ML17059A504
| ML17059A504 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 12/31/1994 |
| From: | NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML17059A503 | List: |
| References | |
| NMP2L-1504, NUDOCS 9411020168 | |
| Download: ML17059A504 (138) | |
Text
Enclosure to NIVIP2L 1504 NINE MILE POINT UNIT 2 SAFETY EVALUATION
SUMMARY
REPORT 1994 Docket No. 50-410 License No. NPF-69 94110201b8 941028 PDR ADOCK 05000410 K PDR
~ ~,
Safety Evaluation Summary Report Page 1 of 136 Safety Evaluation No.: 87-162 Rev. 0 L 1 Implementation Document No.: Mod. PN2Y87lVIX100 USAR Affected Pages: Figures 11.5-2, 11.5-2a, 11.5-6, 11.5-6a System: Digital Radiation Monitoring System (DBMS)
Title of Change: Delete Gaseous Flow Control Boards Description of Change:
Revision 0 of this safety evaluation evaluated a change to 21 gaseous process radiation monitors.
Revision 1 of this safety evaluation reduced the scope of this modification to 4 gaseous process radiation monitors.
This modification replaced the automatic flow control valves and their associated flow control printed circuit boards with manual flow control valves. The changes apply to the following digital radiation monitoring system (DRIVIS) monitors:
2GTS-CAB105 2OFG-CAB13A 2OFG-CAB13B 2HVW-CAB196 Evaluation Summary:
'afety This modification increases the operability and reliability of the DRMS system by eliminating the automatic sample flow control valves and their associated printed circuit boards (on four monitors) and replacing them with hand control valves. The modulated sample flow has been determined not to be required and to be a contributing factor in generating spurious alarms.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 2 of 136 Safety Evaluation No.: 88-009 Implementation Document No.: Mod. PN2Y86MX165 USAR Affected Pages: 9.5-42; Tables 8.3-1 Sh 25, 8.3-5 Sh 5; Figures 8.3-10, 9.5-40a System: HPCS Diesel Air Start System Title of Change: Replacement of Petter Diesel Description of Change:
This modification replaced a nonsafety-related diesel-driven motor in the Division III diesel air start system with a nonsafety-related electric motor. This modification also provided a new power feed between 2NHS-!VICC008 and EGA-MST004.
Safety Evaluation Summary:
The air compressors and drive motors that supply air to the air start system do not provide a safety function and will not impact the safe operation of the HPCS diesel. Due to the current failures associated with the diesel drive motor, this modification will increase air system reliability.
Based on the evaluatio'n performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 3 of 136 Safety Evaluation No.: 89-001 Rev. 2 L3 Implementation Document No.: Simple Design Change SC2-0132-90, SC2-0091-93 USAR Affected Pages: Figures 9.3-1E, 9.3-1F System: Instrument Air Title of Change: Capping of IAS Drain and Test Valves Assemblies on ADS Accumulators Description of Change:
Revision 1 to Safety Evaluation 89-001 was reported in June 1990. Revision 2 addressed Temporary Modification 89-006B as permanent. Thus, the drain valve assemblies for the ADS accumulators 2IAS"TK34 through TK38, the associated instrument line drain valve assemblies for 2IAS"TK32 through TK38, and the test valve assemblies inside primary containment for the IAS supply headers to the ADS accumulators, will be permanently maintained with seal welded plugs or caps.
Revision 3 returned valves 2IAS"V190, 193, 653, 654 to their original design configuration.
Safety Evaluation Summary:
The valve assemblies noted above were found to be leaking excessively during surveillance test N2-IAS-ADS-R106. Temporary Modification 89-006B revised the valve assemblies by either removing/cutting the pipe nipple and cap downstream of the second root valve, and seal welding a plug to the second root valve, or removing/cutting the second root valve and seal welding a cap to the nipple of the first root valve. The drain valves sealed by Temporary Modification 89-006B will be left as is as they are not required for depressurizing the tanks and instrument lines. The pressure transmitter vents are used for this purpose.
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 136 Safety Evaluation No.: 89-006 Rev. 1 Implementation Document No.: Simple Design Change SC2-0450-91 USAR Affected Pages: 9.4-3, 9.4-4; Figures 9.4-1c, 9.4-5 Sh 6 System: Control Building Air Conditioning (HVC)
Title of Change: Control Room/Relay Room Smoke Removal-Electrical Disabling of 2HVC-HVU1 Description of Change:
Revision 0 of Safety Evaluation 89-006 was reported in June 1990. Revision 1 addresses the design change that made Temporary Modification 89-057 permanent. The power leads to makeup air unit 2HVC-HVU1 were lifted from 2NHS-MCC007 CUB7B and the breaker left in the off position.
Safety Evaluation Summary:
This change modifies the control building smoke removal operation. Failure of safety-related air-operated damper 2HVC"AOD142 (fail open) with the inadvertent operation of makeup air unit 2HVC-HVU1 creates a potential for outside air to be introduced into the control room during an accident. To preclude the inadvertent operation of the unit, the power lead cable is lifted until such time as needed.
There is no impact on safe operation or shutdown of the plant. There is no Technical Specification impact.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 5 of 136 Safety Evaluation No.: 89-010 Rev. 1 Implementation Document No.: Simple Design Change SC2-0072-93 USAR Affected Pages: Figure 11.4-1b System: Solid Radioactive Waste Title of Change: Removal of 2WSS-RV320 Description of Change:
Revision 0 of Safety Evaluation 89-010 was reported in June 1990. Revision 1 addresses Temporary Modification 89-010 as a permanent modification. This change deleted a thermal relief valve in the suction piping at the waste concentrate transfer pump which is in the evaporator bottoms handling system. When this system was operating and being flushed, the thermal relief valve caused operational problems. This modification removed the thermal relief valve, and added a blank flange which maintains system integrity without jeopardizing system operation.
Safety Evaluation Summary:
This change enhances the integrity and performance of the evaporator bottoms handling system. When this system was operational and being flushed, the flushing pressures were lifting the thermal relief valve and the valve was not reseating. Currently, the system is used on an infrequent basis for short periods of time. Additionally, when the system is nonoperational it is drained of process fluid, therefore precluding the system configuration that would require a thermal relief valve in the pump suction.
Since a thermal relief valve is not required and the pressure integrity of the system is maintained, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 6 of 136 Safety Evaluation No.: 89-047 Rev. 2 L 3 Implementation Document No.: Mod. PN2Y89MX038 USAR Affected Pages: Figures 9.3-9c thru 9.3-9e, 9.3-12g, 9.3-13 Sh 1, 9.3-16 Sh 5, 6, 8, 9 System: Floor Drain Sumps Title of Change: Replacement of Miscellaneous Canned Sump Pumps Description of Change:
As previously reported in October 1991, Revision 1 of Safety Evaluation 89-047 addressed the replacement of 57 existing floor drain sump pumps with standard off-the-shelf submersible pumps. USAR Revisions 3 and 4 included changes to reflect those portions of the modification which had been completed. Revision 3 of Safety Evaluation 89-047 addresses a decrease in the scope of the modification from 57 to 55 sump pumps.
USAR Revision 7 reflects the completion of Modification PN2Y89MX038.
Safety Evaluation Summary:
As described in the October 1991 Safety Evaluation Summary Report, the equipment involv'ed 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.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 7 of 136 Safety Evaluation No.: 90-021 Implementation Document No.: EDC 2E10032 USAR Affected Pages: Figure 9.2-2 Sh 11 System: Service Water System (SWP)
Title of Change: Revision of 2SWP"TE145ARB QA Category Description of Change:
This change revised USAR Figure 9.2-2 Sheet 11 to change the QA category designation for temperature elements 2SWP"TE145ALB from QA Category I to non-QA Category I. This was accomplished by changing the asterisk (") in the equipment mark number to a dash (-) (i.e., the new designation is 2SWP-TE145ARB).
Safety Evaluation Summary:
Temperature elements 2SWP-TE145ALB are connected to nondivisional (black) cables to provide service water loop A8cB discharge temperature inputs to the plant computer. They do not serve any safety-related function. The associated thermowells 2SWP"TW145ASB are safety related only to provide a pressure boundary on the SWP system piping.
This modification is a documentation only change and does not adversely affect the SWP system or its ability to support plant operation.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 8 of 136 Safety Evaluation No.: 90-048 Rev. 3 54 Implementation Document No.: Mod. PN2Y89MX085 USAR Affected Pages: 3.6A-34, 3.6A-35, 3.6A-36, Tables 1.9-1 Sh 2, 20, 21, 3.6A-1 Sh 1 thru 5, 3.6A-2 thru 3.6A-71, 3.9B-2x Sh 2, 4, 6A.9-3, 6A.9-5, 6A.9-6; Figures 3.6A-12 thru 3.6A-49, 6A.9-1, 6A.9-3 System: Feedwater (FWS), Main Steam (MSS),
Reactor Water Cleanup (WCS), Residual Heat Removal (RHS), Service Water (SWP)
Title of Change: NMP2 Snubber Reduction Description of Change:
Revision 2 of Safety Evaluation 90-048 was reported in October 1991. Revisions 3 and 4 provide additional evaluation of the NMP2 snubber reduction program.
The snubber reduction modification program reduces the number of mechanical'nubbers on NMP2 safety and nonsafety-related piping systems by reanalyzing the piping systems for snubber removal or snubber replacement with struts.
Safety Evaluation Summary:
Due to failure rates associated with snubbers, snubber removal results in piping systems that are more reliable. Other benefits of the program include reduced long-term maintenance, inspection and test requirements. The USAR is being updated to reflect revised analysis criteria and analysis results. NMP2 Technical Specifications are unaffected.
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 136 Safety Evaluation No.: 90-126 Rev. 1 Implementation Document No.: Simple Design Change SC1-0187-91 USAR Affected Pages: N/A System: Technical Support Center Emergency Ventilation System Title of Change: TSC MD-3 HVAC Damper Description of Change:
Temporary Modification 5316 consisted of de-energizing the damper motor actuator and mechanically restraining damper 212-41 (MD-3) at a position for a flow rate of 3000 cfm a 10%. This change was evaluated under Safety Evaluation 90-054, Rev. 0, and reported to the NRC in 1992.
This change (SC1-0187-91) permanently de-energizes the damper motor actuator and mechanically retires damper 212-41 in a fixed open position.
Safety Evaluation Summary:
This design change is consistent with the applicable system design and quality requirements. Based on analysis and performance testing, this change does not affect the ability of the Technical Support Center HVAC system to perform within its design basis.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 10 of 136 Safety Evaluation No.: 90-311 Rev. 0 L1 Implementation Document No.: Mod. PN2Y88MX133 USAR Affected Pages: N/A System: Material Handling - Radwaste Title of Change: Radwaste CCTV Camera Replacement Description of Change:
The original CCTV system installed in the radwaste building required extensive repairs and maintenance. The original equipment became obsolete and spare/repair parts were not available without special tooling by the manufacturer. The original CCTV system was replaced with a new system. Revision 0 also evaluated the addition of a boom-mounted camera/channel to be installed in the truck bay.
However, after the original camera system was replaced, it was determined that the additional boom-mounted camera was not needed. Revision 1 deletes the evaluation for an additional camera.
Safety Evaluation Summary:
The CCTV system in the radwaste building is passive in nature. This system allows remote monitoring of process handling from a central location. The radwaste CCTV system is nonsafety related and there are no seismic or environmental requirements for the installation. The new equipment is installed to the same standards as the original equipment.
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 136 Safety Evaluation No.: 91-074 Rev. 3 I
Implementation Document No.: EDC 2M10304A USAR Affected Pages: Figure 5.4-16b System: Reactor Water Cleanup and Control Rod Drive-Hydraulic Title of Change: CRD to RWCU Pumps Piping Installation Description of Change:
This change installed a new source of pump seal injection water to the RWCU pumps. The original source of seal injection water was from the RWCU pump discharge while the new source is taken from the CRD pump discharge. This provides better mechanical seal performance for the RWCU pumps. The original seal injection source can be valved in for RWCU pump operation when the CRD pumps are not operating.
Safety Evaluation Summary:
The piping installed is seismically supported and routed so as not to impact any equipment required for safe operation and shutdown of the reactor. The impact of unmonitored flow into the RWCU system has been evaluated and determined to be bounded by existing analyses. The flow removed from the CRD system will not adversely impact the system's ability to recharge the scram accumulators. The safety classification, design, performance, and reliability of the CRD and RWCU pumps are not degraded by this change.
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 136 Safety Evaluation No.: 91-076 Implementation Document No.: Mod. PN2Y89MX138 USAR Affected Pages: Figures 1.2-15 Sh 2, 9.5-6 System: Energy Management System (EMS)
Title of Change: Install EIVIS Remote Terminal Unit (RTU)
Description of Change:
This modification installed an energy management system (EMS) remote terminal unit (RTU) at Unit 2. The RTU provides data acquisition and transfer of the current status of the Unit 2 electric system to a digitized data base at the EMS control center.
The EIVIS RTU panel (2CEC-PNL817) houses the necessary equipment to provide signal conditioning and transmission of data via a dedicated phone line to system power control.
An interface panel (2CEC-PNL816) was installed to house the necessary relays, transducers and test devices for the incoming plant signals. Cables were routed from panels P802, P805, P808 and P869 to obtain the station generator and transformer status. The auxiliary relays multiply the contacts for the 115-kV and 345-kV modification status from P852 and P868. All cables were terminated first to test devices to allow for isolation during maintenance and testing activities. All connections to the RTU are made from this panel.
Interface panel P816 and RTU panel P817 have been located on panel module 2CEC"APF733 in the lower control room el 288'-6".
~
Safety Evaluation Summary:
EMS panels P816 and P817 will be installed in a seismic Category I structure. The panels have been classified as 04, whose function is not required during or following a seismic event, but whose failure could impact the function of a safety-related component. The panels have been seismically mounted in accordance with Regulatory Guide 1.29 to prevent this occurrence.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 13 of 136 Safety Evaluation No.: 91-085 Implementation Document No.: Procedures N2-OP-100A 5. N2-OP-100B USAR Affected Pages: Table 4-1 Sh 3 System: Material Handling System (IVIHS)
Title of Change: Replace the Alarm Light System with Administrative Controls for the Emergency Diesel Generator Cranes, 2IVIHS-CRN2, 3, and 4 Description of Change:
The diesel generator cranes'larm light system referenced in the USAR was omitted from plant design configuration. The function of the alarm light system was to alert operators when the crane was out of its stored position. The only time when the crane load is allowed to be over safety-related equipment would be when the diesel generator is declared inoperative. This is based on calculation MS-1917 which seismically qualifies the cranes in the stored position only.
In the absence of this alarm light system, the operators place a markup tag on the disconnect switch for each of the diesel generator cranes, as a means of controlling the use and movement of the cranes during plant operation. In addition to this means of control for these cranes, operational procedures N2-OP-100A and N2-OP-100B have been revised to include a precautionary statement on the use of these cranes.
Safety Evaluation Summary:
The administrative control for the use of the diesel generator cranes is an acceptable and effective means of replacing the alarm light system.
This change involves an alternative method for the control of the stored position of the nonsafety-related diesel generator cranes while the plant is in the normal operating mode (with the diesel generators operable). The alternative method for the control of the stored position of the cranes provides adequate assurance that the operability of the emergency diesel generators will not be adversely affected during a seismic event.
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 136 Safety Evaluation No.: 91-086 Rev. 1 Implementation Document No.: Simple Design Change SC2-0256-91 USAR Affected Pages: Figure 11.5-3d System: Service Water Title of Change: Delete Solenoid Valves from Liquid DRMS Cabinets Description of Change:
This design change removed solenoid valves originally installed on the liquid monitoring skids to facilitate an automatic purge feature. The removal of the solenoid valves eliminates flow restrictions associated with low flow problems.
Safety Evaluation Summary:
The proposed deletion of the solenoid valves does not alter the operating design basis or postaccident monitoring requirements evaluated in the USAR.
Following deletion of the solenoid valves, the liquid monitoring skids will continue to meet system design pressure, flow and temperature requirements. The change will not affect the safe operation of the system.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 15 of 136 Safety Evaluation No.: 91-096 Rev. 1 L3 Implementation Document No.: EDC 2M10135 USAR Affected Pages: Table 3.9A-12 Sh 3; Figure 9.3-20b System: Nitrogen System (GSN)
Titfe of Change: Deletion of 2GSN"RV32A/B and RV34A/B Description of Change:
This change deleted valves 2GSN" RV32A/B and RV34A/B from the nitrogen system. In order to delete these valves, valves 2GSN"V71A/B and V72A/B were locked open so that overpressure protection could be maintained by 2IASASV19A/B.
Safety Evaluation Summary:
With deletion of 2GSN" RV32A/B and RV34A/B, this change will maintain overpressure protection of the automatic depressurization system N~ supply lines through valve 2IAS"SV19A/B with valves 2GSN "V71A/B and V72A/B locked open. Deletion of valves 2GSN"RV32A/B and RV34A/B will reduce the number of valves requiring in-service testing and eliminate valves that could have potential for problems in the future.
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 136 Safety Evaluation No.: 92-003 Implementation Document No.: Calculation 2-SQ-007 USAR Affected Pages: Table 3.10A-1 System: Containment Monitoring System (CMS)
Title of Change: Revise USAR Table 3.10A-1 for Resistance Temperature Detectors (RTDs) of CMS System Description of Change:
Investigation revealed that the dynamic testing performed on certain RTDs supplied by PYCO, under Design Specification C-041D, did not reflect field installed conditions. The subject RTDs were tested with thermowells but were installed without them. This safety evaluation evaluated the installed condition of the RTDs.
Safety Evaluation Summary:
The safety function of the RTDs in the containment monitoring system is to monitor the drywell and suppression chamber air and suppression pool temperatures, and provide analog signals to the control room plant computer and" recorder, which in turn provides a high temperature alarm. The subject RTDs, which were discovered installed without thermowells, have subsequently been dynamically qualified by supplementing the original test report with an analysis.
The analysis demonstrates that the subject RTDs can perform their intended safety function when subjected to the postulated seismic and hydrodynamic loads. The details of the analysis are provided in NIVlP2 Calculation 2-SQ-007. No design or field modification is required as a result of the analysis.
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 136 Safety Evaluation No.: 92-011 Rev. 1 Implementation Document No.: N2-OP-48 Auxiliary Boiler System USAR Affected Pages: N/A System: Service Water (SWP) and Fire Protection-Water (FPW)
Title of Change: Alternate Cooling Water Supply to Auxiliary Boiler Circulating Pump Seal Coolers and Sample Coolers Description of Change:
This procedure change permitted the installation of a temporary fire hose from fire hose reel 2FPW-FHR32 to service water system drain valve 2SWP-V1011 in order to supply the seal and sample coolers of the auxiliary boiler circulating pumps while the normal supply of service water was isolated during Refueling Outage 2.
Safety Evaluation Summary:
This temporary alteration will have no impact on the safe operation or capability to keep the plant in the safe shutdown condition. The portion of service water which normally supplies the cooling water and the service water isolation valve shall both be administratively controlled by a procedure.
