ML17056B478

Unit 2 Safety Evaluation Rept 1991.

ML17056B478
Person / Time
Site:	Nine Mile Point
Issue date:	12/31/1991
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17056B479	List: ML17056B478 ML18038A489
References
NUDOCS 9111070055
Download: ML17056B478 (84)

Text

Enclosure to NMP2L 1324 r,

I.'KNE MILE POINT - UNIT 2 SAFETY EVALUATION

SUMMARY

REPORT 1991 Docket No. 50-410 License No. NPF-69 9iii070055 911030 PDR, ADOCK 05000410 K ~' 8'~ ~

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Safety Evaluation Summary Report Page 1 of 166 Safety Evaluation No.: 86-013, Rev. 1 Implementation Document No.: Mod. PN2Y86MX109 USAR Affected Pages: N/A System: Various Title of Change: Temporary Test Equipment for Power

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Ascension Testing Description of Change:

This change was for the installation and removal (after the first fuel cycle) of temporary test equipment required for power ascension testing.

Potentiometers, thermocouples, electrical cabinets, and associated cables, conduits, and supports were temporarily installed to monitor the following:

Reactor internals vibration Recirculation piping thermal expansion Balance of Plant systems piping thermal expansion (within drywell)

Balance of Plant systems piping vibration (within drywell)

Balance of Plant systems monitored included main steam, feedwater, reactor water cleanup, main steam safety/relief valves, reactor core isolation cooling, residual heat removal, high pressure core spray, and low pressure core spray.

Safety Evaluation Summary:

None of the installed temporary test equipment had any affect on safe operation or shutdown of the plant. The equipment installation was reviewed to assure seismic adequacy.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 2 of 166 Safety Evaluation No.: 87-001 Implementation Document No.: Mod. PN2Y86MX084 USAR Affected Pages: N/A System: Control Room/Remote Shutdown Room Panels Title of Change: Human Pactors Labeling Study

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Implementation Description of Change:

This modification incorporated human factors changes to mimic, marker plates, and INOP window legend inserts in control room power generation control center panels and remote shutdown room panels.

Safety Evaluation Summary:

This modification ensures that the configuration of the control room and remote shutdown room panels is consistent with the human factor guidelines set forth in the applicable sections of the Human Factors Manual, and implements commitments identified in NMPC letters to NRC dated April 14, 1986 (NMP2L

'685) and June 9, 1986 (NMP2L 0737).

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

0 Safety Evaluation Summary Report Page 3 of 166 Safety Evaluation No.: 87-086, Rev. 1 Implementation Document No.: Hod. PN2Y87HX040 USAR Affected Pages: N/A System: Standby Diesel Generator Title of Change: Addition of Emergency Diesel Generator Signals to Permanent Plant GETARs (Divi.sion I, II, III)

Description of Change:

This modification added the following diesel generator (D-G) signals to the permanent plant GE Transient Analysis Recording System (GETARS) for the purpose of having the GETARS as the data recorder of the diesel generator signals for Division I, II, III standby diesel generators: frequency, watts, vars, AC voltage, phase currents, exciter field volts, start indication, and speed indication. For the Division I and II diesel generators, new panels 2EGP*PNL101 and 103 replaced existing panels 2EGS*XC01 and XC03, and reliable 120 Volt AC (Uninterruptable Power Supply) was brought in to externally power the voltage, watt, var, and exciter field voltage transducers. This was performed to ensure that they operate during the 10 second D-G starting time.

This modification simplifies current diesel generator surveillance requirements by eliminating the need for installation of temporary test equipment.

Safety Evaluation Summary:

The design basis is in accordance with IEEE 308 Criteria for Class 1E Electrical Systems for Nuclear Power Generating Stations, IEEE 384 Criteria for Independence of Class 1E Equipment and Circuits, and Regulatory Guide 1.75.

All components added in this modification are for monitoring purposes only and do not result in any functional changes to the control systems. These components are either Class 1E or have appropriate isolating devices in accordance with Regulatory Guide 1.75.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 4 of 166 Safety Evaluation No.: 87-089, Rev. 1 Implementation Document No.: Gale. No. 12177-ES-235, Rev. 1 USAR Affected Pages: N/A System: Secondary Containment (Reactor Building)

Title of Change: Secondary Containment Design Basis (Negative Pressure Drawdown Time)

Description of Change:

An analysis was performed to calculate the time necessary for the standby gas treatment system (SGTS) to activate and re-establish the secondary containment pressure of -0.25 inches water gauge or less. This evaluation was originally reported in letter NMP2L 1177 dated October 26, 1988. This letter stated that the setpoint on the unit coolers was changed to 88'F and the reactor building unit coolers 413A/B changed to simultaneously start on a LOCA signal.

A subsequent revision to the evaluation has determined that the above-stated actions were not necessary. In lieu of the changes, the following alternate actions were implemented:

1 ~ Monitor reactor building temperature 2~ Monitor service water temperature 3~ Verify AT between reactor building temperature and service water temperature is > 16'F

4. Verify building temperature is > 85'F Safety Evaluation Summary:

Verifying that the above-stated conditions were satisfied assured that the secondary containment drawdown analysis was valid. This change did not affect the design function of the SGTS and did not affect safe operation or shutdown of the plant on the evaluation performed,

'ased it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 5 of 166 Safety Evaluation No.: 87-128 Implementation Document No.: Mod. PN2Y87MX125 USAR Affected Pages: Figures 5.4-13d, 5.4-13e, 5.4-13g System: Residual Heat Removal (RHR/RHS)

Title of Change: . To Prevent Annunciator Window RHR Steam Trap Trouble from Always Being in an Alarmed State Equipment 2RHS*PNL100 Description of Change:

Due to common level switches, annunciator window 5601660 (RHR Steam Trap Trouble) was always in an alarm state unless both divisions of RHR were operating in the steam condensing mode. Since this operating configuration is rare, this modification interlocked level switches 2RHS-LS78A&B and 2RHS-LS97A&B with their respective steam line drain SOVs, which are normally closed when not operating. This will prevent either division level switch from setting off the common alarm when not being used.

This modification added/revised internal and field wiring at 2RHS*PNL100, 2RHS-LS78A&B and 2RHS-LS97A&B, and added four isolating MDR relays to provide the necessary control to interlock the level switches. Also, double fuses were added to isolate Category I panel 2RHS*PNL100 from the added MDR relays.

Safety Evaluation Summary:

The added MDR relays in CKT 2RHSB57 are for annunciation only and are not required for safe shutdown of the plant. If one of the four MDR relay coils failed to open or close, there would be no effect on any operating mode or impact on the Division II bus.

