ML20207C097
| ML20207C097 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/08/1999 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20207C104 | List: |
| References | |
| NUDOCS 9903080356 | |
| Download: ML20207C097 (21) | |
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Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 F
uary 8,1999 10 CFR 50.71(e)
U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Submittal of Revision No.16 to the i
Updated Safety Analysis Report (USAR)
Pursuant to 10 CFR 50.71(e) we are submitting one original and 10 copies of l
Revision No.16 to the Updated Safety Anaiysis Report (USAR) for the Prairie Island l
Nuclear Generating Plant. This revision brings the USAR up-to-date as of August j
31,1998 (though some information is more recent), with two exceptions:
Proposed changes identified by the USkR Review Project Team are not included e
in this revision and will be submitted as a separate revision early in 1999.
l The low pressure turbine upgrade has affected the section on turbine generated l
missiles (the probability has been significantly reduced) but the revision to the section has not been prepared yet. It will be submitted sithin the time limit for USAR updates as required by the regulation.
l contains descriptions and summaries of safety evaluations for
. changes, tests, and experiments made under the provisions of 10 CFR 50.59 during the period since the last update. Attachment 1 also contains discussions of changes made to regulatory commitments made within our Regulatory Commitment Change Process.
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,c USNRC February 8,1999 contains the USAR page changes and instructions for entering the pages. Included in Attachment 2 is Revision 21 to the Northem States Power
' Ccmpany Operational Quality Assurance Plan in compliance with 10 CFR 50.54(a).
Changes in Revision 21 to the Plan are described in Appendix D to the Operational Quality Assurance Plan (which itselfis Appendix C to the USAR).
- is Rev.12 of the Fire Hazards Analysis. Submittal of the FHA satisfies a commitment made in our letter to NRC dated Novembar 13,1997 Distribution of is limited to the addressees shown on this letter.
In this letter we have made no new Nuclear Regulatory Commission commitments.
I certify that the information presented herein accurately presents changes made since the last updating submittal of the Prairie Island USAR (with the two exceptions noted above).
Please contact Jack Leveille (651-388-1121, Ext. 4142) if you have any questions related to this letter.
Joel P. Sorensen Plant Manager Prairie Island Nuclear Generating Plant c: Regional Administrator-Region ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments: 1. Safety Evaluation Summaries i
- 2. USAR page changes
- 3. Fire Hazards Analysis (limited distribution)
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ATTACHMENT 1 0
PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS AND EXPERIMENTS - AUGUST 1998 The following sections include a brief description and a summary of the safety evaluation for each of those changes, tests, and experiments which were carried out without prior NRC approval, pursuant to the requirements of 10 CFR Part 50, Section 50.59(b). Also included are discussions of changes made to regulatory commitments made within our Regulatory Commitment Change Process.
Modification 94L483 Part G - Fire Barriers Upgrade-Cabling and Equipment Description of Change Modification,94L483 Part G, implemented changes required by the new Safe Shutdown Analysis to replace or reroute portions of power and control cables, relocate control switches, seal penetration gaps, replace solenoid valves with manual valves, and install local and remote fusing as listed below:
- 1. Pressurizer Power Operated Relief Valves CV-31231,-31232,-31233, and 31234 control cable replacement where routed through the Aux Bldg fire areas 59 and 74.
- 2. 22 SI pump control cable replacement.
- 3. 12 & 22 Charging Pumps control cable replacement and relocate their respective remote / local control switches from the Hot Shutdown Panels in the AFWP Room to their respective charging pump rooms. Replace and reroute and wrap portions of 12 & 22 Charging Pump power cable. Replace solenoid valves with manual valves for local speed control for 12 and 22 charging pumps.12 Charging pump valve is SV-33689,22 Charging pump valve is SV-33837.
- 4. Reroute Unit 2 (2N52) NIS cabling to avoid U2 735' Aux Building, reroute Unit 1 (1N52) to keep the same train routes together.
- 5. Close the gap for the Aux building to containment electrical penetration cabinets 1134,1136,2134, and 2136 perimeter gaps where they join the concrete shield building structure to prevent a fire inside the cabinet from getting out of the cabinet.
6.121 and 123 Air Compressors, add one control power fuse in each compressor's MCC (1 A1 and 2A1) bucket and wire local and remote contacts off the fuses to provide separate fusing for the local and remote control functions.
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- 7. MCC 1K1 replace and reroute power cable to avoid wrapping.
