ML112971251
| ML112971251 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/30/1988 |
| From: | Musolf D Northern States Power Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML112971252 | List: |
| References | |
| NUDOCS 8807190060 | |
| Download: ML112971251 (18) | |
Text
ACCELERATED DIlTIBUTION DEMONSTR 7ON SYSTEM REGULATC INFORMATION DISTRIBUTIOSTEM (RIDS)
ACCESSION NBR:8807190060 DOC.DATE: 88/06/30 NOTARIZED: NO DOCKET #
FACIL:50-263 Monticello Nuclear Generating Plant, Northern States 05000263 AUTH.NAME AUTHOR AFFILIATION MUSOLF,D.
Northern States Power Co.
RECIP.NAME RECIPIENT AFFILIATION Office of Nuclear Reactor Regulation, Director (Post 370411
SUBJECT:
Forwards Rev 6 to updated SAR for Monticello Nuclear Generating Plant & Rev 12 to "O rationa Q
Pn."
DISTRIBUTION CODE: A053D COPIES RECEIVED:LTR EN L SIZE:
TITLE: OR Submittal: Updated FSAR (50.71) and Amendments NOTES:
RECIPIENT ID CODE/NAME PD3-1 LA WRIGHT,R AEOD/DOA/IRB NUDOCS-ABSTRACT RGN3 EXTERNAL: ARM/DCB NRC PDR SAIC LINER,R COPIES LTTR ENCL 1
0 1
1 1
1 1
1 1
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME PD3-1 PD R
'TIA 7E REG FILE 01 LPDR NSIC COPIES LTTR ENCL 1
0 1
0 1
1 1
1 1
1 I
TOTAL NUMBER OF COPIES REQUIRED: LTTR R
INTERNAL:
D S
/
A 13 ENCL 10
Northern States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 Telephone (612) 330-5500 June 30, 1988 Submitted pursuant to 10 CFR 50.71(e)
Director Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Submittal of Revision No. 6 to the Updated Safety Analysis Report (USAR)
Pursuant to 10 CFR 50.71(e) we are submitting 13 copies of Revision No. 6 to the Updated Safety Analysis Report (USAR) for the Monticello Nuclear Generating Plant. This revision updates the information in the USAR for the period from January 1, 1987 through December 31, 1987.
Exhibit A contains a description and summary of the safety evaluation for changes, tests and experiments made under the provisions of 10 CFR 50.59 during this period.
Exhibit B contains the USAR page changes and instructions for entering the pages.
Included in Exhibit B is Revision 12 to the Northern States Power Company Operational Quality Assurance Plan incompliance with 10 CFR 50.54(a).
Changes in Revision 12 to the plan are described in Exhibit A (Item 49, page 15) of this letter.
David Musolf Manager - Nuclear Support Services c:
Regional Administrator-III, NRC NRR Project Manager, NRC Resident Inspector, NRC C Charnoff (w/o Exhibit B)
Attachments PB 07190060 880630 PD ADOCK 05000263 PNU
/ J*
VAF"qw
Exhibit A MONTICELLO NUCLEAR GENERATING PLANT ANNUAL REPORT OF CHANGES, TESTS AND EXPERIMENTS - DECEMBER 1987 The following sections include a brief description and a summary of the safety evaluation for those changes, tests and experiments which were carried out without prior NRC approval, pursuant to the requirements of 10CFR50.59(b).
To bring Monticello's RBCCW system into compliance with the type 'C' testing requirements of 10 CFR 50, Appendix J, a motor operated containment isolation valve, two in-line manual block valves, test and drain connections, and associated pipe supports were added to each of the RBCCW system drywell supply and return lines.
Summary of Safety Evaluation:
Existing system functions were not affected by the RBCCW modifications.
All equipment installed meets the original design requirements of the systems in which they were installed. The new containment isolation valves are operated in the same manner as the original motor operated containment isolation valve. Alternate means of cooling were provided while the RBCCW system was out of service. Provisions were made to contain any loss of RBCCW coolant during the work.