If CKT 2RHSB57 was rendered inoperative because of blown fuses, the solenoid valves 2RHS*SOV71A, 71B, 73A and 73B, which are manually operated open in the steam condensing mode, would immediately be de-energized and closed, thereby isolating the steam drain lines from the reactor building equipment drains.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question. /

Safety Evaluation Summary Report Page 6 of 166 Safety Evaluation No.: 87-133, Rev. 1 Implementation Document No.: Mod. PN2787MX197 USAR Affected Pages: Figure 9.3-9b System: Drywell Drains (DER)

Title of Change: Deletion of 2DER-ED5309, 2DER-ED5312 Description of Change:

As reported in letter NMP2L 1177, dated October 26, 1988, an equipment drain in the secondary containment (2DER-ED5309) was capped to prevent radiation levels in excess of design limits due to shine from the traversing in-core probe cubicle below.

This modification was revised to include capping of equipment drain 2DER-ED5312, and sealing of both drain lines with ICMS product 90. In addition, drain line 2CES-001-101-4 originally draining into 2DER-ED5312 was rerouted to 2DER-ED5307.

The USAR changes associated with this modification were incorporated in USAR Revision 0.

Safety Evaluation Summary:

This modification is non-nuclear safety related and has no interface with safety-related systems. This modification results in a long-term ALARA benefit as it will reduce radiation levels at EL 261'f the secondary containment.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 7 of 166 Safety Evaluation No.: 87-153 Implementation Document No.: Mod. PN2Y87MX231 USAR Affected Pages: Figure 10.1-3h System: Main Steam Title of Change: Bypass Around 2MSS-AOV92A and B Description of Change:

Problems were experienced when placing the moisture separator reheaters (MSRs) in service. When actuation was required, 2MSS-AOV92A failed to open and 2MSS-AOV92B opened unacceptably slow, resulting in a steam hammer event and reactor transient. This modification installed a 2-inch bypass line with a manual isolation valve around each AOV92. The manual valve in the bypass line around AOV92A is labeled 2MSS-V395 and the other 2MSS-V396.

The bypass lines are used to warm up the piping downstream of valves 92A & B and equalize pressure across them. After the steam piping has warmed up, the condensate drained off, and the pressure equalized, 2MSS-AOV92A and B are opened. Once open, the manual bypass valves are closed.

USAR Revision 0 revised Figure 10.1-3h to reflect the installation of the bypass lines and manual valves. This Safety Evaluation Summary was inadvertently omitted from the USAR Revision 0 Safety Evaluation Summary Report, and is therefore being reported at this time.

Safety Evaluation Summary:

This modification is non-safety related and is not required for safe operation or shutdown of the plant. Secondary containment bypass leakage is not increased since the bypass lines are downstream of the MSIVs, and the physical and process characteristics of the bypass leakage path are not changed. The piping downstream of the manual valves is seismically supported,so that failure in the Category II piping will not damage adjacent QA Category I piping or components.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Repoxt Page 8 of 166 Safety Evaluation No.: 87-177 Implementation Document No.: Mod. PN2Y87MX004 USAR Affected Pages: N/A System: Biological Shield Mall (BSM)

Title of Change: Clearance Problem Between Shield Plugs and Pipes Description of Change:

This modification x'emoved a portion of shield plugs 2ISC*SHLD12AP and 2ISC*SHLD12BP, located inside biological shield wall openings. This was necessary to provide adequate clearance between the shield plugs and piping to allow the pipe to move upward during heatup and plant operation without hitting the shield plugs.

Safety Evaluation Summary:

Removing a portion of the shield plugs will not affect safe operation or shutdown of the plant and will not impact other equipment. The increased axea of the two BSW penetrations will have a negligible affect on radiation levels in the drywell.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 9 of 166 Safety Evaluation No.: 88-001, Rev. 1 Implementation Document No.: Mod. PN2Y87MX143 USAR Affected Pages: Figures 5.4-2b, 5.4-2c System: Reactor Recirculation Title of Change: Vibration Sensor Addition to Recirculation Pumps 2RCS*P1A/B Description of Change:

Experience from other plants has shown reactor recirculation pumps failing during power ascension testing due to excessive vibration. To alleviate this concern, a vibration monitoring program for the recirculation pumps was implemented.

Vibration (acceleration) and displacement (proximity) sensors were installed on the recirculation pumps, with the necessary hardware for mounting the associated circuits for the additional vibration monitoring inputs.

This change encompasses a temporary modification that was addressed in Safety Evaluation No.87-106, which was previously reported in letter NMP2L 1239 dated June 11, 1990.

Safety Evaluation Summary:

This modification is non-safety related and is not required for safe shutdown.

The addition of a vibration monitoring program is expected to give early detection of any abnormal variation in pump performance in terms of shaft x, y displacement and x, y acceleration during recirculation flow control valve throttling, thereby reducing the probability of pump failure.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 11 of 166 Safety Evaluation No.: 88-.053, Rev. 1 Implementation Document No.: Mod. PN2Y87MX208 USAR Affected Pages: Sections 6.4, 9.2, 9.4, 9.5, 9A, 9B, 12.3 System: N/A Title of Change: Revised Floor Plan for Rooms Adjacent to Control Room Description of Change:

The floor plan in the Control Room area was revised to reduce the amount of personnel traffic and ease congestion. The following changes were implemented:

1. The present kitchen area was changed to a work release office.

2~ The new kitchen/operational lunch room was located in the area vacated by the I&C shop. (The I & C shop was relocated from El. 306'n the Control Building to El. 261'n the Service Building by Modification No.

PN2Y87MX187.)

3. A Ladies Rest Room was constructed.

Safety Evaluation Summary:

This modification will help to alleviate the personnel congestion problems in the Control Room area. The modifications to the safety related portions of the Control Building HVAC and safety related structural components (i.e.

security doors) are in conformance with current design criteria. These changes do not impact other safety systems or affect safe operation or shutdown of the plant.

Illumination within the modified areas meets or exceeds the requirements of the Illuminating Engineering Society Lighting Handbook.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

0 Safety Evaluation Summary Report Page 10 of 166 Safety Evaluation No.: 88-024, Rev. 2 Implementation Document No.: Mod. PN2Y89MX033 USAR Affected Pages: Figures 1.7-1e, 9.3-20a System: Instrument Nitrogen (GSN)

Title of Change: Abandon 2GSN-SOV154 In-Place and Install New 2GSN-SOV167 Description of Change:

During primary containment inerting, the existing solenoid valve 2GSN-SOV154 malfunctioned and failed to supply nitrogen to operate valve 2GSN-PCV124, thus stopping nitrogen flow to inert the primary containment.

Replacement of the existing solenoid valve would result in extensive skid rework. Therefore, valve 2GSN-SOV154 was abandoned in-place, and new valve 2GSN-SOV167 was installed between valves 2GSN-CV155 and 2GSN-PV124. New stainless steel tubing also was installed from 2GSN-V130 to 2GSN-CV155.

This change was originally installed as a temporary modification, and reported in letter NMP2L1239 dated June 11, 1990. This modification is now permanent.

Safety Evaluation Summary:

The purpose of valve 2GSN-SOV154 is to close the nitrogen header valve 2GSN-PV124 if nitrogen gas temperature drops to the low temperature setpoint. This is a safeguard feature that prevents the flow of low temperature nitrogen gas into the piping distribution system in the event of a trim heater failure.