Fire protection capabilities are not degraded at any time, because system pressure and flow will be maintained in the fire protection system while this procedure is in effect.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 18 of 136 Safety Evaluation No.: 92-035 Rev. 1 Implementation Document No.: Simple Design Change SC2-0098-92 USAR Affected Pages: 9.5-62, 9.5-63; Figures 9.5-52b, 10.1-3h, 10.1-4d System: Main Steam and Auxiliary Boiler Systems 1
Title of Change: Permanent Removal of 2lVISS-IVIOV19A &.
2IVISS-MOV19B Description of Change:
This simple design change was initiated to document and provide engineering direction for the permanent removal of valves 2MSS-MOV19A and 2MSS-MOV19B. Their function was to provide isolation between the main steam and auxiliary boiler systems. The subject valves were normally closed and were never used for providing blanketing steam to their respective moisture separator reheaters during startup, heatup and plant cooldown. The inlet and outlet piping shall remain installed but capped. Cables/interlock wiring will be removed as required.
Safety Evaluation Summary:
This simple design change will have no impact on the safe operation or capability to keep the plant in a safe shutdown condition.
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 136 Safety Evaluation No.: 92-037 Rev. 1, 3 Ec 4 Implementation Document No.: Mod. PN2Y91MX005 USAR Affected Pages: 9.4-50, 9.4-53, 9.4-56, 10.4-17, 10.4-18; Table 9.4-1 Sh 4; Figures 1.2-1, 2.4-1, 10.4-7d, 10.4-7j, 10.4-7k, 10.4-7I System: Circulating Water System (CWS)
Title of Change: CWS Chemical Injection and Analysis Facility Description of Change:
This modification involved construction of a new facility for the storage and injection of sulfuric acid, scale dispersant and copper corrosion inhibitor. The facility also includes a laboratory to be used for the analysis of circulating water, microbiology, and diesel oil.
The original CWS acid injection system, which was not in use, has been abandoned in place. The temporary CWS chemical injection system, which was used to inject the chemical into the circulating water, has also been discontinued.
Safety Evaluation Summary:
This modification provides a permanent, reliable and well-controlled CWS chemical addition system.
This system is not required for safe shutdown of the plant, and any realistic accidents/failures that may be associated with this change will not adversely affect the design function of any safety-related structure, system or component. The change does not adversely impact safe operation and/or shutdown of the plant.
A failure of the injection system could impact plant efficiency over a period of time.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 20 of 136 Safety Evaluation No.: 92-045 Rev. 2 Implementation Document No.: LDCR 2-93-UFS-170 USAR Affected Pages: 9.1-32; Table 9.1-4 Sh 2 System: MHR Title of Change: Acceptability of CAVSPAN Gantry/Manlift at NMP2 Description of Change:
In an effort to improve ALARA and improve outage efficiency, NMPC Operations purchased the CAVSPAN system for refueling outage reactor cavity and storage pool decontamination. The CAVSPAN is a reactor refueling cavity-spanning gantry, designed by Applied Radiological Controls, Inc. (ARC) and used for suspending work cages in support of decontaminating the reactor cavity and internal storage pool.
Safety Evaluation Summary:
The CAVSPAN was reviewed per USAR Appendix 9C "Control of Heavy Loads" criteria for acceptability for use over the reactor cavity. Based on this review, it was determined that the CAVSPAN was acceptable for use via the seismic exception evaluation performed. The use of the CAVSPAN system for decontamination of the reactor cavity and storage pool is an alternate method from the method described in USAR Section 9.1 4.2.7, which describes the use of the decontamination platform for these purposes. This change incorporates a description of the alternate decontamination method into Section 9.1 4.2.7. The temporary change in decontamination technique for the reactor cavity and storage pool, as described in this evaluation, will not affect nuclear safety.
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 136 Safety Evaluation No.: 92-047 implementation Document No.: N/A USAR Affected Pages: N/A System: NRC Emergency Telecommunications System (ETS)
Title of Change: Reflection of NRC/ETS Upgrade in the Site Emergency Plan Description of Change:
Generic Letter 91-14 directed licensees to assist in implementing an upgrade to the NRC's Emergency Telecommunications System (ETS). This included replacement of the ENS (red phone) with a more reliable phone system. This replacement was reflected in the Site Emergency Plan (SEP), which previously described the ENS as a dedicated phone (hotline) that rings at the NRC Operations Center when picked up. The new phone requires dialing a 10-digit number listed on the phone instrument. (Control rooms have speed dial capabilities.)
Safety Evaluation Summary:
Replacement of the ENS will enhance the reliability of the ENS and other phones in the system by utilizing the Federal Telecommunication Systems (FTS) 2000 network. This change to the SEP will not affect the safe operation or safe shutdown of either Unit 1 or Unit 2.
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 136 Safety Evaluation No.: 92-051 Rev. 2 Implementation Document No.: N/A USAR Affected Pages: 13.2-26 System: N/A Tide of Change: Modification of Fire Brigade Continuing Training Description of Change:
The annual Fire Brigade training consisting of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> emergency core cooling system (ECCS) and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> electrical distribution courses has been eliminated.
Fire Brigade members still receive training on ECCS and electrical distribution in order to meet the requirements of 10CFR50 Appendix R. Brigade members meet the requirements by taking Introduction to Boiling Water Reactors (BWR), ECCS, and Electrical Distribution within two years of employment.
Safety Evaluation Summary:
The continued training program for Fire Brigade members does not affect the ability of the Brigade to suppress a fire. Brigade members still receive training on ECCS and electrical distribution in order to meet the requirements of 10CFR50 Appendix R.
Removing the 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of ECCS and electrical distribution continual training will bring the Nine Mile Point Fire Brigade training program in line with the industry for plants with a dedicated fire department, while maintaining 10CFR50 Appendix R compliance.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 23 of 136 Safety Evaluation No.: 92-055 Rev. 2 Implementation Document No.: N/A USAR Affected Pages: Figures 1.2-1, 2.4-1, 9.2-8a, 9.5-1b System: N/A Title of Change: Demolition of Present "Area Complex" Building and Construction of Swing/Unit Two Operations Building Description of Change:
The Unit 2 operations building was constructed where the area complex building was located. The area complex building was demolished and the land used for the installation of the Unit 2 operations building. This change consolidates operating activities from various temporary facilities.
Safety Evaluation Summary:
The construction activity of the Unit 2 operations building does not impact the pertinent licensing issues evaluated in the USAR that are associated with hydrologic engineering. The pertinent issues are flooding, local intense precipitation (probable maximum precipitation), and the impact on the air intake accident X/0 (CHi/0), the atmospheric dispersion coefficient.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 24 of 136 Safety Evaluation No.: 92-069 Implementation Document No.: Simple Design Change SC2-0294-92 USAR Affected Pages: Figure 9.2-17b System: -Condensate Makeup and Transfer
/
Title of Change: Remove Relief Valve 2CNS-RV133 Description of Change:
This simple design change removed relief valve 2CNS-RV133.
Safety Evaluation Summary:
I This simple design change will not impact the safe operation or shutdown of the plant. The condensate storage facility condensate makeup and drawoff system is not required to effect or support safe shutdown of the reactor or to support the operation of any nuclear safety system.
Engineering evaluation indicates that no piping, components or equipment will be overpressurized by the removal of relief valve 2CNS-RV133.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 25 of 136 Safety Evaluation No.: 92-072 Rev. 2 Implementation Document No.: N2-SOP-01, 02 and 03 USAR Affected Pages: NIA System: Station Blackout Title of Change: Station Blackout (Complete Loss of Ac Power)
Description of Change:
Revision 1 of this safety evaluation was reported October 29, 1993. Revision 2 of this safety evaluation reflects a change for RCIC room high temperature isolation bypass "from two hours to twenty minutes" in accordance with NMPC Calculation ES-268 "RCIC Pump/Turbine Room Heat Analysis During Station Blackout."
Safety Evaluation Summary:
NUMARC 87-00 and Regulatory Guide 1.155 specify the station blackout (SBO) duration that plants should be capable of withstanding. Based on a review of the onsite and offsite power distribution systems at NMP2, the plant must meet a 4-hour capability, with a target emergency diesel generator reliability of 0.975 maintained.
The review of the NlVIP2 SBO capabilities addressed the following:
~ The core cooling capability of installed systems.
~ The support systems (i.e., pneumatic supplies and station batteries).
The availability of instrumentation and controls to monitor and carry out required operator actions.
~ Potential limitations that may result from the heatup of certain areas of the plant during the event.
The results of this analysis show that with certain specific operator actions, NMP2 can operate longer than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> without jeopardizing core cooling or the containment capability. Operator actions needed to achieve this capability include
Safety Evaluation Summary Report Page 26 of 136 Safety Evaluation No.: 92-072 Rev. 2 (cont'd.)
Safety Evaluation Summary: (cont'd.)
shedding of nonessential battery loads, maintaining manual RClC flow control and the bypassing of certain RClC isolation logics. These actions are addressed in the Special Operating Procedures (N2-SOP-01, 02 and 03).
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 136 Safety Evaluation No.: 92-073 Implementation Document No.: Simple Design Change SC2-0310-92 USAR Affected Pages: Figure 10.1-6c System: Reactor Feed Pump Seal Water (FWP)
Title of Change: Reactor Feed Pump Seal Water Instrumentation (FWP System)
Description of Change:
This simple design change added flow and temperature monitoring instrumentation to the reactor feed pump seal water lines. This instrumentation allows seal degradation to be monitored and allows seal replacement as a planned evolution and not a dramatic failure with resultant feedwater system transients.
Safety Evaluation Summary:
The addition of temperature and flow instrumentation to monitor reactor feed pump seal water will provide greater reliability of the nonsafety-related pumps by allowing monitoring of the seal's condition in order to prevent pump seal degradation from adversely affecting pump operation.
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 136 Safety Evaluation No.: 92-082 implementation Document No.: Simple Design Change SC2-0143-92 USAR Affected Pages: Figures 9.2-3a, 9.2-3g System: Reactor Building Closed Loop Cooling - CCP Title of Change: Addition of Vacuum Breakers to Expansion Tank Overflow Lines Description of Change:
This simple design change added vacuum breakers to the overflow lines from CCP expansion tanks 2CCP-TK1 and 2CCP-TK2. This was done to prevent negative pressure in the expansion tanks, due to flow in the overflow lines, from causing spurious alarms and/or trips of the CCP pumps.
Safety Evaluation Summary:
This simple design change enhances the reliability of the CCP system which has no safety-related functions. Ties of the CCP system to safety-related components or systems are unaffected by this change. Existing evaluations for spraying=and flooding envelop any spraying which could result from the addition of the vacuum breakers.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 29 of 136 Safety Evaluation No.: 92-084 Implementation Document No.: Simple Design Change SC2-0122-91 USAR Affected Pages: Figure 11.2-1b System: Radioactive Liquid Waste System Title of Change: Add Air Filter 2LWS-FLT3 Description of Change:
This simple design change added air filter 2LWS-FLT3 upstream of 2LWS-SOV'1 69.
Safety Evaluation Summary:
This simple design change will enhance the system reliability. The radioactive liquid waste system is not required to effect or support safe shutdown of the reactor or to support the operation of any nuclear safety system.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 30 of 136 Safety Evaluation No.: 92-088 Rev. 1 Implementation Document No.: N/A USAR Affected Pages: 8.2-24, 8.3-40b System: N/A Title of Change: Allow the Use of Calculated Voltages for AC MOV Starting Description of Change:
This change allows the use of calculated minimum motor starting voltages for safety-related valves. These voltages are used in determining if the valve can develop sufficient torque to operate under design conditions. Calculation EC-154 determines the starting voltage available at the motor terminal for each valve in the GL 89-10 program under degraded voltage conditions. The starting point for the calculation is the steady state degraded voltage relay setpoint from Calculation EC-136.
Safety Evaluation Summary:
The use of the calculated degraded voltage is acceptable because the motor will have this voltage available under the worst possible voltage conditions. The use of the steady state degraded voltage relay setpoint to determine the voltage available at the motor terminals for starting is conservative. If the voltage goes below the setpoint of the degraded voltage relay, the diesel generator for that division would start. The diesel generator would then supply the Class 1E loads. Calculation EC-154 shows that when supplied from the diesel generators, the MOVs would have at least as much voltage available as calculated for the offsite source.
Therefore, it is acceptable to use the calculated minimum motor starting voltage to determine if the valve can develop sufficient torque to operate under design conditions.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 31 of 136 Safety Evaluation No.: 92-091 Implementation Document No.: DER 2-92-0-0865, EDC 2M00266 USAR Affected Pages: Figure 10.1-6e System: Zinc Injection Passivation System (ZIP)
Title of Change: As-Built Update, The Addition of Zinc Injection Passivation System Components Description of Change:
Problem Report PR-08590 and DER 2-92-0-0865 stated that the flush/vent connections located on both the suction and discharge side of zinc injection feed pumps 2ZIP-P1A and P1B (2ZIP-SKD1) did not appear on PAID 6E. In addition, the sight flow glasses on the suction side were not shown. The components being added are original system components inadvertently left off the system P&ID.
Safety Evaluation Summary:
This documentation change will have no impact on safe operation or safe shutdown conditions. No physical changes are being made to the nonsafety-related ZIP system. The ZIP system will continue to function as designed.
Documentation is enhanced by providing NMPC component identification numbers.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary'eport Page 32 of 136 Safety Evaluation No.: 93-002 Implementation Document No.: Simple Design Change SC2-0003-93 USAR Affected Pages: 5.4-9c 'igure System: Reactor Core Isolation Cooling System (ICS)
Title of Change: Deletion of RCIC Test Line Description of Change:
This change is related to the reactor core isolation cooling system (ICS). Due to repeated pin hole leakage at the tee connection to test line 2-ICS-750-65-2, this line has been eliminated.
Safety Evaluation Summary:
This review, which included the effects of the change on the system's operability, reliability, maintainability, structural integrity, and system interactions has found that the deletion of the subject line will improve the system's reliability without causing any significant safety or operability issues.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 33 of 136 Safety Evaluation No.: 93-005 Rev. 0 6 1 Implementation Document No.: N/A USAR Affected Pages: 12.5-1, 12.5-2; Tables 1.8-2 Sh 3, 9A.3-15; Figures 1.2-1, 2.4-1, 9.5-1b, 9A.3-1, 12.3-3, 12.3-18, 12.3-36, 12.3-51 System: N/A Title of Change: New Unit 2 Access Control Building Description of Change:
I The Unit 2 access control building was constructed where the radiation protection trailer (74) was located inside the protected area, south of the Unit 2 reactor building and south auxiliary bay roof.
The building was constructed to facilitate implementation of the single-point control of entry into the restricted area, enhancing the radiation protection measure at Unit 2.
The building is a single-story, nonsafety-related structure consisting of a slab on grade and has a total area of approximately 14,000 square feet. This building-provides office facilities for up to 75 personnel.
Safety Evaluation Summary:
The pertinent safety issues identified in this safety evaluation are flooding, the impact on the control room fresh air intake, radiological atmospheric dispersion coefficient, and,impact of construction and building loads on the Div. 3 duct bank EDB-922. It can be concluded, based upon analysis, that the construction of the Unit 2 access control building does not impact the pertinent licensing issues evaluated in the Unit 1 UFSAR or the Unit 2 USAR.
Revision 0 of,the safety evaluation addressed changes to the facility as described in the Unit 2 USAR. 'Revision 1 to the safety evaluation expanded the scope and evaluation to address changes to the facility as described in the Unit 1 UFSAR.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 34 of 136 Safety Evaluation No.: 93-010 Rev. 2 Implementation Document No.: Mod. PN2Y92MX003 USAR Affected Pages: 9C.8-5; Tables 3-1, 3-2 Sh 3, 3-4, 4-1 Sh 2; Figure 5-1 System:
Title of Change: SRV Removal Phase III Description of Change:
The following are changes that have been implemented by this modification:
- 1. Extended the monorail at el. 305'-9n the primary containment by 4'-3" ~
~
The bus bar that services this monorail was also increased in length by 4I 3ff
- 2. The use of an alternate crane in lieu of existing crane 2MRH-CRN65 is now allowed. The alternate crane is an electrical trolley and chain hoist crane designated as crane 2MHR-CRN65X.
- 3. A monorail system was installed outside of the north hatchway to service the 261'loor elevation, and consists of a 4-ton chain-operated trolley and hoist crane.
Safety Evaluation Summary:
The improvements being made by this modification to the removal and replacement process of the SRVs do not affect any safety-related systems. The extension of the monorail at el. 305'-9" and the use of alternate crane 2MHR-CRN65X in the primary containment are both considered and included in the revised Heavy Load Analysis.
The installation of both crane and monorail in the secondary containment meets the requirements of the seismic evaluation of nonsafety-related components in safety-related areas, and does not affect the safety and reliability of NMP2.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 35 of 136 Safety Evaluation No.: 93-019 Implementation Document No.: Simple Design Change SC2-0004-93 USAR Affected Pages: Figures 9.2-1L, 9.5-52a, 9.5-52c System: Auxiliary Boiler System (ABM), Service Water System (SWP)
Title of Change: Installation of Chesterton Mechanical Seals on 2ABM-P1A & P1B Description of Change:
DER 2-92-3456 describes failures which occurred to the previous mechanical seals manufactured by John Crane. With the previous design, degraded seal water passed through the seal faces, scoring the surface and damaging the seal, thus causing seal failure. Also, during hot standby conditions, the water drained to the boiler raising the water level above acceptable limits. To increase system reliability, a new seal manufactured by Chesterton, designed with an outer seal water cooling jacket, was installed. The cooling water is isolated from the seal faces circumventing intrusion of foreign material. New seal cooling water inlet and outlet piping was also installed.
Safety Evaluation Summary:
This simple design change will have no impact on the safe operation or capability to shut down or keep the plant in a'safe shutdown condition. The ABM system and the affected portion of the SWP piping are classified as nonsafety related, and are not required to effect or support safe shutdown of the reactor or to perform in the operation of the reactor safety features. The ABM and SWP will continue to perform their function as described in USAR Sections 9.5.10 and 9.2.1.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 36 of 136 Safety Evaluation No.: 93-020 Implementation Document No.: Simple Design Change SC2-0049-93 USAR Affected Pages: Table 9.4-4 Sh 4, 5 System: Radwaste Building Ventilation Title of Change: Radwaste Control Room Noise Improvement Description of Change:
This design change was required to reduce excessive noise in the radwaste control room. Under this modification, the existing 7.5 hp return/exhaust air fans (2HVW-FN12A and 2HVW-FN12B) were replaced with new 3.0 hp fans. The new smaller fans have a capacity to provide 10,700 cfm at 1200 rpm. The drive sheave and belt on air handling units 2HVW-ACU2A and 2HVW-ACU2B may also have to be replaced to reduce the speed of supply air fans to approximately.1400 rpm if the desired speed is not achieved by adjusting the motor baseplate. This is necessary to provide a balanced air flow between the supply air and the exhaust air side.
The new fan motors are powered using appropriately sized motor starters and overload heaters to provide proper circuit protection.