- 2. MOD 83-112, Addendum 3, Part B, Appendix J Modifications - Torus Spray System Description of Change:
To bring Monticello's Torus Spray lines into compliance with the type
'C' testing requirements of 10 CFR 50, Appendix J, an in-line manual block valve, test and drain connections, and associated pipe supports were added to each of the Torus Spray lines.
Summary of Safety Evaluation:
All equipment installed meets the original design requirements of the systems in which they were installed. The modifications to the torus spray lines were completed during a refueling outage, thus the lines were not required to be operational. The addition of the valves increase the system head loss.
This increase will not drop the flow rate through the torus spray lines to below the required design flow rate.
- 3. MOD 83-112, Addendum 3, Part C, Appendix J Modifications Recirculation System Description of Change:
To bring Monticello's Recirculation System into compliance with the type
'C' testing requirements of 10 CFR 50, Appendix J, a valved stainless 1
steel test tap assembly was added between the B loop recirculation header and inboard isolation valve CV-2790 in recirculation B loop sample line REW-32-1-3/4".
Summary of Safety Evaluation:
The addition of the test tap does not alter the existing flow path of the sample line, and will have negligible affect on pressure drop. The connection is intended to be used only for draining and testing when the plant is not operating. The addition of the test tap will not have a significant impact on the operating system, process or activity.
The Hydrogen Water Chemistry Crack Arrest Verification System (CAVS) sample water return line was rerouted from a location in the Reactor Water Cleanup (RWCU) pump suction line to a location upstream of the RWCU regenerative heat exchangers. The change allows high temperature water to return to a high temperature process pipe to avoid temperature cycling. The modification also installed oxygen analyzer cells in parallel with the hydrogen analyzers in the offgas system.
Summary of Safety Evaluation:
This modification was designed to the same criteria as the existing systems. The CAVS sample return line rerouting will not affect the operability of the RWCU system. The oxygen analyzer installation allows the monitoring of offgas oxygen concentrations without affecting the operability of the hydrogen analyzers.
The inner filter of all installed and spare Control Rod Drives (CRD's) on site has been modified such that the inner filter is attached to the top of the stop piston. This removes the inner filter from the flow path during scram so that a plugged filter will not affect scram insertion time.
Summary of Safety Evaluation:
The intent of the inner filter is to protect the stop piston seals and CRD internals.
This function is not affected by this modification.
- 6. MOD 85-004, Control Room Ventilation Improvements Description of Design Change:
The logic on the division II control room air conditioning compressor was modified so the it operates in four thermostatically controlled stages as the manufacturer recommends.
Prior to this modification all four stages of the division II compressor came on and went off at the same time.
The division I air conditioning compressor has always operated in four stages.
Annunciation was added on Panel C03, whenever the control switches for emergency service water pumps, P-111C and P-111D, are placed in the pull to lock position.
2
Summary of Safety Evaluation:
The air conditioner compressor logic change improved the cooling capability of the control room habitability system.
The addition of the pull to lock annunciation on the emergency service water pumps meets the general requirements that safety systems be alarmed whenever they are out of service.
- 7. MOD 85-013, Part B, Addendum 1, Reactor Vessel Level Instrumentation Modification Description of Design Change:
Added flanges to excess flow check valves on reactor vessel level instrumentation.
Summary of Safety Evaluation:
The flange installation meets requirements of the reactor vessel level instrumentation design criteria. Fabrication, installation and testing complied with ASME Code Section XI.
The addition of flanges allows accessibility to the excess flow check valves for maintenance without affecting the function or reliability of the reactor water level instrumentation system.
- 8. MOD 85-013, Part C, Reactor Water Level System RPS Power Supply Upgrade Description of Design Change:
Interchanged the Uninterruptible Power Source (UPS) power feeds for the Reactor Protection System (RPS) water level trip cabinets, Panels C304B and C304C, and brought in a second power source to all four RPS water level trip cabinets, C304A, C304B, C304C and C304D.
