increase the probability or the consequences associated with these CLB accidents or events, nor does it increase the probability of a malfunction of equipment important to safety. There are no radiological consequences associated with the change. The electrical power supply poren of MR 97-091 does not pose a USQ nor does it require a change to the TS. (SE 98-052) t Page 67 of 112 i

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89. MR 97-094 and MR-096,(Unit I and 2), Main Steam.

MR 97-094/096 replace the moisture separator reheater (MSR) inlet steam control time pattern transmitter. The new controls are relocated from the rear of the 1&2C-03 main control boards (MCBs) to the front ofit. This allows the unit operator easier access to an important control function. In addition to moving the controller, the control function is provided by two Foxboro H-line controllers; one for each LP turbine, to ensure a more accurate controlling function. His allows one controller to determine the position of MS-2085 and MS-2086 while the other controller determines the position of MS-2087 and MS-2088. His provides individual control for LPI and LP2 crossover temperatures. Turbine crossover temperature recorders I&2TR-2038 are removed to make room for the new controllers. The crossover temperatures are displayed via the PPCS. nese temperatures can be automatically cont.olkd to a temperature setpoint determined by the operator or by manual ,

controf that allows the operator to determine a fixed valve opening. As a result of these changes, loading on '

instrument AC bus I&2Y-02 is reduced by approximately 23 W as a result of the removal of recorder 1&2TR 2038. -

Summary of Safety Evaluation: The modified system performs in the same manner as in the past. The positions of valves 1&2MS-2045 through 1&2MS-2088 determine the crossover steam temperature. By providing an casily accessible control station to the operator and splitting the control functions between two controllers, the crossover temperatures are maintained much closer to the desired value. A single failure only has half the effect on steam demand. Loading changes on the instrument ac buses are negligible. This modification does not affect the turbine protective features. Seismic qualification is maintained for the control board mounting and compor.ents in the system are fully qualified for the application. A seismic design review and post installation walkdown were performed in accordance with NP 7.7.2," Seismic Qualification of Equipment," and the SQUG Generic Implementation Procedure. Installation is in accordance with DG-E07,

" Separation of Electrical Circuits," to maintain electrical isolation and separation. De modification is installed during outage conditions when cresso u temperature control is not required. The modification does not pose a USQ nor does it require a change tc the TS. (SE 98-031)

Summary of Safety Evaluation: The SE revision changes the statements that the MSR steam controls are provided by two controllers. He new control scheme remains consistent with the existing control scheme in that all MSR steam inlet valves 1&2 MS-2085 through l&2MS-2088 are controlled by a single controller. The new controller and recorder have the same power consumption regrirements as that of the existing controller and recorder. Bus loading of 1&2Y-02 is reduced slightly less than previously stated, and 1&2Y-06 loading is increased slightly less than previously stated. Finally, turbine crossover steam temperatures are not routed to the PPCS; instead they continue to be displayed on the main ecntrol board. The SE revision also addresses changes to FSAR Figures 10.1-1 Sheet 2 and 10.' .l A Sheet 2. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-031-01) .

90. MR 97-099,(Common), Auxiliary Feedwater(AFW).

~ '

The modification installs new tr p/ throttle valves on the 1&2P-29 turbine-driven AFPs. This allows the low suction pressure trip to shut these valves on an initiation signal. ne use of the new trip / throttle va'ves for achieving the low suction pressure trip function and removing the function from the 1&2MS-2019/2020 main i of the turbine-driven AFPs. This ensures the protection of these pumps

• steam valves allows a much fe from the loss of the condenr# orat : tank (CST) suction water supply in the event of an earthquake or j tornado missile, which could ; -ntia ly damage the CSTs.

Summary of Safety Evaluation: The low suction pressure trip protects the AFPs in the event of a seismically induced loss of feedwater accident. FSAR design basis accidents are not assumed to be caused by a seismic l event and FSAR analyses are not performed assuming a coincident seismic event. The benefit derived from a more reliable pump protection design on low suction pressure outweighs the marginal increase in failure probability caused by adding an electrical signal to the turbine trip / throttle valves. The overall single failure FSAR accident criteria for the AFW system design is not affected by the changes. Therefore, balancing the benefit and risk associated with the change, the overall probability of an equipment malfunction from the Page 68 of 112

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l modification is not increased. De installation of new trip / throttle valves and the repowering of the low suction pressure transmitter for P 29 does not increase the possibility of occurrence of an accident or event, or '

occurrence of a malfunction of equipment previously evaluated in the CLB.

Evaluations and Calculation 97-0215 show that the volume of water is adequate to protect the two AFPs that do not fail, and the total flow to the SGs is sufficient to show the capability to remove decay heat. Installation of the trip / throttle valve and the separation of the de power supply does not create the possibility of a new accident scenario or new malfunction of a different type than previously evaluated in the CLB.

The modification does not adversely change the fire protection or Appendix R safe shutdown design as

" described in the Fire Protection Evaluation Report (FPER) and associated NRC approved Appendix R exemptions. He modification does not pose a USQ nor does it require a change to the TS. (SE 97-207)

91. MR 97-106,(Unit I), Reactor Coolant System.

MR 97-106 replaces reactor coolant (RC) loop resistive temperature detector (RTD) bypass flow indicating alarm switches I FI A-458 and I FIA-459 with new Rosemount flow indicating transmitters, loop power supplies and alarm units. The new transmitters provide a more reliable and accurate flow indication system to monitor i RTD bypass flow. The new transmitters sense differential pressure electronically with a sealed capacitive

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sensing element that climinates mechanical force transfer and problems with vibration and shock. He new transmitters are provided with a digital liquid crystal display (LCD) meter for local indication and integral drain / vent valves to facilitate filling and venting the transmitter. New loop power supplies and alarm units are mounted locally at the flow switch inside new electrical enclosures designed to meet the weight load capacities l specified in DG-E02," Seismic Conduit Support Design Manual," and the material requirements of  !

specification PB-546," Electrical Installation."

Summary of Safety Evaluation: Unit 1 is defueled or in a cold or refueling shutdown, and depresurized with IP-l A and IP-1B RCPs secured during installation and PMT. Each RCS RTD bypass flow instrument loop is deenergized and the associated instrument root vsives isolated during installation. New cable routed is meggered and continuity tes,ted. Seismic qualification of the new transmitters, including 3/8" instrument tubing, and electrical enclosure mounting, are performed using SQUG methoddogy. The post-maintenance

{

testing portion calibrates the new RCS RTD bypass flow instrument loops and verifies the design function of '

the new instruments by simulating both high and low flow conditions to verify that the control room low flow alerm and status lights correctly respond. A leak check is performed to verify zero RCS leakage resulting from the new insulation.

The new flow instrumentation loops perform the same non-safety-related functions as the old flow instrument loops and provide the same flow range (360 gpm) and low flow alarm setpoint (150 gpm). He RCS RTD bypass flow instrumentation does not provide control or active safety functions. Operator actions to mitigate the effects of an accident do not depend on the information provided by the RTD bypass flow instrumentation.

Postulated pressure boundary failures associated with the replacement transmitters are enveloped or equivalent to the postulated pressure boundary failures associated with the prior flow switches. FSAR Section 14.3,

" Primary System Pipe Rutpures," evaluates a 4" line break associated with the cold leg of the RCS. A single e ,

instrument line break through a 3/8" diameter line is within the normal makeup flow rate for the charging  ;

pumps to maintain pressurizer level without activating the emergency core cooling system. A complete rupture of both RCS RTD bypass flow instrument sensing lines is bounded by the 4" line break analysis. The conclusion is that the high head portion of the emergency core cooling system, together with accumulators, provide sufficient core flooding to keep the calculated peak cladding temperatures below required limits of 10 CFR 50.46. Therefore, adequate protection is afforded by the emergency core cooling system in the event of a small break LOCA. No other system or component pressure or structural boundaries are physically impacted by the modification. No new radiological release mechanisms or paths are created or adversely impacted as a result of the modification. De installation does not challenge safety margins in the TS. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-037 Page 69 of 112

92. M R 97 112' A, (Common), G-01/G-02 EDGs.

De modification removes the manual remote operation circuit for the G-01/G-02 EDG fire damper curtains and installs a protective grating over the ventilation ductwork. De venting consists of standard floor grating, mounted over the existing ventilation openings. An additional piece of plate steel is attached to the vent structure beneath the grating over the area where the solenoid valves used to be. His provides a new attachment point for the fusible links (previously ettached to se solenoid plunger). He old switches, cable and solenoids are removed. He existing fusible links remain in place, providing automatic actuation of the fire dampers. This removes the ability to manually or inadvertently shut the fire dampers.

Summary of Safety Evaluation: G-01 and G-02 EDG room temperature is controlled by ventilation exhaust ~

fans W-012A&B (G-01) and W-012C&D (G-02). There are louvers and fire dampers in the exhaust path of the ventilation fans where the air flow is directed out of the EDG rooms through enclosures on turbine hall El. 26'.

The fire dampers two modes of operation are remote manual and automatic. The remote manual mode of operation is actuated by closing a switch located on Unit I turbine hall El. 8' by the truck access bay. Closing the switch causes the exhaust fans to deenergize, and after a time delay, energizes solenoids, which in turn, drop the dampers in the ventilation exhaust path. The automatic mode of operation is actuated by fusible links rated at 165'F.

Removal of the manual trip circuit and installation of protective grating does not add new radiological release paths, nor does it alter existing release paths. The changes do not increase the probability of a malfunction of EDGs because they continue to perform their safety-related function during a design basis accident.

Additionally, the safety-related function of the fire dampers is to remain open. Removal of the manual switch ,

and circuit removes a potential failure mechanism, making it less likely that the dampers could shut in a design basis accident. Therefore, these changes do not increase the radiological consequences of an accident, event, or malfunction of equipment important to safety previously evaluated in the CLB.

The work occurs with only one damper shut at a time, nis maintains operability of the ventilation system.

He work plan contains proper foreign material exclusion (FME) material controls, and PMT to ensure proper installation and design control and verifies operability of the system after installation. His change does not involve a USQ nor does it r$ quire a change to the TS. (SE 98-011)

Summary of Safety Evaluation: The SE revision documents a different style of floor grating used during this work. De change is based on security barrier requirements its only effect on nuclear safety is the slight increase in ventilation airflow obstruction. In ileu of revising S&L Calculation M-09334-352-VNDG.1, ventilation testing was performed under an EDG ventilation voluntary LCO per TS 15.3.7.B.1. This air flow data is used for acceptance of the modification. Ventilation testing is performed under an EDG ventilation voluntary LCO per TS 15.3.7.B.I. This air flow data is used for f'mal acceptance of the modification. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-011-01)

93. MR 97-l13,(Unit 2), Reactor Coolant System.

The modification installs six relief valves and associated inlet and outlet piping / tubing. The valves relieve a potential thermal overpressure that could occur in RCS isolated piping sections between normally shut valves .

inside containment following, a LOCA or a main steam line break (MSLB). The sections of piping modified are in the reactor vessel level indication system (RVLIS) and postions of the RCS.

Summary of Safety Evaluation: The RCS piping sections are normally isolated when the reactor is operating at greater than cold shutdown conditions. He safety relief valves are selected to relieve excess fluid so that the subject piping and associated components are adequately protected from overpressure due to external heating during a LOC / or MSLB.

Since the instal, ation is performed in Unit 2 containment while the reactor is in cold shutdown and defueled, no operational limitations are necessary during the work. The work is performed on systems that handle the Page 70 of 112

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' reactor coolant; therefore, precautions that are associated with such actions, including FME from the systems,

- are strictly enforced. De defueled reactor is necessary because some of the valves are located at low elevations that require draining of the RCS. A pressure test and an initial service leak test are performed as acceptance for the modification. The valves, piping and tubing are properly qualified and are installed in accordance with approved codes and dtandards, including the original piping specifications. Appropriate seismic requirements are met. He structural integrity of the isolated piping is maintained by this modification and no other equipment is affected. Since the modification does not affect containment integrity, there was no consequences of an accident. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-071) 94I MR 97115,(Unit 1), Service Water.

MR 97-115 installs a hot tap fitting and controls the hot tap process on the SW return header downstream of ISW-2907/2908, the HX-15A-D containment recirculation heat exchanger emergency flow control valves (FCVs). He hot tap isolates ISW-2907 and ISW-2908 from the SW retum header so maintenance to the

. internals of each valve may be performed. ISW-2907 and ISW-2908 are 12" gate valves on the SW outlet from the containment accident fan coolers, lHX-15A-D. De valves open to provide additional flow through the CFCs during accident conoitions.

Summary of Safety Evaluations: He accidents evaluated rely on SW to mitigate their consequences except the fuel handling accident and t,he boron dilution events. The ability of the SW system to mitigate the accidents is not affected. He consequences of the accidents, events and equipment malfunctions previously evaluated are not increased. He possible failures that might be associated with this modification are limited to excessive air inlealcage or a loss of the metal coupon. The SW return header is aligned to the circulating water system on a shutdown unit to ensure that loss of condenser waterbox level associated with excessive air inleakage cannot cause an accident, malfunction or event.

The risk oflosing the metal coupon is minimized by using specialized equipment supplied by the vendor. De cutting tool uses mechanical pins to capture the coupon between the pins and the cutter blade. The vendor has rarely lost a coupon while using this equipment, even in cases where the flow characteristics were much more severe than in the service water retum header. If the coupon becomes detached from the cutter, it is retrieved through SW-2907/2908 during maintenance. Some of the debris created while cutting the 12" hole remains in the system, and falls down to the horizontal piping downstream of SW-2907/2908, where it is retrieved during maintenance. De modification does not affect the operability of the SW headers since both the interim and final configurations are seismically qualified. De ability of the SW system to perform its design basis function is not afTected by the modification. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-003)

Summary of Safety Evaluation: The SE revision allows the SW overboard to be aligned to the circulating water system or a running unit after the plug has be:n verified to be holding well, ne SW ove board must be

• ' realigned to a shutdown unit prior to removing the plug. After the plug is installed and verified to be holding, the risk of air inleakage is minimized. Condenser waterbox level alarms provide notice of significant air inleakage. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-003-01) o 95. MR 97-132,(Unit 1), Reactor Coolant.

MR 97-132 installs three relief valves and associated inlet and outlet piping / tubing, to relieve a potential overpressurization condition that could occur in isolated piping sections inside containment during post-accident MSLB inside containment or LOCA conditions. The modification installs relief valves to relieve the potential pressure buildup between the normally closed valves. De sections of piping modified are in the RVLIS between valves IRC-500J and K and CVCS upstream of valves ICV-285 and 386.

Summary of Safety Evaluntion: nc RVLIS piping is normally isolated when the reactor is operating at greater than cold shutdown conditions. The CVCS lines are nonnally isolated whenever normal letdown line and/or seal water return are available. De safety relief valves are selected to relieve excess pressure so the piping and Page 71 of112 l

l associated components are adequately protected from overpressurization as a result of external heating during a LOCA or MSLB. Installation is performed in Unit I containment while the reactor is in cold shutdown and defueled. No operational limitations are necessary during the work. However, the work is performed on systems that handle reactor coolant. Herefore, precautions associated with such actions, including foreign  !

material exclusion from the system, are strictly enforced. He defueled reactor is necessary because some of the valves are located at low elevations require draining of the RCS.

A pressure test and an initial service leak test are performed as acceptance for this modification.' ne valves, piping and tubing are properly qualified and installed in accordance with approved codes and standards, including the original piping specifications. Appropriate reismic requirements are met. De structant integrity of the itolated piping is maintained and no other equipment is affected. Since this modification does act affect

• 1 containment integrity, there are no consequences of an accident. He modification does not create a USQ nor l does it require a change to the TS.1SE 98-016)  ;

Summary of Safety Evaluation: ne SE revision addresses CR 98-3317 concerns regarding potential flow j induced vibrations. When in service, the flow through the parent piping is at a moderate velocity of I I fps or  :

less. Since no flow induced vibration has been previously identified with these lines and considering a  ;

moderate fluid velocity, the probability of fatigue as a result of flow induced vibrations is unlikely, ne ,

modification does not pose a USQ nonicas it require a change to the TS. (SE 98-016-01)  !