New valve 2GSN-SOV167 will perform the same design function as abandoned valve 2GSN-SOV154, thus allowing primary containment inerting.

This change does not affect safe operation or shutdown of the plant. All work associated with this modification was performed in the yard area.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 12 of 166 Safety Evaluation No.: 88-088, Rev. 0 and 1 Implementation Document No.: Mod. PN2Y87MZ035 USAR Affected Pages: Table 7.5-1, Sh. 1 System: Service Water (SWP)

Title of Change: Expand Scale for 2SWP*FI13A/B and 2SWP*FI201A/B Description of Change:

Flow indicators 2SWP*FI13A/B monitor and indicate the flow of service water to the residual heat removal heat exchangers. To satisfy Regulatory Guide 1.97 requirements to read 110% of design flow, the scales for flow transmitters 2SWP*FT13A/B and flow indicators 2SWP*FI13A/B were expanded to a 0 10,000 gpm range. In addition, to maintain compliance with human factors, the scales on the indicators at the remote shutdown panel for service water flow to the RHR heat exchangers, 2SWP*FI201A/B were also expanded to 0 10,000 gpm.

This change was partially addressed in Safety Evaluation 88U-326, which was previously reported in letter NMP2L1210, dated October 25, 1989. USAR Table 7.5-1, Sh. 1, was revised in USAR Revision 1" to reflect this modification.

Safety Evaluation Summary:

This modification recalibrated and installed new scales and legends on safety related instruments. This change does not affect the safety function of the service water system or affect any other safety systems. Operability of the instruments was affected only during implementation of the modification.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 13 of 166 Safety. Evaluation No.: 88U-329 Implementation Document No.: ECN>>WSS-030 USAR Affected Pages: Figure 11.4-1d System: Radioactive Solid Waste Title of Change: As-Built Revision to the Radioactive Solid Waste System Description of Change:

This change revised USAR Pigure 11.4-1d to reflect the as-build configuration of the radioactive solid waste system, depicting the vendor-furnished programmable controller and associated interface with the flow switch/indicator 2WSS-PS167/2WSS-FI167 in the extruder evaporator lube oil system.

USAR Figure 11.4-1d was revised in USAR Revision 0 to reflect safety evaluation 88U-329. The safety evaluation summary, for this documentation-only change, was inadvertently omitted from the summary report submitted on April 28, 1989.

Safety Evaluation Summary:

This is a documentation-only change to reflect the as-built configuration of vendor-furnished equipment, to show the interface between the system programmable controller and the lube oil system flow monitoring instrumentation on the system P&ID. There was no work associated with this change, since the equipment was supplied by the. vendor.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 14 of 166 Safety Evaluation No.: 89-017, Rev. 1, 2, and 3 Implementation Document No.: Temporary Mods.89-059, 91-005,91-022 USAR Affected Pages: N/A System: Circulating Water Acid Treatment Title of Change: Manual, Direct Acid Injection to Circulating Water Description of Change:

As reported in letter NMP2L 1258, dated October 31, 1990, under Safety Evaluation 89-017, a temporary method of supplying acid to the circulating water system was implemented. Acid injection was performed at the cooling tower through the use of a commercial tank truck and a temporary pumping system delivering acid to the discharge flumes.

Revision 1 to Safety Evaluation 89-017 addressed replacing the tank truck with a temporary storage tank (8000 gal). Revision 2 addressed a Perm-A-Dike containment basin installed around the storage tank and a penetration through the screenhouse west wall. In addition, Revision 3 addressed draining and transfer of the acid in tanks 2WTA-TK1A,B to a tanker truck via temporary routing of piping from the tanks'ischarge to the tanker truck fill connection.

Safety Evaluation Summary:

Acid treatment of the circulating water is utilized to maintain clean heat exchanger surfaces and to prevent biological growth within the circulating water system's condenser.

These temporary modifications provide an alternate method of performing a nonsafety-related function that will not impact the safe operation or shutdown of the plant. The storage tank is designed for the safe storage of hazardous chemicals, and the containment basin will contain the contents of the tank in the event of its rupture. The installation of the storage tank and containment basin will have no impact on site flooding conditions.

Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 15 of 166 Safety Evaluation No.: 89-029, Rev. 1 Implementation Document No.: N2-OP-33 Rev. 4 USAR Affected Pages: 1.12-11, 6.3-18 System: High Pressure Core Spray (CSH)

Title of Change: HPCS Keep Fill System Alternative while Pump (2CSH*P2) is Inoperable Description of Change:

This evaluation was in response to NRC Notice of Violation 50-410/89-05-03 and allows the HPCS system to remain operable when the system pressure pump (2CSH*P2) is inoperative, provided the following requirements are maintained:

The system suction is aligned with condensate storage tank 2CNS-TK1B, The water level in 2CNS-TK1B is 47 feet or greater, The system piping from the pump 2CSH*P1 discharge valve to the system isolation valve is vented (at the high point vents) every 12 hours, and The "high point vent level low" switch (2CSH-LS143) alarm is operable and not in alarm.

Safety Evaluation Summary:

To maintain the HPCS system in an operable condition the system pump (2CSH*P1) discharge piping must be full of water to prevent water hammer upon system initiation. With the system pressure pump 2CSH*P2 inoperable, maintenance of the requirements outlined above ensures that the system piping is full of water up to the outboard isolation valve (2CSH*MOV107).

While the HPCS system is kept full of water by alignment to the CST (2CNS-TK1B) as described above, water hammer and HPCS response time are not changed. Furthermore, this change will not affect Technical Specification requirements.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 16 of 166 Safety Evaluation No.: 89-046 Implementation Document No.: Mod. PN2Y88MX158 USAR Affected Pages: Figure 7.3-2, Sh. 1 System: High Pressure Core Spray (CSH)

Title of Change: Add Keylock Test Switch in HPCS Injection Valve Logic Description of Change:

This modification installed a test switch in circuit 2CSHN05, on control room panel 2CEC*PNL625, to allow functional testing of the HPCS injection valve 2CSH*MOV107 during cold shutdown. This eliminated the need to lift leads to perform the valve test, thus helping prevent operator error. When the test switch is in the test position, a status light will be illuminated in the control room on panel 2CEC*PNL601.

Safety Evaluation Summary:

This modification enhances operation of the HPCS system by eliminating the need for lifting leads to perform surveillance testing, thus preventing the potential for not relanding lifted leads. This modification will have no impact on the safe operation or shutdown of the plant. While the HPCS system is in "test," it will be declared inoperable and the appropriate action of Technical Specification 3/4.5.2 followed.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 17 of 166 Safety Evaluation No.: 89-047, Rev. 1 Implementation Document No.: Mod. PN2Y89MX038 (Partial)

USAR Affected Pages: Section 9.3 System: Floor Drains Title of Change: Replacement of Miscellaneous Canned Sump Pumps Description of Change:

This modification replaced fifty-seven existing floor drain sump pumps with standard off-the-shelf submersible pumps. One additional pump was added in the turbine building. The sump level switches were replaced with fixed level switches that are actuated by pressure and are less susceptible to damage.