Summary of Safety Evaluation:
The interchanging of the UPS power feeds to Panels C304B and C304C prevents the failure of one division of UPS from affecting both channels of the RPS.
Adding a second power source to each RPS trip cabinet keeps a single power failure from putting in a half-scram.
The RPS water level trip cabinets power supply is now single failure proof and the possibility of inadvertent half-scrams has been reduced.
- 9. MOD 85-016, Part D, Appendix R Panel and Cabling Current Transformer Protector Description of Design Change:
Installed protective devices on the current transformer (CT) for the 4KV circuit breakers on #12 emergency diesel generator, load center
- 104, and the 1AR transformer.
Summary of Safety Evaluation:
The protective devices installed on the CT prevents voltage limits from being exceeded should a control/cable spreading room fire cause an open circuit. Prior to this modification, operators were required to install shorting jumpers on these circuits during a fire.
3
- 10. MOD 85-036, Torus Drain System Description of Design Change:
Constructed a permanent hard piped torus drain system to replace the collection of flexible hoses, spool pieces, and pumps previously used to drain the torus during outages.
Summary of Safety Evaluation:
During normal plant operations the torus drain system is inactive with all spool pieces removed and the attachment points blind flanged to preclude inadvertent draining.
The torus drain piping was supported as required to satisfy Seismic Category 2 over 1 requirements.
- 11. MOD 86-017, Emergency Service Water Flow Monitoring Description of Design Change:
Installed flow monitoring instrumentation in each of the emergency service water headers for the emergency diesel generators and in each of the emergency service water headers to the reactor building ECCS motor and room coolers.
Summary of Safety Evaluation:
The flow monitoring instrumentation installed in the ESW system provides operators with information on the operability of the ESW system during operations and accident conditions. The flow instrumentation satisfies the requirements of Regulatory Guide 1.97, Revision 2.
- 12. MOD 86-035, Replacement of Gaseous Chlorination System Description of Design Change:
Replaced the gaseous chlorination system with a liquid sodium hypochlorite/sodium bromide system.
Summary of Safety Evaluation:
The gaseous chlorine storage system was replaced with a liquid bleach system. This was done to eliminate the danger to personnel of a gaseous chlorine release.
The liquid sodium hypochlorite/sodium bromide system is not safety related.
- 13. MOD 86-038, Remove Seal-in from Open Circuit for MO-4085A and MO-4085B Description of Design Change:
Removed the seal-in from the open circuit for the Residual Heat Removal (RHR) intertie valves, MO-4085A and B. This allows the valves to be jogged open to flush the RHR lines slowly when shutting the plant down to minimize any change in reactor water temperature.
4
Summary of Safety Evaluation:
The safety related function of these valves to close on an LPCI initation signal has not been affected. Performance testing during installation verified that the auto-close logic was not affected.
- 14. MOD 86-039, Drywell Air Temperature Thermocouple Replacement Description of Design Change:
Replaced industrial grade drywell air temperature thermocouples and wires with Class 1E Qualified thermocouples and wires.
Summary of Safety Evaluation:
Placement of thermocouple pairs on four drywell levels provides accurate representation of bulk drywell atmospheric temperature to meet system functional requirements during a LOCA with atmosphere stratification. The thermocouple distribution also provides operators with indication of proper drywell cooler operation.
- 15. MOD 86-040, Parts A, B and C, 480 Volt Electrical Distribution Upgrade Description of Design Change:
Part A of this project removed the precipitator tank from the northeast corner of the turbine building on the 931' elevation.
Part B changed the feed to the cooling towers from the plant 4KV buses to the secondary of 1ARS transformer and the tertiary #6 transformer.
This included installation of a 13.8 KV to 4KV substation at the discharge structure to supply buses 17 and 18.
Part C installed a new double ended 4160V to 480V load center in the electrical equipment room. This load center is supplied from breakers 152-306 and 152-406. MCC 114 and Power Panel HP31 power supplies were removed from load center 109 to new load center 107 and 108.