96. MR 97133,(Unit 1), CVCS.

MR 97-133 modifies valve CV 1299 to eliminate the possibility of pressure buildup and disc binding.

I Summary of Safety Evaluation: When the reactor is defueled the valve is disassembled and a 1/8" hol is l drilled in the upstream disc. The hole vents the bonnet cavity to the reactor coolant side, thus eliminating pressure buildup if the valv: is shut and exposed to increased temperatures. The valve is then reassembled and  !

tested for leaks as required by TS 15 A.3. The modification renders the valve operational in one direction. His j does not present a problem since the piping section is designed for one way flow from the reactor to the excess i I

letdown heat exchanger.

MR 97-133 is installed with the reactor defueled and ap propriate :adiological woric controls are implemented. i This ensures that the installition and interim plant conditions do not impact plant or personnel safety.

FSAR Figure 9.2 2 is revised to incorporate a note indicating that valve ICV-1299, although a gate valve, seals in one direction only. The modification does not pose a USQ nor does it require a change to the TS.

(SE 98-015)

97. MR 97-134,(Unit 2), CVCS.

MR 97-134 installs two relief valves and associated inlet and outlet pipinghubing in the CVCS excess letdown line upstream of valve 2CV-285 and in the RCP seal bypass return line upstream of 2CV-386. De CVCS lines are normally isolated whenever normal letdown line and/or seal water return are available. De valves relieve a potential thermal overpressure that could occur in the normally isolated piping sections following a MSLB . I inside containment or LOCA. l t

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l Page 72 of112 j l

Summary of Safety Evaluation: The new relief valves are installed in a dead leg and do not change the flow configuration of the CVCS. When the relief valves operate they relieve a minute amount of contaminated water back to the parent system. Since no new flow paths are created for uncontrolled loss of reactor coolant and operability of the parent systems is maintained, the activity does not alter functions, components or a system. Loss of RCS inventory can only occur if CVCS is in an unusual lineup that induces full RCS pressure across it. In that case, the valve is capable of releasing 0.3 gpm of water. Leakage of 0.3 gpm is readily detectable by instrumentation especially provided for this purpose, water inventory balance, or direct observation. Once the undesired leak is detected, the operator is to follow TS 15.3.1.D that leads to either resolution of the problem or reactor shutdown.

• Since the installation is performed in Unit 2 containment while the reactor is in refueling shutdown : v1 defueled, no operational limitations are necessary during the work. The work is performed on systems that handle the primary coolant, therefore, precautions that are associated with such actions, including the control of exclusion of foreign materials from the systems, are strictly enforced.

A pressure test and an inhlal service leak test are performed as acceptance testing for this modification. The valves and pertinent piping and tubing irtstalled are properly qualified and installed in accordance with approved codes and standards, including the original piping specifications.

When in service, the flow through the parent piping is at a moderate velocity of 11 fps or less and that no flow ,

induced vibration has been previously identified with these lines. Therefore, the probability of fatigue failure due to flow induced vibrations is considered unlikely. The structural integrity of the isolated piping is maintained by the modification and no other equipment a affected. Since the modification does not affect containment integrity, there are no consequences of an accident.

The modification provides CVCE .ad its components with increased reliability / survivability during a LOCA or an MSLB accident by protecting them fron. overpressure conditions. The modification does not pose a USQ j nor does it require a change to the TS. (SE 98-178) '

98. MR 97-135,(Unit 2), CV lS.

l MR 97-135 modifies valve 2CV-1299 to eliminate the possibility of pressure buildup within the valve bonnet and disc binding. 2CV-1299 is disassembled and a 1/8" hole is drilled in the upstream disc. The hole vents the bonnet cavity to the RCS side of the valve when the valve is shut and exposed to increased temperatures. The valve is reassembled and leak tested in accordance with TS 15.4.3.

Summary of Safety Evaluation: When a double disc shut valve is exposed to a pressure differential, it causes the disc toward the pressure to move slightly away from the seat creating a path so the bonnet cavity becomes filled with high pressure fluid. The pressurized fluid then acts on the downstream disc forcing it towards its seat, assuring the valve tightness within the system. The only barrier toward the atmosphere for the pressurized fluid within the bonnet cavity is the valve packing. Therefore, it can be concluded that the valve packing is the only part of the valve that seals toward the atmosphere every time it is exposed to a pressure differential.

Dnlling a hole in the upstream disc changes nothing in the valve function except now the fluid does not lift the upstream disc from its seat to pressurize the bonnet. This method of preventing pressure locking is recommended in NUREG-1275, Volume 9," Operating Experience Feedback Report - Pressure Locking and Thermal Binding of Gate Valves." He modification does not change the original design function or operation of the valve. To ensure that the valve discs are correctly reinstalled in the future, guidance is added to the appropriate maintenance procedures and the design documents are upgraded accordingly.

MR 97-135 can be installed any time the reactor is drained down below El. 43' (less than 68% RVLIS). Plant procedures establish the RCS level when the modification is to be performed. If the modification is performed at other than defueled conditions, the requirements of OP 4F," Reactor Coolant System Reduced inventory,"

must be considered. A pressure test and an initial service leak test are performed as acceptance testing for this modification. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-116)

Page 73 of 112

99. MR 98-010*A and MR 98-010*B,(Common), PAB Ventilation.

MR 98-010 alters the original design of the PAB ventilation to reduce vibration and to eliminate unexpected failures of the W-30A&B PAB exhaust filter fan units caused by design deficiencies. De work includes fan frame mounting, motor frame mounting, and replacement. The replacement of belts, bushings and sheaves in accordance with manufacture /s recommendations and/or calculations prepared for the modification.

Summary of Safety Evaluation: Only one fan is worked at a time so unit availability is not compromised.

Proper operation of the single fan unit is verified prior to declaring a fan unit out of service in order to ensure that a single unit remains av'ailable for operation as required by the CLB. ,

The modified mounting configuration eliminates failure of the W-30A&B fan frame welds and frame bolting.

Elimination of vibration isolators does not adversely affect other equipment. The transmission of vibration between W 30A&B and other equipment is not a concern because of the PAB building slab thickness and relative equipment location.

The W-30 fans are not credited as an accident mitigator in FSAR Chapter 14. In addition, the operation of the W-30 fans described in the CLB remains unchanged As a result, accident consequences described in the CLB remain unchanged. The modification does not pose a USQ nor < foes it require a change to the TS. (SE 98-065) 100.MR 98-012*A,(Unit 1), Service Water.

MR 98-012* A reduces the hydraulic losses and potential sedimentation problems in the CFC SW piping and installs instrumentation that provides a more accurate measurement of SW flow through the CFCs than the old differential pressure measurement system. He results ensure that SW flow through the CFCs meet the design basis.

Summary of Safety Evaluation: MR 98-012*A replaces portions of the old carbon steel SW piping to the four Unit 1 CFCs with stainless steel piping of the same length, nominal diameter and schedule. It also removes old differential pressure flow indicators 1DPI-2804,1DPl.2904,1DPI-2905, and 1DPI-2906, along with their corresponding tubing and supports. The high pressure side root isolation valves for these indicators are removed and the low pressu,re side root isolation valves are left in place with a cap.

The systems affected by this change, containment accident recirculation (CAR) and SW, are used for accident mitigation. Since the modification does not affect the piping flow capacity nor its pressure integrity, there is no increase in the probability of failure to these systems. Also, the CAR system ability to remove heat from containment following an accident is not reduced since there is no adverse affect to the CAR or SW systems.

In fact, replacement of the old pipe improves the system since removing internally fouled piping restores the flow capacity. The use of the new pipe limits future biofouling in these segments ofpipe. The modification '

does not pose a USQ nor does it require a change to the TS. (SE 98-058) 101.MR 98-012*B,(Unit 1), Service Water.

MR 98-012* B installs instrumentation that provides a more accurate measurement of SW flow through the CFCs than the old differential pressure measurement system. The installation provides an improved means of verifying SW flow through the CFCs to ensure that the CFC design basis is met. De flow rate range in which the SW flows to the CFC set is unchanged by the modification. The flow rate range remains consistent with the SW flow model assumptions.

Summary of Safety Evaluation: The modification is installed during UIR24. Work is performed when the reactor is in cold shutdown, refueling shutdown, or is defueled (no fuel in the reactor and no fuel movement), j These modes or conditions do not require the CAR system to be operable for accident mitigation.

l Page 74 of112

1 ne systems affected by this change, the CAR and SW systems, are used for accident mitigation. Since this modification :loes not affect the flow capacity, pressure integrity or electrical and control systems, there is no increase to the probability of failure of these systems. Also, the CAR system ability to remove heat from the I containment following an accident is not reduced since there is no adverse affect on the CAR or SW systems.

installation of the ultrasonic flow instrumentation improves the system since flow measurements are more {

j accurate. This provides better data for setting flow through the CFCs. De modification does not pose a USQ nor does it require a change to the TS. (SE 98-062) ,

Summary of Safety Evaluation: The SE revision accommodates completion of equipment installation with Unit 1 RCS temperature above 200*F and clarifies the acceptability of the ultrasonic couplant used to mount flow

• transducers to the SW piping. Acceptance testing for this modification does not require the manipulation of ,

plant systems and does not affect the operability of the CAR system. Installation of the hardware is performed  !

when the reactor is in cold shutdown, refueling shutdown, or is defueled (no fuel in the reactor and no fuel l

movement). The transducer couplant installation and subsequent modification acceptance testing may be  !

performed in any plant mode. The modification does not affect CAR system operability. The couplant used j between the transducers and the piping is applied in accordance with NP 3.1.1, " Chemical Co stamination Control for Corrosion Resistant Alloys," to assure that the potential for adverse effects on the piping is i minimized. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-062-01)  ;

102.MR 98-014, (Unit 1), Part-Length (P/L) Control Rod Drive Mechanisms (CRDM).

The modification removes the P/L motor housing and installs a threaded and seal welded head adapter plug. l The abandoned head penetration is capped using a threaded and seal welded head adapter plug similar to those j

used at other spare head penetrations with exception of the seal weld design. Other spare penetrations utilize a canopy weld design, while this installation uses a fillet seal weld design. He fillet weld is used for case of  !

installation as well as a history of successful use in this type of application.

Summary of Safety Evaluation: CRDM cooling air baffles and rod travel housing seismic support plates are fabricated and installed to ensure proper air flow and seismic support requirements. He cooling air baffle cans, or " dummy cans," are required to maintain the head cooling air flow balance in the current design condition for cooling flow in this area, he seismic plates are required to provide lateral support to the full length CRDMs and to restore the stiffness to the seismic support.

A calculation verifies the suitability of the installation of the new seal weld on the head adapter plugs. The calculation verifies the seismic and pressure boundary integrity of the installation following removal of the P/L CRDM and replacement with a head adapter plug. RCS flow analysis is performed with the P/L CRDMs removed from the upper guide tubes. The analysis provides the flow and temperature effects of this change on the core bypass flow, upper head temperature, and various reactor system pressure drops. A visual inspection of the fillet seal welds is performed during the RCS leak test at not less than 2085 psig during startup. Analyses show that this modification does not adversely affect Unit 1. His work leads to a decreased probability of leakage from the RCS pressure boundary at the location of the P/L CRDM housings. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-060)

. 103.MR 98-021,(Unit I and 2), RWST/RilR.

MR 98-021 supports upgrades to the AC-151R-3/4/5/8 RWST/SFP recirculation loop small bore piping as well as the Units I and 2 RWST and RilR cross connect branches from the loop. De modification upgrades the loop piping to Seismic Class 1.

Summary of Safety Evaluation: The loop piping and asrociated branch lines are able to withstand the effects of postulated seismic events and remain Code complaint for design basis seismic loads and other design basis loads. He modification ensures that the operability of the existing piping is not compromised during the installation.

Page 75 of112

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Although no functional changes to systems result from this modification, the piping support upgrade expands the Seismic Class I boundaries to include piping that is presently Seismic Class 3. Since there is no functional <

change to the facility, neither the probability nor consequences of accidents or equipment malfunctions are ,

affected. He pipe support upgrades result in an expansion of the seismic boundaries of the facility and changes to the FSAR are made. The modification does not pose a USQ nor does it require a change to the TS.

(SE 98-103) 104.MR 98-037* A,(Unit I and 2), St.

SI-885A&B check valves are currently installed as a means of pressure lock prevention for double disk gate valves SI-852A&B. The existing configuration vents excessive bonnet pressure to the header downstream of SI-852 via 3/4" pipe and SI-885 check valves. In order to ensure the SI-852 valves are capable of opening and meeting timing requirements, proper operation of the SI-885 valve is required. nc 1986 edition of ASME Section XI IWV requires check valves with a safety-related function be tested or actuated quarterly. He existing configuration does not provide a practical means of performing fimetional testing of SI-885.

Summary of Safety Evaluation: ne new pipe installation is in accordance with the original specificatMn, to Seismic Class I requirements. The new valves are Anchor / Darling 1878# Class 3/4" globe valves, that exceed design temperature and pressure requirements of the system. / Ithough valve NDE is not performed as described in FSAR Section 6.2.2, the new valves are manufarured to ASME Class I per 1983 ASME Section til with Summer 1983 Addenda, less N-Stamp. Installation occurs with the unit in a cold or refueling shutdown, with SI-852A&B headers drained down and isolated. FSAR Figure 6.2-1 is updated, to reflect the modified configuration. Return-to-service leak testing is performed using routine alignment of plant equipment, with DHR capabilities as required by TS 15.3.1. ,

The modification does not pose a USQ nor does it require a change to the TS. (SE 98-176) 105.MR 98-049,(Unit I), SI.

MR 98-049 installs modified motor operators on valves IRH-700,1SI-841 A and ISI-841B. The slower motors preclude motor shaft damage due to high inertial forces.

Summary of Safety Evaluation: Work on the modified operators is performed by using state-of-the-art techniques that are either equal to or better than the original techniques. The parts are obtained from the same manufacturer and meet tolerances, material quality and compatibility of the original parts. Increase in the valve actuator weight was evaluated for its seismic impact on the existing systems and structures. It was concluded that there are no seismic concerns caused by the modification. MOV overload Calculation P-94-017 provides the existing actuator, such as motor size, actuator gearing, gearing orifices, and the equation necessary to detennine the effects the new motor gearing has on the actuator capabilities. MOV overload sizing Calculation P-94-004 provides the methodology for sizing the new thermal overloads to be used with the new motor-gearing.

The modification is installed in Unit I containment when the reactor is defueled. De defueled reactor is necessary because RHR must remain operable if the fuel is in the reactor. Since work is performed on systems -

that handle reactor coolant, precautions that are associated with such actions, including exclusion of foreign materials from the systems, are strictly enforced. A pressure test, an initial service leak test and valve stroke test are performed as acceptance for this modification. He modification does not pose a USQ nor does it require a change to the TS. (SE 98-066)

Page 7(, of 112

l06.MR 98-069, (Common), G-05 Combustion Turbine.

MR 98-069 installs an overfrequency relay for monitoring the output of the gas turbine and provides a trip signal to the combustion turbine generator output breaker and field breaker when the frequency limit is exceeded for a specified period of time. The overfrequency relay includes an alarm output to the combustion turbine local annunciator panel at a lower frequency than the breaker trip frequency setpoint. Relay final connections and PMT is performed while G-05 is out of service.

Summary of Safety Evaluation: The modification affects the operation of G-05. The combustion turbine is not an initiator of accidents or events previously evaluated in the CLB. G-05 overfrequency could result in

, overspeed of connected safety-related motor loads, resulting in overload trip of the associated supply breakers.

The new overfrequency relay prevents tripping of safety-related loads caused by combustion turbine overspeed.

Tripping of G-05 on overspeed allows safeguards loads to be sequenced onto the EDGs, which is the normal

• response to the loss of offsite power. Installation controls and periodic calibration of the overfrequency relay ensure the required 95% reliability of G-05 is maintained.