Also, the check valves which prevent backflow into the sumps from the drain header were replaced with ball check valves. In line strainers and piping supplying cooling water to the bearings were deleted since the new pumps do not require these lines. Floor drain sumps in the reactor building, turbine building, radwaste building, and other miscellaneous buildings are affected.

USAR Revision 3 reflects those portions of the modification completed prior to April 30, 1991.

Safety Evaluation Summary:

The equipment involved in this modification serves no safety related function, and its operation or failure to operate does not affect safety related equipment. The function of the existing floor drain system and the parameters under which it operates are not changed. The only change is in the equipment manufacture and model that carries out the functions of the system. The new equipment will be more reliable which will enhance the performance of the system.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

4 Safety Evaluation Summary Report Page 18 of 166 Safety Evaluation No.: 89-062 Implementation Document No.: DRF L12-0785 USAR Affected Pages: 15.0-9, 15.0-12, 15.1-8 and Table 15.0-3 (Changes incorporated in USAR Rev. 2, under Safety Evaluation 90-066 Rev.1)

System: Various Title of Change: Transient Re-Analysis to Disposition Test Data Description of Change:

The licensing-basis transient analyses documented in USAR'Chapter 15 used nominal parameters to calculate the change in critical power ratio (delta CPR). Based upon the analyses in Chapter 15, the operating limit minimum critical power ration (OL MCPR) was established in the plant Technical Specifications. Some of these parameters as measured during the startup tests deviated significantly from the nominal values used in the Chapter 15 transient analyses. The purpose of this safety evaluation was to demonstrate the acceptability of the deviation. The parameters of specific concern are (1) main stream line pressure drop, (2) turbine steam bypass capacity, and (3) feedwater controller runout flow.

Safety Evaluation Summary:

The USAR-documented transient analyses were performed using the previously approved REDY/ODYN methodology. In order to either revise or confirm the licensing-basis OL MCPR, the USAR was reviewed to identify those transients which required reanalysis. Five limiting transient events which determined the OL MCPR Technical Specifications were re-analyzed with the actual measured values for main steam line pressure drop, turbine steam bypass capacity, and feedwater controller runout flow. The re-analysis used the GEMINI methodology. GEMINI has been approved by the NRC for generic application. In all cases for core flow up to 100%, the current Technical Specification operating limits were shown to be bounding. The OL MCPR is adequate in assuring that the MCPR during any event is no lower than the safety limit of 1.06.

The effect of reduced main steam line pressure drop on steam lines stresses was also evaluated. It was determined that steam line stresses are acceptable with the restriction that the plant cannot exceed 75% power with any steam line isolated.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 19 of 166 Safety Evaluation No.: 89-064 Rev. 1 Implementation Document No.: Mod. PN2Y86MX122 USAR Affected Pages: Figures 9A.3-5, 9A.3-6, 9A.3-7, 9.5-1d, 9.5-1f, 9.5-2a, Tables 9.5-3 Sh 3, 9.5-3a System: Fire Protection Water Title of Change: Installation of New Fire Protection Standpipes Description of Change:

This modification installed seven new hose reel stations at locations within the screenwell building/turbine building and access passageway. Five hose reel stations were installed (2 each in the screenwell building, 3 each in the turbine building) which, along with the presently installed hose reel stations, effectively covered the following fire zones: 752 NZ, 727 SW, 728 NZ, 722 NZ, 723 NZ, 724 NZ, and 725 NZ. The remaining two hose reel stations were installed in the access passageway to effectively cover fire zones 611 NW and 715 NZ. With the implementation of this modification, NMPC meets the requirements of NFPA 14 Section 3.2.1.

In addition, an exemption to NFPA 14 Section 4-7.1 has been taken regarding mandatory pressure limiting devices on new and existing hose reel stations.

This exception is described on USAR Table 9.5-3 Sh. 3.

Safety Evaluation Summary:

The installation of these hose reels will not affect the heavy loads criteria, Technical Specifications or the Environmental Protection Plan. The installation complies with existing design and construction specifications and codes. Testing of piping and components is performed as required by Engineering installation instructions and specifications. Fire Protection training and procedures are reviewed and revised as necessary to insure that they address this modification.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

/

Safety Evaluation Summary Report Page 20 of 166 Safety Evaluation No.: 89-065, Rev. 1 Implementation Document No.: Mod. PN2787MX032 USAR Affected Pages: Table 3.9A-12, Figures 9.1-5a, 9.1-5b System: Spent Fuel Pool Cooling (SFC)

Title of Change: . Elimination of 6 SFC Anti-Siphon Check Valves Description of Change:

This modification eliminated six anti-siphon check valves originally installed on return lines to the spargers in the Spent Fuel Pool (2SPC*V300 A&B),

Reactor Refueling Cavity (2SFC*V301 A&B), Cask Handling Area (2SFC*V302) and Reactor Internals Storage Pit (2SFC*V303). The valves were removed and the small bore nipples capped. Anti-siphon protection for the Spent Fuel Pool was provided by drilling through a side wall on each downcomer and installing half-couplings, short one-inch pipe nipples, and downturned elbows to form "down spouts".

Anti-siphon protection for the Cask Handling Area fill and drain line was provided by changing valve 2SFC*V255 from normally closed to normally opened and removing the plug. The anti-siphon function for the refueling cavity and the internals storage pit was provided by requiring operator action to open existing manual vent valves 2SFC*V306, V307, or V308 when required.

Safety Evaluation Summary:

The anti-siphon protection required for the spent fuel pool is being provided by passive piping systems. In the event the vent or downspout were to become plugged, the SAR allows for a failure of the anti-siphon device by recognizing that the system incorporates control room alarmed pool water level, water temperature, and building radiaiton level monitoring systems that will initiate operator corrective action.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 21 of 166 Safety Evaluation No.: 89-067, Rev. 1 Implementation Document No.: Mod. PN2Y86MX188 USAR Affected Pages: Figures 10.1-7f, 10.1-7g, 10.1-7p, 10.1-7q, 10.1-7r, 10.2-3, 10.4-12 System: Extraction Steam (ESS)

Title of Change: Closure of Extraction Steam Isolation Valves from Separate Relays Description of Change:

This modification added two (2) additional master turbine trip relays in the EHC Cabinet 2CEC-PNL848 and six (6) turbine trip auxiliary relays in each of the relay cabinets 2CEC-PNL856 and 2CEC-PNL857. Turbine trip auxiliary relays will be energized by their dedicated Master Turbine Relay and control power supply. All extraction steam isolation valves, non-return valves and other equipment controls associated with the master turbine trip were regrouped by their strings so that in the event of the failure of one Master Turbine Trip Relay or turbine trip auxiliary control circuit the other strings would not be affected.

This change replaced a temporary modification that was addressed in Safety Evaluation No.87-045 (Modification No. PN2Y87MX041).