Summary of Safety Evaluation:
Part A:
The precipitator tank and associated feed tanks are no longer in use for raw water, so the makeup deminerializer system is not affected by their removal.
Part B and C:
Results of an electrical load study, the electrical protection coordination review, and an updated computer model of the Monticello electrical distribution system provide assurance that there has been no degradation of the system by the new installation.
- 16. MOD 86-041, 24 VDC Battery Charger Replacement Description of Design Change:
Replaced all four 24-VDC chargers with new chargers of identical capacity. Installed new fuse panel to each division to replace the main panel breaker, to correct main panel breaker and load breaker protection miscoordination.
5
Summary of Safety Evaluation:
The new chargers were procured and installed to specifications better than or equal to the original specifications. The new charger provide better filtered and regulated outputs, under steady state or transient conditions.
Installation of new fuse panels provides proper electrical protection coordination for faults downstream of the load breakers.
Installed an alternate Safety/Relief Valve (SRV) nitrogen system on two of the SRV's in response to NUREG-0737, Item II.K.3.28.
Summary of Safety Evaluation:
The new N2 supply line ties into the instrumentation air lines serving the SRV's between the existing check valve and accumulator. All equipment downstream of the check valves are seismically supported.
Primary containment isolation is provided by doubled valving the lines outside containment as near the containment penetration as possible.
Because these lines are considered to be small diameter instrumentation lines, an exception to USAR Section 5.2.1.2.2 requirements is allowable.
This system is designed to withstand a hostile environment and still perform its function for 100 days following a LOCA.
Replaced eight Class 1E DC Rotork motor operators with qualified DC Limitorque motor operators.
Summary of Safety Evaluation:
The new motor operators have no effect on the function of the existing equipment except for changing the stroke times of the Reactor Core Isolation Cooling (RCIC) torus/condensate storage tank suction valves.
The change in stroke time has been evaluated and has been determined acceptable. The stresses induced by the heavier Limitorque motor operators on the piping and support system have been evaluated and determined to be acceptable. Changes in electrical loads incurred by the new motor operators have been evaluated and determined acceptable.
- 19. MOD 87-004, Generator Synchronizing Protection Description of Design Change:
Installed Beckwith M-0193 and M-0188 synch check relays at the substation for each of the generator breakers, 8N4 and 8N5, to provide protection against inadvertent faulty synchronization of the main generator to the grid.
6
Summary of Safety Evaluation:
The procedure for 8N4 and 8N5 breaker closures for connecting various sources remains unchanged. However, any close signal initiated outside a window of safe operation by an operator, the breaker will not close and would require a second attempt.
- 20. MOD 87-005, Generator Field Protection Description of Design Change:
Installed microprocessor-based field programmable volts/hertz (V/Hz) relay in place of existing GE V/Hz relay to provide improved protection for excessive V/Hz excitation of main generator. Added undervoltage relays for detecting blown fuses or open circuits in potential metering phases.
Summary of Safety Evaluation:
The new programmable relay provides six setpoints of V/Hz versus time and is matched to the limit curve of the main generator. The undervoltage relays eliminate the potential for generator damage resulting from overexcitation due to an open metering phase. These changes provide improved protection to the main generator without compromising other normal and protective functions.
- 21. MOD 87-009, 3D Monicore Core Monitoring System Description of Design Change:
Remove the process computer code which was used to monitor core thermal power and thermal limits. Replaced old code with modified version of GE design code PANACEA that runs on a modern 32 bit minicomputer.
Summary of Safety Evaluation:
A new Core Calculation and Monitoring System has been designed, tested and installed at the Monticello Plant by General Electric. The central portion of the new monitoring code used in this calculation is PANACEA which is also used by GE for core design and licensing.
The PANACEA code has been benchmarked and the accuracy verified. The code characteristics have been documented in GE report NEDO-30130 which has been reviewed and approved by the NRC.
The adaptive methodology used to provide PANACEA with monitoring capability had also been benchmarked and is discussed in detail in GE report NEDE-20340-3 which is available for review upon request.