The modification does not affect the design or functions of safety-related components or systems. Therefore, it does not affect the consequences of malfunction of safety-related equipment or the ability of safety-related components to mitigate the cons:quences of an accident or event. G-05 gas turbine is used as an Appendix R power supply and station blackout power supply. It is required to maintain 95% reliability and can power loads within one hour of a station blackout. The modification does not adversely affect these requirements.

The modification does not affect system operating margins or redundancy of safety-related systems or components. TS 15.3.7.A.I.b and 15.3.7.B.I.b that place limits on reactor operation when one or both j 345/13.8 kV auxiliary transformers are out of service when the gas turbine is in operation are unaffected. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-100) 107.MR 98-074,(Unit 1), Reactor Vessel.

The modification addresses the design change associated with the new acceptable bolting pattern and the new bolt design of the RV bame plates, it is not intended to encompass the means and methods of the inspection and bolt replacement as these issues are addressed separately.

Summary of Safety Evaluation The new bame-former bolt design as presented in WCAP-15133,

" Determination of Acceptable Bame-Barrel Bolting for Point Beach Units I and 2,"is functionally equivalent to the original design. The plant specific WCAP is based upon WCAP-15029, Westinghouse Methodology for Evaluating the Acceptability of Bame-Fonner-Barrel Bolting Distributions Under Faulted Load Condition" and approved NRC SER " Safety Evaluation of Topical Report WCAP-15029 " Westinghouse Methodology for Evaluating the Acceptabilty of Baffle-Former-Barrel Bolting Distribution Under Faulted Load Conditions,"

• dated November 10,1998. The new bame-former bolt design includes a new acceptable bolt pattem, leaves bolts in place with indications at non-critical locations, and addresses the bolt material.

Bame-former bolts are not considered an accident initiator. The new baffle former bolt design and acceptable bolting pattern have been determined to maintain the functionality of the lower internals. The change does not pose a USQ nor does it require a change to the TS. (SE 98-112) 108.MR 98-082,(Unit 2), Reactor Coolant System.

MR 98-082 replaces the existing reactor coolant loop RTD bypass flow indicating alarm switches 2FIA-458 and 2FIA-459 with new Rosemount Flow Indicating transmitters, loop power supplies, and alann units. The modification provides a more reliable and accurate flow indication system with improved calibration features to monitor RTD bypass flow.

Page 77 of 112

Summary of Safety Evaluation: Unit 2 is defueled or in cold shutdown, and depressurized condition with 2P-1 A and 2P-IB RCPs secured during the installation and post maintenance testing portion of the modification. Each RTD bypass flow instrument loop is deenergized and the associated instrument root valves isolated during inst,alation. New cable routed is meggered and umtinuity tested as documented on the l associated cable pull tickets. j l

The new flow loops perform the same non-safety-rdated function, as the existing flow instrument loops and  !

provide the same flow range and low flow alarm setpaint. The RTD bypass flow instrumentation does not {

provide control or active safety function. Operator actions to mitigate the effects of an accident do not depend j on the information provided by the RTD bypass flow instrumentation. Postulated pressure boundary failures i associated with the replacement transmitters are enveloped or equivalent to the postulated pressure boundary

. l failures associated with the existing flow switches. l FSAR Section 14.3," Primary System Pipe Ruptures," evaluates a 4" line break associated with the cold leg of  !

~

the RCS. A single instrument line break through a 3/8" diameter line is within the normal makeup flow rate for t the charging pumps to maintain pressurizer level without activating the emergency core cooling system. A complete rupture of both RC RTD bypass flow instrument sensing lines is bounded by the four accumulators, provide sufficient core flooding to keep the calculated peak cladding temperatures below required limits of 10 .

CFR 50.46. Therefore, adequate protection is afforded by the emergency core cooling system in the event of a {

small break LOCA. No other systems or components, pressure or structural boundaries are physically impacted .j t

by this modification. No new or existing radiological telease mechanisms or paths are created or adversely impacted as a result of this modification. De installation does not challenge the existing safety margins found I in the TS. De modification does not pose a USQ nor does it require a change to the TS. (SE 98-123)  !

109.MR 98-093*B,(Unit 2), Containment Electrical Penetrationsi j i

ne modification removes the purge valves and associated tubing from the monitor rings of the following j containment electrical penetration assemblies (EPAs) and installs a welded pipe cap over the hole on the .

monitor ring left by the removed valve and tubing: Q-2, Q-3, Q-4, Q-5, Q-6, Q-7, Q-8, Q-9, Q-10, Q-II, Q-12, Q-13, Q-14, Q-16, Q-17, Q-18, Q-19, Q-23, Q-24, Q-25, Q-26, Q-27, Q-38, Q-40, Q-2, Q-47, Q-49, Q-51, j Q-53, Q-55, Q-56 and Q-57. The cap is fabricated from S A-105 carbon steel.  !

t Summary of Safety Evaluation: Equipment important to safety is not adversely affected by the change. De l pipe cap meets the requirements of ASME Section XI, Subsection IWE and the original design requirements of l ASME Section ill,1965.' The pressure and seismic integrity of the EPAs are not adversely affected by the  !

modification. Removing the valves and installing the pipe caps improves the pressure integrity of the EPAs. )

ne processes used to remove the existing valves and install the new pipe caps does not damage the conductors  ;

in the EPAs. The EPAs are purged with dry nitrogen to remove moisture from its internal components such  !

that the manufacturer's dryness requirements are maintained. The control room is notified prior to welding so j operators are aware of potential false data and alarms. Equipment that could be actuated or . damaged as a result ,

of the false signals is secured.

The containment system is used for accident mitigation. Since the modification does not adversely affect the containment pressure boundary, there is no increase in the probability of failure of this system. Electrical and

~

instrumentation functions remain unaffected. Removal of the valves and installation of the pipe caps ensures no adverse affect on containment integrity since the possibility of a valve leakby or a valve left open is eliminated. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-110)

Summary of Safety Evaluation: The SE revision reflects that EPA Q-2 is added to the scope of MR 98-093*B.

The scope addition is required because the intent of MR 98-093*B is to remove the nitrogen purge valves that are installed on the containment side of Westinghouse canister type EPA that has a nitrogen purge valve installed on its containment side. De modification does not pose a USQ nor does it require a change to the TS.

(SE 98-110-01)

Page 78 of112

l 110.MR 98-109,(Unit 2), Reactor Internals Lifting Rig.

f i

MR 98-109 replaces the existing carbon steel platform on the Unit 2 reactor internals lifting rig with a new I platform fabricated of stainless steel. De existing Unit 2 platform is similar to the platform on the Unit I reactor internals lifting rig, that was identified during UlR24 having an unqualified coating. He existing  !

Unit 2 platform is assumed to have the same unqualified coating. I Summary of Safety Evaluation: Removal of the existing platform and installation of the new platfonn takes place while the lifting rig is positioned at containment El. 66.' Unit 2 is in either the cold shutdown or refueling  ;

shutdown while installation work is performed. No work is performed while the lifting rig is in or over the reactor cavity.

Neither the existing platform nor the replaument platform have an adverse impact to the reactor intemals lifting rig. De previously established design margins that are required by NUREG-0612," Control of Heavy l

Loads," ANSI N14.6-1978 and the AISC Manual of Steel Construction are still met. Since the plationns do not l

interfere with or affect other plant systems or equipment, no new accidents or malfunctions are created and the -

i consequences of previously evaluated accidents and malfunctions are not increased. De modification does not l pose a UsQ nor does it require a change to the TS. (SE 98-170) '!

Ii1.MR 98-113,(Unit 2), RHR/SI. I i

I i

The modification replaces the 3600 rpm motor operators with 1800 rpm motor operators and an actuator gear  !

ratio change (currently 34.941:1 to 55.846:1). De modifications result in a significant reduction in inertial j loading at the motor shaft by slowing the valve stroke from 10 seconds to approximately 30 seconds while maintaining an equivalent motor torque (60 ft-lb). The self-locking characteristics of the worm gear set are maintained with the new gearset, j

- Summary of Safety Evaluation: De only functional or operational change is a minor increase in valve stroke '

time for the valves. Design basis stroke time requirements are not defined for the valves and minor increases in valve stroke time has no impact on the plant ability to operate safely or to respond to transients. The modification improves the overall reliability and availability of the valves by eliminating the potential for motor .

. shaft cracking and related maintenance activities. I e

There are no seismic implications associated with this change since the new components weigh approximately l

80 lbs. less than the original components. His reduction in mass does not impact the structural piping analysis  ;

since the net difference in mass is well within the assumed tolerance for the subject valves. His modification  ;

does not pose a USQ nor does it require a change to the TS (SE 98-134) i i 12.MR 98-116, (Unit 2),4160 vac System.  !

I MR 98-116 replaces the 4160 vac air magnetic breakers in buses 2A-01,2A-02,2A-03,2A-04, and 2A-05 that

. were not previously replaced by SPEED 97-019. Non-safety-related vacuum breakers were procured and _

l installed in 2A-01,2A-02,2A-03, and 2A-04. Qualified vacuum breakers were procured for installation in bus l

2A 35. The air magnetic breakers are replaced while loads are de-energized or while loads are supplied from I an alternate source.

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Page 79 of i12 I

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I Summary of Safety Evaluation: The new vacuum breakers are an improvement in design and technology over the air magnetic breakers. The new breakers are less subject to wear and are more easily maintained than the I old breakers. For these reasons, the failure rate of the new vacuum breakers is expected to be as low as or  !

I lower than that of the air magentic breakers. Herefore, the probability of the loss of power to safety-related equipment supplied by the vacuum breakers is not increased. Testing and qualification of the breakers ensures that no common failure modes exist. Herefore, installation of the vacuum breakers does not introduce new accidents nor do they increase the consequences of an accident. He vacuum breakers are manufactured to IEEE C37.59-1991 specifications and are seismically qualified by testing and analysis in accordance with IEEE 344-1987.

The replacement vacuum breakers have epoxy bushings that are less susceptible to moisture absorbency rod ,

chemical degradation than the ceramic bushing insulation on the DH air magnetic breakers, resulting in a lower probability of primary insulation failure with the new breakers. The new breakers are as reliable as the original breakers in performing their design functions under normal and accident conditions. He primary failure mechanism is unchanged for the existing and replacement breakers, namely, the inability to interrupt a current are during contact separation. The failure in vacuum applications is localized to the contacts withic. the vacuum interrupter enclosure compared to a more catastrophic failure for air circuit breakers, nerefore, a fallure of a new breaker results in less severe consequences than the failure of an existing breaker. Other breaker failure modes including failure to open or close when required result in identical consequences to failures of the existing air magnetic breakers.

De safety-related functions of the new breakers are the same as the original. Control logic functions remain the same. De vacuum retrofit is mounted on an original air magnetic breaker truck which allows proper interface with the original cell interfaces. The breaker racking mechanism interlocks are maintained to ensure that a breaker is not repositioned in the cubicle while the main contacts are closed. Shutter operation is accomplished in the same manner as with the original breaker / cell interface. The major difference between the new and existing breakers is that the replacement breaker contacts are completely enclosed in a sealed vacuum enclosure while the existing air magnetic breaker contacts operate in ambient air accompanied by a blast of air from a puffer assembly. Due to the limited ion particles to support the arc plasma and the rapid dielectric recovery rate of vacuum technology there has been an industry concem that use of vacuum switchgear can result in large transient overvoltages during switching operations. EPRI has performed an in-depth study of this phenomenon and has determined that the vacuum interrupters do not cause new failures in the equipment supplied by them. The vacuum breakers do not introduce new breaker failure modes. The modification does not pose a USQ nor does it require a change to the TS. (SE 98-111) i 13.MR 98-123,(Unit 2), Turbine Building Crane.

The modification provides a source of power to the turbine building crane that is not routinely isolated during refueling outages. De crane presently is powered from 2B-04, breaker 2BS2-308, as shown in FSAR Figure 8-9,"480V One Line Diagram." 2B-04 gets down-powered during ORT 3 during each refueling outage.

• Because of this, power to the crane is isolated for a day or two during the performance of ORT 3. l The modification supplies power to the turbine bu,lding crane from power panel PP-54, which gets its power i from altcmate shutdown bus B-09, breaker B52 e /C. The disconnect switch for the crane is also moved from -

l I

its location in the southwest corner of the work control center to a location above the work control center.

Summary of Safety Evaluation: The final configuration is an improvement over the current configuration with respect to its effect on 2B 04. A fault in the turbine building crane cable (which is routed in the turbine building) challenges the c vercurrent protection on the crane supply breaker, and coordination between this breaker and the 2B-04 supply breaker to prevent losing B-04 bus. B;cause the crane in a non-linear load, moving the turbine building crane from 2B-04 to PP-54 will slightly decrease the total harmonic distortion on 2B-04 and slightly increase the total harmonic distortion on B-08/B-09. His has no impact on the loads on 2B-04. The increase in harmonics on B-09 is not expected to have an effect on loads connected to B-08 or B-09. He only potentially sensitive load is the backup power supply and sync signal for the instrument bus Page 80 of 112

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3 inverters, which is powered from B-09 via power panel PP-54 and transformer XY-08. The XY-08 instrument bus ahernate source transformer is a regulating-type transformer that filters out harmonics; preventing it from

" affecting the instrument buses. The modification does not pose a USQ nor does it require a change to the TS.

(SE 98-178)

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

r e, ,

TEMPORARY MODIFICATIONS -

1

. The following temporary modifications (TMs) were implemented in 1998:

1 E< l. TM 97-013,(Unit 1), Safety injection System (SI); Refueling Water Storage Tank (RWST) Silica Removal.- l The TM addresses the removal of silica from 1&2T-13 refueling water storage tanks (RWSTs). De 'IM is installed on one unit at a time while the unit is in cold shutdown.

( Summary of Safety Evaluation: Applicable work plans maintain adequate shutdown margin during refueling .,

and cold shutdown and prevent a dilution event, by having Chemistry and Operations ensure that the RWST is

• J

. isolated from the RCS if RWST boron concentration falls below 1800 ppm. De work plans include direction for Operations to pump the fa9ade sump to the waste hold up tank (WHUT) should it contain radioactive j leakage from the reverse osmosis (RO) system. The TM does not pose a USQ nor does it require a change to j

~

the TS. (SE 97-210-01) 'i Summary of Safety Evaluation: Since each RWST is processed only while c unit is in cold shutdown, the >

volume and boron concentration required by the TS need not be maintained during processing ar long as  :

another source of borated water is available for injection to the core (e.g., a boric acid stwage tar k) when fuel is  !

I in the reactor. When the core is offloaded, boric acid flow paths are not required and 1800 ppm boron i requirement does not apply' when the core is offloaded. (SE 97 210-02) _;

i 2." TM 97-013,(Unit 1), Sl; Temporary Installation of Contractor Equipment for Radioactive Liquid Waste .;

Processing.

Processing of the liquid in the T-19 waste holdup tank (WHUT) via the HX-40 blowdown evaporator is not possible when ooth units are shut down so an alternate means ofprocessing the waste liquid is needed. A TM >

allows the use of contractor-operated RO equipment for processing waste liquid contained in the WHUT. De 1 power supply and positioning of the equipment remains the same as described in SE 97-210-02. He RO j system is configured so it receives water from either the RWST or the WHUT and sends the permeate to either ,

the RWST or T 104A/B waste distillate tank.

i Summary of Safety Evaluation: The RO system does not change the overall function of the waste disposal  !

system. De TM does not cause the liquid waste disposal system to become an initiator of a previously analyzed accident or equipment malfunc: ion. Existing heavy load lifting procedures are used to transport j materials associated with the TM over the spent fuel pool (SFP), and the RO system equipment is not located [

near safety-related components. Therefore, the movement and location of the TM does not increase the ,

t probability of a malfunction of the SFP or other SSCs. Other potential causes of accidents or malfunctions are flooding seismic events, and fires. Fiooding is addressed in our response to INPO SER 50-84 and Supplement 1,"Intemal Flooding of Power Plant Buildings." Possible hose failures in the PAB as well as a f failure of the temporary process feed tank or a demineralizer pressure vessel in the PAB are bounded by the flooding evaluation. Since the waste liquid system piping to which the temporary RO system is connected is ,

not seismic and since the RO system components are not located near any safety-related equipment, the fact i that the RO system is not seismic is acceptable. Hoses running between the RO system and plant piping are ,

restrained sufliciently so they cannot become seismic rcissiles during an earthquake The FPER does not allow l

. intervening combustibles in the PAB during unit operation. However, the hose arrangement is acceptable for  !

this applica: ion since the hose is only routed through the open hatches when both units are in cold shutdown. ,

The effects on MCC 18-31 by the RO system were previously addressed.