Safety Evaluation Summary:

This modification does not change the function of master turbine trip nor the parameters under which it operates. This modification increases the overall reliability of ESS feedwater heating during normal operation and minimizes the potential turbine overspeed and water induction during Master Turbine Bus tripping. None of the safety related structures, systems or components are impacted by this modification, and there is no affect on safe operation or shutdown of the plant.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 22 of 166 Safety Evaluation No.: 89-075, Rev. 0, 3, 4 and 5 Implementation Document No.: Mod. PN2Y87MX038 (Partial)

USAR Affected Pages: Figures 9.5-26, 9.5-35 System: Communications Title of Change: Addition of Communication Equipment Description of Change:

This modification added Gaitronic/Communication capabilities in various plant areas by adding phone jacks, speakers, speaker volume controls, handsets, strobe lights, associated wiring and conduit, and administrative controls as required. The modification satisfied the commitment addressed in LER 87-025, and incorporated improvements identified from system verification testing, site operating experience, and NRC Emergency Preparedness Exercise Inspection (10/29/86).

One of the power sources for the Gaitronic communications system is 2VBB-UPS1C. This UPS is currently loaded to full capacity. Therefore, portions of this modification will not be made permanent until power is made available from 2VBB-UPS1C.

Safety Evaluation Summary:

This modification enhances communication capabilities for the performance of surveillance testing, enables personnel to respond to alarms in areas with inherently high noise levels, and adds communication equipment in areas that have been identified as needing communication capabilities. These changes do not diminish the capability of the plant communication systems to provide effective and reliable communications capability necessary for plant personnel during times of: 1) plant accidents and transients combined with total loss of offsite power and 2) use of the remote shutdown panel for a plant shutdown.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 23 of 166 Safety Evaluation No.: 89-076, Rev. 3 Implementation Document No.: Mod. PN2Y88MX190 USAR Affected Pages: Figure 1.2-15, 7.7-36, 7.7-37 System: Process Computer Title of Change: 3D Monicore Core Monitoring System Description of Change:

This modification involved installing a new Digital Equipment Corporation Computer System including a Microvax 3800, two Vax 3100 workstations with printers (one in the computer room and one in the control room) and an RS232 Link between the existing PMS computer and the new Microvax 3800 with 3D Monicore Software. Six additional power cables were run from 2VBS-PNLC102 (UPS1G) to the computer room for new receptacles. A printer cable and a communications cable were run from the computer room to the control room. The 3D Monicore software was installed on the new Microvax 3800 computer. Two new control room annunciators "PMS-3D Core Margin Alarm" and "PMS-3D PCRAT Alarm" replaced existing annunciators "PMS-NSS LPRM Alarm", and "PMS-NSS Program Alarm" on Panel 842. Also, the spare disk drive was daisy-chained to the existing disk drive to act as a back-up.

Safety Evaluation Summary:

General Electric has performed an analysis that concludes that the 3D Monicore model is more accurate than the allowances made for in the previous process computer Pl software nuclear model, thereby justifying its use with current margins. Further analysis is presented in the General Electric Report NEDE-20340-3 Class III, April 1986 Rev 1 Process Computer Evaluation Accuracy.

The new 3D Monicore software on the PMS computer will not adversely impact the existing PMS computer performance, and has no affect on safe operation or shutdown of the plant. In the event that the Microvax 3800 computer is lost, an alternate method has been provided that will allow core monitoring calculations to be performed within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period following the loss of the main system.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

0 Safety Evaluation Summary Report Page 24 of 166 Safety Evaluation No.: 89-077, Rev. 2 Implementation Document No.: EDC 2M00328C USAR Affected Pages: N/A System: Liquid Radioactive Waste (LWS)

Title of Change: Lining of Regenerant Evaporator Reboiler (2LWS-E7) Tubes Description of Change:

Thin-walled tubing was installed within the existing deteriorated tubes of the regenerant evaporator reboiler (2LWS-E7). The new tubing, made of ASME SA 286 TP446, was inserted and then expanded mechanically such that the inner tube is held firmly in place by compressive forces caused by the original tube, ensuring metal-to-metal contact, and rolled at the tube sheets for sealing.

The purpose of this repair was to allow the reboiler (2LWS-E7) to be placed in service in a safe and reliable manner without cross-contamination from the radioactive side process to the clean side steam.

Safety Evaluation Summary:

The new lining of the regenerant evaporator reboiler tubes assures pressure integrity between the radioactive working fluid, tube side, and the clean shell side steam. The overall tube strength is still derived from the original tubes, while the lining serves to provide the pressure boundary for areas where deterioration by corrosion pitting has caused random wall thinning. Possible contamination of the "clean steam" side of the reboiler due to tube failure has been evaluated per the requirements of IE Bulletin 80-

10. Precautions have been implemented to limit contamination should the reboiler be operated with leaking tubes.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

0 Safety Evaluation Summary Report Page 25 of 166 Safety Evaluation No.: 89-078, Rev. 1 Implementation Document No.: Mod. PN2Y86MX105 USAR Affected Pages: 9.1-7, 9.1-23 Appendix 9C, 9C.8-1; Figure 5-4 System: Reactor Building Cranes and Elevators Title of Change: Permit Travel of Polar Crane Main Hoist Over Spent Fuel Pool Restricted Area Description of Change:

During normal fuel handling operations the spent fuel pool gates are moved from their normal position to their stored position on the side of the fuel cask storage pool. Eventually, spent fuel in the fuel pool will extend into the safe load path that the spent fuel gate must travel. This modification added a restriction area bypass switch to allow bypassing of the interlocks that control movement of the polar crane main hoist over the spent fuel storage pool, thereby enabling the Reactor Building polar crane (RBPC) to operate over the spent fuel pool and the spent fuel if necessary.

Safety Evaluation Summary:

This modification was addressed in a license amendment request submitted to the NRC by NMPC in letter NMP2L 1203 dated July 26, 1989, and supplemented by letter NMP2L 1221 dated December 14, 1989. The NRC accepted this change with the issuance of License Amendment No. 20 on July 17, 1990.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 26 of 166 Safety Evaluation No.: 90-002 Implementation Document No.: Mod. PN2Y86MZ084 USAR Affected Pages: 9.2-8, 9.2-9, 9.2-10; Figure 9.2-2, Sh. 4 System: PGCC/Service Water Title of Change: - Human Factors Changes Description of Change:

This modification changed name plates and annunciator window engravings on main control room panel 2CEC*PNL601 for service water supply header instrumentation. These changes were made to maintain consistency between the system logic diagram "condition" descriptions, panel drawings, and actual field-installed marker plates and annunciator windows.