There is no safety significance to either the current or the former system as they are used as an aide to the operator in monitoring the status of the plant, but not required for safe operation. However, the overall system has been designed to high standards and a complete design basis and benchmarking verification is maintained by GE in their system and software design specifications. The required system input data has been verified correct by GE and checked by the plant.
The system has been tested in static and dynamic states and its operational output verified correct by GE and the plant staff.
7
- 22. MOD 87-012, Instrument Air Drver Replacement Description of Design Change:
Installed a Pall Pneumatic Instrument air dryer Model No. DHA-400. The electrical power source, system annunciation and service water requirements were modified to accommodate the new instrument air dryer.
Summary of Safety Evaluation:
This modification reduced the load on load center 109 by 12KW and increased the load on load center 104 by 30 watts. A plant load study shows the additional 30 watts on load center 104 is acceptable. Demand for plant service water was reduced because the new equipment is not water cooled. The new alarm configuration was approved by the Human Factors Committee.
Due to the deterioration of the positive post battery seals, the replacement of the Division 1 250V DC battery was required. In addition, the battery capacity was increased to satisfy the requirements of the proposed NRC Station Blackout Rule and future loading concerns.
Summary of Safety Evaluation:
The replacement battery was sized in accordance with IEEE standards to ensure adequate capacity existed for a design basis accident and to satisfy requirements of the proposed NRC Blackout Rule.
The battery cells and racks ate seismicly qualified under the seismic accelerations committed to in USAR Appendix A. The electrical load study and electrical protection coordination review provide assurance that there has been no degradation of the system by the new installation. A Capacity/Modified Service test was performed to verify the battery meets it design requirements and satisfies plant Technical Specifications.
Replaced 8" Fisher Control Valves with 12" Valtek Anti-Cavitation Control Valves in the RHR Service Water (RHRSW) system.
Summary of Safety Evaluation:
The RHRSW was improved by this modification. Its purpose was to replace eroded valves caused by cavitation with valves of improved design.
- 25. MOD 87-019, PCB Transformer Replacement 1987 Outage Description of Design Change:
The 1987 outage portion of this modification replaced the transformers on load centers 101 (XlO), 103(X30), 105(X50 and X80), and 106(X60 and X70).
The #11 and #12 diesel generator neutral grounding transformers were also replaced. These transformers were replaced to meet our goal of removing all PCB contaminated transformers from service by 1990.
8
Summary of Safety Evaluation:
The replacement transformers were manufactured to match, as closely as practical, the characteristics of the original transformers.
Engineering reviews determined any deviations from the original characteristics to be acceptable.
Modified the Standby Liquid Control System (SBLC) to meet the NRC's ATWS Rule (10CFR 50.62) SBLC injection rate requirements The boron solution was replaced with a solution enriched in the B10 isotope.
Instrumentation setpoints were changed to reflect concentration changes.
Summary of Safety Evaluation:
The SBLC system performance has been improved by effectively doubling the rate the system can insert negative reactivity. B10 enrichment also permitted solution concentration reduction reducing the potential for precipitation.
Removed manual isolation valve for RV-2470, rerouted pipe to bypass manual isolation valve for RV-2471, added system isolation valves, thus enabling "A" and "B" Loop Core Spray and RHR to be individually isolated.
Summary of Safety Evaluation:
Removal and bypass of manual isolation valves CST-150 and CST-151 brings the system piping into compliance with ANSI B31.1, "Power Piping Code".
The analysis performed to ensure seismic qualification, concluded that existing supports were adequate.
Replaced original 3/4" RCIC turbine exhaust drain line and steam trap with a new 1" steam drain line and flow orifice.
Summary of Safety Evaluation:
The evaluation for determination of system operation impact, compliance with ANSI B31.1, and installation process impact concluded that the replacement was acceptable.
Installed crossties between the Emergency Filtration Train Emergency Service Water (EFTESW) and the Emergency Service Water (ESW) systems to reduce the potential effects of ESW piping corrosion and corrosion buildup products.