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The electrical aspects of the RO system installation do not increase the probability of occurrence of a malfunction of equipment. The TM does not cause an increase in offsite dose because spills from equipmmt in the PAB are collected in the WHUT via Boor drains and because the processed WIIUT water will be sampled, analyzed, and discharged according to normal practice using approved PBNP procedures. Although reverse osmosis is not included in the TS 15.7.1 Pan I list of components and devices used to reduce liquid radioactive material in effluents, RO is an industry-accepted method of processing radioactive liquid wastes and its use at PBNP will not reduce the margin of safety defined in the TS.

The TM does not increase the probability of occurrence of an accident, event or malfunction of equipment important to safety nor will it reduce the margins of safe *y. No new release paths for radioactive maGrial are created that may impact the safety and the health cf the public. Therefore, use of the RO system to process radioactive liquid wastes does not involve a USQ nor does it require a change to the TS. (SE 98-040)

~ Summary of Safety Evaluation: The SE revision allows routing of hoses between elevations in the PAB when a unit is above cold shutdown. The change is needed because the blowdown evaporator may not be operational when a unit is above cold shutdown (e.g., in hot shutoown), yet radioactive liquid wastes will continue to accumulate in the WHUT and may reach a level such that processing is required. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-040-01)

3. TM 98-006, (Unit 1), Lube Oil System; Turbine LO Flush Electrical Support.

i The TM installs and removes temporary test instrumentation required to support the performance of PBTP 85,

" Turbine Perfonnance Test, Unit 2."

e Summary of Safety Evaluation: Temporary main generator metering instruments are connected to existing plant instrumentation at interface points. He associated equipment is specifically designed to be used without interruption of the current transfer (CT) or potential transfer (PT) signals and is commonly used during relay replacement and calibration l The generator output metering and protective relaying circuits are not safety-related and are used in the electrical hydraulic (EH) system as input to the load drop anticipatory circuit  !

(closure of the governor valves via the overspeed protection circuitry if a ~30% load mismatch is sensed i

between MW output and crossover pressure and the main generator breakers are open). The high accuracy

]

transducer is a passive element in the circuit and does not produce feedback signals to other portions of the metering or protective circuitry such that a protective function could be disabled or a malfunction of any '

equipment important to safety could occur. The TM does not pose a USQ nor does it require a change in the TS. (SE 98-064)

4.  !

TM 98-008, (Unit 2), Containment Spray; Installation and Removal of the Temporary Na24 Injection System l for Unit 2 Moisture Canyover Testing

, The TM installs and removes the temporary Na24 injection system used during PBTP 086," Moisture Carryover Testing, Unit 2."

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• Summary of Safety Evaluation: The temporary injection system is connected to the main feedwater pump suction header at 2CS-137A. Demineralized water is supplied to the temporary system from either 2HA-148 i near the secondary chemical injection skid or at WT 534 in the water treatment building. TM 98-008 connects the temporary injection 1/4" tubing to 2CS-137A via an appropriate reducer. The 1/4" tubing from the

)

i temporary injection system is fitted with an isolation valve and a check valve between the system isolation and )

2CS-137A. The check valve maintains the feedwater system integrity in the even the temporary pump is not  !

running or a leak develops in the temporary injection system. The temporary injection system is isolated from l the feedwater piping except when in use. Following prefabrication, each of the 1/4" tubing sections are  ;

pressure tested at the maximum pressure expected prior to installation into the temporary system. Once the '

system is installed, the system is leak checked using demineralized water at normal operating pressures.

Page 83 of 112

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t j- %c area is surveyed for potentialleaks prior to installation of the temporary injection system. Instructions are

' prov'ided to protect the floor area under and around the temporary injection system to prevent contamination of

' the floor in the event of a leak from the temporary system.

1 Following injection of the Na24, the temporary injection system is flushed for an hour to reduce the radiation and contamination levels. The Na24 is allowed to decay off(90-150 hours) befm the temporary injection system is dismantled. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-075)

5. TM 98-016,(Unit 1), VNCC; Flush Containment Accident Fan Cooler (IHX-015DI) to Remove Flow ,

J Restrictions in the inlet Plenum.

~

TM 98-016 removes the supply and return connections of I HX-015DI to enable a cross-connect to be l f established. His cross-connect, tying the supply header to the retum connection of the heat exchanger and the

< j i retum header to the supply connection on the heat exchanger, allows a backflush of the coil in order to remove j

$ accumulated debris (zebra mussels) from the inlet plenum. In addition, a strainer is attached to the return i header connection to capture material dislodged by the flush to ensure it is not retumed to the system. ];

Summary of Safety Evaluation: EX-015DI removes heat from the containment following a LOCA or MSLB }'

in order to limit containment pressure and temperatures to less than the containment design !imits. The fan and ,

its associated heat exchangers are not mentioned in CLB documents as initiators of accident scenarios, ne TM j has no adverse effect on the intended function of the unit, but may enhance its capabilities. De fan and coolers  !

are required to be operable during power operatien. Maintenance is performed during refueling or cold {

shutdown; therefore, there are no unforeseen challenges to its ability to perform its intended design functions. l There are specific instructions included in the TM and work plan to restore the cooler to its original conriguration. Post-maintenance SW flow testing is performed. He TM does not pose a USQ nor does it require a change to the TS. (SE 98-072) ,

TM 98-017,(Unit 1), Fuel Handling; Change Location of the Frame-up Upender Switch. i 6.

The proximity switch to stop upender motion is mounted underwater. This proximity switch has failed and is ,

replaced with a new switch that actuates off of the upender cable travel. De switch is set to actuate i approximately l" before the previous back-up switch actuated. The new switch stops upender motion and provides frame-up light indication.- Two switch redundancy will be maintained. His new switch is only temporary until plant conditions allow the proximity switch mounted in the cavity to be replaced. ,

Summary of Safety Evaluation: The upender system is provided with a spring loaded kicker assembly. As the  !

upender reaches the frame-up position, the kicker assembly is compressed. The kicker allows the upender to  !

move when no fuel assembly is mounted in the upender. The kicker assembly also provides a cushioned stop  ;

vchen the upender reaches the frame-up position. De cable and pulleys are rigidly mounted and attached, so by

• monitoring rable position the upender position is effectively monitored.

{

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The electrical circuit and its features remain unchanged by this TM. Appropriate operation of the upender equipment is verified by PMT. De PMT includes verification that both upender frame-up switches are functional, and Operations removes and installs a dummy fuel assembly in the upender. He TM does not pose  ;

a USQ nor does it require a change to the TS. (SE 98-061;  !

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7. TM 98-019,(Unit 1), Service Water; Backflush IHX-015CS Containment Fan Cooler to Remove Zebra Mussel Shells.

The TM removes the supply and return connections for llX-015CS to enable a cross-connect to be established.

His cross connect, tying the supply header to the retum connection of the heat exchanger and the return header to the supply connection on the heat exchanger, allows backflush of the coil to remove accumulated debris (raussel shells) from the inlet plenum. A basket strainer is installed in the return line to collect debris that is flushed. The TM is installed while Unit I is in cold shutdown and containment integrity is not required.

Summary of Safety Evaluation: The fan and its associated heat exchangers are not mentioned in CLB

• documents as initiators of any accident scenasios. The TM has no adverse efTect on the intended function of the unit. The fan and coolers are required to be operable is during power operation. Maintenance is performed during refueling or cold shutdown. Failure of the temporary connections would not divert enough SW flow to adversely affect other safety-related SW loads.

The TM allows the inlet plenum and tubes on liiX-015C5 to be backflushed separately from the remaining cooling coils on the fan. This ensures removal of potential flow restricting debris. Removal of the debris enhances the performance of the unit. Specific instructions are included in the TM and work plan to restore the cooler to its original configuration and post maintenance SW flow testing is performed. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-088) 8.

TM 98-029, (Unit 2), CV, Reactor Coolant Pump "A" No. I Seal Spiking licise Gauge Installation.

The TM gathers data to help analyze the cause oflow flow spiking on Unit 2 RCP "A" when the gas stripper is offline. The connection is made from the low side pressure test connection at PT-173, RCP "A" No. I seal DP transmitter. This shows if there is a pressure spikejust downstream of the No. I seal when the seal return flow spikes low. Connection ofrecorders on PT-131, labyrinth seal DP and PT-139, volume control tank (VCT) pressure test connections in the control room. These test points are the outputs from the pressure transmitters.

Connection is non-intrusive to the system. This shows if there is a pressure spike in the VCT or at the labyrinth seal when seal return flow spikes low.

Summary of Safety Evaluation: A licise gauge and associated flex hose and fittings are connected to the test connection of PT-173 RCP "A" No. I seal DP to monitor seal return pressure. There is no safety function associated with PT-173 other than for pressure boundary integrity. The transmitter and sensing lines make up part of the RCS pressure boundary and therefore are QA scope and safety-related. De transmitter is used to verify adequate pressure across the No. I seal during depressurized RCS conditions. The pressure in this portion of the system is expected to be approximately 40 psig during installation of the TM. This is normal system pressure. Pressure in this line only increases to full RCS pressure if the downstream MOV CV-270A is shut, or if there is a Nc. I seal failure on either RCP. In the unlikely event that the line failed and it could not be immediately isolated, the leak through the 3/8" lines are within the capability of the charging system to allow shutdown of the unit without actuating engineered safety features.

The pressure indication provided by these transmitters is not mentioned in TS nor does the connection of the 11eise gauge increase the probability of an RCS leak or change the TS limits associated with RCS leakage. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-109)

Page 85 of 112

9. TM 98-041,(Unit 2),19 kV; Temporary Disconnection of 2X-02 Unit Auxiliary Transformer for Repair.

He 2X-02 unit auxiliary transformer is isolated from the 19 kV system temporarily for repairs. The Unit 2 main generator is taken off-line to allow removal of the disconnect links. ne isolated phase bus is covered to prevent entry of foreign materials and to maintain bus duct cooling system integrity. He 2X-02 protective relay inputs to the main generator lockout circuit are disabled and the main generator is synchronized to the grid and reloaded. To facil; tate movement or repair, the low side winding connections and supporting equipment are disconnected if necessary in accordance with plant procedures. After the transformer is repaired, the connections are restored.

Summary of Safety Evaluation: He 2X-02 transformer,19 kV system, and 4160 V system components

• connected to the low side transformer windings are not safety-related. Continued operation with the 4160 V non-safety-related buses (2A-01 and 2A-02) aligned to buses 2A-03 and 2A-04 were previously evaluated in SE 98-133. Dir, connection of the transformer high side links does not degrade the 19 kV system. ne transformer is not credited in any analyzed event for the mitigation or monitoring of design basis or radiological event. No new failure modes in the 19 kV,4160 V, or other plant systems important to safety are created.

Isolation of the transformer from the 19 kV system and disabling the transformer protection inputs to the main generator lockout circuit do not increase the probability of occurrence of an analyzed event, nor does it create the potential for an accident of a different type than previously analyzed. Disconnecting the unit auxiliary transformer temporarily for repair does not reduce the margin of safety defined in the basis TS. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-137)

10. TM 98-052,(Unit 2), Service Water; install Pipe Cap at 2SW-02923 Relief Valve.

TM 98-052 installs a 1",3000 lb. ASTM A-105, QA Scope pipe cap at 2SW-02923 relief valve. His allows performance of a SW flow and pressure check of the 2HX-12D CCW heat exchanger. Unit 2 CCW is not required during this portion of U2R23 since the Unit 2 core is offloaded and the radwaste loads supplied by Unit 2 CCW are shut down.

Summary of Safety Evaluation: The valve is a thermal relief valve that provides protection for the CCW 211X 12D in the event that SW is isolated to the heat exchanger and a CCW heat load is present. Dere is no CCW heat load on thit heat exchanger; therefore, relicf protection is not required. He Unit 2 CCW pumps are isolated while the cap is installed to ensure that CCW heat load is not inadvertently reestablished. In addition, the work plan ensures that ambient heating of the SW does not result in a significant pressure increase that could challenge the heat exchanger pressure boundary. The 2HX-12D CCW heat exchanger can only serve Unit 2 CCW. nat equipment is not required to be cperable per TS since the core is offloaded and the heat exchanger does not need to support normal, shutdown or accident cooling loads. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-188)

I1. TM 98-053,(Unit 2),480 V; Temporary Power to B-21.

The change supplies temporary power to B-21 from a spare breaker located in 2B-02 for approximately 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />. This involves routing a temporary 500 MCM cable between B-21 and 2B-02 cubicle 41D. A spare breaker is used with the amptector set at 400 A. -

Summary of Srsfety Evaluation: B-21 and 2B-02 are non-safety-related buses. The temporary supply bus 28-02 supports a currently defueled unit. This does not change during the durit%f this temporary power.

There is equivalent load tagged out on 28-02 (575 A) that allows 2B-02 to temporarily supply power to B-21 (less than 400 A).

The activity ensures freeze protection is supplied to fire mair.s during winter temperatures. This maintains potable water to the site and supports outage work currently in progress. The TM does not pose a USQ nor does it require a change to the TS. (SE 98-189)

Page 86 of 112

MISCELLANEOUS EVALUATIONS The following evaluations were implemented in 1998:

1.

Dechtel Drawing Change Notices, Drawings M-201, Sheets 1-3; M-202 Sheets 1-3; M-203 Sheets 1-2; M-206 Sheet 1; M-2201 Sheet 3; M-2202 Sheet 1; and M-2202 Sheets 1-2.

Discrepancies to the noted drawing.s were discovered during plant walkdowns. He majority of the changes involved adding pipe caps to vent valves; adding valve numbers to existing valves and showing them on the drawings; and correcting typographical errors. Several changes involve moving piping connections and deleting instrument references and valves that no longer exist.

Summary of Safety Evaluation: The changes ensure that the drawings reflect the actual plant configuration.

No changes to the physical plant layout or to plant operating procedures were made as a result of these drawing changes. The main steam and condensate system (systems represented on the drawings being changed) design basis functions are not affected by the changes. The change does not pose a USQ nor does it require a change to the TS. (SE 98-002)

2. Bechtel Drawing Change Notices, Drawings M-207 Sheet 2, and M-2207 Sheet 1.

He drawings were changed to reflect the as-built configuration of the plant. CR 98-3478 identified two flow control valves associated with the IP-41 vacuum priming pumps that are not shown on the drawings. The condition repott also discussed the probability of the same valve configuration on 2P-41.

Summary of Safety Evaluation: The air removal f unction and the seal water supply to the vacuum priming pump are non-safety related and non-QA. The piping is Seismic Class III, non-safety-related and components are non-QA. Neither the SW system nor its components are identified as an initiator of accidents identified in the CLB. The seal water supply to the vacuum priming pumps where these flow control valves are located is supplied from SW to each of the turbine buildings. The turbine building load off the SW system is considered a non-essential load and is automatically or manually isolated for the respective unit when either an Si or undervoltage condition occurs. Isolation of the non-essential service loads isolates this flow control valve and the potential of the seal water supply line to the vacuum priming pump to compromise the SW system during an accident. The flow control valve is considered as part of the original equipment. The change does not pose a USQ nor does it require a change to the TS. (SE 98-128) 3.

DCN 98-2775, T-24B Condensate Storage Tank (CST) Drain Line As-Built Configuration Change.

De change documents the installation of a 2" diameter drain downstream of valve AF-14 and upstream of the existing fire hose c.:nnections. The 2" diameter line is socket welded to the 4" line downstream of AF-14. De

' 2" line contains an isolation valve and a hose connection. The change facilitated heating of the secondary system via the CSTs prior to reactor startup and heat exchanger maintenance.