Safety Evaluation Summary:

This modification does not affect the design or function of the service water system. The marker plate and annunciator window changes were made to maintain the consistency requirements of the NMPC Human Factors Manual and recommendations of NUREG-0700.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 27 of 166 Safety Evaluation No.: 90-006 Implementation Document No.: Mod. PN2YSSHX159 USAR Affected Pages: Figures 1.2-1, 9A.3-1 System: Security Fence Title of Change: Free-Standing E-Field and Nuisance Fence Description of Change:

, This modification included relocating (approximately 1,000') "E-Field" mounted on the north security fence to a free standing "E-Field" midway between the inner security fence and the outer nuisance fence on the north perimeter of Unit 52; adding a new nuisance fence (approximately 1,100') on the east perimeter of Unit 52, from the existing Security Building to the north perimeter fence; relocating (approximately 1,100') "E-Field" mounted on the east security fence to a free standing "E-Field" midway between the inner security fence and the new outer nuisance fence on the east perimeter of Unit 52; relocating existing ditch, manhole, and concrete drain pipe approximately 12'o the east of their present location; filled in existing areas with the material removed for the new drainage arrangement; relocated existing fire hydrant No. 703 to a location outside of the new nuisance fence on the east perimeter; removed insulated storage building, and removed additions on the west wall of the main pipe fabrication building.

Safety Evaluation Summary:

The construction activities and site changes do not result in significant elevation changes. The relocation of the existing drainage ditch does not affect the general site grading and will not affect the drainage of the surrounding area. The site changes also do not adversely affect the exterior barriers around the plant buildings that are used to divert the PMP flood from the immediate watershed encompassing the site.

The relocated fire hydrant 703 will continue to provide coverage to the same area as was previously protected without causing hydrant spacing problems or new unprotected hazards. Hydrant 703 is not protecting safety-related equipment.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 28 of 166 Safety Evaluation No.: 90-008 Rev. 1 and 2 Implementation Document No.: Mod. PNZY86MZ041 USAR Affected Pages: Figures 6.2-71a, 6.2-71b System: Containment Atmosphere Monitoring System (CMS)

Title of Change: Modification and Relocation of Primary Containment Humidity Analyzers Description of Change:

During performance of the April 1986 Integrated Leak Rate Test (ILRT), three of the six humidity analyzers failed due to the compounded effects of high particulate (dust) levels in the drywell, high humidity in the suppression chamber, lack of a "track and hold" feature on the analyzers, and the inaccessibility of equipment to the test engineers during the test.

To correct these deficiencies, the analyzers were separated from the sensors and relocated outside the primary containment. To do this, four new junction boxes were added outside the primary containment, and the existing data cable to the sensors inside the primary containment were replaced with a new 14 conductor cable. Also, a remote operated vacuum pump with a 5 to 10 micron non-hygroscopic filter was mounted at the inlet to each sensor. This modification also provided a 0 to 10 VDC output to utilize the "track and hold" feature of the control unit.

The modification also added primary fuse protection to cable numbers 2CMSNNC501 and 510 going to primary containment penetrations 2CES-Z40E and 35E, respectively. This was done to comply with Regulatory Guide 1.63.

Safety Evaluation Summary:

This modification, which reconfigures the non-safety related humidity analyzers used for the ILRT, has no impact on the safety operation or shutdown of the plant. Work on the affected primary containment electrical penetrations was performed with approved governing procedures, and appropriate leak rate testing was performed to ensure primary containment integrity. The added electrical cables in each penetration are energized only during shutdown conditions; therefore, electrical penetration overcurrent protection during an accident is of no concern.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

0 Safety Evaluation Summary Report Page 29 of 166 Safety Evaluation No.: 90-011 Rev. 1 Implementation Document No.: Mod. PN2Y89MZ089 USAR Affected Pages: Figure 10.1-5e System: Condensate Booster Pump Lube Oil (CNO)

Title of Change: Condensate Booster Pump Lube Oil System Modifications Description of Change:

This modification replaced the existing three-way valves with separate isolation valves for each filter. This arrangement allows the operator to place the clean filter in operation prior to isolating the other filter.

To prevent overpressurization, in the event both filter paths are inadvertently closed, a relief valve was installed upstream of the filter isolation valves.

e In addition to the above, the following modification:

changes were made as part of this Vent and drain valves were installed on each filter. Previously, only vent and drain plugs were provided to the filter housing.

The change improved the maintenance of the filters.

2~ The setpoint for pressure differential indicating switches 2CNO-PDIS3A/B/C were calculated and revised consistent with the other changes.

3. To prevent actuation of the PDIS due to oscillations in the pressure readings, snubbers were installed on each PDIS.

4~ Added access covers to the oil sumps (2CNO-TK1A/B/C) to facilitate maintenance of the strainers without dismantling any piping.

5. A 3/4-inch valved connection was provided on the sump for connecting a portable purifier in the future. The existing sump drain connection will be utilized.

6. Editorial Changes Check valves 2CNO-V2A/B/C and V3A/B/C were incorrectly shown in FSAR Figure 10.1-5e as normally closed (i.e.,

area is darkened).

Safety Evaluation Summary Report Page 30 of 166 Safety Evaluation No.: 90-011 Rev. 1 Description of Change: (continued)

This modification was also needed to prevent nuisance alarms in the Control Room caused by the actuation of PDIS3A/B/C. This condition was corrected by the changes described above.

Safety Evaluation Summary:

The changes implemented by this modification do not adversely impact the design function of the CNO system. The modification improves the overall reliability of the system by preventing the filter gasket failure. The elimination of nuisance alarms helps alleviate the human factors concerns.

None of the safety-related structures, systems or components are impacted by this modification. The modification does not adversely impact the capability to shutdown the plant safely and to maintain the plant in a safe shutdown'ondition.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

0 Safety Evaluation Summary Report Page 31 of 166 Safety Evaluation No.: 90-014 Xmplementation Document No.: Mod. PN2788MX166 USAR Affected Pages: Figures 1.2-28, 12.3-21, 12.3-54 System: N/A Title of Change: Provide Emergency Egress from the Roof of Service Water Pump Bays Description of Change:

This modification installed a standard detail egress ladder on the outside wall of the service water pump room extending from elevation 280'own to the ground floor of the screenwell building (elevation 261').

The addition of this ladder provides free and unobstructed egress from the roof of the service water pump bays in case of an emergency.

Safety Evaluation Summary:

This non-safety related minor modification does not impact the safe operation or shutdown of the plant. Failure of the egress ladder would not affect either the floor at elevation 261'r the east wall, which are safety related.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 32 of 166 Safety Evaluation No.: 90-017 Implementation Document No.: DRF AOO-835-5 USAR Affected Pages: N/A System: Feedwater Title of Change: Feedwater Pump Loose Part Analysis Description of Change:

On December 26, 1989, NMP2 was shut down due to excessive feedwater system vibration. Upon investigation, it was discovered that the feedwater pumps, 2FWS-P1A, B, and C were damaged. This damage was the result of the dislocation in each pump of a rectangular piece of the flow splitter approximately 5" X 3" X 7/8" thick in size. The parts separated from pumps 2FWS-P1C and B were recovered essentially intact, but were damaged. The part from 2FWS-P1A was not recovered.

This analysis assumed that the part/parts were in the feedwater piping and could travel with system flow. The purpose of this evaluation was to determine the possible safety consequences of continued operation of the feedwater system with the loose part.