Safety Evaluation Summary:
This design change will improve environmental working conditions in the radwaste control room by reducing noise levels. The proposed change does not affect or involve any safety-related equipment in the plant. All changes under this modification are associated with the nonsafety-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 37 of 136 Safety Evaluation No.: 93-022 Implementation Document No.: Simple Design Change SC2-0261-91 USAR Affected Pages: Figures 11.3-1b, 11.5-2 System: Offgas (OFG)
Title of Change: Offgas/DRMS Process Flow Element/
Transmitter(s) Replacement Description of Change:
This simple design change replaced the 6-52.5 scfm range offgas process flow element/transmitter units with similar units capable of monitoring flow over a 15-150 scfm range of process flow (as well as providing local indication). The required changes to the offgas cabinet microprocessors to properly communicate with the replacement units is also included in this change. USAR Figures 11.3-1b and 11.5-2 have been revised to show the flow transmitters as flow indicating transmitters. Additionally, USAR Figure 11.3-1b was corrected to show the as-built locations for the flow elements (which were physically relocated from the offgas charcoal absorber tank(s) room to a lower elevation, in June 1986 by Stone 5 Webster; change package notice CPM-Y152).
Safety Evaluation Summary:
An engineering review of the proposed changes has dete'rmined that the implementation of this change will improve the offgas system's reliability/
operability without causing any significant safety or operability issues.
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 136 Safety Evaluation No.: 93-023 Implementation Document No;: Mod. PN2Y92MX004 USAR Affected Pages: 1.2-31, 9.3-2 thru 9.3-6, 9.3-8 thru 9.3-10; Tables 8.3-1 Sh 31, 8.3-2 Sh 30, 9A.3-6 Sh 6; Figures 1.2-19 Sh 2, 1.2-23 Sh 1, 8.3-1, 9.2-1n, 9.2-3b, 9.2-3g, 9.3-1a thru 9.3-1c, 9.3-2 Sh 1, 2, 5, 12, 13, 9.3-3a, 9.3-3c, 9.3%, 9.3-10g, 12.3-14, 12.3-47 System: Instrument Air System (IAS), Service Air System (SAS), Breaking Air System (AAS)
Title of Change: IAS Upgrade Description of Change:
This modification upgraded the existing instrument air system (IAS) by replacing the existing compressors and drain traps and replacing the carbon steel piping, between the compressors and dryers with stainless steel pipe. Other improvements included: relocation of the pressure sensor for pressure indicator 2IAS-PI101 from its position upstream of the IAS prefilters to a point downstream of the IAS air receiver tanks; provided a tie-in to the breathing air system (AAS);
provided a means to repressurize the service air system (SAS) gradually after a SAS system outage; and installed a new supply line from the IAS system for the SAS system downstream from the IAS dryers. These improvements reduced moisture carryover, improved air quality, increased compressor availability, reduced system maintenance, eliminated the need to upgrade major components in the AAS system, and improved service air quality and system design.
Safety Evaluation Summary:
This modification improves the IAS system by providing additional capacity to meet compressed air requirements. The SAS system is improved by the addition of a second tap from IAS that supplies dry compressed air to the SAS system. Also, provision is made to gradually repressurize the SAS system after a SAS system outage. The AAS remains functional with no changes to capacity or pressure.
The AAS compressor is deleted and the IAS supplies sufficient air to the AAS system to allow it to function as designed.
Safety Evaluation Summary Report Page 39 of 136 Safety Evaluation No.: 93-023 (cont'd.)
Safety Evaluation Summary: (cont'd.)
The lAS provides air to both safety-related and nonsafety-related components.
However, the system is not considered a safety-related control air system since all of the safety-related components that it supplies perform their safety functions without air or are provided with safety-related accumulators capable of supplying the required quantities of air. The SAS and AAS are used only for nonsafety-related equipment and components during normal plant operation.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 40 of 136 Safety Evaluation No.: 93-027 Rev. 1 Implementation Document No.: N2-FHP-14 USAR Affected Pages: 9.1-23, 9.1-41 System: FNR Title of Change: Lower or Remove Fuel-Preparation Machine Full-Up-Stops Description of Change:
This safety evaluation addresses the temporary change of the position of fuel preparation machine (FPM) full-up-stops. This change reduces the time/exposure spent during the transfer of new fuel to the spent fuel pool. Additionally, this reduces the potential for personal contamination and plant contamination.
It is desirable to load new fuel assemblies into the spent fuel pool using a FPM to avoid contamination of the crane. This is accomplished by transferring a fuel assembly from the new fuel inspection stand to a FPM.
The in-use FPM has its full-up-stops set (or removed) such that a new fuel assembly loaded into its carriage has its bail handle above the spent fuel pool water level. (Positive stopping of the FPIVI carriage is performed by the end stops on the roller chain mechanism). After the crane is disconnected from the new fuel assembly, which is sitting in the FPM, the assembly will be transferred by the refueling platform to its temporary storage location in the spent fuel storage rack.
When all the new fuel has been loaded into the spent fuel pool, the FPM mechanical stops shall be repositioned in their original location.
Safety Evaluation Summary:
The purpose of the mechanical stops is to prevent the carriage on the FPIVI from lifting a fuel assembly, channel, or bundle to a height where water shielding is less than 7 ft. above the fuel bundle, channel, or assembly. The mechanical stops should only prevent the lifting of an irradiated fuel assembly or bundle to a height where water shielding is less than 7 ft. above a fuel bundle or assembly-not new unirradiated fuel.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 41 of 136 Safety Evaluation No.: 93-028 Implementation Document No.: Simple Design Change SC2-0085-93 USAR Affected Pages: 11.3-6; Figures 11.3-1a, 11.3-1b System: Offgas System (OFG)
Title of Change: Retire Offgas Flow Switches 2OFG-FS17A/B and 2OFG-FS140 Description of Change:
This simple design change defeated the control, alarm, and annunciation functions of flow switches 2OFG-FS17A/B and 2OFG-FS140, which have a history of spuriously de-energizing hydrogen analyzers 2OFG-AT16A/B and 2OFG-AT115.
This change made Temporary Modification 89-053 permanent.
Safety Evaluation Summary:
These flow switches are not utilized in the determination of hydrogen analyzer operability required per Technical Specification 3.3.7.10. The circuitry to be removed will not impact any safety-related or 0-related equipment, nor will it have any adverse impact on any plant effluents or the effluent monitoring capabilities of the environmental protection plan..
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 42 of 136 Safety Evaluation No.: 93-029 Implementation Document No.: Simple Design Change SC2-0086-93 USAR Affected Pages: Figures 9.3-3a, 9.3-4 System: Breathing Air System (AAS)
Title of Change: Removal of Valve 2AAS-SOV125 Description of Change:
This simple design change removed valve 2AAS-SOV125 from the closed loop cooling subsystem for the AAS air compressor. Removal of the valve allows cooling to circulate through the compressor at all times in lieu of only when the compressor is running as per original design.
Safety Evaluation Summary:
An engineering review of this change, which included the effects of the change on the system's operability, maintainability, structural integrity, and system interactions, has found that the implementation of this change will not adversely impact the AAS system or cause any significant safety or operability issues, while at the same time reducing maintenance associated with the valve.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 43 of 136 Safety Evaluation No.: 93-030 Implementation Document No.: Simple Design Change SC2-0434-91 USAR Affected Pages: Figure 9.4-23 Sh 2 System: Hot Water Heating (HVH)
Title of Change: Expansion Tank 2HVH-TK1 Level Change for Makeup Control Description of Change:
This simple design change made Temporary Modification 88-231 a permanent change to the plant. The change involved the addition of a jumper wire in local panel 2CES-IPNL203 (in the turbine building) to defeat the switch for the lower level float of level switch 2HVH-LS114. The level switch controls the hot water heating expansion tank level control band, and defeating the lower float switch resulted in narrowing the band to the upper switch trip and reset differential. This change reduced the excessive usage of nitrogen gas in the tank due to the previous wide band (12 inches) causing the relief valve to lift. This would result from the makeup pump refilling the tank water and compressing the nitrogen enough to lift the relief during normal system makeup cycle. The narrow band implemented by the Temporary Modification jumper addition has been proven (by functional testing of the hot water makeup) to reduce the excessive nitrogen usage.
Safety Evaluation Summary:
An engineering review of the proposed change to make Temporary Modification 88-231 permanent was performed and documented in the analysis section of the safety evaluation. The review, which included operability, availability, constructability and potential system interactions has determined that this change will improve the hot water heating system operation (by reducing the excessive usage of nitrogen) without causing any significant safety or operability issues.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 44 of 136 Safety Evaluation No.: 93-032 Implementation Document No.: Simple Design Change SC2-0084-93 USAR Affected Pages: Figure 9.2-6a System: Makeup Water System Title of Change: Permanent Makeup Water System Connection for the Portable Demineralizer Description of Change:
Temporary Modification 87-2008 controlled installation for the use of the portable demineralizer as provided within the ecolochem trailer. The temporary modification also described hose connections to valve 2WTS-V261 and a new connection installed upstream of valve 2MWS-V42. Temporary Modification 87-2008 is being cleared.
This simple design change made permanent the new hose connections with isolation and sample valves, and revises the affected drawings accordingly. The use of the demineralizer trailer and intermittent attachment of the hoses to both the makeup water and water treating systems will now be procedurally controlled as described in N2-OP-15.
Safety Evaluation Summary:
This simple design change will have no impact on the safe operation or capability to shut down or keep the plant in a safe shutdown condition. The makeup water system is classified as nonsafety related, and is not required to effect or support safe shutdown of the reactor or to perform in the operation of the reactor safety features. The makeup water system will continue to perform its function as described in Sections 9.2.3 of the USAR.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 45 of 136 Safety Evaluation No.: 93-035 Implementation Document No.: N2-SOP-03 USAR Affected Pages: N/A System: Service Water System 6 Offsite Reserve Ac Power Title of Change: Service Water System Divisional Separation Associated with 2SWP" MOVGOA/B, 3A/B, 19A/B, 93A/8, and 599 and 2SWP" FV47A/B and FV54A/B During a Loss of Offsite Power Description of Change:
N2-SOP-03 Sections 1.3 and 2.3 allow operator action to override open service water valves 2SWP"MOVSOA/B, 3A/B, 19A/B, 93A/B, and 599, and 2SWP" FV47A/B and FV54A/B during a loss of offsite power (LOOP).
Safety Evaluation Summary:
The purpose of this safety evaluation is to justify overriding the service water system divisional separation isolation valves 2SWP MOV50A/B, 3A/B, 19A/B, 93A/B and 599, and 2SWP" FV47A/B and FV54A/B during a LOOP. These valves are being overridden open, during a LOOP, to minimize the plant transient that would result if service water remains isolated to the balance of plant systems. The plant transient that would result if service water remained isolated to the balance is described in USAR Section 15.2.6, Loss of Ac Power. Subsequent operator action is acceptable for realigning the service water system in the event a loss-of-coolant accident or a loss of an emergency diesel generator occurs. These actions maintain the service water system and its ability to perform its safety functions.
The justification for these actions is: 1) they are the safest actions for the plant,
- 2) the applicable Technical Specifications Limiting Conditions for Operation 3.0.3 and 3.8.1.1.f are implemented, and 3) a plant shutdown is commenced on the initiating event.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 46 of 136 Safety Evaluation No.: 93-038 Rev. 0, 1 L2 Implementation Document No.: Simple Design Change SC2-0287-92 USAR Affected Pages: 8.3-42a, 8.3-70 System: RCIC Title of Change: Install Shorting Bars in Lieu of Overload Heaters for Dc Motor-Operated Valves Description of Change:
This simple design change installed shorting bars in the thermal overload heater (TOL) circuits for the dc IVlOVs in the GL 89-10 program. The trip contacts of the TOL are bypassed by an automatic safety signal or for manual operation by holding the switch in the operate position. Although the overload heater's trip contacts are bypassed, the overload heater physically remains in the power circuit. This increases the voltage drop from the power source to the valve motor. Replacing the TOLs with shorting bars will also remove the resistance of the TOLs from the power circuit.
Safety Evaluation Summary:
Replacing the TOLs with shorting bars will ensure that sufficient voltage will be available at the dc MOVs as shown in electrical calculation disposition EC-154-01C. As a precautionary measure for valve protection, the valve's stroke time will be monitored during periodic testing. After any maintenance on the valves, appropriate postmaintenance testing will be performed. If it is shown that the motor operator was subjected to an overload condition, then procedures will require that it be analyzed for operability.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 47 of 136 Safety Evaluation No.: 93-046 Implementation Document No.: Mod. PN2Y91MX027 USAR Affected Pages: Figures 5.4-13d, 5.4-13e System: Residual Heat Removal (RHS)
Title of Change: Add Valves to RHS Condensing Lines Description of Change:
This modification installed manual isolation valves (one valve in each loop) in the RHS steam condensing supply lines between the loop heat exchanger and the pressure control valves PV21A(B).
This change was made during the R-3 refuel outage.
Safety Evaluation Summary:
The proposed change to the RHS steam condensing supply lines will enhance the system in the following ways:
The installation of the manual isolation valves will eliminate the need to in-op the respective loop heat exchanger during valve seat leakage test (VSLT) and maintenance of the line valves. The time previously required to drain and refill the heat exchangers will be saved.
- 2. Having both loop heat exchangers available during VSLT of one loop increases the operator's flexibility by allowing use of both shutdown cooling loops.
- 3. Maintenance can be performed on supply line valves much more easily and man rem exposure rates, will be lower.
Installation of new supports and/or removal of snubbers will be performed in accordance with all applicable procedures, codes, and criteria established in Safety Evaluation 90-048 (NMP2 Snubber Reduction Modification Program).
Safety Evaluation Summary Report Page 48 of 136 Safety Evaluation No.: 93-046 (cont'd.)
Safety Evaluation Summary: (cont'd.)
The proposed change is safety related; however, there will be no impact on the safe operation or shutdown of the plant. The RHS system design bases specified in USAR Section 5.4.7.1, and the functional design basis of the reactor steam condensing mode specified in USAR Section 5.4.7.1, will not be changed.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 49 of 136 Safety Evaluation No.: 93-047 Implementation Document No.: Simple Design Change SC2-0070-92 USAR Affected Pages: 5.4-36, 6.3-15; Figures 5.4-13a, 5.4-13b, 6.3-7a System: Residual Heat Removal (RHS), Low Pressure Core Spray (CSL)
Title of Change: Valve Modification to Prevent Pressure Locking Description of Change:
This change added small bore piping to three valves in the RHS system and one valve in the CSL system. This piping prevents bonnet pressure locking in these valves.
These changes were made during the R-3 refuel outage.
Safety Evaluation Summary:
The proposed change to valves 2RHS"MOV24A, 24B, 24C and 2CSL" MOV104 will allow these valves to perform their intended design functions.
The proposed changes are made to safety-related components. A sentence is added to the valves'escription to indicate that the valves have been modified to prevent bonnet pressure locking.
Based on the evaluation performedit is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 50 of 136 Safety Evaluation No.: 93-048 implementation Document No.: N/A USAR Affected Pages: 9A.3-53b System: Carbon Dioxide Fire Protection System Title of Change: Elimination of Puff Test Requirements Description of Change:
This change removed the requirement to perform a flow test (commonly known as a "puff" test) as part of the CO, system functional test. Removal of this part of the functional test is in compliance with NFPA 12, Carbon Dioxide Extinguishing Systems.
Safety Evaluation Summary:
NFPA Code 12-1993, Section 1-10, does not suggest that a CO~ system nozzle puff test be performed at regular intervals as part of a system functional test.
NMP1 performed this flow, or puff, test every 6 months as a conservative action in functional testing of the CO~ system. Additionally, the systems are routinely (monthly) visually inspected as described in NFPA 12, Section 1-10.3A. These functional tests and monthly visual inspections provide sufficient system operability checks. To date, puff testing has found no system nozzle or piping to be blocked.
The removal of the puff test remains in compliance with NFPA codes which are appropriate per NRC Branch Technical Position (BTP) CMEB 9.5-1, position C.6.a.
This change does not change the reliability of the carbon dioxide extinguishing system or the ability to achieve and maintain safe shutdown in the event of a fire.
No field work is required as a result of this change, and no operability concerns or procedures are affected.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 51 of 136 Safety Evaluation No.: 93-050 Implementation Document No.: Simple Design Change SC2-0033-93 USAR Affected Pages: 5.4-25, Table 6.2-56 Sh 2, 3,4, 5, 6, 7, 11 System: CCP, DFR, ICS, RHS Title of Change: Gear Set Changes for Various Motor Operated Valve (MOV) Actuators Description of Change:
This simple design change changed operator gear sets of safety-related MOVs to provide a sufficient thrust window to use the valve operation test and evaluation system (VOTES) diagnostic equipment for testing of MOVs within the scope of NMP2's Generic Letter (GL) 89-10 program.
As a result of operator gear set changes, valve closure stroke times have been increased. A review of mechanical, radiation protection and engineering safeguard calculations ensured that these higher stroke times do not adversely impact plant safety.
Safety Evaluation Summary:
This simple design change installs new operator gear sets in safety-related MOVs to accommodate GL 89-10 testing. This will ensure that the MOVs will perform their intended safety functions during and/or following design basis accidents.
Actuator sizing calculations performed during design basis reviews in accordance with the GL 89-10 program verify that the MOVs will not operate outside of their design or testing limits due to operator gear set changes.
This simple design change will have no adverse effect on the safe operation or shutdown'of the plant. Postinstallation MOV stroke time testing in accordance with site procedures will verify valve operability.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 52 of 136 Safety Evaluation No.: 93-051 Implementation Document No.: Simple Design Change SC2-0096-93 USAR Affected Pages: Figures 5.1-2a, 7.7-8 System: Reactor Vessel Instrumentation (ISC)
Title of Change: Enhanced Monitoring of RPV Level Instrument Reference Legs Description of Change:
This change provided for two additional analog RPV level signals to the Emergency Response Facility (ERF) computer system from channels C and D of the narrow range series of vessel instruments. These two signals, along with the two existing channel A and B signals in the system, allow operations to better monitor, as required per NRC Bulletin 93-03, for discrepancies, inaccuracies or notching in RPV level indications caused by degassing in the instruments'eference legs when the vessel is depressurizing.
Safety Evaluation Summary:
The level signals provided to the ERF computer by this change are either electrically isolated from their safety-related source or will have no adverse loading effect on the trips, isolation functions of their associated components, or circuits.
The proposed change will not impact any systems or components important for safety of the facility nor will this change have any impact on the plant's effluents or effluent monitoring capabilities.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 53 of 136 Safety Evaluation No.: 93-052 Rev. 1 Implementation Document No.: N/A USAR Affected Pages: 1.10-139, 1.10-140 System: ICS, RHS, GTS Title of Change: Deletion of Helium Leak Test Description of Change:
Previously, the reactor core isolation cooling (RCIC), residual heat removal (RHR),
and standby gas treatment (GTS) systems outside the primary containment were leak tested by helium. The RCIC and RHR helium tests were replaced by a system walkdown/inspection for leaks during system operational or hydrostatic testing.