Page 87 of 112

Summary of Safety Evaluation: The CSTs are equipment important to safety since they are required to provide an AFP suction source for the station blackout event, and are needed prior to switching over to the SW suction for various other accidents. He CSTs are vented to the atmosphere. He static head of the T-24B CST is maintained by normally shut valve AF-14. De piping change does not affect the ability of AF-14 to maintain the pressure boundary of the CST. The CS's is equipped with an existing fire hose connection downstream of AF-14. Therefore, the change provides a function already supported by the CST design. He piping material used consisted of carbon and stainless steel. The connections were socket welded and threaded. He change does not prevent the CST nor AF-14 from functioning in accordance with the CLB. The failure of the CSTs and subsequent flooding of the area surrounding the CSTs was evaluated via NRC SER dated September 16,1986. The SER stated that adequate drainage is available on turbine building El. 8' and no safety-related equipment was located in the areas that would be exposed to water. He change does not aster the conclusions of the NRC SER. The change does not pose a USQ nor does it require a change to the TS.

(SE 98-139)

4. ECR 98-0288, WVSC-24-04 MTC Cover Plate Bolt Substitution. -

During dry cask loading operations for WVSC-24-04, one of the 16 specified cover p ste bolts was identified as missing. The bolts fasten to the MTC lid (cover plate) to the top of the MTC. ne lid i fastened during conduct of RP 7 Part 7, when the MTC/MSB is lifted to the top of the open VCC, and the i ISB is lowered into the VCC. ECR 98-0288 describes substitution of a stud and nut combination for the spec' sed cover plate bolt.

This substitution is only applicable to WO 9609951 during lo*.fing of WVSC-24-04.

10 CFR 72.48 Evaluation Summary: As a substitute fct the specified bolt, for use only during execution of WO 960"951 for loading of WVSC-24-04, a 1" stud made from ASME SA-193 B7 is used with a 1"-8 heavy hex, ASME SA-194 2H. Both materials are QA.

ISFSI SAR Section 3.4.3.3 describes the material for the cover plate bolt as being A325 material. ECR 98-l 0288 provides technicaljustification for this substitution. Since the safety function of the bolt as defined in the ISFSI SAR remains valid by this one-time substitution, the validity of the SAR analysis is preserved. The l

l change does not involve a USQ nor does it require a change to the H. (SE 98-115)

5. Establishment of a Radioactive Materials Storage Area in an Area Located Outside the RCA for the Storage of a Spare Reactor Coolant Pump (RCP) Motor.

The SE addresses the transport to and the long-term storage of a contaminated RCP motor in a radioactive materials storage area established in an area outside the radiation control area (RCA). Areas where a new storage area may be established include the operating deck of the turbine building; areas on the east side of the turbine building that are immediately adjacent to the truck access bays, El. 8' warehouses (both inside and outside the protected area), and outside yard areas (both inside and outride the protected area). De SE does not address storage of the motor in the south bay of the steam generator storage building.

Summary of Safety Evaluation: The activity was reviewed against FSAR Section 11 " Waste Disposal rmd Radiation Protection System," Section 14, " Safety Analyses," and Appendix 1,"10 CFR 50 Appendix 1 Evaluation of Radioactive Releases from Point Beach Nuclear Plant." The activity was also assessed against -

10 CFR 20, " Radiation Protection." A calculation demonstrated that the effluent concentrations at the boundary of the unrestricted area are below the regulatory limits contained in 10 CFR 20, Appendix B, Table 2. He calculation showed that tin dose to the maximally exposed person at the site boundary control center is 2.0E-04 mrem, if the release is to the air, and 2.7E-08 mrem if the release is to the lake. The calculated doses are below the dose limits of 10 CFR 50, Appendix 1. Although it has been demonstrated that the release of the loose contamination from the motor will not exceed regulatory limits, constraints are imposed to foiiow the intent of 10 CFR 50, Appendix 1 to minimize releases of radioactive materials to unrestricted areas and to ensure levels are as low as reasonably achievable.

I Page 88 of 112 {

ne capability of the fire protection system in the warehouses and the turbine building was reviewed to determine if combustible loading presented by the oil in the motor was within the design capabilities of the present system. Storage of the motor is within the design capabilities of those areas and does not invalidate the Appendix R evaluation of the warehouses or of the turbine building. Storage areas selected outside the protected area have fire protection provided by the Town of Two Creeks.

The following constraints are implemented if the motor is transported or stored outdoors: 1) ne motor shall be enclosed in strong, tight packaging ifloose contaminates may exist; 2) ne packaging shall not support combustion; 3) The motor must be stored off the floor with access to the storage area limited to those persons authorized for entry by plant supervisors and heahh physics personnel; 4) The storage area shall be surveyed, classified and conspicuously posted with appropriate radiation caution signs; 5) A procedure shall be established addressing the safe storage of the contaminated motor in an area outside the RCA; 6) The procedure shall address grab samples of effluents to assess radioactive material releases should releases be i

- suspected; and 7) The procedures shall verify that the activity limits in 10 CFR 20 for unrestricted areas are met and an assessment is performed to determine the impact on the public health and safety.

Evaluations and additional constraints validate the transport and long-term storage of the contaminated RCP motor in an area outside the RCA. The change does not pose a USQ nor does it reqs e change to the TS.

(SE 98-096)

6. FPER, Fire Protection Evaluation Report.

Summary of Safety Evaluation: The changes made to the fire protection program resolve identified conditions needed to make the document accurate. Evaluations of the changes were conducted to ensure that the design basis for fire protection was not compromised. The reviews revealed that either minor adjustments in the r ogram were necessary, including implementing procedures, or the changes were enhancements to the program that provided clarification. Therefore, the equipment important to safety is not adversely affected by the changes to the FPER. The change does not pose a USQ nor does it require a change to the TS. (SE 98-138) 7.

FSAR Table 6.2-8(a), Single Failure Analysis - Safety injection System.

The change revises the Si sy' stem single failure analysis in FSAR Table 6.2-8(a). Specifically, the FSAR change revises two table sections (C and D) that evaluate SI active valve failures. The changes clarify where acceptable single failure consequences rely on redundant SI trains, rather than on parallel flow paths within a common line (as implied by current table wording).

n Summary of Safety Evaluation: The overall conclusions of the SI active valve single failure analysis are N ~

unchanged by this FSAR revision. He changes to Table 6.2 8(a) demonstrate that the fallare of single Si active valves to reposition properly during either injection or switchover to recirculation does not prevent the SI

' system from performing its safety function to mitigate a LOCA. Here are no hardware modifications, procedure changes, tests, or experiments involved in making this FSAR change. The change does not pose a USQ nor does it require a change to the TS. (SE 98-008)

8. FSAR Section 7.7, Control Systems - Rod Position Indication.

Changes to FSAR Section 7.7 allow the plant process computer system (PPCS) rod position indication (RPI) t be used to meet TS surveillance requirements and verify compliance with LCO.

Page 89 of112

I Summary of Safety Evaluation: ne RPI system is an indication system. He rod deviation alarms -.

f generated by PPCS based on the analog signal after it is conditioned with the " compensating poly.a,.nials."

l He circuitry that inputs to the meters is the same as that which inputs to PPCS. No alarms or automatic control runctions are provided by the C-04 main control board (MCB) meter circuit. If there was a malfunction of the RPI on C-04, it could not increase the probability of accidents, increase the consequences of accidents, nor l cause a malfunction of equipment which is important to safety. If there are failures of the detector or its

( associated circuitry, it is indicated on all display devices.

TS Table 15.4.1-1, item 19 contains the shiftly surveillance requirement for RPI, as well as the requirement for PPCS to monitor rod position. Allowing PPCS RPI to be used to meet the shiftly surveillance requirement is consistent with Table 15.4.1 1, Jtem 19. The change does not pose a USQ nor does it require a change to the TS. (SE 98-131)

9. FSAR Section 8.2.1, Network Interconnections.

The change to Section 8.2.1 removes the statement," any one line is capable of carrying the full output of the p two gene-ating units," and clarifies statements uncerning the potential for cascading failure of the 345 kV l

transmission system because a fault on one of the four transmission lines, or bus sections at PBNP, or the loss l of both units.

Summary of Safety Evaluation: Analysis shows that the possibility exists for a cascading power failure if only three 345 kV lines are connected to PBNP and failure of one transmission line, or a fault on one transmission line, or 345 kV bus section occurs. The potential for this cascading failure is dependent on the power genera 6n level at PBNP, the load on the transmission system at the time, and the configuration of transmission lines at PBNP. Operating restrictions are in place that limit PBNP operation so the potential for a cascading failure of the transmission system is greatly minimized. This helps ensure that offsite power is available to PBNP.

Total loss of the 345 kV system is bounded by FSAR Section 14.1.9. He anal) sis evaluates t!.e loss of electrical load and loss of all offsite power. He FSAR changes maintain the probability of such an occurrence the same as that of the original design and licensing basis. Given the failure of a single transmission line or a fault t,n a single transmission line or 345 kV bus section, the potential for loss of offsite power is greatly minimized. Since there are no equipment changes made, there is no potential for increased probtbility of or a different type of malfunction of equipment important to safety. The radiological consequences of the accident evaluated in FSAR Section 14.1.9 are not affected by the change. He change maintains the margin of safety associated with a potential loss of all electrical load and loss of offsite power included in the original basis of TS. The change does not pose a USQ nor does it require a change to the TS. (SE 98-095)

10. FSAR Figure 9.2 2, Westinghouse Drawing 684H41, Sheet 3.

He FSAR figure change corrects the position of a valve located in the IP-1 A RCP bypass vent line from open

• to shut. De current valve lineup shows only one shut valve. The proper valve alignment shuts a manually operated valve further up the bypass vent line.

Summary of Safety Evaluation: The valve position indication change illustrates the integrity of the RCS pressure boundary by placing a secondary isolation valve to the bypass vent line. The second isolation valve ensures unwarranted leakage into the containment building is minimized. Procedure checklists ensure the valve is shut.

The accident analysis is limited due to the currently evaluated LOCA events in the CLB. Dual isolation of the line helps maintain the primary pressure boundary and ceres an uncontrolled leak be minimized. ne change in valve position reflects the procedure control used to maintain the RCS pressure boundary. He change does not pose a USQ nor does it require a change to the TS. (SE 98-113)

Page 90 of112 i

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11. FSAR Section 1b.2.2, Page 10.2-7, Paragraph 2, Crossover Steam Dump Valve Closure Pressure Source.

The teYt of FSAR Section 10.2.2, Page 10.2 7, Paragraph 2 was revised to read,"The dump valves are rescated by applying rescat steam pressure following a time delay after the required blowdown. Setwice air may be used as an alternative administrative pressure source to assist in closing a stuck open dump valve."

Summary of Safety Evaluation: The current FSAR description of the crossover steam dump valves reseat pressure source as service air was incorporated as a FSAR update in June 1990 as a result of MRs 88139/062.

The modifications added an isolation valve with a capped pipe nipple to each crossover steam dump valve bonnet. SER 89-047 summarized the addition as a supplemental source to resent steam for closing a stuck open

, dump valve. Service air may administratively be connected to the valve bonnet isolation valve, as an alternative motive force, to assist in closure of a stuck open crossover steam dump valve. Previous FSAR text descriptions in 1982 and 1988 described the pressure source for rescating the dump valves as rescat steam.

~

The change correctly describes the methodology of the crossover steam dump valves as currently installed and does not impact or revise the facility design. The change does not pose a USQ nor does it require a change to the TS. [SE 98-107)

12. FSAR Figure i1.1-1 Sheet 2, Waste Disposal System Process Flow Diagram.

CR 97-3019 discusses a valve mispositioning event in which WL-1698A, RCDT to PAB sump divert isolation, was found shut during an attempt to drain the RCDT. WL-1698A is a normally open manual isolation valve.

FSAR Figure 11.1-1, Sheet 2 shows this valve as a normally open valve and is revised to designate the valve as a normally locked open valve.

Summary of Safety Evaluation: The accidental release of recycle or waste liquid is defined in F3AR Section 14 as," Accidents in the Auxiliary Building which would result in the release of radioactive liquids are those which may involve the rupture or leaking of system pipe lines or storage tank:s." The installation of a red lock seal on WL-1698A assists in maintaining the valve in its normally open position. This change is not an accident initiator. Installation of a red lock on valve WL-1698A does not increase the source term, affect the release rate or duration, nor create a new release path or release mechanism. WL-1698A is designated non-safety-related and non-QA.

The change installing a red lock seal on WL-1698A is an administrative control to assist in maintaining the valve in its nonnally open position. Installation of the red lock is not an accident initiator. No physical modifications are made to the valve. The normal position of WL-1698A is not changed by the FSAR change.

The change does not pose a USQ nor does it require a change to the TS. (SE 98-105)

13. FSAR Section 11.2, Radiation Monitoring and Waste Disposal.

FSAR Section 11.2 did not accurately reflect how the waste gas system is presently operated. Three statements indicated that the waste gas compressors either are run continuously, or that they can be started automatically on high waste gas system vent header pressure. This is not the case. The waste gas compressor controls are maintained in pullout and the compressors must be manually started.

Summary of Safety Evaluation: The changes involve the waste gas compressors, gas decay tanks and the waste gas system, none of which are safety-related systems. The components are not important to safety, and do not have an impact on safety-related equipment. A possible accident or event related to the waste gas compressors, gas decay tanks or the waste gas system are bounded by existing analyses described in the CLB. The charge does not pose a USQ nor does it require a change to the TS. (SE 98-174)

Page 91 of 112

\

14.L FSAR Section 14.1.4, Chemical and Volume Control System Malfunctions.

FSAR Section 14.1.4 changes the charging pump operational requirements for the dilution during cold shutdown accident analysis by limiting the charging operation to one charging pump in service while the plant

' is in a reduced volume condition. This limits the number of charging pumps in service consistent with new FSAR Figure 14.1.4-1 while the plant is in a reduced volume condition.

Summary of Safety Evaluation: In addition to the FSAR change, limit charging pump operation to one

- charging pump in service while in cold shutdown limiting the number of charging pumps in service. His is consistent with new FSAR Figure 14.1.4-l.

The practice oflimiting charging pump operation to one pump in service during reduced inventory is inadequate. De use of new FSAR Figure 14.1.4 1 to control the number of charging pumps in service based on required shutdown margin for redaced volume conditions is consistent with the FSAR analysis. When not in a reduced (or effectively reduced) volume condition,1% shutdown margin as specified in TS and provided in ROD 9," Reference Boron Concentration," ensures the 15-minute criteria for charging pump configurations to recover from an inadvertent dilution. Providing an allowance for shifting pumps ensures minimal plant impact without challenging the assumptions or outcome of the accident analysis. The change does not pose a USQ ner does it require a change to the TS. (SE 98-187)

15. G.01 and G-02 EDG Derating.

Operations procedures are revised to reflect the derating of G-01 and G42 EDGs. De 2000-hour rating is changed by 1.4%,(changing the kW rating from 2850 kW to 2810 kW). The derating is based on the 120*F design basis limit of EDG room temperature.

Summary of Safety Evaluation: The CLB states that the EDG 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating as 2850 kW. His statement is correct provided EDG room temperature (combustion air temperature) remains sil5'F. His is not accurate if EDG room temperature exceeds 115'F per the revised vendor technical manual and supporting documents. '

He maximum EDG room temperature per TS 15.3.7 basis is 120*F. The 2000-hour EDG rr'.ing is lowered to

' 2810 kW when the EDG room temperature is 120*F.

He electrical capacity of the EDG remains sufficient to provide the necessary power required during all phases of an Si during a LOCA, which is the most limiting accident and bounds all other accidents, per calculation N-91 016, Table 7.1. The most limiting condition per FSAR Table 7.1 is a Unit 1 LOCA and Unit 2 in cold shutdown during the injection phase. Under this condition, maximum EDO loading is 2808 kW. De change does not pose a USQ nor does it O rei a change to the TS. (SE 98-068) 16.' IST IT-007C, IST Flow Acceptance Criteria.