Safety Evaluation Summary:

The existence of the lost part from feedwater pump 2FWS-P1A does not present a safety concern. The effect of the loose part(s) on feedwater heating, feedwater flow, primary containment isolation and RPV water level was analyzed and found to be acceptable and bounded by existing USAR analysis. The lost part was additionally analyzed by General Electric (GE) in terms if its effect once inside the vessel. GE's analysis concluded that the lost part will not be a concern in terms of:

1. The potential for fuel bundle blockage and subsequent fuel damage.

2. The potential for control rod interference.

3. The potential for corrosion or other chemical reaction with reactor material.

Based on the evaluation performed, it is concluded that the continued operation of the feedwater system with the lost part does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 33 of 166 Safety Evaluation No.: 90-018 Implementation Document No.: Mod. PN2Y88MX197 USAR Affected Pages: Figures 9.3-11e and 10,1-8c System: Radwaste Auxiliary Steam Title of Change: 2ASR-V34 and V35 Drain Lines Description of Change:

The radwaste auxiliary steam header supplies steam to the radwaste evaporators as part of radwaste processing.

This modification provided drain line runs from the two steam header drain valves, 2ASR-V34 and 2ASR-V35, to drains 2DFW-DNF0801 and 2DFN-ED3402 respectively, to drain condensate from the steam header.

Safety Evaluation Summary:

This modification does not affect safe operation or shutdown of the plant.

Should a failure of a drain line occur, the effluent would be recovered by the floor drain system, the same system being utilized by the piping. This would not cause a release of any radioactive material. No potential safety hazards or system interactions are created by this change.

Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 34 of 166 Safety Evaluation No.: 90-019, Rev. 0, 1, and 2 Implementation Document No.: Mod. PN2Y89MZ018 USAR Affected Pages: Pigures 9.1-5c, 9.1-5d, 9.1-6, Sh. 4 System: Spent Fuel Pool Cooling and Cleanup (SFC)

Title of Change: ~

SPC Filter Discharge Strainer Description of Change:

This modification installed one "Y" type strainer and bypass in the common 8" diameter outlet line from the SFC filter/demins (2SFC-FLT 1A, B) outside the equipment cubicle. This strainer traps resin powder that may escape the demineralizer resin beds during pre-coating operations. A differential pressure transmitter initiates a local alarm and a system alarm in the control room for excessive pressure drop across the strainer. Strainer blowdown is directed to the radwaste system, and a condensate water connection was provided for strainer backwash.

Safety Evaluation Summary:

This modification was made to the nonsafety-. related cleanup section of the SFC system, outside of the Category I isolation valves 2SFC*AOV153, 154. It will help ensure that SPC pool water quality is maintained after filter/demin backwash and pre-coat operation.

Revision 1 to the safety evaluation addressed relocation of sprinkler pipe and heads to provide clearance for the bypass line. Revision 2 addressed an additional change to USAR Figure 9.1-5c to depict the blowdown valve air operator, and a change to Figure 9.1-5d to depict the tie-in to the existing filter/demin drain line.

This modification has no impact on the safety-related fuel pool cooling function of the SFC system.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 35 of 166 Safety Evaluation No.: 90-025 Rev. 1 Implementation Document No.: Mod. PN2Y89MZ040 USAR Affected Pages: Figure 10.1-6B System: FWS Peedwater Title of Change: Removal of Test Connection Valves 2PWS*V16C and 2PWS*V18C Description of Change:

This modification removed a test connection in the feedwater system (PWS) that was no longer required and was in the removal path of the main steam (MSS) safety relief valves (SRV). The connection was cut off at the small bore piping nipple coming off the coupling on large bore line 2FWS-012-34-1, capped and socket welded. A pipe support was removed as it was no longer required (BZ-409EE). Metal reflective insulation of the PWS system piping was modified to suit the new configuration.

This safety evaluation also corrected the code class break shown on USAR Figure 10.1-6b for five remaining test connections that were not removed by

~

~

~

this modification.

~ ~ ~

~

For small lines (equivalent to one inch size), the first intervening valve between the small bore line and the large bore has the lower code class designation (i.e., Class 2 rather than Class 1). This is consistent throughout NMP2 on the lines one inch and less that are vents, drains, tests, samples, and fill connections off of the RCPB.

Safety Evaluation Summary:

The installation meets system design pressure and temperature requirements and does not affect the safe operation of the system. This installation has no impact on other safety-related systems. This change will not affect any electrical circuits, fire protection, electrical or lighting circuits. This installation was accomplished using site-approved procedures and included normal QA/QC involvement.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 36 of 166 Safety Evaluation No.: 90-027 Implementation Document No.: Mod. PNZY88MX069 USAR Affected Pages: Figure 9.2-17a, 9.2-18, Sh. 1; 15.7-6 System: Condensate Makeup and Storage (CNS)

Title of Change: Condensate Storage Tank Level Setpoint Change Description of Change:

This modification cleared nuisance alarm 851518 (Condensate Storage Tank 1 A/B Level High) by recalibrating the alarm setpoint. The operating band was revised by the changeout of 4 level switches from float type switches to pressure type switches. These pressure type switches expanded the control range of the condensate storage tank (CST) makeup valve 2CNS-AOV123 and raised the CST low level alarm setpoint to be consistent with normal operational makeup requirements.

Safety Evaluation Summary:

The condensate storage tanks (2CNS-TK1A/B) capacities are designed to meet the requirements of makeup water for safeguard (RCIC & HPCS preferred source),

normal, and refueling conditions. The CSTs are non-safety related, non-seismic tanks and no credit is taken in any transient or accident analysis for CST water inventory for HPCS or RCIC operability. The CSTs are the preferred source of water for RCIC and HPCS due to the superior water quality compared to the suppression pool. Raising the CST high level setpoints will not affect the safeguards water storage requirement. Also, the condensate storage analysis described in USAR Section 15.7.3.1.5 is based on release of tanks'upture the full volume (900,000 Gal) of both tanks as calculated from overall tank dimensions, and is therefore unaffected.

Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 37 of 166 Safety Evaluation No.: 90-028 Implementation Document No.: Mod. PN2YSSMX156 USAR Affected Pages: Table 3.9B-2V, Figure 9.3-17a System: Standby Liquid Control (SLS)

Title of Change: Addition of Test Connections for Valves

. 2SLS*V12 and 2SLS*V14 Description of Change:

This modification installed test connections to provide a leak path during reverse flow testing of check valves 2SLS*V12 and 2SLS*V14. The test connections allow the valves to be tested quarterly, as required by ASME Section XI, and fulfill a commitment made in the ISI/IST plan to have these installed by the first refuel outage (Relief Request SLS VRR2). The test connections are located in a flanged pipe spool which replaced an existing spool at the discharge of each SLS pump. The test connection for each check valve consists of a sock-o-let, piping, 2 valves, coupling, and pipe plug.

Safety Evaluation Summary:

The test connections are installed in an ASME III Class 2, safety related system. The materials of construction are compatible with the existing installation.