The GTS system helium leak test was modified to test only the portion of the system which operates at positive pressure.
Safety Evaluation Summary:
The NMP2 commitment provided for leak testing either by physical testing or system walkdown for leaks during operation testing. The walkdown alternative, however, is limited to liquid systems only. Therefore, RCIC and RHR are tested using helium. By redefining our commitment for the steam portion of the system, it is possible to delete the helium test and use visual inspection for leaks as a valid method for leak detection for RCIC and RHR systems.
The bulk of the GTS system operates under vacuum, and integrity of this portion of the system is indirectly verified by the GTS surveillance tests (drawdown and in-leakage). Therefore, the helium leak testing is not required for the portion of the system which operates under vacuum. However, leak detection (visual inspection with a leak detection agent) for the portion of the GTS system which operates at positive pressure is required to meet the existing commitment. This change meets the current regulatory requirement.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 54 of 136 Safety Evaluation No.: 93-053 Implementation Document No.: N2-PM-W3 USAR Affected Pages: N/A System: Main Turbine Title of Change: Alternate Method of Performing the Main Turbine Mechanical Overspeed Trip Test Description of Change:
This procedure change allows the use of a rig external to the front standard to spray turbine oil into the concentric ring of the mechanical overspeed device. This is being performed to compensate for a broken oil line to the nozzle of the oil trip located in the front standard. This change also altered the frequency of the test from weekly to bi-monthly.
Safety Evaluation Summary:
The use of this rig performs the function of spraying the oil into the ring as adequately as the normal method. The frequency for performance of this was changed from weekly to bi-monthly for ALARA, heat stress, risk of incident, and human factors concerns and is based on sound engineering judgement. Changing the method of performing the test has no impact on the mechanism for it is not the delivery of the oil to the concentric ring that causes the action, but the oil in the ring which then causes centrifugal forces to overcome spring tension. Due to this reason the ability of the mechanical overspeed protection device to perform its function is not altered or impaired. Changing of this. frequency does not impact the failure probability of the mechanism, only its detection. The frequency change has no detrimental effects on the alternate test methodology so, therefore, has no effect on the mechanical overspeed mechanism.
In addition, the USAR states that the overspeed protection control has no impact on nuclear safety, so the implementation of this change will in no way adversely affect the capabilities of the turbine generator overspeed protection system to perform its function.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 55 of 136 Safety Evaluation No.: 93-054 Rev. 0, 1 62 Implementation Document No.: Mod. PN2Y89MX135 USAR Affected Pages: 9.5-64, 9.5-66; Figures 9.5-52a, 9.5-52c, 9.5-53 Sh 5 System: Auxiliary Boiler System (ABF, ABH, ABM)
Title of Change: Auxiliary Boiler System Upgrade Description of Change:
This modification made the following changes to the auxiliary boiler system:
Modified the feedwater pump suction piping to eliminate pump cavitation problems.
Relocated the sodium sulfite injection point from the auxiliary boilers to the deaerator in order to protect the piping between the deaerator and the auxiliary boilers from oxygen attack and to reduce the burden on the operators.
Replaced the feedwater discharge piping due to corrosion.
Added a manual flush line for the phosphate feed line in order to prevent future plugging.
Added a bypass line around auxiliary steam pressure control valves 2ABM-PV11A/B to facilitate startup.
Added isolation valves to the auxiliary boiler pressure instrumentation trees to allow for maintenance of the devices.
Abandoned in place pH instrumentation loops 2ABD-10A and 2ABD-10B.
Safety Evaluation Summary:
These changes were made to upgrade the auxiliary boiler system to improve system reliability and its capability to support the plant operations. These changes will have no impact on the safe operation or shutdown of the plant since the affected system is nonsafety related, is not required to achieve or maintain safe shutdown, is not required for safe reactor operations, and does not directly interface with any safety-related systems. Furthermore, the changes are confined to the nonseismic auxiliary boiler building which does not contain any safety-related components; therefore, indirect interface between the proposed changes and safety-related systems is precluded by physical separation. Finally, all
Safety Evaluation Summary Report Page 56 of l36 Safety Evaluation No.: 93-054 Rev. 0, i &. 2 (cont'd)
Safety Evaluation Summary: (cont'd.)
piping changes have been designed in accordance with the original design basis and do not introduce piping to an area which did not previously contain this type of piping.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question:
Safety Evaluation Summary Report Page 57 of 136 Safety Evaluation No.: 93-057 Rev. 0, 1 L2 Implementation Document No.: Mod. PN2Y93MX001 USAR Affected Pages: 3.9B-S, 3.9B-23, 9.1-27, 9.1-30, 9.1-31, 9C.3-4, 9C.3-5; Tables 3.9B-2n Sh 3, 9.1-2, 9.1-4 Sh 1; Figure 9.1-7, 9.1-25 System: N/A Title of Change: Refuel Bridge Upgrades Description of Change:
This safety evaluation evaluated refuel bridge upgrades implemented under Modification PN2Y93MX001. This modification addressed the service pole caddy system (SPCS) scope and the readout assembly scope.
Servi e Pol ad The SPCS is an attachment to the existing refueling bridge. The SPCS consists of a handling platform on the refueling bridge and a set of hi-torque service poles. The platform stores and also provides a motorized hoist to handle the hi-torque poles. The handling platform is attached to a rigid frame which is attached to the south end of the refueling bridge.
Readou Assembl Sco e The position readout assembly was relocated from the east side of the control console to the west side of the control console. The support bracket is hinged to allow for position adjustment. In the previous location, the readout assembly display became an obstruction to the operators during various in-core evaluations.
Safety Evaluation Summary:
The improvements being made by this modification to the refueling bridge do not affect any safety-related system. The electrical activities are not safety related.
The refueling platform with the SPCS structure has been seismically/dynamically re-evaluated and is considered Quality Category Q5. The service pole caddy hoist capacity is 1/4 ton; therefore, the Heavy Load Criteria specified in NUREG-0612 and referenced in USAR Appendix 9C do not apply. The USAR analysis in Chapter 9 and Appendix 9C are still bounding.
Safety Evaluation Summary Report Page 58 of 136 Safety Evaluation No.: 93-057 Rev. 0, 1 L 2 (cont'd.)
Safety Evaluation Summary: (cont'd.)
The service pole caddy system is an auxiliary platform attached to the refueling platform. It is used to carry service poles to desired locations over the reactor cavity to perform underwater servicing activities on reactor equipment, such as shroud head bolts removal and installation, steam line plugs installation and removal, and underwater camera manipulation.
The relocation of the readout assembly from the east side to the west side of the control console does not affect the structural analysis of the refuel bridge.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 59 of 136 Safety Evaluation No.: 93-058 Rev. 0 L 1 Implementation Document No.: Calculation H21C-027 Procedure N2-MPM-GEN-R901 USAR Affected Pages: 9.1-36 System: FHS Title of Change: Removal of Reactor Cavity Shield Plugs A, B, C, and D at 40% or Less Reactor Power Description of Change:
This safety evaluation evaluated the USAR changes and the impact of the NMP2 design as a result of the removal of reactor cavity shield plugs A, B, C and D at 40 percent or less reactor power.
A review of the structural considerations regarding the removal of the top four shield plugs indicated that the removal of these plugs will not affect the structural integrity of this shield plug barrier. Specifically, the seismic and tornado design loads that the original eight plug barriers were designed for will not be exceeded by having only the lower four plugs in place when at 40 percent or less reactor power.
The radiological considerations regarding the removal of these four plugs (A, B, C and D) have been evaluated via calculation H21C-027 and result in a projected accumulated dose rate per outage worker far below 10CFR20 limits. The potential increase in cumulative dose rate due to this shield plug change was reviewed and approved by the NlVlP2 ALARA Committee on June 17, 1993.
Safety Evaluation Summary:
The removal of reactor cavity shield plugs A, B, C and D at 40 percent or less reactor power does not affect the structural integrity of the shield plug barrier.
The radiological effects of the proposed change have been calculated and determined to be negligible for radiological consequences to the refueling operators during normal refueling operations.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 60 of 136 Safety Evaluation No.: 93-059 Implementation Document No.: Temporary Mod.93-042 USAR Affected Pages: N/A System: High Pressure Core Spray (HPCS)
Title of Change: Jumper Control Signal for 2CSH "MOV118 Description of Change:
A temporary jumper was installed in the control circuit of the HPCS suppression pool suction valve 2CSH "MOV118 to simulate a closed valve signal from the HPCS test return valve 2CSH "MOV112. This provided a permissive signal for 2CSH" MOV118 to open even though valve 2CSH"MOV112 was deenergized and/or stroked open (not closed). With 2CSH "MOV118 capable of opening, the HPCS is capable of transferring water from the suppression pool to the reactor vessel and meets the requirements of Technical Specification 3/4.5.1.c. The HPCS can be declared operable without 2CSH"MOV112 functioning, which allows it to be repaired and tested prior to the refueling outage.'afety Evaluation Summary:
The HPCS system can be considered operable since this jumper installation will allow it to perform its designed functions without any impact from 2CSH"MOV112 on the system's flow rates, pressures, response times, flow paths, or setpoints.
The jumper will not affect any other components or systems. The repairs and testing of 2CSH" MOV112 can be performed safely prior to the refueling outage.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 61 of 136 Safety Evaluation No.: 93-062 Rev. 0 8c 1 Implementation Document No.: Simple Design Change SC2-0038-93 USAR Affected Pages: 9.1-28, 9.1-29, 9.1-29a, 9.1-32, 9.1-37 thru 9.1%0, 9C.6-1; Tables 3.2-1 Sh 7, 9.1-2, 9.1-4 Sh 1, 3-3 Sh 1, 2, 3 (App. 9C);
Figures 1.2-10 Sh 2, 9.1-19a, 9.1-20a, 9.1-25, 9.1-27, 5-2 (App. 9C), 12.3-12, 12.3-45 System: Fuel Handling System (FHS)
Title of Change: Refuel Floor Improvements - RPV Strongback Carousel, Wetlift System, Main Steam Line Plugs Ec Kevlar Slings Description of Change:
This safety evaluation evaluated the implementation of four key improvements in the refuel floor equipment handling and servicing methods used for vessel opening and closing activities during refueling operations. Specifically, the four changes were as follows:
Use of reactor pressure vessel (RPV) head strongback carousel capable of lifting the vessel head. (The simultaneous lifting of vessel head, closure stud hardware, tensioning equipment, and the carousel and all associated equipment has been evaluated and will be implemented following additional related modifications to the plant.)
- 2. Use of new wetlift system that consists of a dryer/separator spreader beam with air-actuated lift pins, four turnbuckles for level adjustment, and two pairs of matched-length Kevlar slings each with a special shackle to match the turnbuckles.
- 3. Use of new GE REM" light steam line plugs to permit the local leak rate testing (LLRT) of the MSIVs to be performed in the direction of the normal steam flow with the reactor vessel and cavity flooded.
- 4. Use of lightweight Kevlar slings for lifting reactor shield plugs, WCS filter and demineralizer plugs, storage pool plugs, drywell head, insulation frame, storage p'ool gate, and fuel transfer bridge.
Safety Evaluation Summary Report Page 62 of 136 Safety Evaluation No.: 93-062 Rev. 0 5. 1 (cont'd.)
Description of Change: (cont'd.)
The design of this new handling and servicing equipment meets the single-failure proof criteria of NUREG-0612.
The use of this new equipment reduces the time required to manually perform the corresponding rigging attachment activities associated with each component during refueling. This time savings results in a net reduction in personnel exposure to radiation/contamination. The new wetlift system will allow the wet transfer of the dryer which provides additional shielding via the flooded reactor cavity. The new main steam line plugs are designed to be installed underwater from the refueling bridge.
Safety Evaluation Summary:
The alternate lifting equipment for vessel opening and closing activities associated with refueling operations has been provided consistent with the commitments to NUREG-0612, which is referenced in USAR Section Appendix SC. The new main steam line plugs have been provided as safety related and are bounded from an accident standpoint by the fuel rod drop accident as described in USAR Section 1 5.7.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 63 of 136 Safety Evaluation No.: 93-063
)mpiementation Document No.:
USAR Affected Pages: 3.1-27, 4.1-2a, 4.6-25 System: Control Rod Drive Title of Change: Control Blade Replacement Description of Change:
The replacement of 21 control blades previously installed in the Unit 2 reactor was required in order to meet shutdown margin requirements in Cycle 4. In addition to these blades, the accelerated replacement of an additional 12 blades reduced the spread of activated cobalt throughout the plant and allowed for some radioactive decay of cobalt in these blades prior to burial. Some radioactive decay of cobalt in these blades reduces their disposal costs. The Stellite pins and rollers used in the previous design were major contributors to increasing radioactive dose rates in the plant. The control blades to be replaced were selected based on nuclear considerations and the blades'oncentration of activated cobalt.
A total of 33 All-B4C control blades previously installed were replaced with eight General Electric Duralife 215 and 25 Marathon control blades.
Safety Evaluation Summary:
Certain control blades must be replaced to ensure that the reactor can be shut down next cycle as defined in the Technical Specification basis for shutdown margin. The high cobalt stainless steel All-B4C control blades with Stellite pins and rollers currently used will be replaced with low cobalt stainless steel control blades with Inconel pins and rollers. This will reduce the spread of activated cobalt throughout the plant and reduce future control blade disposal costs. The replacement of certain All-B4C control blades with Duralife 215 and Marathon control blades will have no impact on the safe operation or shutdown of the plant.
The NRC performed a safety evaluation review of the Marathon control blade and concluded that it is acceptable for use as a control blade in a BWR. All of the features in the Duralife 215 control blade have been reviewed and approved by the NRC on other control blade designs.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 64 of 136 Safety Evaluation No.: 93-064 Implementation Document No.: Simple Design Change SC2-0153-92 USAR Affected Pages: 9.5-43, 9.5-44 System: Div. III Diesel Air Start System (EGA)
Title of Change: "
Revise the Setpoints for the Division III Diesel Air Start Air Compressors and Air Receivers Description of Change:
This simple design change reduced the Division III diesel air start system air compressor start pressure from 230 psig to 215 psig. Previously, the compressors started at 230 psig and stopped at 240 psig. This change maximizes the differential between the start and stop pressures to reduce compressor cycling and reduce compressor wear. The new setpoint maintains sufficient margin to the 190 psig minimum pressure requirement in Technical Specification Section 4.8.1.1.2.a.8. This change also reduced the air receiver low pressure alarm from 225 psig to 200 psig to coordinate alarm setpoints with operating pressure setpoints.
Safety Evaluation Summary:
This change will increase the differential between the start and stop pressure to reduce compressor cycling and reduce compressor wear. This change will also reduce the air receiver low pressure alarm to coordinate alarm setpoints with operating pressure setpoints. The new setpoints remain conservative in relation to system operating requirements and will maintain sufficient margin to ensure Technical Specification requirements are met.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 65 of 136 Safety Evaluation No.: 93-066 Rev. 1 Implementation Document No.: N2-OP-72 USAR Affected Pages: 8.3-11 System: EHS Title of Change: Momentary Paralleling of Motor Control Center (MCC) Main or Tie Feeder Breakers with Kirk Key Interlocks Description of Change:
This change allows the momentary paralleling of power supplies to the motor control centers (MCC) during a power transfer, in lieu of a dead bus transfer by the use of Kirk Keys.
Safety Evaluation Summary:
The MCCs are designed with dual feeds; each feeder is sized to carry the entire load of the MCC. The momentary paralleling would have no effect on the operation of the equipment as the power supplies would be synchronized because both power feeders are connected to the same power and the existing coordination is adequate. This procedure change will enhance the availability of the MCCs.
Based on the evaluation performed, it is concluded'that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 66 of 136 Safety Evaluation No.: 93-067 Implementation Document No.: Temporary Mod.93-046 USAR Affected Pages: N/A System: Offgas (OFG)
Title of Change: Temporary Bypass of the Offgas Charcoal Filters Description of Change:
The offgas filters shown on PAID 42C have experienced high temperatures beyond normal limits. This temporary modification installed an alternate bypass line to bypass and isolate the charcoal filters to mitigate the consequences of a charcoal fire.
Safety Evaluation Summary:
An engineering review of the effects of this change on the offgas system and the radiological release limits has been performed. The alternate temporary bypassing of the filters will allow quenching of the charcoal fire without causing the offgas radiological release limits through the main stack to exceed the Technical Specification limits.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 67 of 136 Safety Evaluation No.: 93-068 Rev. 0, 1 Ea 2 Implementation Document No.: Temporary Mod.93-047 USAR Affected Pages: N/A System: Offgas (OFG)
Title of Change: Nitrogen Insertion into the Offgas Charcoal Filters Description of Change:
The offgas filters shown on PSID 42C have experienced high temperatures beyond normal limits, with indications of fire in the charcoal beds. This temporary modification introduced nitrogen into the filters to mitigate the high temperature condition. The nitrogen supply source was a tanker truck with tubing, valves, and instrumentation. Note that the charcoal filters were isolated via Temporary Modification 93-046 prior to nitrogen insertion.
The purpose of Revision 2 was to verify that the charcoal fire was completely extinguished and that re-ignition would not take place when the filters were placed back into service. This was accomplished by gradual admission of dry air into the filters after having reasonable assurance that the charcoal fire had been put out as monitored by measuring the carbon monoxide concentration and charcoal filter temperature.,
Safety Evaluation Summary:
An engineering review of the effects of this change on the offgas charcoal filters and the system has been performed. This change is to mitigate the consequences of the charcoal high temperature conditions. Since the charcoal filters will be isolated prior to the use of this temporary modification, no adverse impact on the system's functional capability is introduced.
Based on the evaluation performed, it is concluded that this-change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 68 of 136 Safety Evaluation No.: 93-069 Implementation Document No.: GAP-POL-01, QAP-POL-1.01, NLAP-POL-01, NTP-POL-500 USAR Affected Pages: 13.1-3, 13.1-5, 13.2-1; Figures 13.1-1, 13.1-2, 13.1-5; Appendix B System: N/A Title of Change: Nuclear Quality Assurance, Licensing, and Training Organizational Reporting Structure-Revised Procedures GAP-POL-01 and QAP-POL-1.01 Description of Change:
The Nuclear Quality Assurance, Licensing and Training Branches were reorganized as follows: the position of Vice President Nuclear Quality Assurance has been eliminated and the new position of General Manager Safety Assessment, Licensing, and Training established. The organizational structure of the Quality Assurance, Licensing, and.Training organizations has changed such that the Managers Quality Assurance Units 1 and 2, Licensing, and Training report directly to the General IVlanager Safety Assessment, Licensing, and Training. The Manager Quality Assurance Support reports administratively to the Manager Quality Assurance Unit 2, but retains functional responsibilities for both units. Prior to this change, the IVlanagers Quality Assurance Units 1 and 2 and the Manager Quality Assurance Support reported to the Vice President Nuclear Quality Assurance; the Manager Licensing reported to the Executive Vice President Nuclear; and the Manager Training reported to the Vice President Nuclear Generation. Functions currently performed by the Quality Assurance, Licensing, and Training organizations are not affected by the revised reporting structure.