SW pump IST flow acceptance criteria, SW and CCW heat exchanger operating instructions SW valve inservice tec procedures, SW strainer periodic callups, and Unit 2 turbine hall logs are revised to ensure the SW system configuration meets the assumptions in the SW design flow calculations.

• Page 92 of 112

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t Summary of Safety Evaluation: The changes do not affect the SW system ability to perform its important to safety functions. The revisions ensure the SW system configuration meets the assumptions in the SW design flow calculations. De calculations were performed in accordance with the PBNP QA program. The calculation that determined the new IST flow values incorporated the effects ofinstrument inaccuracy in the acceptance values. The calculations demonstrate that the SW system can respond to the worst case design basis accident under the SW system valve and pump alignment conditions allowed by TS, as supplemented by DCS 3.1.7," Service Water Pump Operability."

The change does not pose a USQ nor does it require a change to the TS. (SE 98-097)

, 17. License Amendments 176 and 180, Unit 2 LP2 Inspection interval Extension.

The change extends the inspection interval of the Unit 2 low pressure (LP2) turbine rotors beyond the existing

' interval specified by the FSAR. FSAR Appendix T. " Technical Requirements Manual," requires a low pressure turbine rotor inspection every 5 years, or 60 months. This requirement is expressed in terms of calendar-time.

The extension is based on the use of the operating period, as opposed to the calendar period. The change is manifested in an FSAR change request that revises the FSAR to describe the inspection interval in terms that relate to the operating period of the rotor, rather than a calendar period.

Summary of Safety Evaluation: The change is justified based on the fact that the 5-year rotor inspection does not identify defects that lead to high energy missiles. It is not a physical change to the machine. Since there are no physical plant changes related to this activity, there is no possibility for the activity to introduce some new material, new fluid or new equipment interface that lead to an equipment malfunction not previously analyzed.

The change does not pose a USQ nor does it require a change to the TS. (SE 98-010)

18. License Amendments 176 and 180, Low Pressure Turbine inspection Interval Extension.

De change involves a one-time extension of the FSAR Appendix T inspection interval for the Unit 2 low pressure (LP2) turbine rotors beyond the existing interval of 5 operating years (60 months as approved via SE 98-010). He change extends the LP2 turbine rotor inspection to 66 operating months. De requirements and NRC commitment for tlje UT disc bore inspection are not changed.

Summary of Safety Evaluation: De vendor provided information that allows for an extension of the LP2 rotor inspection beyond the allowable 5 operating years. The rotor inspection could be extended an additional 6 months past the currently-approved 60 operating months.

Extending the LP2 turbine rotor inspection interval to 66 operating months allows for continuous operation of the Unit 2 turbine generator until February 7,1999, without entering a surveillance extension period. Operation until February 7,1999, accommodates U2R23 schedule.

The change isjustifiable because the 5 operating years low pressure turbine rotor inspection does not identify defects that lead to high energy missiles. It is not a physical change to the machine. Since there are no physical

• plant changes related to this activity, there is no possibility for the activity to introduce new material, new fluid, or new equipment interfaces that ould lead to an equipment malfunction not previously analyzed. The change does not pose a USQ nor does it require a change to the TS. (SE 98-102) e Page 93 of112

19. Monitor Containment Temperatures in Accordance with the EQ Program.

De change provides additional ambient temperature monitoring inside containment by installing ACR Smart Reader data loggers in Unit I or Unit 2. De temperature loggers are installed at various locations outside the RCS loop compartment wall for a duration not to exceed one refueling cycle. Multiple data loggers and multiple remote temperature probes (devices with hard wire connections to one data logger used for additional temperature points) may be used concurrently. Data loggers are battery-powered and are not part of plant systems. ney are attached to existing equipment structural supports using Terzel Ty-Raps. Data loggers are not iocated in a position such that they interfere with the function of systems, structures or components.

Remote temperature probes are attached to equipment supports or conduits in a similar fashion.

Summary of Safety Evaluation: ne installation procedure ensures separation of electrical circuits are maintained per DG-E07," Separation of Electrical Circuits." Compliance with Appendix R is not compromised. The addition of a data logger does not affect the seismic analysis because of the insignificant weight of the logger. ne only other identified potential impact on safety-related equipment is to partially -

block the sump screen in the event of a LOCA. The postulated effects of post-LOCA temperature, chemical spray, and radiation were used to determine if the components in the installation would affect the ability to recirculate coolant following a LOCA. From test data, it was concluded that the logger case and Tefzel Ty-Raps are resistant to post-LOCA temperatures, chemical spray and radiation. He remote probes are resistant to post-LOCA temperatures and chemical spray. The Teflon insulation on the probes degrade because of post-LOCA radiation and might fall oft the wires; however, the Teflon does not reach the containment sump because ofits relatively high specific gravity. Derefore, the ability to recirculate coolant following a LOCA is not affected by installation of data loggers with probes. ne change does not pose a USQ nor does it require a change to the TS. (SE 98-143)

20. MWR 931832, functional Change for Service Water Isolation Valve ISW-2880.

MWR 931832 relocated a control signal wire from one time delay relay (TDR) to another which was believed at the time, to restore it to original system design. The logic change affects the conditions for automatic closure of the Unit I nonessential turbine loads. The valve is one of eight isolation valves in the SW system that receives automatic signals to shut based on the occurrence of one or more train (s) of Si actuation signals, number of operating SW pumps, and power source availability. Specifically, the logic for valve ISW-2880 was revised from a 1B-04 bus voltage input to a 2B-04 bus voltage input for automatic closure with a Unit i Si, less than 4 SW pumps operating and a 30-second TDR.

Summary of Safety Evaluation: Valve ISW-2880 used to shut on a Unit i Si actuation, less than 4 SW pumps operating, and Train B power availability on Unit 1. The change resulted in a change from the Unit i Train B power availability to Unit 2 Train B. The differeaces in circuit operation can only affect system or plant operation for events or conditions that are beyond the licensed design basis,(c. g., multiple failures or coincidence of events with failures during LCO conditions).

The design basis requirements for nonessential SW isolation were evaluated. The valves' control logic is not specifically discussed in the FSAR or the CLB. The wiring change is limited to the engineered safety features l Train B rack. Circuit components are properly qualified and rated for the service. Operation of the valve .

during design basis events including single failures remains unaffected. The potential failure modes of the isolation valve and circuit components, and the consequences of failure are unchanged. Design basis calculations suppon that adequate SW is provided to essential loads given the stated conditions. The SW isolation valve is not an event initiator, and does not create the possibility of a new type of event. The change does not pose a USQ nor does it require a change to the TS. (SE 98-093)

Page 94 of112

21; SPEED 98102, New Automatic Lubricators for Containment Accident Recirculation Fans.

The type of automatic lubricators that are used on the containment accident recirculation fans is changed. The present lubricator uses an electro-chemical reaction to generata nitrogen gas. The gas build up slowly in a bellows and injects grease into the bearing. ne change is a lubricator that uses a small mechanical pump to inject the grease. The new lubricator prevents the possibility of oil separation from the grease base, while being under constant gas pressure.

Summary of Safety Evaluation: Use of the new lubricator does not increase the probability of an accident since it only affects the CPCs. The CFCs are not initiators of accidents or events described in the licensing basis.

, The CFCs are equipment important to safety. Use of the lubricators results in an increase in the overall reliability of the fans. Failure to deliver grease and injecting all grease at one time are not likely to occur.

Testing has demonstrated the device to be very reliable under typical industrial conditions. Ilowever, failure of the device under accident conditions is possible because of temperature limits and radiation effects associated with the materials and components. Batteries could fail at conditions above 150*F. In this case, there are no consequences, as long as the lubricator stays in place.

The other accident failure mode is if the lubricator broke and fell off the bearing. This is possible since the materials in the base of the device start to soften at about 158'F as well as suffer from a significant strength decrease at post-accident radiation doses. The weakest point is a threaded connection between the base and a brass adapter for connection to the bearing housing. Failure at this point allows the memolub to fall and remain intact, if the device broke and fell from the bearing, it opens a path for containment atmosphere directly into the bearing. The passage remains filled with grease, which has a dropping point (melting point) of 465*F.

Pressure equalizes through the bearing labyrinth seal. Water or debris is not injected into the bearing because of ailure. The lubricator also does not affect CFC operation. The bearing housing is located downstream of the fan blades. If the lubricators fail during a severe accident, the CFCs would be in better condition than if using the mar.ual lubrication, since they would have had fresh grease injected up until the time of the accident.

If a less severe event or accident occurred, the lubricators would most likely continue to function and provide continuous lubrication during the recovery. The use of the new lubricator does not increase the radiological consequences of an accident since the CFCs continue to function in the same way as that without the lubricators. The use or failure of the lubricator does not introduce a new type ormalfunction into the CFCs.

The most likely failure mode of the fan itselfis because of a lack oflubrication or contamination of the lubricant. The automatic tubricator provides a continuous fresh supply of grease and reduces the probability of failure due to lubricant problems. The change does not pose a USQ nor does it require a change to the TS. [SE 98-180)

22. TS Amendments 126 and 130, Storage of Fuel Assemblies with Enriched Axial Blankets in New Fuel Vault and Spent Fuel Pool.

This evaluation addresses the modification of our CLB from the use of axial blankets of non-enriched fuel to axial blankets enriched to any weight percent U-235 up to the sr.me enrichment as tie center, active region of the fuel assembly.

O Summary of Safety Evaluation: TS 126/130 increased the allowed U-235 enrichment level for storage in the SFP and new fuel vault (NFV) to 4.75 wtM for optimized fuel assemblies (OFA). The amendment also allowed the use of axial blankets. The NRC requested addi'.ional information during review of this TS change request. The WE response stated, "This daign [the adcl blankets] presently incorporates pellets of natural, non-enriched, uranium in the top and c' ottom six inches of the fuel bearing regions of all fuel rods in the fuel assembly." The NRC SER associated with Amendments. 126/130 states, "The licensee has confirmed the staffs understanding that its use of an optional axial-zoned core-loading scheme refers to the use of a single i enrichment throughout the active portion of the fuel assembly with axial blankets of natural (non-enriched) fuel above and below tin active portion of the fuel assembly." Unit 1, Cycle 24 feed assemblies utilized axial blankets enriched to 2.6 wtM U 235.

Page 95 of 112

Analysis performed assumed a fuel assembly with a single enrichment throughout the entire length of the fuel assembly (c. g., the " axial blanket region" is enriched to the same level as the active fuel region) when performing the analysis to determine the Keffand Koo for fuel stored in the SFP and the new fuel vault. De analysis determined the values for Kefrand Koo for both the SFP assuming fuel enriched from 0.0 wt/% up to 4.75 wt/% U-235with or without axial blankets and for the new fuel storage vault assuming fuel enriched from 0.0 wt/% U-235 up to 5.5 wt/% U-235. The analysis also reviewed the use of natural uranium blankets or lower enriched fuel in the blanket region. The analysis states," Replacing any amount of fuel with a lower enrichment will reduce the multiplication factor. Thus, in all cases, the use of natural uranium ends will in rease the subcriticality margin of the storage racks beyond that previously calculated for both the spent fuel pool and the new fuel storage racks." ,

Additionally, the NRC issued Amerdments 179/183, that increased the enrichment allowed to be stored in the new fuel vault and SFP from 4.75 wt/% U-235 to 5.00 wtM4 U-235. The issue of enriched vice non-enriched blanket regions was not addressed in these amendments. Westinghouse evaluated the storage of fuel assemblies in the SFP by incl.tding calculational and methodological biases and the sum of uncertainties to obtain a maximum Keff of 0.94876 compared to the acceptance criterion of Keffless thar 0.95 for enrichments up to 4.6 wtr4 U-235. Fuel assemblies containing U-235 enriched to greater than 4.6 wu% must contain sufficient integral fuel burnable absorber to obtain an equivalent 4.6 wt/% enrichment for criticality pmposes. For these calculations, the use of axial blankets enriched to less than 5.0 wt/% U-235 was not considered and the conclusions regarding the use of axial blankets contained in the PLG analysis remain valid. The change does not pose a USQ nor does it require a change to the TS. (SE 98-005) l l 23. Unit I Cycle 25 (UlC25) Extended Fuel Cycle Implementation.

He SE evaluates the implementation of an extended fuel cycle (nominal 18 months) for UIC25. A nominal 18-month fuel cycle is being implemented for UIC25 during the Spring 1998 refueling outage.

Summary of Safety Evaluation: Three potential impacts of the UIC25 extended fuel cyde include increased boron concentration, increased surveillance intervals and revised radiological source term. The increased TS boron concentration requirements for extended fuel cycles approved by the NRC via License Amendment 180, Unit I, is not implemented during the UIC25 reload core design because of the greater use of burnable poisons in this design. Therefore, there can be no increase in probability or consequences of accidents, events, or malfunctions of equipment previously evaluated, the creation of new accidents, events, or malfunctions, or a redaction in the margin of safety defined in the basis of TS because of this aspect of the UIC25 extended cycle.

Increased surveillance, calibration, test, and maintenance intervals were evaluated for the UIC25 operation.

De evaluations indicate no expected adverse impact on the reliability or performance of systems or components that could initiate or mitigate accidents, events, or malfunctions or equipment previously evaluated.

TS requirements for calibration ofinstrumentation, radiation monitors, electrical relays, and miscellaneous equipment must be met by August 1999 or the TS require a revision. NRC commitments regarding reactor trip breaker maintenance, EQ replacements, ECCS system leak checks, specific valve cycling tests, CCW heat exchanger cleaning and inspection, fire hose flushing and testing. and AMSAC must be performed or the commitments appropriately revised by February 1999.

The radiological source teim in the reactor core for the extended UIC25 fuel cycle was compared to previous cycles and is not changed. In addition, the radiological releases offsite and control room dose for design basis accidents were reanalyzed using source terms from a nominal 18-month cycle reload core design. The results were reviewed and accepted by the NRC via License Amendment 178.

I 4

Page 96 of 112

l Operation of the UIC25 extended fuel cycle does involve several TS changes related to specified maximum I calibration intervals of 18 months for certain instmments, radiation monitors, and relays; but cycle operation is consistent with the TS provided the calibrations are performed or the TS changed to extend the interval prior to August,1999. He change does not pose a USQ nor does it require a change to the TS. (SE 98-087)

24. Unit 1 Cycle 25 (U1C25) Reload Core, Refueling Shutdown Mode Only.

The SE evaluates the UIC25 core loading pattern as described in letter 98WE-G-0019 dated May 1,1998. The SE covers mechanical design, nuclear design, thermal hydraulic design, FSAR accidents and TS changes that apply to the UIC25 reactor core during its refueling shutdown mode only.

Summary of Safety Evaluation: The UIC25 core contains 121 upgraded OFAs arranged in a specified core loading pattern. Forty-five feed assemblies are supplied. The core also contains 12 peripheral power suppression assemblies. The new Region 27 feed assemblies incorporate several minor fuel design changes. A grooved ring on the top and bottom end-plugs allows weld material to flow into and fill the groove, allows greater weld penetration and improved visual indication of weld characteristics. The IFBA pellets in Region 27 assemblies increase the enriched boron loading from 1.0X to 1.5X B10. The backfill pressure remains unchanged at 200 psi and the IFBA length has decreased from 132.0" to 120.0." UIC2$ is planned to be operated for approximately 15.5 months for a design burnup of 12,565 MWD /MTU. The cycle length is greater than the 12 months nominal cycle length for previous cycles.