9
Additional benefits include the elimination of the need for "normally" closed High Energy Line Break (HELB) valves in the ESW system and dedicated pumps (P-111A & B) for Emergency Diesel Generator cooling.
Summary of Safety Evaluation:
The piping supports have been designed to meet Seismic Category I requirements. A HELB evaluation was performed, the new configuration improves the HELB situation for the ESW system. A review of Appendix R requirements required the installation of a new hand switch (Addendum
- 1) on the Alternate Shutdown System panel to remotely start ESW pump P 111D. Flow adequacy for the new configuration has been evaluated and found acceptable.
The cold reference leg safeguards level instruments exhibit a large change in indicated versus actual level when going from cold shutdown to rated temperature. The feedwater control system makes the change more apparent because it provides temperature compensated level indication over a broad range of temperatures.
The combination of these effects requires special considerations by the operators during startup and shutdown.
To correct this situation, the temperature compensation function was removed from the feedwater control system and the system adjusted to provide accurate level indication at rated operating conditions.
Summary of Safety Evaluation:
Removal of the temperature compensation function does not create any new failure modes nor increase the probability of any previously analyzed event. Level signal failures would be less likely because of the reduced number of components contributing to the level calculation.
Feedwater control system failures analyzed in the USAR are assumed to occur at rated conditions where there would be no change in level instrumentation operation as a result of this modification.
- 31. MOD 87-041, Offgas Preheater Drain Line Description of Design Change:
The offgas preheater level control valves were replaced with steam traps and the low level annunciators were removed from Control Room panels.
Summary of Safety Evaluation:
This modification was reviewed with respect to the probability that a hydrogen explosion could occur in the recombiner or associated piping.
It was concluded that this modification was acceptable. The pressure temperature integrity of the drain line was not compromised by the change.
10
- 32. MOD 87-046, Mechanical Vacuum Pump Trip and/or Air Operated Pump Suction Valves Close on Auto Initiation of the Standby Gas Treatment System (SBGT)
Description of Design Change:
Installed necessary trip logic to isolate the Mechanical Vacuum Pump (MVP) if it would be running when an auto initiation signal is received to start the Standby Gas Treatment System (SBGT).
This logic change eliminates a SBGT low flow condition when the SBGT and MVP operate simultaneously.
Summary of Safety Evaluation:
The SBGT system, in conjunction with secondary containment, is designed to minimize any ground level release of radioactive materials which might result from a serious accident. Because operation of the SBGT system while the MVP is operating reduces system flow, the potential for building exfiltration exists. Therefore, tripping the MVP on SBGT auto initiation will minimize the potential for any ground level releases due to decreasd building vacuum.
- 33. MOD 87-048, Replace RC-6-2 Description of Design Change:
Replace lifting check valve RC-6-2 in the 3" Reactor Water Cleanup (RWCU) return line crosstie to the RCIC line with a tilt disc check valve. The valve was replaced because it had failed closed and could not be replaced with the same type valve.
Summary of Safety Evaluation:
The use of a tilt disc rather than a lift check valve was reviewed and found acceptable. The installation complied with ANSI B31.1, Power Piping Code, using compatible material. The modification does not affect the function or the reliability of the RWCU system.
- 34. MOD 87-051, Part B, Correction of Breaker/Fuse Miscoordination Description of Design Change:
To achieve proper protection coordination on the plant electrical systems, several fuses were changed and circuit breaker trip setting adjustments were made.
Summary of Safety Evaluation:
Fuse changeouts were performed using a fused jumper, or when the system equipment was parallel to an alternate power source, or when equipment was not affected by de-energization. If fusing of lower current carrying capacity was installed, it was functionally tested to verify current carrying adequacy in addition to continuity testing for all fuse changouts.