NOTE: See summary for Safety Evaluation 93-127, Rev. 1, for subsequent organization changes.
Safety Evaluation Summary:
The changes made to the organizational structure of Quality Assurance, Licensing, and Training continue to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Unit 1 and Unit 2. Clear management-control and effective lines of communication and authority between the organizational units involved in the management, operation, and technical support for the operation of Nine IVlile Point Unit 1 and Unit 2 continue to be
Safety Evaluation Summary Report Page 69 of 136 Safety Evaluation No.: 93-069 (cont'd.)
Safety Evaluation Summary: (cont'd.)
provided. The Managers Quality Assurance Units 1 and 2 retain overall authority and responsibility for the QA Program for their respective units, and the General Manager Safety Assessment, Licensing, and Training will have senior management responsibility for Quality Assurance, Licensing, and Training/Emergency Preparedness activities, allowing for the elimination of the Vice President Nuclear Quality Assurance position. The Managers Quality Assurance Units 1 and 2, Licensing and Training, will have direct access to responsible corporate management at a level where action appropriate to the mitigation of quality assurance, licensing and training/emergency preparedness concerns can be accomplished, and sufficient independence from cost and schedule is maintained.
Based on this evaluation, the organizational structures of the Quality Assurance, Licensing, and Training/Emergency Preparedness organizations continues to satisfy the acceptance criteria of SRP 13.1.1, SRP 13.1.2-13.1.3, SRP 17.1, SRP 17.2 and Unit 1 and 2 Technical Specification 6.2.1, and does not constitute an unreviewed safety question.
Safety Evaluation Summary Report Page 70 of 136 Safety Evaluation No.: 93-070 Implementation Document No.: Simple Design Change SC2-0039-93 USAR Affected Pages: 9.1-37; Figure 3.8-8 System: PCB Title of Change: Addition of Jacking Blocks to the Drywell Head Description of Change:
This safety evaluation evaluated the implementation of four jacking blocks on the drywell head for the purpose of providing a means via hydraulic jacks to assist the polar crane in removing the drywell head. The previous method of removing the head with the polar crane resulted in the swaying of the drywell head as the binding between the head and the chimney section is suddenly relieved as the crane lifts the head.
The four hydraulic jacks have been equally spaced around the base of the drywell head and are connected to a common pump unit to ensure that they act simultaneously and evenly lift the head.
The design of the four jacking blocks envelops the case where the drywell head' weight is carried by the four jacks. The intent of the jacks is to provide the additional controlled uplift force to overcome the frictional force between the head and the chimney section as the polar crane lifts the head out of the chimney section. Through the use of these jacks, which assist the polar crane in lifting the drywell head, the swaying of the head during the lift is eliminated.
The polar crane will remain as the primary lifting device for the drywell head.
Therefore, the Jacking blocks and jacks are not required to meet the single-failure proof criteria of NUREG-0612 for heavy loads. The welding of the jacking blocks to the exterior of the drywell head is not considered pressure boundary welding and welding was done in accordance with ASME Section III, 1971 through Summer 1973 Addenda.
Safety Evaluation Summary:
The addition of the four jacking blocks on the drywell head to improve the present method of drywell head removal does not affect the commitment to Guideline for the Control of Heavy Loads tNUREG-0612) as described in USAR Appendix 9C.
Safety Evaluation Summary Report Page 71 of 136 Safety Evaluation No.: 93-070 (cont'd.)
Safety Evaluation Summary: (cont'd.)
The welding of the blocks to the drywell head does not affect the drywell head' function as a primary containment boundary.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 72 of 136 Safety Evaluation No.: 93-073 Implementation Document No.: Procedure N2-OP-62 Procedure N2-OSP-CNT-M001 USAR Affecte'd Pages: NIA System: DBA Hydrogen Recombiner Title of Change: Deactivation of 2HCS" MOV1A, 18, 3A and 38 Description of Change:
10CFR50 Appendix J requires Type C testing of 2HCS"MOV4A, 48, 6A and 68 for primary containment penetration leakage and requires the test pressure be applied in the same direction the valve must function to preclude leakage in an accident. However, reverse direction testing is permitted if it can be demonstrated that such test results are equivalent or more conservative than forward postaccident direction testing.
The HCS inboard isolation valves 2HCS"MOV4A, 48, 6A and 68 are flexible disc gate valves and were reverse direction tested in a way that test results are not equivalent or more conservative than results obtained using forward postaccident direction testing.
Therefore, outboard isolation valves 2HCS" MOV1A, 18, 3A and 38 were deactivated in the closed position to maintain primary containment integrity. This safety evaluation evaluates the operability of HCS following implementation of the actions associated with Technical Specification 3.6.3.
Safety Evaluation Summary:
Since the leak-tightness of HCS inboard isolation valves 2HCS"MOV4A, 48, 6A and 68 cannot be confirmed, these valves are declared inoperable. Therefore, HCS outboard valves 2HCS" MOV1A, 18, 3A and 38 need to be temporarily deactivated in the closed position to maintain primary containment integrity until such time that the inboard isolation valves can be Type C tested satisfactorily to ensure primary containment isolation requirements are met.
. The HCS outboard isolation valves will be deactivated in the closed position by opening their respective breakers at 2EHS"MCC102 and 2EHS"MCC302. These breakers will be administratively controlled and be allowed to be closed only when
Safety Evaluation Summary Report Page 73 of 136 Safety Evaluation No.: 93-073 (cont'd.)
Safety Evaluation Summary: (cont'd.)
HCS operation is required per Operating Procedure N2-OP-62. However, it is recognized that in accordance with Technical Specification LCO 3.6.3, and applicable footnote, that the outboard isolation valves may be activated on an intermittent basis with administrative control. Required surveillances will be maintained current to assure HCS operability.
Closing the breakers as required by N2-OP-62 will ensure that primary containment pressures and temperatures are consistent with the design pressure and temperature of HCS.
The MCCs have been evaluated for accessibility. Walking paths for an operator from the control room to the MCCs and from the OSC to the MCCs have been analyzed to determine the radiation doses that the operator would receive. It is expected that the time required for an operator to walk from the control room to the MCCs, close the breakers and return to the control room is 10 minutes or less and from the OSC to the MCCs, close the breakers and return to the OSC is 15 minutes or less. A separate operator will walk to each MCC. The resulting doses have been evaluated and determined to be within the acceptable limits as defined in the Standard Review Plan (NUREG-0800) Section 6A and meet the requirements of GDC 19 in 10CFR50 Appendix A.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 74 of 136 Safety Evaluation No.: 93-074 Implementation Document No.: Temporary Mod.93-040 USAR Affected Pages: N/A System: Fire Protection - Water (FPW)
Title of Change: Connecting Fire Protection System to Condenser Tube Flush Rig Description of Change:
This temporary modification permitted the installation of a temporary fire hose from fire hose reel 2FPW-FHR14 for the flushing of the main condenser tubes during refueling outage 3. The connection at the fire hose reel required the installation of an angle Y-gate with two valves; one to isolate the firefighting hose and another to isolate the temporary fire hose used for tube flushing.
Safety Evaluation Summary:
This temporary modification will have no impact on the safe operation or capability to keep the plant in a safe shutdown condition.
Fire protection capabilities are not degraded at any time, because system pressure and flow will be maintained in the fire protection system while this temporary modification is in effect.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 75 of 136 Safety Evaluation No.: 93-076 Rev. 1 Implementation Document No.: Calculation No. A10.1-E-116 USAR Affected Pages: 3.9A-27a System: RHR, RCIC, LPCS, HPCS Title of Change: Valve Air Leakage to Water Leakage Correlation Description of Change:
This change allows some of the reactor coolant system pressure isolation valves to utilize their Type "C" air test leakage data to satisfy the high-pressure water leak rate test requirement.
Safety Evaluation Summary:
An analysis was made of the air leakage and water leakage data taken at NMP2 from 1986 to 1992, and from this analysis an empirical correlation was developed to convert the Type "C" air test leakage rate to a high-pressure water leak rate in order to meet the requirements of Technical Specification 3.4.3.2.
This analysis will have the practical effect of eliminating the high-pressure water leakage test for check and globe valves that have been Type "C" air tested, and the leakage rate converted to water leakage using the correlation meets existing acceptance criteria.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 76 of 136 Safety Evaluation No.: 93-077 Implementation Document No.: N/A USAR Affected Pages: N/A System: 115-kV Offsite Power Sources Title of Change: Replace 115-kV Circuit Breakers (R50, R60, R115 and R225) at Scriba Station Description of Change:
This modification replaced the four 115-kV circuit breakers at Scriba Station with breakers of the same electrical rating. The replacement breakers are of different design so that previous breaker trips associated with the breaker hydraulic system were eliminated. The work was scheduled such that both the main and feeder breaker for one 115-kV line position were replaced at the same time so that only a single line outage was required.
Safety Evaluation Summary:
This modification enhances operation of the 115-kV offsite supply system by replacing existing circuit breakers with more reliable units. This modification will have no impact on the safe operation or shutdown of the plant. The work will be performed while the unit is shut down; therefore, the requirements of Technical Specification 3.8.1.2 will be met by ensuring that one other offsite supply line and at least one other diesel generator will be available.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 77 of 136 Safety Evaluation No.: 93-078 Rev. 0 8c 1 Implementation Document No.: Temporary Mod.93-024 USAR Affected Pages: N/A System: IAS, AAS Title of Change: Temporary Compressors for the IAS Upgrade Modification Description of Change:
This change was related to the instrument air system (IAS) and the breathing air system (AAS) as described in USAR Section 9.3. It was required to be implemented to facilitate the implementation of permanent Modification PN2Y92MX004, Instrument Air System Upgrade. As part of the above-referenced permanent modification, the plant's existing air compressors have been replaced and the plant's permanent breathing air compressor has been eliminated. The purpose of this temporary modification was to install two temporary air compressors to supply the plant's instrument and breathing air needs while the new permanent compressors were being installed. Based on a compressed air availability/reliability review, Operations requested to have check valves installed on the temporary air lines to reduce the load on the permanent IAS compressors in the event of a line break or loss of temporary air while the plant is operating. In addition, a temporary hose was installed from the outlet of 2IAS-DRY1B to the inlet of 2IAS-FLT3B, bypassing existing plant piping. This change facilitates permanent tie-ins to be implemented as directed by the permanent modification.
Safety Evaluation Summary:
An engineering review of the effects of this change on the IAS, AAS, and other interfacing systems has been performed. Nuclear safety will not be compromised as the safety-related components supplied by these temporary air compressors do not rely on air to perform their safety function, or are provided with safety-related accumulators capable of supplying the amount of required air for the performance of their safety functions.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 78 of 136 Safety Evaluation No.: 93-079 Implementation Document No.: Temporary Mod.93-025 USAR Affected Pages: N/A System: SAS, IAS Titfe of Change: Temporary SAS Compressor for the IAS Upgrade Modification Description of Change:
This change was related to the service air system (SAS) and the instrument air system (IAS) as described in USAR Section 9.3. It was required to be implemented to facilitate the implementation of permanent Modification PN2Y92MX004, Instrument Air System Upgrade. As part of the above-referenced permanent modification, the plant's air compressors have been replaced. The purpose of this temporary modification was to install a temporary air compressor to supply the plant's service air needs while the new permanent compressors were being installed.
Safety Evaluation Summary:
An engineering review of the effects of this change on the SAS, IAS, and other interfacing systems has been performed. Nuclear safety will not be compromised as the loss of service air will not affect the plant's safe shutdown capability or the plant's ability to maintain the plant in a safe shutdown condition.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 79 of 136 Safety Evaluation No.: 93-080 Implementation Document No.: Simple Design Change SC2-0102-93 USAR Affected Pages: 9.1-7a; Figures 9.1-3, 5-2 (App. 9C); Table 3-3 Sh 3 (App. 9C)
System: FNS (Nuclear Fuel Storage)
Title of Change: Control Blade Storage Frame Description of Change:
This change provided a design for a safety-related storage frame (2FNS" FRM1) which supports three nonsafety-related control rod blade racks (2FNS-RAK19, 2FNS-RAK20, and 2FNS-RAK21). The frame was placed on the spent fuel pool floor and provides storage for an additional 30 control rod blades. The additional storage location allows for replacement of old control rod blades with new control rod blades without delay to the refueling process. Placement of the storage frame and the three control rod blade racks within the spent fuel pool was accomplished by use of the main hoist of the reactor building crane.
Safety Evaluation Summary:
The addition of the control blade storage frame and three control blade racks enhances the refueling process for the replacement of control rod blades. The frame also provides storage for the control rod blades until arrangements for offsite storage are made. The storage frame shall be classified as a nuclear safety-related component, designed and analyzed to the same technical criteria as specified in USAR Section 9.1 for the spent fuel storage racks.
The lifting of the frame and racks shall be accomplished utilizing the main hoist of the reactor building crane, which is of single-failure proof design. The lifting rigging shall meet the criteria provided in NlVIPC's commitment to the guidelines for the Control of Heavy Loads (NUREG-0612) and the requirements of ANSI N-14.6.
This change to the arrangement in the spent fuel pool shall have no impact on the safe operation or shutdown of the Unit 2 reactor.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 80 of 136 Safety Evaluation No.: 93-082 Implementation Document No.: Simple Design Change SC2-0011-.93 USAR Affected Pages: Figures 1.2-10, 9.1-25, 5-2 (App. 9C),
12.3-12, 12.3-45 System: SCA, MHF Title of Change: Access Control Building, Refuel Floor Description of Change:
This simple design change installed an access control building on the refuel floor of the reactor building at el 353'-10". The access building is used as an office
~
primarily by Radiation Protection for controlling access to and activities on the refuel floor. The designated storage location for two jib cranes, one storage pool plug, and reactor shield plugs was revised to accommodate the new building and avoid physical interference. Electrical power required for building lighting, recepticals, exhaust fan, and heating and cooling units is supplied by the normal plant ac distribution system.
Safety Evaluation Summary:
The refuel floor structure, used to support the access building, along with systems used to support the operation of the building, have been evaluated and determined not to be adversely impacted by this change. The building, which is classified as nonsafety related, has been evaluated in accordance with Nuclear Engineering Administrative Procedure NEP-DES-243 since it is being located in a safety-related area, and it was determined that no components related to safety will be impacted by the building during a seismic event.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 81 of 136 Safety Evaluation No.: 93-083 implementation Document No.: N2-EPM-GEN-V582 USAR Affected Pages: Table 7B-1 Sh 5 System: SCA Title of Change: Changing the Testing Cycle for Non-Class 1E Redundant Protective Devices Within PGCC From Each Refueling Outage to 18 Months Description of Change:
USAR Table 7B-1 was revised to indicate testing will be performed on the redundant protective circuit breakers within PGCC on an 18-month cycle in lieu of the current "each refueling outage." There were no equipment, installation or operating changes. This change allows testing of these protective devices on 18-month cycles instead of each refueling outage. There is no need to test the breakers during refueling outages; they can be tested during normal plant operations.
Safety Evaluation Summary:
Non-Class 1E cables routed in divisional ducts are installed in grounded flexible conduit. Flexible conduit is not considered as a separation barrier. In order to satisfy the requirement the redundant circuit breakers were provided for protection of Class 1E circuits in PGCC divisional ducts from non-Class 1E circuit runs in flexible conduit. The respective circuits provide lighting and convenience outlet power to control room and relay room PGCC panels. There is no specified frequency for testing of the circuit breakers within Regulatory Guide 1.75 or IEEE Standard 384-1974. NMPC self imposed "each refueling outage" in a commitment to the NRC in 1986.
The 18-month cycle is within a general refueling outage schedule of 18 months.
Therefore, changing of the test frequency from each refueling outage to 18 months has no impact on the breaker performance because the time interval is not changed.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 82 of 136 Safety Evaluation No.: 93-085 Implementation Document No.: Procedures N2-FHP-13.1, N2-FHP-13.2, and N2-FHP-3 USAR Affected Pages: 15E.2-1 System: Refuel, Nuclear Fuel, Reactivity Control Title of Change: Modification of the NMP2 Reload/Offload Procedures Description of Change:
The NMP2 core offload/reload procedures were revised to include the following changes:
~ Four exposed fuel bundles around each SRM will be unloaded after the spiral offload pattern has been completed. Conversely, four fuel bundles will be loaded around each SRM before the spiral reload pattern has begun.
Added a contingency for an inoperable SRM in the precautions section of the offload/reload procedures.
During core offload/reload, with all rods inserted, the RPS circuitry may be in coincident logic (i.e., shorting links installed). Before a control rod is withdrawn, the RPS circuitry shall be placed in a noncoincident configuration.
The instructions for removing/installing the RPS shorting links during reload was moved to Procedure N2-FHP-13.2, Attachments 3 and 4, respectively.
During core reload, a holdout tag will be placed on the control rod withdrawal push button. This provides administrative controls to assure all control rods remain fully inserted.
~ A core offload/reload sequence map was added to procedures N2-FHP-13.1 and N2-FHP-13.2, respectively.
Safety Evaluation Summary Report Page 83 of 136 Safety Evaluation No.: 93-085 (cont'd.)
Description of Change: (cont'd.)
~ Instructions to defeat the SRM period alarms were moved from N2-FHP-3 to the end of N2-FHP-13.1. Likewise, the instructions to restore the alarms were moved to the beginning of N2-FHP-13.2.
~ Instructions will be added to N2-FHP-3, the "Refueling Manual," to place the RPS in noncoincident mode prior to performing the shutdown margin test, and it shall be in a noncoincident configuration for all control rod manipulations prior to the demonstration of shutdown margin. RPS will be restored to coincident mode following shutdown margin testing.
Several of these changes save time during core offload and reloading. Other changes eliminate problems that could become critical path for the outage.
Additionally, by placing the RPS in coincident mode, surveillance and maintenance activities that previously suspended core movement or that had to wait until refueling was complete can now be done concurrent with offload/reload activities.
Safety Evaluation Summary:
The proposed changes to the refuel procedures will not result in an unreviewed safety question. Each change follows the NMP2 licensing basis refueling requirements. In addition, industry recommendations have been reviewed and, where appropriate for NMP2, have been incorporated into the suggested changes.
Specifically; the requirements of Technical Specifications 3/4.1 and 3/4.9, GE SILs 068 and 372, GE RICSIL 039, GE PRC 89-10 and NSAC/164L will be met by these changes.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 84 of 136 Safety Evaluation No.: 93-087 Implementation Document No.: Simple Design Change SC2-0105-93 USAR Affected Pages: Figure 5.4-9c System: Reactor Core Isolation Cooling (ICS)
Title of Change: Replacement of Steam Trap 2ICS-TRP1 Description of Change:
This simple design change was initiated to document and control the work activities associated with the replacement of reactor core isolation steam trap 2ICS-TRP1. The previous trap did not perform its intended design function.
Replacement was required with a type better designed, thus more reliable. The replacement trap is not provided with a drain connection; therefore, the existing drain line was removed.