Completed calculations ensure that with the reactor in the refueling shutdown condition, the operators have sufficient time to respond to a boron dilution event. This ensures that the probability of an accident or event previously evaluated is not increased. The design changes to the fuel are minor, and do not affect the structural integrity of the fuel assemblies. Thus no increased probability of occurrence of a malfunction of equipment important to safety occurs. Maintaining the tenctor in refueling shutdown conditions, and performance of the boron dilution at refueling shutdown condition calculations, ensures that the probability of an inadvertent criticality event is not increased, and thus the radiological consequences of an accident, event, or malfunction of equipment is not increased. A number of administrative controls, including revising RP-IC, " Refueling,"

1 RMP-9096, " Reactor Vessel Head Removal and Installation," and a hold point in the refueling schedule, ensure the proper plant conditions are maintained to satisfy the boron dilution at refueling condition calculations. TS requirements to load the core and maintain it in a refueling shutdown condition are met, and thus the margin of safety in the basis for any TS is not reduced. The change does not pose a USQ nor does it require a change to the TS. (SE 98-073)

25. Unit I Cycle 25 (UlC25) Reload Core, Cold / Refueling Shutdown Modes.

This supersedes SE 98-073, Unit I Cycle 25 (UlC25) Reload Core, Refueling Shutdown Mode Only." The SE applies to the refueling shutdown and the cold shutdown conditions. Reactor vessel water level must be

• maintained above the centerline of the loops, which is controlled using present Operations refueling procedures.

This SE does not address heatup above 200* F, reactor startup, nor power operation. This reduced scope SE necessary to allow refueling operations to continue while Westinghouse safety evaluation activities are being completed.

o

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Page 97 of 112

Summary of Safety Evaluation: The fact that full core offload capability does not exist once these fuel assemblies are loaded in the RV does not affect the probability of an accident or event, since the RV is an acceptable fuel storage location. With refueling or shutdown conditions maintained, fuel can be stored in the RV for an extended period of time without affecting the probability of an accident or event. Contingency plans are in place to ensure Unit I does not leave the cold shutdown condition. A hold point was placed in the Unit I refueling schedule to ensure cold shutdown (<200*F/<0.99 Keff) is maintained until a SE is written and approved to allow Unit I to go above 200*F. A temporary procedure change was also placed on OP 1 A," Cold Shutdown to llot Shutdown," to ensure the cold shutdown conditions are maintained. Other aspects of SE 98-073 apply to this SE as well. The change does not pose a USQ nor does it require a change to the TS.

(SE 98-079)

26. Unit I Cycle 25 Reload.

This SE permits RCS heatup above 200*F, reactor physics testing, reactor startup, and the operation of UIC25 to end-of-life. The SE includes the UIC25 reload design and safety analyses performed by Westinghouse for reload fuel supply and core design supported by additional evaluations performed by WE personnel.

Summary of Safety Evaluation: The UIC25 design does not cause safety limits to be exceeded, provided the following conditions are met: UIC24 end-of-cycle burnup is 12,567 MWD /MTU; UIC25 burnup does not exceed the end-of-full power capability, plus up to 1,500 MWD /MTU of power coastdown operation; and, adherence to plant operating limitations as specified in the TS.

U IC25 reload core design meets applicable design criteria or has shown to maintain the same levels of safety as considered in design basis evaluations. Though fuel and core design are not directly related to the probability of previously evaluated accidents or events, the demonstrated adherence to applicable standards and acceptance criteria precludes new challenges to components and systems. No new performance requirements are imposed on systems or components such that design criteria are exceeded nor do the changes cause the t ,re to operate in excess of pertinent design basis operating limits. The change does not pose a USQ nor does it reqwre a change to the TS. (SE 98-089)

27. WO 900198, Rework T-015 Chemical Drain Tank Bubbler Tubing.

The T-015 chemical drain tank level indicator (LI-1002) bubbler tube failed. It is suspected that corrosive chemicals in the tank ate a hole through the bubbler tube. LI-1002 is providing erroneous readings at low

, levels. The tank, bubbler, Swagelok fitting, and the valve are non-safety-related and non-QA. A new 3/8" stainless steel tube is slipped inside the existing bubbler tube and routed to LI-1002. The existing diaphragm valve is replawd with a new diaphragm valve and mounted to existing tubing supports. s t

Summary of Safety Evaluation: The chemical drain tank sees nearly every type of chemical used in the plant, including radioactive chemicals. Therefore, all parts potentially in contact with chemicals are stainless steel.

The level indicator has a low lesel pump stop for the transfer pumps. The level mdicator also has high and low level alarms. The new tube is 3/8" rather than 3/4" as described in FSAR Figure 11.1 1, Sheet 2. The new bubbler tube is made of the same material and wall thickness as the existing bubbler tube, i The bubbler tube samples the pressure near the bottom of the tank. The smaller diameter bubbler provides slighter higher resistance to air flow, resulting in a slightly higher pressure reading at the instrumentation. The slight difference does not affect the operation of the level indicator. The chemical drain tank and the level indicator are not discussed in accidents or event evaluations. However, there are accidents that involve chemical and gas release in the PAB. The change does not increase the amount of radioactive liquids or gases in the tank. After installation, LI-1002 is calibrated to verify proper operation. The change does not pose a USQ nor does it require a change to the TS. (SE 98-173) 1 Page 98 of 112  !

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, 28. WOs 940226,940227,940228,940229,940230,and 940231, Unit I and 2 IIDTP Vent Line Installation.

The WOs install steam vent lines to the pump pits of the Unit I and 2 heater drain tank pumps. Steam is produced wFen an out of service pump is started. This warms the ground water accumulated within its pit.

Summary of Safety Evaluation: The vent lines are used to vent this steam from the pump pits to the 1&2T-71 atmospheric blow off tank associated with the respective unit. The vent configuration allows the steam from the pit to be routed to a location safe for plant operation and operations work crews.

The vent line from each pump pit runs to a common header, and is then routed to the blow off tank. The Unit i

, vent header has a check valve, a pressure indicator, and two isolation valves. He check valve prevents back flow of fluid / steam irom the atmospheric blow off tank. He pressure gauge provides a me:ns to observe header pressure (Unit I only). Isolation valves are provided to accommodate maintenance. The isolation valves are added to the valve checklist (CL-13D). The change does not pose a USQ nor does it require a change to the TS (SE 97-198-01)

29. WOs 9604133,9802354, and 9802360, Temporary Installation, Use, and Removal of Tygon Tubing.

Temporary installation and use of tygon tubing, attached to RC-00522B, RV local level LI-447B variable leg vent provides an alternate reduced inventory local level indication. The tygon tubing is used for local level indication in lieu of LI-447B because of a damaged glass tube. LI-447B was installed under MR 89-023/024 to replace tygon tubing that was previously used for the alternate reduced inventory local level indication. The temporary change controls the design, installation (WO 9802354), and removal (WO 9802360) of the tygon tubing when the tygon level indication is no longer required.

Summary of Safety Evaluatica: An alternate level indication to LI-447B is required because of the unavailability of LI-447B due to a cracked glass sightglass housing. The tygon tubing is installed on valve RC-522B to provide the local indication in a similar fashion as used prior to the installation of LI-447B under MR 89-023. The top of the tygon tubing is vented to the containment atmosphere to provide the reference leg.

The tygon tuW is used in operating procedures in the same manner as LI-447B. Existing level markings are used to determine the level.

The modifkation occurs while the RCS is depressurized. The tygon tubing to bstalled has a pressure rating well in excess of the maximum possible pressure that could be seen by the venteo byel indicator. In the unlikeV event of a leak, it gives a conservative indication of a lower than actual level. Ba ckup/ indirect indications of adequate core cooling (core exit thermocouples, RiiR system parameters) are not impacted by the change. The change does not pose a USQ nor does it require a change to the TS. (SE 98-025)

30. WOs 9606829,9606830, Replace Control Power Transformers in Supply Breakers for SW Motor-Operated A

Valves SW-2869/2890.

The control power transformers for the supply breakers for SW ring header motor-operated valves SW-2869 and SW-2890 are replaced. This SE addresses the interim conditions during the installation work.

Page 99 of 112

Summary of Safety Evaluation: De valves are maintained open with electrical power removed during installation. He ring header valves are normally open and a TS LCO entry is required when a valve is shut.

The valves are shut one at a time for post-maintenant e stroke testing under the TS LCO for interruption of the SW ring header. If an accident occurs during the installation work or post-maintenance stroke testing, the SW

system performs its function as designated. There is no need to reposition a ring header valve to respond to an accident. He rir g header valves are installed to allow isolation of a SW header section in case of a malfunction, such as a leak in the header. The ability to shut the valves electrically from the control room is not a safety-related function as documented in the SW design basis docurnent. He motors on the SW ring header motor-operated valves are classified as non-QA scope. If a SW malfunction occurs, the valves could be shut manually, if needed, via steps in AOP-9A," Service Water Malfunction." The installation activity does not increase the probability of a malfunction of equipment important to safety. He change does not pose a USQ ,

nor does it require a change to the TS. (SE 98-106)

31. WOs 9704153,9704154, Capping of LeakofTLines of ISI-841 A and ISI-841B MOVs.

The stem leakofflines on valves ISI-841 A and ISI-841B were capped per WO 9704153 and WO 9704154 respectively. The valves were repacked in accordance with M132.2, " Valve Packing," to utilize a more effective packing arrangement. The MI however, did not discuss capping of the stem leakoff Mich is required for the new packing configuration. The capping is required to prevent liquid at system pressure from traveling out the leakofiline connection and flowing to the reactor coolant drain tank (RCDT).

g Summary of Safety Evaluation: The new packing arrangement has shown to be more effective at minimizing stem leakage per SPEEDS 9704153/9708640. The old deep stuffing box packing configuration incorporated a a

leakoffline between an upper and lower 5-ring packing set. Leakage past the first sets of packing is diverted to the RCDT via the leakoffs. Due to excessive packing leakoff, however, the new packing configuration replaces the lower packing with a carbon spacer. Capping the packing leakoff line prevents liquid at system pressure from traveling out the leakoffline connection and flows to the RCDT. The remaining tubing leading to the RCDT (located inside containment) was also capped to prevent gases or liquids from the RCDT from transferring back up to the leakoffline into containment. The leakofflines were capped a. the end of the installed 1/2" leakoff connection (about 6" from the valve body) by use of a Swagelok phig. Since the valves are located inside containment, stem leakage utilizing the new packing configuration are .nside containment.

The accumulator and its associated discharge path are accident mitigators. Installation of a tubing cap on the valve leakoffline does not create the possibility of an accident of a difTerent type nor does it iwease the probability of previously evaluated accidents in the CLB. The leakoff caps are rated for the temperatures and pressures of the accumulator discharge piping system and do not effect the seismic requirements of the system.

Neither the shut nor open function of the MOVs are effected by the capping of the leakofiline. The valves are normally open and not repositioned during accident scenarios, except for the cooldown following a small break LOCA. The valves are shut to prevent nitrogen addition into the RCS during depressurization. Their function is not efTected. In the event of a total failurr of the leakoffline cap, a 1/2" leak in the accumulator discharge piping is created. This is bounded by the FSAR single failure analysis outline in FSAR Table 6.2-8a that includes an accumulator discharging to a broken loop, spilling the entire contents of the accumulator to the containment floor. The change does not pose a USQ nor does it require a change to the TS. (SE 98-067)

Summary of Safety Evaluation: De stem leakofflines on the accumulator discharge isolation MOVs were capped per WO 9704153 and 9704154 in order to utilize an alternative packing configuration. The accumulator and its associated discharge path are accident mitigators and the installation of a tubing cap on the valve leakoffline does not create the possibility of an accident of a difTerent type or increase the probability of previously evaluated accidents in the CLB. The probability of a malfunction of the valves is not increased as a result of capping the leakoff connections. The leakoff caps are rated for the temperatures and pressures of the accumulator discharge piping system and do not effect the seismic qualifications of the valves. In addition, the valves are part of the Class 2 piping pressure test program and are tested accordingly. These valves are normally open and are not required to be shut during accident scenarios. The valves are shut, however, during the cooldown and depressurization following a small break LOCA. This prevents nitrogen addition into the Page 100 of 112 1

1 RCS. Neither the open or shut function of these valves is affected with the capping of the packing leakofflines.

The chaage does not pose a USQ nor does it require a change to the TS. (SE 98-067-01)

32. WOs 9708134,9707051, WMSB-02 Shell and Shield Lid Arc Strike Removal.

Summary of Safety Evaluation: %e SE revision addresses the comments of the PBNP Off-Site Review Committee as stated in CR 97-3709. His revision removes the underline and strikeout annotations, and makes no other changes. Therefore, the argument and conclusions of SE 97-143 remain unaffected by this revisbn.

The change does not pose a USQ nor does it require a change to the TS. (SE 97-143-01)

, 33. WO 9801733, Maintenance Verification of P-32F SW Load Sequence Timer.

The WO verifies that contacts 1-7 for relay 2-271X4/B03 and 2-272X4/B03 are operable and satisfy TS requirements. The other contacts and the relay armature were verified during other work activities. This work places P-32F SW pump out of service and enters an LCO per TS Section 15.3.3.D.2.a. One SW pump may be out of service for 7 days, or two or three pumps may be out of service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Summary of Safety Evaluation: Voltagt; and continuity readings are taken across the applicable contacts prior to and after int.erting a insulating sleeve between the contacts during this WO. This, along with visual inspections and prior relay operation, shows that the contacts are not stuck closed, and thus are operable. When the sleeve is inserted between the contacts, the effect on the other contacts of the relay is minimal because there is adequate dktance between sets orcontacts to work on one without afTecting the others. He relays are of a sturdy construction and the contacts are spring-loaded so that one set can be moved without moving the others or the relay armature. Concurrent checks are used to ensure the correct relay and contacts are selected.

Since the work only affects the P-32F SW pump (as it is placed out of service), and the applicable LCO is entered per TS 15.3.3.D.2.a, there is no afTect on the probability of occurrence or radiological consequences of a previously evaluated accident, event or malfunction. The change does not pose a USQ nor does it require a change to the TS. (SE 98-014)

34. WO 9812294,2T-34A, SI Accumulator Leakage Troubleshooting.

WO 9812294 troubleshoots the source of water leakage from 2T-34A Si accumulator. The work plan isolates the accumu!ator drain line by shutting normally open manual isolation valve 2SI-892A, T-34A Si accumulator drain to T-16 RCDT. Accumulator level is then monitored for at least 7 days, or a conclusive level change of 2 2% is needed. If no change in level occurs in 7 days,2SI-892A is returned to its normally open position and a WO is initiated to repair seat leakage from AOV 2SI-844A, T-34A SI accumulator drain to T-16 RCDT. If a level change is noted within 7 days, the accumulator fill line is isolated by shutting normally open manual isolation valve 2SI-886C, T-34 A Si accumulator fill from cold leg S: second-offisolation. Accumulator level is then monitored for at least 7 days, or a conclusive level change of 2 2%.

Summary of Safety Evaluation: The SI system and the Si accumulator remain fully functional and operable during the performance of the troubleshooting activity. The accumulator is maintained at its TS-required volume and pressure through the entire activity. The design capacity of the accumulators is based on the assumption that flow from one of the accumulators spills onto the containment floor though the ruptured loop, and the flow from the remaining accumulator provides sufficient water to fill the volume outside of the core barrel below the nozzles, the bottom pienem, and one-half the core. Each accumulator is separated from the RCS by two check valves and as such, can not initiate accidents or events.

The troubleshooting activity does not alTect the accumulators ability to perform its design function of discharging into the RCS. Shuttiag the normally open manual fill and drain line isolation valves has the same impact on the accumulator Gring a design basis accident as the normally shut air-operated fill and drain isolation valves since the air-operated fill and drain line isolation valves remain shut during the design basis

~

accidents. De troubleshooting activity affects the ability to remotely fill or drain the accumulators; however, f

Page 101 of 112

there is no requirement to fill or drain within a specified time period. If accumulator level adjustment is necessary during the performance of the WO, the accumulator is adjusted to ensure volume and pressure are maintained within the TS limitations. The change does not pose a USQ nor does it require a change to the TS.

(SE 98-136)

P I

?

I 4

1 l

Page 102 of 112

r V. COMMITMENT CilANGE EVALUATIONS 1.

Reactor Trip / Bypass Breakers. Our response to GL 83-28 dated December 28,1984 that states the semi-annual preventive maintenance ac*ivities might be extended to 9-12 months and are not to exceed this in*erval, requires an update to reflect 1986 Westinghouse Owner's Group recommendations.