The test connection design is similar to that used on other safety related systems throughout the plant, but incorporates the use of stellite free valves as an effort to reduce cobalt sources in the plant. The additional weight of the test connections changed the as-built nozzle loads on pumps 2SLS*P1A and 2SLS*P1B, but is within the allowable limits. The installation meets system design pressure and temperature requirements.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 38 of 166 Safety Evaluation No.: 90-030, Rev. 2 Implementation Document No.: Mod. PN2Y87MX035 USAR Affected Pages: Chapters 1, 6, 7, 8, 9, 10, 11, 18 System: Various Title of Change: Detailed Control Room Design Review (DCRDR) Modifications Description of Change:

Changes were made in the control room, remote shutdown room, and relay room to resolve human engineering issues consistent with NUREG-0700 guidelines. This safety evaluation evaluated human engineering discrepancies (HED's) which were required to be completed during the first refueling outage consistent with the requirements of license condition 2.C.(9), as stated in the NRC's SSER 5, Section 18.1. Human factor issues discovered subsequent to the issuance of SSER 5 were also evaluated.

Appropriate changes were made to site operating procedures to reflect the resolution of HED's (annunciator window label changes, addition of new instrumentation, computer point ID changes, etc.).

The effect these changes have on plant equipment during the implementation phase was addressed via the signoff of equipment out-of-service tags by operations. The operability of equipment addressed by plant technical specifications and any required compensatory action (entering of action statements) during the implementation phase was also addressed by operations.

Upon completion of the implementation of the resolution of the HED's, appropriate retesting was performed to ensure that plant equipment was properly restored to its original configuration prior to declaring associated equipment/systems operable.

Safety Evaluation Summary:

The resolution of the HED's is consistent with NUREG-0700 guidelines and will enhance the ability of the operator to respond to transients and accidents by improving the operator/machine interface. A mechanism for the NRC staff to monitor and review changes to HEDs is described in License Amendment No. 24 (issued on December 18, 1990), which deleted License Condition 2.C(9).

Based on the evaluation performed, it is concluded that this temporary change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 39 of 166 Safety Evaluation No.: 90-031, Rev. 1 Implementation Document No.: EDC 2M10064 USAR Affected Pages: Figures 9.5-1d, 9.5-1e, 9A.3-6 System: Pire Protection Water (FPW), Pire Protection Foam (PP)

Title of Change: Change Foam/Water Pire Hose Reel Water Supply Isolation Valves from Normally Open to Normally Closed Description of Change:

The Unit 2 Turbine Building is equipped with foam/water fire hose reel (F/W FHR) stations that protect the turbine generator at various elevations. These FHR stations are used in manual firefighting around the turbine generator and have the capability to provide either water or water with 3X foam concentration. The fire protection water (PPW) and the fire protection foam (PPF) systems meet at the P/W PHRs and are separated by a normally closed foam blocking valve and check valve. A problem was identified whereby low expansion protein foam from the FPP system had leaked into the FPW system by leaking past the foam blocking valves, check valves and through the normally open P/W FHR water supply isolation valves (WSIVs) at each of the 14 F/W FHRs.

To alleviate this problem, the F/W FHR WSIVs were changed from the normally open position to the normally closed position. In addition, drawing EB-22E (USAR Figure 9A.3-6) was corrected to show fire hose reel FHR-7 as a water hose reel rather than a foam system hose reel.

Safety Evaluation Summary:

Changing the F/H FHR WSIVs from the normally open to the normally closed position will ensure that foam from the PPP system will not enter the lower pressure FPW system, which could cause possible adverse affects on system reliability and operability. None of the fire zones protected by the F/W FHRs and FHR 7 contain safe shutdown equipment. The closed WSIVs are in close proximity to the FHR angle valve, and are easily opened manually. Pire brigade personnel were trained regarding the change in normal WSIV position.

Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 40 of 166 Safety Evaluation No.: 90-033 Implementation Document No.: Mod. PN2Y88MX174 USAR Affected Pages: Figure 8.3-6 Sh 16 & 17 System: Standby Diesel Generators Title of Change: Modify Div. I and II Diesel Generator Starting Circuit Description of Change:

Diesel Generators 2EGS*EG1 and 2EGS*EG3 utilize dual class IE emergency start/run circuitry (primary and back). Also, each diesel generator utilizes a non-1E start/stop circuit for manual start/stop and testing. During FMEA (failure modes and effect analysis) update, it was discovered that the loss of offsite power (LOOP) signal to diesel generator emergency start/run circuit is not sealed in; a spurious energization of non-1E devices could shutdown the diesels.

This design deficiency was uncovered during a FMEA update and was addressed in LER 88-44, Revision 1.

This modification performed the following:

Backed out temporary modifications88-192, 88-193,88-206, and 88-207, as described in LER 88-44 Revision 1.

2. Added a class 1E HFA latching relay in both 4160V switchgears. This relay is dedicated to start the diesel generator upon loss of offsite power. The LOOP signal will be sealed-in by the HFA latching relay and a continuous signal will be available for diesel emergency running circuitry. The relay is reset by a closed signal of either the normal or alternate offsite power feeder breakers. This modification ensures that common mode failure of non-1E devices will not shutdown diesels when they are running in emergency mode with offsite power not available.

Safety Evaluation Summary:

This modification was performed to ensure compliance to Regulatory Guide 1.53 for single failure criteria for standby diesel generators. The modification improves the safety-related function of the standby diesel generators.

The seal-in function of this modification for the LOOP signal is similar to the LOCA signal previously analyzed.

Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.

Safety Evaluation Summary Report Page 41 of 166 Safety Evaluation No.: 90-034, Rev. 1 Implementation Document No.: Mod. PN2Y88MX110 USAR Affected Pages: Figures 5.4-2b, 5.4-2c, 9.2-3b System: Reactor Recirculation (RCS)

Title of Change: Reactor Recirculation Pump Stuffing Box Modification Description of Change:

Reactor recirculation pumps at LaSalle 1 and WNP-2, with basically the same stuffing box design as Nine Mile Point 2, have experienced failures of the mounting hardware securing the upper wear ring and bearing assembly to the stuffing box. As a result of these failures, the stuffing box design was revised to increase the number, change the material and thread type, and change the size of the wear ring mounting capscrews.

In order to facilitate motor removal to gain access to the stuffing boxes, several small bore pipes and conduits, as well as their supports, were redesigned.

Additional changes included the following:

Replaced the existing single plane vibration monitors with 2 monitors installed in different planes, 90'rom each other.

Added a thermocouple to measure the seal water inlet temperature for each recirculation pump.

Relocated an expansion joint closer to the cooler on one of the lines to a cooler on the "B" recirculation pump motor.

Made minor changes to the saddles used to remove and install the personnel air lock.

Made minor revisions to removable hand rails and grating to reduce exposure time during installation and removal of the removable seal under the RCS pumps.

Modified the reactor recirculation pump motor handling cart to provide redundant lifting capacity in accordance with NUREG-0612.