- 35. SRI 87-008, Chem-Nuclear RDS-1000 Radwaste Dewatering System Description of Design Change:
The primary method of processing filter media at Monticello is now the Chem-Nuclear RDS-1000 rapid drying system. This equipment replaces the cement solidification equipment, although a portable solidification 11
trailer will still be available for use. The RDS-1000 system allows more waste volume per shipping container, thus reducing the overall waste volume shipped for burial.
Summary of Safety Evaluation:
The RDS-1000 process is NRC approved and the resulting waste is accepted for burial at two burial sites.
The mechanical and electrical demands for the RDS-1000 equipment have proven to be less that that of the original solidification equipment.
- 36. SRI 87-015, Pressure Test Temperature Reduction for Reactor Vessel Description of Design Change The Appendix G pressure-temperature limit curves were revised by General Electric to reflect an improved beltline weld material toughness estimate. The Original RTNDT of the weld material was +400 F based on the original purchase specification for the weld material.
Based on better information, this RTNDT was lowered to -400 F which makes other material in the vessel more limiting than the weld material and reduces the minimum temperature required for hydrostatic tests.
Summary of Safety Evaluation General Electric Company's report concludes that adequate margin exists to RTNDT limits.
- 37. EE 86-044, Fire Barrier Application to Conduit Run from Battery Room to Turbine Building Description of Design Change Installed 3-hour fire-proof covering on an electrical conduit which runs through the hallway of the administration building (Fire Zone 10),
which is a Division I area. The cables contained in the conduit are Division II cables.
Covering this cable with fire-proof material will maintain the required separation between shutdown divisions.
Summary of Safety Evaluation The fire proof material has been tested to ASTM E-119 fire endurance requirements and is fully qualified by American Nuclear Insurers.
The conduit supports were reanalyzed and it was determined that the added weight would not exceed design margins.
Derating of the electrical conduit was also reviewed and it was determined that the derating of 9.72% would be acceptable.
- 38. EE 87-005, Investigation for Construction Startup Strainers in Plant Pump Suction Piping Description of Design Change Verified construction start-up strainers had been removed from plant pump suction piping. Corrected plant Piping and Instrumentation Drawings (P&IDs) to reflect that the start-up strainers had been removed.
12
Summary of Safety Evaluation The investigation involved removal of spool pieces from the suction side of the RCIC pump, HPCI booster pump, and the waste sample pump to verify the start-up strainers had been removed. All strainers were confirmed to have been removed from the pump suction piping prior to plant startup.
- 39. EE 87-34, Wiring Discrepancy on MO-2021 Description of Design Change Due to a shortness of a lead in control cable to MO-2021, Loop 12 Drywell Spray Inboard Isolation valve, connections to the motor operator could not be made exactly as depicted in the connection diagram. The connection was made in a configuration slightly different than shown in the connection diagram but functionally the same.
The connection diagram was revised to show the as-left configuration.
Summary of Safety Evaluation The as left connection arrangement is functionally the same as depicted in the connection diagram. Operation of the valve is not affected in any way. Operational testing of the valve has verified correct operation.
- 40. EE 87-060, MSIV Packing Change Description of Design Change The stuffing box configuration for the main steam isolation valves was changed to use die-formed graphite packing, incorporate a bushing to reduce stuffing box depth, and to incorporate live loading of the packing gland. In addition all stem leakoff lines were cut and either capped or plugged.
Summary of Safety Evaluation The safety evaluation addressed valve operability, material compatibility, leakoff line capping, installation, and testing. The change has no adverse impact on valve operability. Testing was performed to verify valve operability and packing integrity.
- 41. EE 87-078, Reroute Jet Pump Instrument Line JP-1-1"-DCA Description of Design Change Rerouted jet pump instrument line JP-1-1"-DCA. JP-1-1"-DCA was within 3/4" of safety relief valve line RV-24A-10" and potentially subjected to elevated stresses.
Though the elevated stresses were within acceptable limits as documented in SRI 86-017, it was decided it would be prudent to reroute the jet pump instrument line away from the safety relief valve line.