Safety Evaluation Summary:
This simple design change will have no impact on the safe shutdown or the capability to keep'the plant in a safe shutdown condition.
The function or the method of performing the function of the ICS system is not degraded at any time. Materials of construction, design and installation requirements are consistent with the original design of the system.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 85 of 136 Safety Evaluation No.: 93-088 Rev. 0 L 1 Implementation Document No.: N/A USAR Affected Pages: 9A.3-3, 9A.3-4, 9A.3-31 System: N/A Title of Change: Adoption of NFPA-600 Physical Fitness Requirements for Fire Brigade Description of Change:
This change modified the Fire Brigade physical fitness requirements to reflect the current consensus standard as published by the National Fire Protection Association (NFPA 600).
Safety Evaluation Summary:
The physical fitness requirements for Fire Brigade members included a requirement for an annual agility test. This requirement was contained in NFPA 1001, a standard which was applicable to municipal fire department firefighters. In 1991, NFPA approved the issuance of NFPA 600 to address the requirements of industrial Fire Brigade organizations, such as the one at Nine Mile Point. This evaluation adopts the physical fitness requirements from NFPA 600 for application to the Fire Brigade. Since the qualifications outlined in NFPA 600 satisfy requirements delineated in 10CFR Appendix R,Section III.H, no degradation will result from this change.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 86 of 136 Safety Evaluation No.: 93-089 Implementation Document No.: GAP-POL-01, Rev. 04 USAR Affected Pages: 13.1-13, 13.6-1; Figure 13.1-2 System: N/A Title of Change: Nuclear Security and Procurement Organizational Structures - Revised Procedure GAP-POL-01 Description of Change:
The Nuclear Security and Procurement Branches of Site Support were reorganized as follows:
1
~ The Manager Nuclear Security's new organization is comprised of the following Direct Reports/Sections:
General Supervisor Nuclear Security Operations Supervisor Nuclear Security Support Nuclear Security Investigators Supervisor Nuclear Security Administration Supervisor Access Authorization/Fitness-For-Duty
~ The Manager Procurement's new organization is comprised of the following Direct Reports/Sections:
Supervisor Procurement Engineering General Supervisor Inventory Management Supervisor IVlaterial Receipt, Test, and Inspection Supervisor Warehouse and Storeroom Operations General Supervisor Purchasing The discussion and depiction of positions reporting directly to the IVlanager Nuclear Security were deleted from Section 13 of the USAR. The NRC-approved Physical Security Plan includes an organization chart and functional descriptions of responsibilities and relationships for key personnel positions in the Nuclear Security Branch. Changes to the Physical Security Plan are implemented per the provisions of 10CFR50.54(p).
Safety Evaluation Summary Report Page 87 of 136 Safety Evaluation No.: 93-089 (cont'd.)
Description of Change: (cont'd.)
NOTE: Subsequent organization changes were evaluated under Safety Evaluations93-127, 94-010 and 94-016.
Safety Evaluation Summary:
The changes made to the organizational structures of Nuclear Security and Procurement continue to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Unit 1 and Unit 2. Clear management control and effective lines of communication and authority between the organizational units involved in the management, operation, and technical support for the operation of Nine Mile Point Unit 1 and Unit 2 continue to be provided.
Based on this evaluation, the organizational structures of Nuclear Security and Procurement continue to satisfy the acceptance criteria of SRP 13.1.1 and SRP 13.6 and do not constitute an unreviewed safety question.
Safety Evaluation Summary Report Page 88 of 136 Safety Evaluation No.: 93-090 Rev. 0 L 1 Implementation Document No.: Mod. PN2Y93MX003 USAR Affected Pages: 4.6-7, 4.6-8, 4.6-8a, 7.2-5, 7.2-5a, 7.2-5b; Figures 4.6-5a, 5.1-2a System: Control Rod Drive, Reactor Instrumentation Title of Change: Reactor Vessel Instrumentation Reference Leg Backfill Description of Change:
This modification added a flow-controlled backfill to four reactor vessel level instrumentation reference legs. The backfill was intended to prevent dissolved noncondensable gas from collecting in the reference legs and creating a false high level measurement. Backfill flow was taken from control rod drive pumps discharge, filtered, metered, and passed through the reference legs. This modification was made in response to NRC Bulletin 93-03.
Safety Evaluation Summary:
An engineering review of the proposed changes was performed. The review, which included design, operability, and potential system interactions, has determined that the implementation of Modification PN2Y93MX003 will improve the response of the level instrumentation on reduction of reactor pressure without causing any significant safety or operability issues.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 89 of 136 Safety Evaluation No.: 93-092 Implementation Document No.: Calculations PR(C)-26-G Rev. 0, 2-93-004, H21C-029 USAR Affected Pages: N/A System: FHS Title of Change: Allowing Work to Proceed in the Upper Elevations of the Drywell While Removing Reactor Core Control Blades, LPRMs, and Other Nonfuel-Irradiated Components Description of Change:
This safety evaluation evaluated the USAR and the impact on the NMP2 design as a result of allowing personnel to work in the upper areas of the drywell during the movement of nonfuel-irradiated hardware. The function of this review was to assure that radiation workers could safely carry on their duties during the refuel outage. The NMP2 unit may be operated safely with the implementation of the change.
The change has been evaluated with respect to ALARA considerations, allowing work to proceed in the drywell during irradiated component movement. Although this change could have resulted in an increased dose rate to the individual radiation worker, engineered (fuel transfer shield bridge) and programmatic controls were in place to minimize any impact. The projected accumulated dose per outage worker remains far below 10CFR20 limits.
Section 15 of the NMP2 USAR provides the design basis analysis for the radiological consequences of postulated accidents. Calculations supporting these analyses were reviewed. The radiological conditions of any of the analyzed accidents, as required in NUREG-0800 (Standard Review Plan), are not affected by this change.
Safety Evaluation Summary:
Based on the review and analyses performed, all anticipated effects of this change on plant systems and setpoints have been reviewed and found to be bounded within existing USAR evaluations. This change allows the movement of irradiated hardware without adversely impacting permanent plant monitoring systems.
Safety Evaluation Summary Report Page 90 of 136 Safety Evaluation No.: 93-092 (cont'd.)
Safety Evaluation Summary: (cont'd.)
Radiation will continue to be monitored by existing plant instrumentation. The activities have no active function relative to plant safe shutdown or to quantitative monitoring of releases of radioactive material to the environment. The proposed activities have no impact on any effluent streams. The basis for controlling movement of irradiated components is to limit radiation exposure to operators during the normal operation of NMP2 in compliance with 10CFR20. Irradiated hardware removal will not create the possibility of a different type of accident or equipment malfunction different than currently evaluated in the NMP2 USAR.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 91 of 136 Safety Evaluation No.: 93-093 Implementation Document No.: Procedure N2-Tl P-HVR-5001 USAR Affected Pages: N/A System: Reactor Building Ventilation (HVR)
Title of Change: Removing Unit Coolers 2HVR"UC408A &. B or 2HVR"UC409A & B from Service Simultaneously for Testing per N2-TTP-HVR-5001 Description of Change:
Generic Letter 89-13 requires that heat exchangers cooled by service water be regularly tested to evaluate their actual capacity against design.
Testing of the divisional switchgear unit coolers 2HVR"UC408A & B and 2HVR"UC409A &. B, located in the auxiliary bays, required that both unit coolers in one switchgear room be out of service for short periods of time.
This safety evaluation addresses the ac and dc divisional switchgear equipment operability concerns raised by this testing method.
Safety Evaluation Summary:
The unit coolers for the ac and dc divisional switchgear rooms may both be removed from service for testing in accordance with this procedure. A dedicated operator will ensure that the operable unit cooler is returned to service when the
'room temperature reaches 95'F. This ensures that cooling is established prior to the room temperature reaching the 104'F ac and dc divisional switchgear equipment qualification limit. Therefore, this test method does not impair the safe operation of the plant, safe shutdown, fire protection, jet impingement, Category II Over I, ALARA, equipment qualification, control room habitability, fuel analysis, equipment clearances or seismic qualifications.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 92 of 136 Safety Evaluation No.: 93-094 Implementation Document No.: Temporary IVlod.93-050 USAR Affected Pages: N/A System: Service Water System (SWP)
Title of Change: 2SWP"PSX1003A Internal Diaphragm Replacement Description of Change:
This temporary modification repaired switch 2SWP"PSX1003A with noncertified parts (diaphragm and 0-ring) and returned it to operation.
Safety Evaluation Summary:
The safety-related function of valve 2SWP"FV47A is to isolate CWS from SWP under a loss of offsite power condition. The proposed change does not affect the safety-related parameters of the valve since the mechanical portion of this switch is isolated from the portions of the valve providing the safety-related function.
Failure of the switch does not inhibit the valve's ability to perform its safety-related function. However, since the switch is powered from a safety-related Class 1E source and is not isolated from it, the electrical portions of the switch along with the associated structural parts are classified as safety related to meet Regulatory Guide 1.75 r'equirements. The proposed diaphragm changeout will have no impact on the structural or electrical properties of the switch and, therefore, no new failure scenarios are introduced.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 93 of 136 Safety Evaluation No.: 93-095 Implementation Document No.: N/A USAR Affected Pages: 5.4%2a, 10.4-22a System: N/A Title of Change: Methods of Chemical Analysis Change Description of Change:
This change allowed for the identification of additional methods to be used for the determination of chlorides, pH, and conductivity in the reactor vessel water.
Safety Evaluation Summary:
The identification of ion chromatography for the determination of chloride ion, and techniques providing adequate sensitivity to meet the limits specified in Regulatory Guide.1.56 Revision 1, Table 1, for the analyses of chlorides, conductivity, and pH in reactor vessel water is consistent with industry practices, provides a suitable technique(s) to measure at or below acceptable reactor water chemistry limits, is in compliance with regulatory requirements, and maintains the design basis for the reactor water cleanup system.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 94 of 136 Safety Evaluation No.: 93-096 Implementation Document No.: GAP-DES-03 USAR Affected Pages: N/A System: N/A Title of Change: Installation and Control of Temporary Communications Equipment Per GAP-DES-03 Description of Change:
Procedure GAP-DES-03 was revised to include an exclusion for temporary communications installed in accordance with a new Technical Support Administrative Procedure. The new procedure was developed to allow the installation and control of temporary communications equipment (GAI-TRONICS) in facilities at Nine Mile Point (e.g., temporary trailers installed in support of refueling outage activities).
Safety Evaluation Summary:
Procedure GAP-DES-03 is being revised to allow temporary communications equipment to be installed in temporary facilities at Nine Mile Point utilizing a Technical Support Administrative Procedure. This procedure controls the installation and removal of the temporary communications equipment and ensures that no additional electrical load is added to the normal plant communications system power source. Page and party line signals from the plant communications system will be fed to the temporary equipment but the 120V ac power shall be supplied from the individual facility's wall outlets.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 95,of 136 Safety Evaluation No.: 93-098 Implementation Document No.: Simple Design Change SC2-0147-93, Procedure N2-ESP-RCS-R737 USAR Affected Pages: Figure 8.3-88 Sh 12 System: Reactor Recirculation System Title of Change: Recirculation Pump Motors Primary Containment Penetration Protection Setpoint Change of Change: 'escription This change increased the setpoint of the instantaneous overcurrent protection of the penetration which provides power to reactor recirculation pumps. The setpoint increase from 52A to 64A, or from 12 times to 15 times the full load current of the motor, was needed to eliminate inadvertent trips of the pumps during upshifting from low to high speed.
Safety Evaluation Summary:
After increase of the setpoint, the protective relays still provide adequate protection of the penetration. The worst condition of the short circuit for the penetration which is allowed by the protective relays is still below the penetration capability.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 96 of 136 Safety Evaluation No.: 93-100 Implementation Document No.: SAFER/GESTR LOCA Methodology USAR Affected Pages: Sections 6.3, 15B, 15C, Appendix A System: Various Title of Change: Operation of NMP2 Reload 3/Cycle 4 Description of Change:
Due to the introduction of fuel of a new design (GE11) in Reload 3 (9x9 lattice versus 8x8 lattice used in previous reloads), various issues not normally considered in previous reloads have been evaluated. The Reload 3 fuel bundle design is known as GE11-P9CUB332-13GZ-120M-146-T and is an approved fuel design in GESTAR. The Reload 3 bundles have an average enrichment of 3.32 wt lo U-235.
The GE11 design consists of 74 fuel rods (8 being part length rods) and 2 large central water rods covering 7 fuel rod lattice positions. The maximum allowable peak LHGR for the GE11 fuel is 14.4 kW/ft. The Cycle 4 core loading will insert 196 fresh GE11 bundles and 32 twice burned GE6B bundles that were discharged at the end of Cycle 2.
The GE11 fuel design was licensed under Amendment 22 of the GESTAR process approved by the NRC in July 1990. Amendment 22 contained fuel design criteria against which a new fuel design could be compared and judged acceptable, thus allowing the introduction of new fuel without prior NRC review and approval.
The SAFER/GESTR LOCA methodology was used for the Reload 3/Cycle 4 loss-of-coolant accident (LOCA) analysis. A Technical Specification change was obtained which allowed the use of the SAFER/GESTR methodology. SAFER/GESTR is the GE improved LOCA methodology, which has been approved for use by the NRC.
The NRC approved the use of SAFER/GESTR at NMP2 in Amendment No. 52, dated November 10, 1993.
The limiting transient for Reload 3 is the rod withdrawal error event. The change in MCPR for this event set the OLMCPR at 1.36 for the GE11 fuel and 1.31 for the GE6B and GE9B fuel. The limiting event for the vessel overpressurization analysis is the MSlV closure (flux scram). The peak steam line and vessel bottom pressures are 1252 and 1283 psig, which is well below the safety limit of 1375 psig.
Safety Evaluation Summary Report Page 97 of 136 Safety Evaluation No.: 93-100 (cont'd.)
Description of Change: (cont'd.)
As part of the reload, additional analyses were performed for equipment out of service. This analysis determines the allowable combinations and their operating impact for various equipment out of service.
Safety Evaluation Summary:
Based on the evaluation performed, it is concluded that Nine Mile Point Unit 2 can be safely operated during Reload 3/Cycle 4. Operation in accordance with this safety evaluation does not involve an unreviewed safety question nor is a Technical Specification change required.
Safety Evaluation Summary Report Page 98 of 136 Safety Evaluation No.: 93-101 Implementation Document No.: Temporary Mod..93-056 USAR Affected Pages: N/A System: Rod Sequence Control System (RSCS)
Title of Change: Alternate Power Source for RSCS Power Supply Description of Change:
This temporary modification provided for an alternate power source for RSCS PS¹6 by disconnecting cable 2RDSNNK523 (fed from 2VBS-PNLB101) in 2CEC" PNL701 and connecting cable 2RDSNNK504 (fed from 2VBS-PNLA101) in its place.
Safety Evaluation Summary:
There is no adverse nuclear safety significance to this change since the normal.
power provided to this device is black (nondivisional, nonsafety related) from the vital bus u'ninterruptible power supply 2VBB-UPS1B. The alternate power source utilized in this temporary modification is from 2VBB-UPS1A which is also part of the plant's normal 120-V ac system. The RSCS is not a safety-related system and is not required during the refueling mode.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 99 of 136 Safety Evaluation No.: 93-102 Implementation Document No.: Simple Design Change SC2-0158-93 USAR Affected Pages: Table 3.9A-12 Sh 9; Figures 5.4-13b, 5.4-13c, 5.4-13e System: Residual Heat Removal System (RHS)
Tide of Change: Removal of Internals from 2RHS"V7, V8 and V9 Description of Change:
This simple design change removed the internals from the check valves, 2RHS"V7, V8, and V9, located in the minimum flow test lines of the RHS pumps (P1A, B, C).
Safety Evaluation Summary:
This design change will have no impact on the safe operation or capability to keep the plant in the safe shutdown condition.
The RHS system will not be degraded at any time since system pressure and flow will be maintained during pump minimum flow operation, and that system pressure integrity will.be maintained during all modes of system operation.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 100 of 136 Safety Evaluation No.: 93-103 Rev. 1 Implementation Document No.: N2-OP-38, Rev. 6 USAR Affected Pages: N/A System: SFC Title of Change: Backup Spent Fuel Pool Cooling During Refueling Outage Description of Change:
Four methods were proposed for the spent fuel cooling and mixing as backups to the SFC Loop B during this outage. Of these, one is currently approved and the remaining are being evaluated. The SFC Loop A pump was used by restoring Div. I power to it and by removing some unavailable pump protective functions. The disabling of the protective features allows the availability of the spent fuel cooling system that otherwise would be unavailable.
The two methods of circulating pool and cavity water were acceptable because they did not interfere with the existing SFC and helped reduce the spent fuel pool heatup rate.
Safety Evaluation Summary:
The disability of the low flow pump trip will be replaced by manual trip upon exceeding predetermined motor current. Therefore, pump low flow protection will still be available. The other two protective functions (low suction and discharge pressures) are indirectly available through the motor current.
The temporary cavity cleanup system and sump pumps are being proposed for enhancing the mixing between the spent fuel pool and reactor cavity.
Based on the evaluation performed, it is concluded that these changes do not involve an unieviewed safety question.
Safety Evaluation Summary Report Page 101 of 136 Safety Evaluation No.: 93-104 Implementation Document No.: N2-TTP-RHR-5001T USAR Affected Pages: N/A System: Residual Heat Removal System Title of Change: Defeat Interlock of 2RHS"P1B Associated with Valves 2RHS" MOV112 and 2RHS" MOV113 Description of Change:
This change defeated the interlock that prevents pump 2RHS"P1B from starting unless both valves 2RHS"MOV112 and 2RHS"MOV113 are in the full open position. This change was done to facilitate dynamic testing of valve 2RHS" MOV40B. This is a temporary bypassing of the pump protection while the testing is being done.
Safety Evaluation Summary:
In order to perform the dynamic testing of 2RHS"MOV40B, a suction path for pump 2RHS"P1B is provided by opening 2RHS"MOV113. Since Division I dc power for 2RHS"MOV113 is unavailable, the valve control logic will maintain the pump in a trip condition and prevent the pump from starting. The interlock associated with 2RHS"MOV113 to prevent starting pump 2RHS"P1B without 2RHS"MOV113 full open must be defeated. This will be accomplished by lifting leads'in the valve control circuit. The similar interlock for valve 2RHS"MOV112 will also be bypassed due to the physical arrangement of the wiring. This will allow the pump to be started without position indication for 2RHS"MOV113. The function of the defeated interlocks will be provided by procedural control. This temporary alteration will be removed upon the completion of the 2RHS"MOV40B dynamic test procedure. All other design functions remain unchanged. The SFC B system will be available for heat removal and is unaffected by this change.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 102 of 136 Safety Evaluation No.: 93-105 Implementation Document No.: N2-OP-38 USAR Affected Pages: N/A System: SFC Title of Change: Natural Circulation of Reactor Vessel During Refueling Phase of RFO3 Description of Change:
Natural convective cooling of the reactor vessel is evaluated as an alternate method of compliance with Technical Specification 3/4.9.11.1 in case the primary cooling method (RHR loop) is lost. To increase the effectiveness of the natural circulation, it was necessary to change the SFC flow arrangement (cold water is routed to both reactor cavity spargers).