Justification: The Westinghouse new reactor trip switchgear maintenance program extended the recommended task interval for switchgear assembly energized breakers from 6-12 months to every refueling /18 months (provided 200 breaker c>perations are no't exceeded during the interval). The switchgear assembly deenergized breakers remained essentially the same (every refueling /18 months). No more than 40 breaker operations are expected between refueling outages. An evaluation of reactor trip / bypass breaker work order history identified no adverse trends in the corrective maintenance of these breakers. Functional testing of these breakers to perform their safety-related feature is demonstrated on a monthly basis. (CCE 98-001) 2.

Engineered SSfety and Auxiliary Systems. Ourimplementation letter for NUREG-0578 dated March 14,1980 that states systems in our leakage reduction and preventive maintenance (LRPM) program are to be leak checked once per year requires an update.

Justification: NUREG-0578, item 2.1.6.a established the requirements for a LRPM program to assist in the reduction of leakege to an as-low-as practical level for systems that would or could contain highly radioactive fluids during a serious transient or accident. The establishment of a LRPM program and the testing frequency of the LRPM program tests is not related to the ability of SSCs to perform their safety function and has no impact on the ability of personnel to ensure the SSC is capable of performing its safety function and has no impact on the ability of personnel to ensure the SSC is capable of perforraing its safety function. The NUREG requirements ensure licensees minimize leakage and dose to plant personnel and me public in the event of a serious transient or accident. The LRPM program evaluates new leakage identified to ensure total leckage is within allowable limits. If significant leakage is discovered, or allowable leakage limits are exceeded, our operability determination process addresses the ability of the SSC to perform its safety function. A change in the LRPM test frequency has no impact on the ability of the SSC to perform its safety function. (CCE 98-004)

3. Charging Pump Operation. Our March 31,1981 letter discussing postulating uncontrolled boron dilution at cold shutdown conditions requires clarification.

Justification: The original intent of the commitment made that limits charging pump operation in cold shutdown was to ensure that adequate shutdown margin is available to provide at least 15 minutes for operator action to terminate an inadvertent dilution during cold shutdown conditions. FSAR Figure 14.1.4-1 depicts the relationship between the number of running charging pumps, RCS boron concentration, and the required shutdown margin necessary to ensure the 15 miere criteria is met for the worst case condition.' This figure can be used to coatrol RCS boron concentration and/or charging pump operation during a reduced volume

• condition. This is adequate to meet the accident analysis assumptions and acceptance criteria. ROD 9 " Boron Concentration " can be used for charging pump configurations to ensure adequate boron concentrations exist when not in a reduced volume condition. Therefore, a restriction that limits charging pump operation during

• cold shutdown to a single pump is overly conservative and unnecessarily hinders plant operation. The change is administrative, and therefore does not impact the safety function of SSCs. (CCE 98-006) l Page 103 of 112

VI. NUMBER OF PERSONNEL AND TOTAL DOSE BY WORK GROIJP AND JOB FUNCTION - 1998 I

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 SI 13.582 10.182 1360 0.270 1.770 Maintenance 49 19.000 13.480 0.130 0.890 4.500 Chemistry & Health Physics 33 8.839 7.419 1.420 Instrumentation & Control 6 1.740 1.290 0.450 Administration Engineering,&

Regulatory Services 12 3.930 3.090 0.840 Utility Employees 30 13.520 0.910 12.610 Contractor Workers & Others 237 108.552 3.600 1.760 13.000 90.002 0.190 GRAND TOTALS 418 169.163 25.201 29.140 15330 90.002 2.770 6.720 1150 individuals were monitored exempt from the provisions of 10 CFR 20.

Page 104 of112 .

l VII. STEAM GENERATOR INSERVICE INSPECTIONS STEAM GENERATOR EDDY CURRENT TESTING The following abbreviations are used throughout this report section.

xil #x Tube Support Plate flot Leg xC #x Tube Suppon Plate Cold Leg

, AVx Anti-Vibration Bar #x TSli Tubesheet Hot Leg TSC Tubesheet Cold Leg

, FBH Flow Distribution Baffle llot Leg FBC Flow Distribution Baffle Cold Leg DNT Dent (condition where the tubing inside diameter is less than nominal)

DSI Distoned Support Indication (condition where a discrete flaw signal forms)

DTI Distorted Tubesheet Indication NQI Non-Quantifiable Indication FSI Free Span Indication BLG Bulge Unit i Inspection Plan: Unit I steam generators were inspected from March 25,1998 to March 31,1998. Rows 3 and above were inspected full length using a bobbin coil (3027 tubes in each steam generator). Rows I and 2 were inspected over the straight length using a bobbin coil and a pluspoint was used for the U-bend (183 tubes in 'A' steam generator and 182 tubes in 'B' steam generator). Additionally, the hot leg expansion transition area (-2",

+3")in 20 percent of the tubes was in inspected using a pluspoint (643 tubes in each steam generator). Special interest pluspoint exams were performed on all 1-codes and a sample of dent indications (55 tubes in steam generator "A" and 68 tubes in steam generator "B"). Four tubes were previously plugged in steam generator"A,"

and 5 tubes were previously plugged in steam generator "B."

Inspection Results: The following table summarizes the number of tubes found with indications. Some tubes have more than one indication (numbers in parentheses are the total number ofindications).

Indication 'A'SG 'B' SG DSI 3 1 DTI 3 2 NQi i i FSI O 1 0 to 19 % 9 (13) 7 (12) 20 to 29 % 0 1(2)

BLG 0 1 DNT 86 (120) 74 (91) r l Page 105 of 112

The following table lists each indication in the steam generator"A."

Itow Column Indication Location inch Mark 1 29 DNT TSC 4.99 2 88 DNT FBli 1.97 3 4 DNT 06H 5.37 3 34 DNT 03C 25.12 3 34 DNT OlC 19.06 4 8 DNT 06H 9.98 4 28 DNT 0611 10.24 4 48 DNT OlH 21.89 4 48 DNT 0211 10.3 4 59 DNT 06H 10.17 5 48 DNT 06H 12.49 5 48 0611 12.56 DNT_

5 51 DNT 01H 21.91 5 51 DNT 021{ 8.3 5 51 DNT 02H 13.31 5 62 DNT 0111 45.36 6 29 DNT 02C 2.24 6 50 DNT TSC 17.37 7 1 DNT 02H 25.27 7 25 DNT TSH 4.13 7 79 DNT 0 111 47.93 7 91 DNT TSC 18.18 8 85 DNT 03H 27.59 8 86 DNT 0311 27.48 9 23 DSI 03H 0 9 39 DNT TSC 5.87 9 43 DSI 05H 0 9 80 DNT 0111 35.31 10 22 DNT 04H 49.91 10 65 DNT 06H 7.44 10 66 DNT 06H 8.3 11 42 DNT 06H 14.85 II 87 DNT TSH 1.5 12 78 DTI TSH 0.84 13 67 DTI TSli -0.16 14 4 DNT TSH 1.19 14 9 DNT 05C 41.55 ,

14 24 DNT 06H -0.12 14 45 DNT OIC 42.69 15 89 DNT TSH 10.88 15 89 DNT TSH I1.99 16 80 DNT FBC 17.96 16 88 DNT FBC 9.79 16 89 DNT TSH 1.09 16 89 DNT TSH 2.18 19 28 DNT TSil 16.39 Page 106 of112

Row Column Indict. tion Location Inch Mark 19 54 12 AV4 0 19 54 13 AV2 0 19 57 DNT 06H 15.56 19 61 10 AVI O 19 61 13 AV2 0 19 71 6NT AVI 9.93 19 85 DNT OSC 44.18 20 13 DNT 03C 20.16 21 6 DNT AVI O 21 27 DNT AV3 3.03 22 8 DNT 04H 45.06 23 7 DNT 06H 19.86 23 7 DNT 06H 21.35 23 79 DNT OIC 9.87 24 8 DNT AVI -0.11 24 68 DNT OIC 48.98 24 85 DNT AV4 0.69 26 57 NQI 0$H 17.1 26 63 DNT TSH 2.92 26 84 DNT AV2 20.78 26 84 DNT AV4 1.63 27 29 DNT 06H 21.67 27 44 DNT FBC 1.3 27 83 DNT AV3 -0.09 28 46 DNT 03H 3.48 28 56 DNT OlH 24.37 28 57 DNT 04H $0.45 31 37 DTI TSH 0.02 31 71 DNT 05H 34.86 31 71 DNT 05H 36.67 3! 71 DNT 05H 37.68 31 71 DNT. 05H 44.11 31 72 DNT 03C

~' 44.43 31 80 DNT AV3 -0.05 32 14 14 AV3 0 32 45 DNT OlH 22.16 37 35 DNT 06C 0.37 38 22 12 AV3 0

, 38 43 12 AVI O 38 43 12 AV2 0 38 54 14 AV3 0 39 29 DNT TSH 13.66 41 43 DNT 05C 7.02 42 31 DNT AV2 0 42 33 DNT AV2 0 42 44 DNT AV2 18.67 42 47 DNT 06C 0.37 42 62 DNT 06C 0.55

(

l Page 107 of i12

Row Column Indication Location inch Mark 42 63 DNT 06C 0.55 43- 33 DNT AV2 0.12 43 51 DNT 02C 10.08 44 40 DSI 06H 0.56

) 44 41 DNT AV3 0.39 44 42 DNT AVI -033 44 42 DNT AV2 -0.28 44 42 DNT AV3 039 j 44 42 DNT AV4 035 44 43 DNT AVI -038 "

44. 43 DNT AV2 -037 44 43 DNT AV4 035 44 44 DNT AVI -0.4 44 44 DNT AV2 -0.23 44 45 DNT AVI- -0.47 44 45 DNT AV2 -0.23 44 45 DNT AV3 0.32 44 46 DNT AV3 037 44 46 DNT AV4- 032 j 44 47 . DNT AVI -0.53 44 47 DNT AV2 -0.45 i 44 47 DNT AV3 039

44. 48 DNT AVI -0.62 44 48 DNT AV2 -0.47 44 48 DNT AV3 039 44 48 DNT AV4 039 44 49 DNT AV3 032 44 . 49 DNT AV4 039 44 50 DNT. AV1 -0.62 44 50 DNT. AV3 037

-44 50 DNT AV4 037 44 51 DNT AVI -0.46 44 51 DNT AV2 -037 44 51 DNT AV3 028 44 51 DNT AV4 0.35 l

44 52 DNT AV2 -035 45 41 7 AVI -0.15 45 41 7 AV4 0 45 43 15 AVI -0.16 ,

45 44 DNT AVI -0.09 45 44 DNT 06C -0.19 45 44 DNT 06C 0.49 45 46 DNT AVI 0 45 48 DNT AVI -0.13 45 49 13 AVI O 45 49 DNT AV3 0.14 l

Page 108 of 112 r

L . . . . . . . . . _ .

1 I

The following table lists each indication in the steam generator"B."

)

Row Column !ndication Location Inch Mark I 51 DTI TSli -0.23 1 73 DNT 0411 7,48 2 II DNT 03H 32.7 2 62 DNT 03H 32.11 3 26 DTI TSH -0.2 3 30 DNT 06C 16.04 3 30 DNT OSC 36.14 4 3 DNT FBli 14.27 4 3 DNT FBH 14.32 4 56 DNT OlH 8.57 4 78 DNT OIC 24.69 5 4 DNT FBC 10.45 5 32 DNT 03C 28.34 5 35 DNT 04H 16.63 5 36 DNT 03C 26.9 6 3 DNT 03H 20.5 6 91 DNT FBC 1.28 7 39 DNT TSil 45.4 7 58 DNT TSC 26.27 8 14 DNT 01C 22.77 8 35 DNT 021I 17.65 8 35 DNT 05C 37.4 8 4? DNT OlH 20.87 9 2 DNT TSC 2.55 9 2 DNT TSC 3.65 9 62 DNT 0311 20.06 9 80 DNT 02C 4.87 11 45 DNT TSH 47.82 11 77 DNT 0511 26.73 11 77 DNT OSC 28.18 II 89 DNT 05H 32.14 11 89 DNT 05H 32.62 11 89 DNT 05C 31.54 12 5 DNT 01C

• 28.62 12 18 DNT 02C 31.78 13 43 DNT OIC 32.75 14 36 DNT TSC 12.02 14 43 DNT 03H 20.15 14 43 DNT 0$H 35.54 14 44 DNT 05C 7.73 15 4i DNT 02C 14.89 16 4 DNT AVi -0.44 16 89 DNT AV1 0.02 17 86 DNT 0311 35.22 18 88 DNT AV4 0 20 62 NQ1 TSH 0.62 Page 109 of 112

i Row Coluren Indication Location Inch Mark 21 6 DNT AVI -0.39 21 6 DNT AVI 0.45 21 12 DNT OIC 46.72 21 87 DNT AV4 -0.09 22 58 13 AV3 0 22 58 16 AV2 0 23 14 DNT OIC 37.19 23 33 16 AV2 0 23 33 19 AV3 0 r 24 8 DNT AVI O 24 63 DNT 0511 35.67 26 59 DNT OIC 29.15 26 84 DNT AV4 0.14 27 10 DNT TSC 2.49 28 20 DNT OIC 28.88 29 54 DNT AV2 -0.87 29 54 DNT AV2 2.35 29 75 DNT 0411 21.18 29 76 DNT 0611 25.08 29 81 DNT AV2 0 30 61 DNT 03C 10.68 31 80 DNT AV2 0.74 32 38 13 AVI O 32 38 16 AV2 0 32 38 25 AV4 0 32 38 28 AV3 0 32 46 11 AV2 0 32 46 17 AV3 0 32 49 10 AVI O 32 49 7 AV2 0 33 15 DNT AV2 -0.41 33 15 DNT AV3 -0.27 33 67 DNT OlC 26.15 33 71 13 AVI 0 33 77 DNT AV2 0.23 33 78 DNT AV2 0.42 34 36 DNT 0511 47.34 34 36 DNT 0611 0 34 #

46 8 AV3 0 34 66 DNT 02C 46.27 34 67 DNT Olli 17.03 35 {

36 DNT 06H 0.44 35 62 DNT 05C 47.29 35 75 DNT AV2 0.28 36 37 DNT 0511 0.3 36 59 DNT 0211 34.85 37 36 DNT 06C 0.42 37 65 DNT FBli 5.1 Page 110 of 112

Row Column Indication Location inch Mark 37 65 DNT TSC 15.47 37 68 DNT OSC 46.22 37 73 DNT AV2 0 38 30 DNT FBH 14.26 38 69 DNT v6H -0.58 38 69 , DNT 06C -0.68 39 65 DNT AV4 15.36 39 67 DNT AV4 16.11 41- 56 DNT 06C 0.75 42 37 DNT 06H 0.09 44 49 BLG TSH 0.3 45 41 DNT FBH 0.77 45 41 DNT FBH 2.72 45 41 DNT FBH 3.93 45 41 DNT 06H 0.34 45 42 FSI 0$H 41.15 45 49 DSI 06H -0.75 Repaired or Plugged Tubes: No repairs or plugging was required.

UNIT 2 Eddy current testing was performed on Unit 2 during December 1998 and was completed on January 8,1999.

Results will be reported in the 1999 report.

Page111of112 l_ _ . .. . . . . .. .

Vill. REACTOR COOLANT SYSTEM RELIEF VALVE CH ALLENGES OVERPRESSURE PROTECTION DURING NORMAL PRESSURE AND TEMPERATURE OPERATION There were no challenges to the Unit I or 2 reactor coolant syrtem power-operated relief valves or safety valves during nonnal pressure and temperature operation in 1998.

OVERPRESSURE PROTECTION DURING LOW PRESSl'RE AND TEMPERATURE OPERATION There were no challenges to the Unit 1 or 2 reactor coolant system power-operated relief valves or safety valves during low pressure and temperature operation in 1998.

VIV. REACTOR COOLANT ACTIVITY ANALYSIS There were no indications in 1998 where reactor coolant activity exceeded that allowed by Technical Specifications in either Unit I or Unit 2.

Page 112 of 112

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