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Rockbestos Firezone R cable is used for part of the recabling work. - An exemption :
l request has been submitted to the NRC to credit this cable for Appendix R
't purposes. This is in accordance with GL 86-10..
i Summary of Safety E*taluation I
i All permanent results of the recabling and rerouting portions of this modification are designed to improve the plant's ability to respond in the case of a fire while maintaining the current safety function by improving the cabling without changing -
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the operation of electrical equipment or its power source. Therefore, the operation j
' of these components remains unchanged by this modification and no new-j malfunctions are introduced.
i The permanent results of the replacement of the local control solenoid valves with l-manual valves for 12 and 22 charging pumps and the additional fusing of the air 1
i compressors change functional aspects of these components; however, these functions are not relied upon in any accident analysis. Charging pump local control -
valves cannot affect the remote controls which are relied on for operation from the.
control room in analysis unless the local / remote switch is positioned in local. Local -
l position is alarmed in the control room as before the modification; therefore no new malfunctions are introduced. Local and remote fusing does not change the operation l
of the air compressors; it just makes local operation easier since no manual fuse replacement is needed if the remote control circuit fuse is blown.
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Structural, seismic, electrical loading, and cable routing criteria assumed in Section l
8.7 of the USAR are not changed by this modification. Also, operation remains the j
same as previously analyzed for PRA and accident analysis.
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Further, the probability of a malfunction of these loads or its related wiring is decreased as a result of this modification since the new wiring meets or exceeds
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existing wiring abilities. Structural, seismic, electrical loading, and cable routing criteria assumed in Section 8.7 of the USAR are not changed by this modlocation.
Also, operation remains the same as previously analyzed for PRA and accident analysis.
This Safety Evaluation concludes that this modification will not result _in an
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unreviewed safety question and does not affect or change Plant Technical.
Specifications. Existing compensatory measures (fire watches) will remain in effect 1
1 where required until the final Safe Shutdown Analysis and exemption request are
' approved.
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February 8,1999.
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' Modification 95L519 Part A - Unit 1 Low Pressure Turbine Replacement i
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. Description of Change e
I Low pressure turbine rotors utilizing shrunk-on discs were removed from service and '
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replaced with fully integral nuclear LP rotors. Newer forging technology has allowed -
i the rotor to be a single forging and therefore eliminated the disc keyways and bores.
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Rotor peak stresses are significantly reduced, leading to a large reduction in probability of a rotor burst.
Summary of Safety Evaluation.
l The safety evaluation concluaes that the activity is safe and does not involve an unreviewed safety question. Two turbine <jenerator malfunctions are discussed:
turbine overspeed and turbine missiles. The overspeed discussion in the USAR is not affected by the new rotor design. Turbine' missiles may result from rotor failure at operating speed. Probability of rotor burst for the new rotors does not exceed 104 evin after 30 years of running time.
i Modification 96SG04 - Removal of Tube / Sleeve Samples from 11 and 12 i.
Steam Generators Description of Change in order to implement the voltage based repair criteria on Unit 1 under Design Change 97SG05, to gain additional knowledge about the degradation mechanisms
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of the Prairie Island steam generator tubing, and to evaluate the effects on steam 1
generator tube structural and leakage integrity of non-quantifiable eddy current indications, removal of tube samples and sleeve sample was required during the j-October 1997 refueling outage.
This design change:
- 1) Evaluated the effects of the removal of tube samples from Unit 1 Steam Generators including the remnant tube stub left in place in the steam generator and the boring of the tubesheet hole to accommodate the outside diameter of the upper sleeve weld joint for sleeve samples.
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- 2) Evaluated the Combustion Engineering welded tubesheet plug used for the nominal and oversize tubesheet holes.
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Page 4 of18 February 8,1999 I
- 3) Evaluated the Combustion Engineering mechanical tube plug used in the intact tube end.
- 4) Removed 3 tube and 1 sleeve samples.
- 5) Installed Combustion Engineering welded tubesheet plugs in the tubesheet bore holes left by the removed tubes and mechanical plugs in the opposite tube end.
Summary of Safety Evaluation in order to remove the sleeved tube, the diameter of the tubcsheet hole was enlarged by boring. The effect of the enlarged bore is to decrease the ligament thickness (the shortest distance between adjacent tube holes in this square pitched tube bundle) of the four bore holes adjacent to the bored hole. The primary stresses in the thinned lic ; ment and the effect on the tubesheet fatigue usage factor as required by ASME Code Section ill were determined to be acceptable.
The portion of the tube remaining in the steam generator was evaluated for the potential of becoming a vibration hazard to adjacent tubes. Turbulence vibration amplitudes and related forces were shown to be small enough to preclude the generation of a loose part through fretting wear or cyclic fatigue failure at the tube support plate.
The plugs maintain the same pressure boundary integrity and leak tightness of the original tube by plugging the opened tubesheet bore ho's and open tube end.