In addition, to comply with the Technical Specification requirement of temperature monitoring, a method is recommended to monitor the reactor coolant temperature.
The reactor coolant temperature will be monitored at two locations: (1) core outlet temperature (hot plume) within -5'f the top of core, and (2) near the vessel wall in downcomer region, near the jet pump inlet area. The arithmetic average of these two temperatures is the average reactor coolant temperature.
Safety Evaluation Summary:
The above changes have been evaluated for their impact on the fuel assembly cooling and local boiling within the spent fuel pool. The new flow arrangements will continue to provide adequate cooling to fuel assemblies in the spent fuel pool.
The current margin (-35 F) between the local saturation temperature and peak clad temperature will not be significantly affected.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 103 of 136 Safety Evaluation No.: 93-106 Rev. 1 Implementation Document No.: NLI NCR-02-37, 9-22-1992 USAR Affected Pages: 8.3-58 System: BYS Title of Change: Reduction in One IVlinute Discharge Rate for Division I and Division II Emergency Dc Batteries Description of Change:
One minute discharge rate of Division I and Division II emergency dc batteries has been reduced by the vendor due to more accurate battery test. This datum is a part of the dc system description in the USAR. The USAR has been revised to reflect this change.
One minute discharge rate also affects the battery discharge curves. These curves describe the performance of the batteries. The impact of the new one minute discharge rate and new battery discharge curves has been analyzed.
Safety Evaluation Summary:
The analysis perfo'rmed revealed that new one minute discharge rate and new battery discharge curves do not affect the required size of Division I and Division II emergency dc batteries. The batteries still can support adequately two-hours emergency discharge cycle and station blackout. At any point of postulated emergency discharge the voltage of the batteries does not drop below the limits identified in the USAR. The change has no impact on dc safety-related cables sizing. Short-circuit current of the battery is reduced; therefore, the dc short-circuit current margin is increased. This analysis is applied for existing batteries and for replacement cells.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 104 of 136 Safety Evaluation No.: 93-108 implementation Document No.: Procedures N2-OSP-EGS-R003 5 R004 USAR Affected Pages: N/A, System: . Spent Fuel Pool Cooling (SFC)
Title of Change: Overriding Operation of 2SFC" HV6A or 6B Description of Change:
Operations requested the capability of overriding the closure of 2SFC"HV6A or 6B during Operating Conditions 4 or 5 when conducting Procedure N2-OSP-EGS-R003 and R004.
These valves are the suction cross-connects between the respective skimmer surge tanks in Division I and II of the SFC system. As designed, the valves may either be operated manually to open or close or will close automatically upon a loss of power.
The design bases of SFC permits either the system to be operated cross-connected or split, with the division cross-connect isolation valves closed, when either one or both pumps are in operation. N2-OSP-EGS-R003 and R004 tests are conducted on the SFC division which does not have its pump in operation. When power is lost to that division its associated suction cross-connect valve will go closed. Prior operating experience has indicated that the running pump may trip on low suction pressure with the resulting condition that all SFC is lost.
Safety Evaluation Summary:
These procedure changes have no impact on the safe operation or capability to keep the plant in the safe shutdown condition or preclude removing the decay heat from the spent fuel pool.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 105 of 136 Safety Evaluation No.: 93-109 Rev. 0 L 1 Implementation Document No.: Simple Design Change SC2-0176-93 USAR Affected Pages: N/A System: High Pressure Core Spray Title of Change: Replacement of HPCS Valve Motor Contractors Description of Change:
This change installed new starter contactors into the control circuit of valves 2CSH"MOV107, 2CSH"MOV110, 2CSH"MOV111 and 2CSH"MOV112.
Safety Evaluation Summary:
The new starter contactors to be installed "pull in" at a lower voltage and require less "hold in" current. This will ensure that even under degraded voltage conditions the control circuit will perform its intended function.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 106 of 136 Safety Evaluation No.: 93-111 Implementation Document No.: N/A USAR Affected Pages: Figures 1.2-1, 2.4-1, 9A.3-1 System: N/A Title of Change: Construction of the Hazardous Material Storage Building Description of Change:
The hazardous material storage building was constructed to the south of the main warehouse and west of the bottled gas storage building, outside the protected area.
The storage facility is a single-story, nonsafety-related structure having a slab on grade and provides a total area of approximately 6,000 square feet. This building is designed to store hazardous material currently stored in the temporary construction building (warehouse C annex) and meets all federal/local code and environmental requirements.
Safety Evaluation Summary:
Based on the evaluation performed, it is concluded that the construction of the hazardous material storage building does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 107 of 136 Safety Evaluation No.: 93-112 Rev. 0 &. 1 Implementation Document No.: N/A USAR Affected Pages: N/A System: Various Title of Change: Revision 5 of the NMP2 Emergency Operating Procedures Description of Change:
This revision of the EOPs changed some operating limits as a result of the new GE11 fuel. The limits which have been revised are the following: minimum alternate RPV flooding pressure, minimum number of SRVs required for emergency depressurization, minimum RPV flooding pressure, minimum core flooding interval, maximum core uncovery time limit, minimum steam cooling, minimum zero injection RPV water level, heat capacity temperature limit, heat capacity level limit and the pressure suppression pressure.
Safety Evaluation Summary:
Although some changes have been made to the EOPs by Revision 5, it was verified that:
the operator actions prescribed in this new revision are in accordance with the BWROG EPGs, and when applied to licensing basis accidents and transients, the EOPs will not increase the consequences of these events as depicted in the USAR.
None of these changes has altered the philosophy, logic, or validity of the NMP2 EOPs.
Based on the evaluation performed, it is concluded that Revision 5 of the EOPs does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 108 of 136 Safety Evaluation No.: 93-114 Implementation Document No.: NIP-DES-04 USAR Affected Pages: Table 8.3-16 System: Reactor Building Cranes (MHR)
Title of Change: Revision to the NMP2 USAR Table 8.3-16 and Attachment (3) of NIP-DES-04, Primary Containment Electrical Penetrations and Penetration Conductors which are not Required During Reactor Operation and are Protected by De-energization Description of Change:
This change updated USAR Table 8.3-16 and Attachment (3) of NIP-DES-04 to reflect as-installed plant conditions for those circuits associated with electrical penetrations and electrical penetration conductors which were required to be de-energized during reactor operation. This change removes circuits for 2MHR-CRN3, 4, 65, 67 and 66, which have primary and backup circuit overcurrent protection at 2NHS-MCC005 cubicle 7B, 7C, 7D, 7E and 7F, respectively. New circuit for 2RDS-PLAT1 is added to the table which was inadvertently left in the list of lists for de-energized circuits.
Safety Evaluation Summary:
This change amends NMP2 USAR Table 8.3-16 and Attachment (3) of NIP-DES-04 to include a circuit for 2RDS-PLAT1, which is required to be de-energized for electrical penetration and penetration protection during reactor operation, and removing those circuits which are no longer required to be de-energized due to operability of primary and backup circuit overcurrent interrupting devices at 2NHS-MCC005 cubicle 7B, 7C, 7D, 7E and 7F for reactor building cranes 2MHR-CRN3, 4, 65, 67 5 66, respectively. The 2SCA-PNL406 breaker ¹9 provides power at the platform (2RDS-PLAT1) for lighting purposes only. The platform 2RDS-PLAT1 is used to rotate and position CRD handling equipment during refueling outage only. This circuit shall be included in USAR Table 8.3-16 and Attachment (3) of NIP-DES-04. The outboard cables (2RCSANC528 6 2RCSBNC538) from 2SCI-PNLC104 shall be spared in place.
Safety Evaluation Summary Report Page 109 of 136 Safety Evaluation No.: 93-114 (cont'd.)
Safety Evaluation Summary: (cont'd.)
There are no Technical Specification changes required to Sections 3/4.8A.1 or 3/4.8.4.2 as a result of these documentation only changes.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 110 of 136 Safety Evaluation No.: 93-115 Implementation Document No.: DER 2-93-2511 USAR Affected Pages: Figures 5.4-13a, 5.4-13b System: Residual Heat Removal (RHS)
Title of Change: Orientation of 2RHS" MOV40A/B Description of Change:
This change accepted as is the installation of outboard isolation globe valves 2RHS"MOV40A/B. These valves were throttled to control the cooldown rate in the shutdown cooling mode. These valves were installed such that LOCA and reactor vessel pressure could be applied to the above seat side of the globe valve should the inboard isolation check valves 2RHS"AOV39A/B fail open, thus creating a potential leak path through the valve packing.
Safety Evaluation Summary:
Though the installed orientation of 2RHS" MOV40A/B enhances the throttling capabilities of the valve, the orientation also creates a potential leak path through the valve packing. However, the design of the valve packing arrangement is based on pressures that exceed LOCA pressure of 39.75 psig and peak reactor vessel bottom pressure of 1279 psig. In addition, valves 2RHS"MOV40A/B are local leak rate tested to at least 39.75 psig and hydro pressure tested every 18 months.
Therefore, there is a high degree of assurance that these valves will maintain leak-tight integrity.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 111 of 136 Safety Evaluation No.: 93-116 Implementation Document No.: DER 2-93-2511 USAR Affected Pages: Figure 5.4-13c System: Residual Heat Removal (RHS)
Title of Change: Orientation of 2RHS" MOV33B Description of Change:
This change accepted as is the installation of containment isolation globe valve 2RHS"MOV33B. This valve is throttled to control suppression chamber spray flow. This valve was installed such that peak suppression chamber could be applied to the above seat side of the globe valve, thus creating a potential leak path through the valve packing.
Safety Evaluation Summary:
Though the installed orientation of 2RHS" MOV33B enhances the throttling capabilities of the valve, the orientation also creates a potential leak path through the valve packing. However, the design of the valve packing arrangement is based on pressure that exceeds the calculated peak suppression chamber pressure of 33.98 psig. In addition, valve 2RHS"MOV33B is local leak rate tested at 40 - 42 psig every 18 months. Therefore, there is a high degree of assurance that this valve will maintain leak-tight integrity.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 112 of 136 Safety Evaluation No.: 93-118 Rev. 0 5. 1 Implementation Document No.: Temporary Mod.93-062 USAR Affected Pages: N/A System: Hot Water Heating 5. Glycol (HVH)
Title of Change: Temporary Hot Water Heating Plant Description of Change:
This safety evaluation covered the installation and operation of a temporary hot water heating plant, which is used to provide an alternate supply of hot water to the reactor building glycol heat exchanger, 2HVG-E2. The normal supply of hot water from the HVH hot water heat exchangers may not be available during normal plant operation and a source of hot water is needed for glycol heating. The temporary hot water heating unit is located at least 50 feet from any plant structure. Demineralized water for system filling is furnished via drain valve 2CCP-V915. Electric power for the unit is provided from the existing construction power loop.
Safety Evaluation Summary:
As discussed in USAR Section 9.4.11, the plant glycol heating system functions in conjunction with the plant hot water heating system (discussed in USAR Section 9.4.12) to heat outdoor makeup air used for ventilation. As described in USAR Section 9 4.11.3 for the glycol system and in USAR Section 9.4.12.5 for the plant hot water heating system, the failure or malfunction of either or both system(s) will not compromise any safety-related system or component or prevent safe reactor shutdown.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 113 of 136 Safety Evaluation No.: 93-120 Implementation Document No.: N2-ISP-LRT-R5058A USAR Affected Pages: Table 6.2-65 Sh 1 System: Reactor Core Isolation Cooling (ICS)
Title of Change: Appendix J Testing of 2ICS"MOV126 Description of Change:
This change allowed reverse testing of containment isolation 2ICS"MOV126. This change resulted in a change to testing procedure N2-ISP-LRT-R5058A which controls the testing of this valve.
Safety Evaluation Summary:
This change, reverse testing of 2ICS"MOV126, 'is a more conservative test because both the LOCA and non-LOCA seat will be subjected to the LLRT pressure. This change will have no impact on the safe operation 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 114 of 136 Safety Evaluation No.: 93-121 Implementation Document No.: EDC 2F00874, DER 2-93-0501 USAR Affected Pages: Figure 11.2-1J System: LWS Title of Change: Liquid Waste Management System As-Built Condition Description of Change:
This change provided as-built conditions on the affected drawings to show removal of flushing lines (2-LWS-002-464-4 and 2-LWS-002-625-4) and valves (2LWS-V194 and 2LWS-V203).
Safety Evaluation Summary:
Due to ESDCR C94489B, and in accordance with ECN LWS-631, pressure switches 2LWS-PS42A and 2LWS-PS42B were deleted from the system. The flushing lines (2-LWS-002-464-4 and 2-LWS-002-625-4) and valves (2LWS-V194 and 2LWS-V203) were the flushing connections for the pressure switches and have been removed. To maintain plant configuration control, said flushing lines and valves will be removed from affected drawings to show as-built condition.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 115 of 136 Safety Evaluation No.: 93-122 Implementation Document No.: Temporary Mod.93-067 USAR Affected Pages: N/A System: Main Steam System Title of Change: Leak Repair 2MSS-PV29V Description of Change:
This temporary change was related to the main steam system (MSS). The reheating steam low load isolation valve to moisture separator reheater 2MSS-E1B, valve 2MSS-PV29B, was experiencing a steam leak through the valve packing area. The change involved the temporary injection of leak repair nuclear grade sealant type 2X to isolate the leak until permanent repairs were made.
Safety Evaluation Summary:
An engineering review of the requested change has been performed. This review, which included the effects of this temporary change on the system's operability, reliability, maintainability, structural integrity, and system interactions, has found that the injection of the sealant and possible seizure of valve 2MSS-PV29B in the open position will not cause any adverse safety or operability issues.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 116 of 136 Safety Evaluation No.: 93-126 Rev. 0 & 1 Implementation Document No.: N/A USAR Affected Pages: 12.3-8 thru 12.3-11, 12.3-19, 12.3-22, 12.5-1 thru 12.5-4, 12.5-6, 12.5-10 thru 12.5-14, 12.5-16, 12.5-17; Tables 1.8-2 Sh 3, 4, 1.9-1 Sh 49, 50 System: N/A Title of Change: 10CFR20 Revision Description of Change:
To incorporate the changes necessary to implement the revised 10CFR20, "Standards for Protection Against Radiation."
Safety Evaluation Summary:
Revision 1 provides a basis for changing the TLD processing frequency that was incorporated in the Radiation Protection Program based on, but not specifically addressed in, Revision 0. Also, the Technical Review Committee's comments of May 25, 1994, concerning the use of 0-200 mR pocket dosimeters and clarification of radiation zone terminology are included in Revision 1.
The changes are fundamentally administrative in nature. The changes necessary to support the revised Part 20 are consistent with the philosophy of maintaining dose As Low As is Reasonably Achievable (ALARA).
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 117 of 136 Safety Evaluation No.: 93-127 Rev. 1 Implementation Document No.: GAP-POL-01, Rev. 05; NSAS-POL-01, Rev. 00; QAP-POL-1.01, Rev. 04 USAR Affected Pages: 9A.3-1, 9A.3-1a, 9A.3-2, Section 13.1, 13.2-1, Appendix B System: N/A Title of Change: Nuclear SBU Organizational Structure and Responsibilities - Revised Procedures GAP-POL-01 5 QAP-POL-1.01, and New Procedure NSAS-POL-01 Description of Change:
The positions of General Manager Site Support and General IVlanager Safety Assessment, Licensing, and Training were combined under the single position of Vice President Nuclear Safety Assessment and Support. Senior management responsibilities for the Vice President Nuclear Safety Assessment and Support include the present "Safety Assessment" functional areas of Quality Assurance (QA), Licensing, and Training; and the "Support" functions previously implemented within the Site Support organization. The Manager Quality Assurance reports directly to the Executive Vice President Nuclear for all QA activities within the Nuclear Safety Assessment and Support organization (to ensure sufficient authority and independence for effectively implementing QA responsibilities within the Nuclear Safety Assessment and Support organization).
The Unit 1 Operating Organization (IVlaintenance, Technical Support, and Work Control/Outage Branches only) was revised to include the following Branch Manager direct reports:
~ Maintenance - General Supervisor l&C Maintenance, General Supervisor Mechanical/Electrical Maintenance (combined), Supervisor IVlaintenance Procedures, Lead Maintenance Support, and Program Director 89-10 Implementation.
~ Work Control/Outage - General Supervisor Maintenance Planning, Supervisor Outage Management, and Supervisor Maintenance Planning Programs.
Safety Evaluation Summary Report Page 118 of 136 Safety Evaluation No.: 93-127 Rev. 1 (cont'd.)
Description of Change: (cont'd.)
~ Technical Support - Lead System Engineers (2) and Administrative Support Coordinator(s) (for SORC, NPRDS, Technical Review, and Modification activities).
Safety Evaluation Summary:
The proposed organization continues to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Units 1 h 2.
Clear management control, effective lines of authority, and communication between the organizational units irivolved in the management, operation, and technical support of Nine Mile Point Units 1 L 2 are maintained. The organizational changes alter the reporting structure of existing positions but do not affect the performance of functions or responsibilities.
Lines of authority, responsibility and communication for "onsite" and "offsite" organizational elements which function under the cognizance of the QA Program are established in the form of revised organizational charts. Functional descriptions of the Nuclear Safety Assessment and Support Organization and the revised Unit 1 Operating Organization, and job descriptions, relationships, and responsibilities for key personnel positions are documented in Procedure GAP-POL-01, NSAS-POL-01, and QAP-POL-1.01.
Based on this evaluation, the revised organizational structure of the Nuclear SBU continues to satisfy the acceptance criteria of SRP 9.5.1 (BTP CMEB 9.5-1), SRP 13.1.1, SRP 13.1.2-13.1.3, SRP 13.6, SRP 17.2, and Unit 1 and 2 Technical Specification 6.2.1.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 119 of 136 Safety Evaluation No.: 93-128 Implementation Document No.: DER 2-93-2396 USAR Affected Pages: 8.3-62, 8.3-70 System: BYS Title of Change: IEEE Standard Issue Change for Class 1E Battery Testing Description of Change:
During the replacement of Division I dc battery it was discovered that there was a discrepancy between the USAR and Technical Specifications. Technical Specification Bases, pg. B3/4 8-2 states that the surveillance requirements for demonstrating the operability of the batteries are in accordance with the recommendations of Reg. Guide 1.129 and IEEE-Std-450-1980. However, USAR Section 8.3.2 states that the regular inspection and maintenance of the batteries is performed in accordance with Reg. Guide 1.129 (IEEE-Std-450-1975). Engineering analyzed this discrepancy and came to the conclusion that the USAR was in error and needed to be corrected to agree with Technical Specifications.
Safety Evaluation Summary:
Engineering performed a comparison between 1975 and 1980 issue of IEEE-Std-450 to ensure that this change is not adversely affecting the Class 1E batteries. This comparison revealed that in the Maintenance and Corrective Actions sections, the 1980 issue of the standard is equivalent to the 1975 issue or more restrictive.