Summary of Safety Evaluation The rerouted jet pump instrument line was analyzed and found to meet ANSI B31.1 allowable stress requirements. The use of secondary stress allowables in no longer required.
13
- 42. EE 87-088, Replace MSIV 2-Way Valves Description of Design Change The Main Steam Isolation Valve (MSIV) 2-way air pilot valves were replaced with current model valves. The previous valve is no longer manufactured.
Summary of Safety Evaluation The replacement valves are identical in form, fit, and function.
Operability testing was performed to verify correct operation.
- 43. EE 87-090, Machine Flange Facings for RHR Decon Connections Description of Design Change The flange facings for the RHR decontamination connections were machined to improve the leaktightness of the joint and to make proper boltup of the joints easier to achieve.
Summary of Safety Evaluation The change to the flange facings improved the leaktightness of the bolted joints. The strength of the flange fittings was not affected.
Hydrostatic testing of the joints was done to verify a leak tight joint.
- 44. EE 87-91, Core Reload for Cycle 13 Description of Design Change The reactor core was loaded with 120 new GE8B fuel bundles, 16 bundles previously removed from the reactor at EOC-11, and 348 bundles retained from cycle 12.
Four control blades were replaced with new control blades. These blades use boron carbide as an absorber material.
Four LPRM strings were replaced with fresh NA-200 detector strings.
Six IRM/SRM dry tubes were replaced.
The Core Monitor Software database was updated.
Summary of Safety Evaluation The fuel was arranged in a pattern established by General Electric Company. Reload analysis methodology approved by the NRC was used.
The new control blades are identical to the blades loaded at BOC-12.
The design has been approved by the NRC.
The fresh detector strings are identical to the strings loaded at BOC
- 12.
The dry tubes were replaced with dry tubes that are identical to the replaced tubes in all important dimensional respects. They are fabricated from material that has a higher resistance to IGSCC than the old tubes.
14
The database was updated to reflect cycle 13 core changes to enable proper monitoring of the cycle 13 reactor core.
- 46. EE 87-102, Replace #11 RHR Pump Motor Description of Design Change Installed a new motor on # 11 RHR pump to replace a temporary motor installed under DC 82MO98. The cooling water lines were reconnected to the motor since the new motor requires cooling water while the old motor did not.
Summary of Safety Evaluation The new motor is identical in form, fit and function to the originally installed RHR pump motors. The new motor is environmentally and seismically qualified to current IEEE standards. The plant electrical load study was reviewed and installation of the motor was found to be acceptable. Testing was performed after the installation of the motor to verify correct operation and protective relay settings.
- 47. EE 87-112, Repair of MO-2078 Description of Design Change Removed the old valve seat ring. Installed and seal welded a new seat ring. A new weld was added to the the RCIC system as a result of this repair.
Summary of Safety Evaluation The impact of the new weld on the stress level of the RCIC steam supply line was analyzed. This analysis concluded that the new weld was in compliance with ANSI B31.1, "Power Piping Code".
- 48. EE 87-117, Replacement Oil Seal for MO-2076 Description of Design Change Replaced Garlock Klosure type #21168-0052 oil seal with a Chicago Rawhide type #5066 in the valve actuator motor for MO-2076.
Summary of Safety Evaluation Replacement of the original oil seal with the Chicago Rawhide model was analyzed with respect to form, fit and function. The impact of this change on the equipment qualification status of MO-2076 was evaluated and it was concluded that this replacement was acceptable.
- 49. Revision 12 to the Operational Quality Assurance Plan Revision 12 to the NSP Operation Quality Assurance Plan was internally reviewed and approved on May 26, 1988. We have concluded that this revision does not reduce the commitments of NSP's Operation Quality Assurance Program and does not adversely impact the safe operation of the nuclear power plants.
Specific changes with reason for the change and basis for concluding no reduction in commitments [per 10 CFR 50.54(a)(3)] are presented in Appendix D to the plan. The Operational Quality Assurance Plan, Revision 12, is included in Appendix C to the USAR.
15
3L cA