The increase in bore hole size did not significantly reduce the tubesheet stress and fatigue margins.
Section 14 of the USAR was reviewed for impact. The function of the steam generator tubing is to maintain the primary system pressure boundary and to transfer heat from the reactor coolant system to the secondary side. This modification does not change the Tsilure modes or failure impact of the Steam Generator tubing. No impact on tt a USAR Section 14 accidents was identified.
Modification 97FH02 - Fuel Examinations for the incomplete Rod Insertion issue Description of Change Several plants containing Westinghouse high burnup fuel have experienced incomplete rod insertion (IRI). Westinghouse undertook an extensive study to
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February 8,1999 j
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determine the root cause of the IRI issue. Site testing was performed at several '
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plants including South Texas and Wolf Creek.' Some of the tests performed included j
visual examinations, drag testing, fuel rod growth, and single thimble guide tube -
probe.l Hot cell examinations were performed on irradiated fuel rods obtained from j
the Wolf Creek plant. Manufacturing effects were also investigated by sampling u
i assemblies which had thimble guide tubes that were manufactured before, during, j
and after the time period in which the thimble guide tubes for Wolf Creek were -
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manufactured.
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In order to support an industry effort to collect more data on fuel assembly growth, Westinghouse performed an examination campaign at Prairie Island. Testing j
Included visual examination of fuel assemblies, drag testing, assembly length j
measurement, fuel rod oxide measurement, thimble guide tube oxide measurement j
i and removal of instrument tubes so hot cell exams could be performed.
l The fuel inspections were performed on assemblies with different cladding and i
thimble guide tube material (Zircaloy-4/ Improved Zircaloy-4/ZlRLO). The testing j
was performed on discharged fuel. Drag testing was performed on the Region 15 fuel that may be reinserted as center assemblies in the next several fuel cycles.
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In summary, the root cause of the insertion anomalies at Wolf Creek and South l
Texas is thimble guide tube distortion caused by compressive loads on the I
assemblies. Westinghouse also concluded that 'or the 14x14 and 15x15 fuel designs (Prairie Island cores are in this group), no indications of an insertion i
anomaly have appeared. This is because these plants run at a relatively low I
temperature and therefore creep and corrosion do not occur to the degree seen in the higher temperature plants.
l Summary of Safety Evaluation Removal of the instrument tube from a Westinghouse 14X14 fuel assembly has no impact on fuel handling or criticality analysis. The other fuel examinations are non-j intrusive and non-destructive and result in no impact on subsequent fuel handling.
4 The assemblies with removed instrument tubes will not be used in future core designs.
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4 Modification 97FH03 - Unit 1 Cycle 19 Reload -
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j Description of Change
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This modification replaced depleted Unit 1 fuel assemblies with a fresh reload ofl44 Westinghouse VANTAGE + fuel assemblies allowing another cycle of power operation.-
The new fuel assemblies are enriched to a nominal 4.95 w/o U235 and results in a.
projected cycle length of 18717 mwd /MTU.
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The Unit 1 Cycle 19 reload was developed by NSP Nuclear Analy,is & Design (NSPNAD) using approved methodology addressed in NSPNAD-8101-A, j
Qualifications of Reactor Physics Methods for Application to PI Units. More details on the operational parameters can be found in NSPNAD-97004, Rev. O, Prairie Island Unit 1 Cycle 19 Startup and Operations Report, and NSPNAD-97003, Rev. O, Prairie Island Unit 1 Cycle 19 Final Reload Design Report.
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l Summary of Safety Evaluation l
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The following safety concerns were addressed in the safety evaluation.
I A.
Thermal Hydraulic Analysis l
B.
Accident and Transient Analysis C.
Main Steam Line Break / Containment Response Analysis D.
LOCA-ECCS Analysis E.
Rod Ejection Analysis 1
F.
Fuel Handling Accident G.
Refueling Shutdown Margin -
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Heatup/Cooldown Curves - Reactor Vessel Radiation Surveillance Program
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Fuel Rod Design Performance J.
Soent Fuel Pool Heat Load j
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New Fuel Rack / Spent Fuel Rack Criticality L.
Core Exposure Limits /Off-site Dose Calculations i
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Peak Linear Heat Generation Rate -
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Fuel Assembly Design Change j
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Startup and Operations P.
Validity of Safety Evaluation j
4 All results were acceptable and are presented in NSPNAD-97003, Rev. O, Prairie
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lsland Unit 1 Cycle 19 Final Reload Design Report. The LOCA analysis was performed by Westinghouse and is documented in the Unit 1 Cycle 19 LOCA i.