In the Battery Capacity Test section, there is a slight difference between these two issues. The nature of this difference depends on the battery test temperature.
However, the value of this difference is insignificant to compare with the excess of the capacity of the dc batteries. Based on the analysis performed, the conclusion is made that this change does not compromise the ability of the batteries to perform the safety-related function as designed and as described in the USAR.
C%
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 120 of 136 Safety Evaluation No.: 94-002 Implementation Document No.: Temporary Mod.94-003, N2-OP-35 USAR Affected Pages: N/A System: Reactor Core Isolation Cooling (ICS)
Title of Change: Procedure Change Evaluation, N2-OP-35 Description of Change:
This change was related to the reactor core isolation cooling system (ICS) as described in USAR Section 5 4. The change throttled open 2ICS-LV132 as required to decrease its cycling (opening/closing) frequency to an acceptable level.
Previously, to control the level in the ICS steam supply drain line drain pot, this valve was cycled at approximately 5-minute intervals causing annunciator 601302 to alarm. This change has been procedurally implemented and is controlled via Temporary Modification 94-003 to minimize annunciation until the underlying cause of the problem is evaluated and resolved (reference DER 2-93-2805).
Safety Evaluation Summary:
An engineering review of the effects of this change on the ICS and other interfacing systems has been performed. Nuclear safety will not be compromised as the change will continue to provide an adequate level control system for the steam supply drain line drain pot.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 121 of 136 Safety Evaluation No.: 94-003 Implementation Document No.: Temporary Mod.94-008 USAR Affected Pages: N/A System: Reactor Protection System (RPS)
Title of Change: Group Four Scram Pilot Valve Power Source Description of Change:
Inadvertent cable terminations at hydraulic control unit (HCU) 2RDS-L704 caused reactor protection system trip channels A1 and A2 to control the "8" pilot scram valve solenoid instead of the intended "A" solenoid. Accordingly, trip channels B1 and B2 control the "A" solenoid instead of the intended "B" coil. The wiring error was located at the first HCU of a 24 HCU daisy chain, thereby affecting the entire chain. All group four HCUs on side two of the reactor were being affected by. the described condition. The change documents the wiring error as a temporary modification since restoration will not be performed until the next scheduled plant shutdown.
Safety Evaluation Summary:
The safety-related function of the RPS is to scram the reactor when specific predetermined variables are exceeded. The dual trip protective system employs a one-out-of-two taken twice logic to control fail-safe solenoids. The function of the RPS is not adversely affected by the reversed wiring at the group four HCUs. Two separate power sources controlled by two separate trip systems are still provided for each control rod as delineated by system eliminators. This deviation introduces a configuration control concern but has no functional impact on the RPS or the affected control rod drives.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 122 of 136 Safety Evaluation No.: 94-004 Implementation Document No.: Simple Design Change SC2-0170-91 USAR Affected Pages: 9.3-6 System: Service Air System (SAS)
Title of Change: 1/2 Inch Hose Coupling Replacement Description of Change:
This change made permanent the 1" hose couplings installed at stations 2SAS-V264 and 2SAS-V265 per Temporary Modification 90-069. The increase in hose coupling, valve and associated components from 1/2" to 1" provides increased airflow to equipment as needed.
Safety Evaluation Summary:
An engineering review of the effects of this change on the SAS and other interfacing systems has been performed. Nuclear safety will not be compromised as the change will continue to provide an adequate supply of service air to equipment on demand. This change does not cause any equipment important to safety to be dependent on SAS to perform its safety function.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 123 of 136 Safety Evaluation No.: 94-005 Implementation Document No.: Temporary Mod.94-018, N2-OP-35 USAR Affected Pages: N/A System: Reactor Core Isolation Cooling (ICS)
Title of Change: Procedure Change Evaluation, N2-OP-35 Throttle Open 2ICS-LV132 Description of Change:
This temporary modification throttled open 2ICS-LV132 as required to decrease its cycling (opening/closing) frequency to an acceptable level. Currently, to control the level in the ICS steam supply drain line drain pot, this valve is cycling at frequent intervals causing annunciator 601302 to alarm. This change has been procedurally implemented and is controlled via Temporary IVlodification 94-018 to minimize annunciation until the underlying cause of the problem is evaluated and resolved.
Safety Evaluation Summary:
An engineering review of the effects of this change on the ICS and other interfacing systems has been performed. Nuclear safety will not be compromised as the change will continue to provide an adequate level control system for the steam supply drain line drain pot.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 124 of 136 Safety Evaluation No.: 94-007 Implementation Document No.: Calculation MS-2162-OOC USAR Affected Pages: Table 3.9A-10 Sh 1, 4 System: Service Water System Title of Change: Update USAR Table 3.9A-10 "Summary of Seismic Stress Analysis Results" Description of Change:
Calculation disposition MS-2162-OOC was performed to reflect the latest results of the seismic stress analysis of the service water pumps (2SWP" P1A, B, C, D, E 5 F) at NMP2.
Safety Evaluation Summary:
The calculation was reviewed and approved to provide the technical justification of the revised values in USAR Table 3.9A-10. The calculation disposition shows that the results of the latest seismic stress analysis are all within their corresponding allowables. This change does not result in any physical change to the equipment nor will it affect the function, operability and structural integrity of the subject pumps. Therefore, this change does not have an impact on the 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 125 of 136 Safety Evaluation No.: 94-008 Implementation Document No.: DER 2-93-1896 USAR Affected Pages: Table 6.2-56 Sh 19, 20 System: Residual Heat Removal (RHS)
Title of Change: Revise Type C Testing of 2RHS"MOV26A, 26B, 27A and 27B from Air to Water Description of Change:
During an independent audit of the NMP2 10CFR50 Appendix J program it was noted that USAR Table 6.2-56 indicated that valves 2RHS"MOV26A, 26B, 27A and 27B were Type C tested. These valves were on lines whose discharges were below suppression pool minimum water level and, therefore, were not exposed to primary containment atmosphere.
These valves were previously Type C tested with air in accordance with Appendix J,Section III.C.2.a. In answer to the audit item, testing of these valves has been changed from a Type C test using air to a hydrostatic test.,
Safety Evaluation Summary:
Valves 2RHS"MOV26A, 26B, 27A and 27B tie into lines that penetrate the primary containment and discharge into the suppression pool at 193'-2 3/8". The lowest minimum suppression pool water level is 199'-6". These valves do not become exposed to the primary containment atmosphere during a DBA-LOCA since their discharge is below minimum suppression pool water level and, therefore, do not.represent potential containment atmospheric leakage paths. In addition, minimum suppression pool water level is maintained per Technical Specification such that the water seal is assured.
These valves will be hydrostatically tested at a pressure of 1.1 Pa.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 126 of 136 Safety Evaluation No.: 94-010 Implementation Document No.: NSAS-POL-01, Rev. 01; POL, Rev. 5 USAR Affected Pages: Section 13.1 System: N/A Title of Change: Alter Organizational Structure and Responsibilities Within the Nuclear Strategic Business Unit - Revised Procedure NSAS-POL-01 and Nuclear Division Policy, llPOI ff Description of Change:
Administrative responsibility for the Fitness for Duty Program was transferred from the IVlanager Nuclear Security to the Director Human Resource Development. The position of the Supervisor Access Authorization/Fitness for Duty reporting to the Manager Nuclear Security was eliminated. Responsibility for administering the .
Access Authorization Program was assumed by the Supervisor Nuclear Security Support, who also assumed the responsibilities of the Supervisor Nuclear Security Administration leading to the elimination of that position.
Safety Evaluation Summary:
The new organizational structure will continue to provide for the integrated management of activities that support operation of Nine IVlile Point Units 1 and 2.
Clear management control and effective lines of authority are continued for Nuclear Security within the Nuclear Safety Assessment and Support organization and for Fitness for Duty within Human Resource Development. Although organizational changes alter the reporting structure as it applies to management of the Fitness for Duty Program and supervision of the Unescorted Access Authorization Program and Security administrative services, the actual functions within these programs will not be affected.
Lines of authority, responsibility and communication relating to Human Resource Development are currently established in organizational charts in the Unit 1 and Unit 2 UFSARs. Lines of authority, responsibility, and communication relating to Nuclear Security are also shown in the organizational charts in the UFSARs, as well as in the Physical Security Plan. Revised job descriptions and responsibilities for
Safety Evaluation Summary Report Page 127 of 136 Safety Evaluation No.: 94-010 (cont'd.)
Safety Evaluation Summary: (cont'd.)
the Manager Nuclear Security and the Director Human Resource Development will be documented in NSAS-POL-01 and in POL, "Nuclear Division Policy,"
respectively.
The revised organizational structure of the Nuclear SBU meets the acceptance criteria of SRP Sections 13.1, 13.1.1, 13.6, and 17.2, as well as Unit 1 and Unit 2 Technical Specification 6.2.1.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 128 of 136 Safety Evaluation No.: 94-011 Implementation Document No.: Simple Design Change SC2-0146-93 USAR Affected Pages: Figure 8.2-7 System: SPG, YXC, CEC Title of Change: Remove Control Room Recorder RFM-2SPGN02 (Frequency) and RV-2YXCN10 (Voltage) from Panel 2CEC"PNL852 Description of Change:
Frequency recorder RFM-2SPGN02 and voltage recorder RV-2YXCN10 were inoperable and repair parts were no longer available from the manufacturer. The recorders monitored the 345/25-kV Generator Scriba Station Line ¹23 potential circuit. The information is available from other existing instrumentation on the panel and as output from GETARS. Therefore, the recorders were removed.
Safety Evaluation Summary:
The design change has removed inoperable and unnecessary equipment from the control room panel to enhance operator visual focus on other operating instrumentation. This change will have no impact on the 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 129 of 136 Safety Evaluation No.: 94-014 Implementation Document No.: NSAS-POL-01, Rev. 01; NIP-FPP-01, Rev. 03 USAR Affected Pages: 9A.3-2 thru 9A.3-4, 9A.3-31, 13.2-21; Table 13.1-1; Figure 13.1-5 System: N/A Title of Change: Fire Protection Organizational Structure-Revised Procedures NSAS-POL-01 and NlP-FP P-01 Description of Change:
The Fire Protection organization was restructured from unit specific to a site organization reporting to a site Supervisor Fire Protection. The "Site" Supervisor Fire Protection reports to the Manager Technical Services, and the Manager Technical Services continues to maintain overall responsibility for site implementation of the Fire Protection Program.
The Fire Brigade is comprised of at least three members from the Fire Protection staff and up to two members from other site organizations, thereby satisfying the minimum site Brigade complement of five. Brigade members do not include the SSS or other members of the minimum shift crew necessary for safe shutdown of the unit, or any other personnel required for other essential functions during a fire emergency. All members of the Fire Brigade continue to be trained/qualified per existing Fire Brigade Training Program requirements.
Safety Evaluation Summary:
The proposed organizational changes alter the reporting structure of existing Fire Protection staff positions and the composition of the Fire Brigade, but do not affect the performance of Fire Protection staff functions or responsibilities. The "site" organization continues to provide for integrated. management of fire protection activities to support the operation and maintenance of Nine Mile Point Units 1 and 2, and to achieve and maintain safe shutdown in the event of a fire.
Clear management control and effective lines of authority and communication between the organizational units involved in the management, operation, and technical support for the operation of Nine Mile Point Units 1 and 2 are maintained, and the response capability of the Fire Brigade is not affected by the reorganization.
Safety Evaluation Summary Report Page 130 of 136 Safety Evaluation No.: 94-014 (cont'd.)
Safety Evaluation Summary: (cont'd.)
Functional descriptions of the Fire Protection organization, and job descriptions, relationships, and responsibilities for key personnel positions responsible for implementation of the Fire Protection Program, are documented in Procedures NSAS-POL-01 and NIP-FPP-01. Based on this evaluation, the revised structure of the Fire Protection organization continues to satisfy the acceptance criteria of SRP 9.5.1 (BTP CMEB 9.5-1), SRP 13.1.1, and Unit 1 Technical Specification 6.2.1, and does not constitute an unreviewed safety question.
Safety Evaluation Summary Report Page 131 of 136 Safety Evaluation No.: 94-015 Implementation Document No.: Procedures N2-OP-29, N2-OP-101A, N2-OP-101 C, N2-OP-101D USAR Affected Pages: N/A System: Reactor Recirculation System (RRS)
Tile of Change: Revise Procedures N2-OP-29, N2-OP-101A, N2-OP-101C, and N2-OP-101D to Add Owners'roup Stability Guidance When Operating Near the Stability Region Description of Change:
This safety evaluation evaluated the impact of adding the revised BWR Owners'roup guidance concerning operation near the stability region. This change increased the size of the stability exit region. The change also added a region of heightened awareness between the 65% and 70% rod line and less than 45% core flow, and the area greater than the 65% rod line and between 45% and 50% core flow.
Safety Evaluation Summary:
Current practice is to upshift the recirculation pumps at the lowest rod line possible to maximize the margin to the stability regions. These changes will proceduralize this operating philosophy and ensure that recirculation pump upshift and downshift during a normal startup or shutdown will occur at less than the 65% rod line. ~
Previously the guidance would allow recirculation pump upshift up to the 80% rod line. Any entry into the heightened awareness zone will require continuous monitoring for thermal hydraulic oscillations. These changes are conservative measures which help to provide a greater margin to the stability exclusion region.
The operating procedures which are impacted by these changes are as follows:
N2-OP-29 Reactor Recirculation System N2-OP-101A Plant Startup N2-OP-101 C Plant Shutdown N2-OP-1 01 D Power Changes
Safety Evaluation Summary Report Page 132 of 136 Safety Evaluation No.: 94-015 (cont'd.)
Safety Evaluation Summary: (cont'd.)
Based upon this evaluation, it is concluded that Nine Mile Point Unit 2 can be safely operated in accordance with these procedure changes. The changes evaluated in this safety evaluation will serve to increase the operating margin to the region of the power/flow map in which reactor stability is a concern.
Based on the evaluation performed, it is concluded that these changes do not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 133 of 136 Safety Evaluation No.: 94-016 Implementation Document No.: GAP-POL-01, Rev. 06; NSAS-POL-01, Rev. 01; GAP-OPS-01, Rev. 03; NIP-TQS-01, Rev. 04 USAR Affected Pages: 12.1-7, 12.1-8, 12.5-12, Section 13.1, 13.2-22 System: N/A Title of Change: Nuclear SBU Organizational Structure and Responsibilities - Revised Procedures GAP-POL-01, NSAS-POL-01, GAP-OPS-01, and NIP-TQS-01 Description of Change:
This change analyzed the impact of proposed rightsizing and organizational changes within the Nuclear Generation and Nuclear Safety Assessment and Support organizations. The changes reflect an overall reduction in site staffing levels and a reduction in the management layers of certain groups within the Operations, Maintenance, Work Control/Outage, Radiation Protection, and Technical Support Branches of Nuclear Generation; and the Training Branch, Occupational Safety &. Health, Construction Services, and Office Administration/
Facilities groups of Nuclear Safety Assessment and Support.
Responsibilities for certain functions were consolidated within branches or transferred between branches, and several General Supervisor and Supervisor positions were abolished resulting in an increase in the number of direct reports to applicable Branch Managers.
After rightsizing, the total Nine Mile Point site staff is approximately 918 people.
This staff level is consistent with NUREG-1047, Section 13.1.2.1, which identifies the anticipated Nine Mile Point site staff of about 900 people as being within the range normally expected for a two-unit site.
Safety Evaluation Summary:
The rightsized Nuclear Generation and Nuclear Safety Assessment and Support organizations continue to provide for the integrated management of activities to support the operation and maintenance of Nine Mile Point Units 1 and 2. Clear
Safety Evaluation Summary Report Page 134 of 136 Safety Evaluation No.: 94-016 (cont'd.)
Safety Evaluation Summary: (cont'd.)
management control and effective lines of authority and communication are maintained. Functional descriptions of the Nuclear Generation and Nuclear Safety Assessment and Support organizations, and job descriptions, relationships, and responsibilities for key personnel positions are documented in Procedures GAP-POL-01, NSAS-POL-01, GAP-OPS-01, and NIP-TQS-01.
Based on this evaluation, the revised organizational structures of the Nuclear Generation and Nuclear Safety Assessment and Support organizations continue to satisfy acceptance criteria from SRP 13.1.1, SRP 13.1.2-13.1.3, Unit 1 and 2 Technical Specification 6.2.1, ANSI N18.1-1971 (Unit 1), and ANSI/ANS 3.1-1978 (Unit 2); and site staff total is consistent with the staffing range expected for two-unit sites (per NUREG-1047). The organizational changes are in compliance with NRC standards and do not constitute an unreviewed safety question.
Safety Evaluation Summary Report Page 135 of 136 Safety Evaluation No.: 94-041 Implementation Document No.: DER 2-94-0624 USAR Affected Pages: Table 8.3-16 System: RSC Title of Change: Removal of Maintenance and Calibration Jacks from the List of De-energized Circuits USAR Table 8.3-16 Description of Change:
This change revised USAR Table 8.3-16 to remove maintenance and calibration (jacks 124, 128, 134 and 137) circuits that provided low energy dc signal to communication jacks in the primary containment.
Safety Evaluation Summary:
The maintenance and calibration circuits for jacks 124, 128, 134 and 137 carry low-energy dc signals on the order of milliamperes. The output current from the power supply at RCS-88 is limited so that it will deliver no more than 230 ma under shorted line conditions. The short-circuit current value is insignificant as compared to the current-carrying capability of the penetration assembly (2CES-Z38E). Moreover, the Technical Specification 3.8.4.1 requirements are applicable to ac circuits only. Therefore, maintenance and calibration circuits that carry low-energy dc signals for these jacks can be removed from the list of de-energized circuits in USAR Table 8.3-16. Procedures NIP-DES-04 and N2-OSP-LOG-D001 were revised accordingly.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.
Safety Evaluation Summary Report Page 136 of 136 Safety Evaluation No.: 94-042 Implementation Document No.: N/A USAR Affected Pages: 13.5-1, 13.5-3, 13.5-5; Tables 13.5-1 Sh 1, 13.5-5, 13.5-6 Sh 5, 6, 13.5-7 System: N/A Title of Change: Implementation of Operating Procedure Improvements Description of Change:
This change was made to incorporate a description of the scope and format of Special Operating Procedures within the Unit 2 procedure program structure.
Other changes to the USAR were to improve internal consistency and to more dearly describe the Unit 2 procedure program.
Safety Evaluation Summary:
The safety evaluation addresses the acceptability of using Special Operating Procedures to satisfy the requirement of the NMP2 Technical Specifications and licensing basis. The changes to the USAR simply clarify the method of implementing the guidance of ANSI/ANS 3.2-1982 and provide another procedural mechanism for addressing events that have a regulatory specified procedure requirement. Based on th'e evaluation performed, these changes meet the requirements of the Technical Specification, the licensing basis, and industry standards, and are in compliance with NRC standards.
Based on the evaluation performed, it is concluded that this change does not involve an unreviewed safety question.