Confirmation Letter 97NS-G-0062, August 15,1997. This letter confirms that Unit 1 Cycle 19 will continue to conform to the acceptance criteria of 10CFR50.46.
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Page 7 of 18 February 8,1999 Since all transient analyses meet the acceptance criteria, there are no unreviewed safety questions for the Unit 1 Cycle 19 Core Reload Design Change.
Modification 97FP02 - Reactor Coolant Pump Lube Oil Collection Description of Change This modification instsl led lube oil collection spray shields around the oil lift pumps and associated piping for the reactor coolant pumps on both units. This change brings the units into compliance with Section Ill.O of 10CFR50, Appendix R.
j Summary of Safety Evaluation The spray shields will not adversely affect operations with containment, and will improve collection of potentially leaking oil.
Modification 97RV04 - Reactor Coolant Gas Vent System Jumper Description of Change The design change provides a vent path without a loop seal from the reactor vessel head to the pressurizer without using the RCGVS solenoid valves. The intent is to eliminate cold shutdown cycling and operations that maintain the valves energized for long periods of time, reducing the chance of startup problems with the valves. The jumper functions include: release gas accumulation in the head during reactor coolant system pressurizer relief tank float operation through cooldown; dedicated head vent path to atmosphere during reactor coolant system draindowns impacting reactor vessel water level; communication path reactor vessel *o pressurizer for Reduced Inventory; air vent reactor vessel to pressurizer during reactor coolant system filling; reactor vessel head bypass flow to pressurizer for reactor coolant pump jogs during filling and venting to improve reactor vessel head venting.
This modification does not involve a Technical Specification change and is administratively controlled for use only during Mode 5,6 (Cold Shutdown) operation with design values based on postulated Mode 5,6 events.
Page 8 of 18 February 8,1999 Summary of Safety Evaluation The design change was evaluated aguinst postulated xld shutdown events of low temperature overpressurization and loss of residual heat removal capability with the reactor coolant system partially filled. The safety evaluation concluded that the activity is safe and does not involve an unreviewed safety question.
Modification 97ZH02 - Backup Compressed Air Supply for the Control Room Chillers
, Description of Change As a result of an engineering self-assessment of the Service Water System, it was determined that the control room chilled water system would not function without ir.strument air. With the chilled water system not functional, the temperature in the control room and relay room would exceed equipment qualification temperatures in a short time. This modification provides a reliable backup source of compressed air '
for operation of the chillers and associated valves.
Summary of Safety Evaluation Health and safety of the public are not adversely affected by this modification. In
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fact, this modification increases the margin of safety related to habitability of the Control Room.
Modification 97ZN02 -Isolation of Control Room to Relay Room Vent Descri) tion of Change This design change isolates the control room ventilation system from the relay room ventilation by installing blanks in the supply and retum ducts for the relay room.
Summary of Safety Evaluation
. actrothermal links in the system hold the dampers open until a CARDOX actuation signal initiates a pyrotechnic reaction to melt the solder holding the fusible link togethsr. Melting of the fusible links allows the dampers to spring closed. It was discovered that the electrothermal links were not installed correctly and therefore the dampers would have failed to close on CARDOX actuation, immediately upon i
discovery, the dampers were placed in a safe condition, that is, closed to prevent
.l carbon dioxide entry into the control room. This design change isolates the control
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Page 9 of 18 February 8,1999 room ventilation system from the relay room ventilation ' system. The safety.
evaluation concludes that the design change constitutes an improvement in safety.
Safety Evaluation 476 - Abandon Radiation Monitors R-34 and R-43 through:
R 47 Description of Change i
i The Safety Evaluation provides justification to permanently remove several radiation monitors. These monitors are low use, high maintenance equipment not required by.
Technical Specifications. Radiation protection coverage will be provided by portable or other installed instrumentation.
-Summary of Safety Evaluation i
The monitors provide no automatic actuation or control function, only indication.
Their purpose is to provide general area radiation monitoring. All design basis accidents in Section 14 of the USAR were reviewed. Deletion of these monitors will not cause an undetected increase in exposure to plant personnel for any accident i
listed in that section. De!stion of the monitors will not affect liquid or gaseous i
releases as listed in the USAR or constrain any fuel handling activities.
i Safety Evaluation 491 - Fire Protection / Appendix R USAR Revision 3
j Description of Change i
The safety evaluation implements a revision to the USAR to update the information related to fire protection and to reference a document that maintains the approved '
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fire protection progiam current with plant conditions.
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i Summary of Safety Evaluation The revision was conducted in accordance with guidance provided in Generic Letter 86-10.
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Safety Evaluation 502 - F5 Appendix F Revision 12, Fire Hazards Analysis Description of Change The safety evaluation addresses changes made to F5 Appendix F in Rev.12; F5 I
Appendix F contains the Prairie Island Fire Hazards Analysis.
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1 Page 10 of 18 February 8,1999 Summary of Safety Evaluation The revision was conducted in accordance with guidance provided in Generic Letter 86-10.
License Amendments 123 and 116 - Pressurizer Safety Valves and Main Steam Safety Valves Lift Setting Tolerance Change and Safety Limit Curve.
1 Changes Description of Change The ameridments revise the Technical Specifications by changing the pressurizer and main steam safety valve lift setting tolerance from i1 percent to 13 percent (as found only), revising the safety limit curves, reformatting Section 2, and correcting typographical errors.
Summary of Safety Evaluation The License Amendments were issued May 21,1996.
License Amendments 130 and 121 - Spent Fuel Special Ventilation Technical Specifications
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Description of Change The amendments change Technical Specifications limitations on crane operations in the spent fuel pool enclosure relating to spent fuel pool special vantilation system operability. These changes are necessary to allow movement of loads over spent fuel stored in the spent fuel pool enclosure with the spent fuel pool special ventilation system inoperable.
Summary of Safety Evaluation The License Amendments were issued September 15,1997.
License Amendments 131 and 122 - Cooling Water System Technical Specifications 3
Description of Change The amendments revise Technical Specifications governing the cooling water system. The changes improve plant operation based on operational experience with
Page 11 of 18 February 8,1999 the vertical motor-driven cooling water pump. The changes also incorporate information obtained during the Service Water System Operational Performance inspection (SWSOPI).
Summary of Safety Evaluation The License Amendments were issued October 21,1997.
License Amendments 132 and 124 -Incorporation of Combustion Engineering Steam Generator Welded Tube Sleeve Designs Description of Change The amendments incorporate Combustion Engineering steam generator tube sleeve designs and instaliation and examination techniques.
Summary of Safety Evaluation The License Amendments were issued November 4,1997.
License Amendments 133 and 125 Incorporation of Voltage-Based Steam Generator Tube Repair Criteria 1
Description of Chege The amendments revise certain Technical Specification limitations on reactor coolant system leakage and steam generator tube surveillance, and implement
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voltage-based repair criteria. The amendments also add a license condition to Appendix B of the licenses.
5 Summary of Safety Evaluation The License Amendments were issued November 18,1997.
License Amendments 134 and 126 -Turbine-Driven Auxiliary Feedwater Operability during Unit Startup Description of Change The amendments revise Technical Specification 3.4.B to provide specific guidance for conducting post-maintenance operational testing of the turbine-driven auxiliary feedwater pump and associated system valves to meet limiting conditions for operation and establish system operability during unit startup. An additional change
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permits, during Mode 2 when the main feedwater pumps are not required to be operated, the bypassing of the auto-start feature of the auxiliary feedwater pumps that results from the trip of both main feedwater pumps.
Summary of Safety Evaluation The License Amendments were issued November 25,1997.
4 License Amendments 135 and 127 - Use of a Pressure and Temperature Limits Report l
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Description of Change 4
The amendments update the Technical Specification heatup and cooldown rate curves and extend their reactor vessel fluence limit from the current 20 effective full 4
power years (EFPYs) to a new value of 35 EFPYs, incorporate into Technical Specifications the use of a Pressure and Temperature Limits Report, and change the power-operated relief valves temperature requirement for operability.
Summary of Safety Evaluation The License Amendments were issued May 4,1998.
License Amendments 137 and 128 - EF* Steam Generator Alternate Repair Criteria Description of Change The amendments revise the Technical Specifications to allow use of alternate steam generator tube repair criteria (elevated F-star or EF*)in the tubesheet region when used with the repair method of additional roll expansion. The amendments incorporate revised acceptance criteria for tubes with degradation in the tubesheet region and will enable NSP to avoid unnecessary plugging and sleeving of steam generator tubes.
Summarc/ of Safety Evaluation The License Amendments were issued August 13,1998.
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Page 13 of 18 February 8,1999 P.lcense Amendments 136 and 129 - Modification to ATWS Mitigating System Actuating Circuitry Description of Change The amendments authorize a design modification of the existing Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC).
3 The design modification would install a Diverse Scram System (DSS) designed to meet the requirements of a DSS described by 10CFR50.62 and make major modifications to the existing AMSAC.
Summary of Safety Evaluation The License Amendments were issued September 22,1998.
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License Amendments 139 and 130 -Inoperable Rod Position Indicator Channels Description of Change The amendments clarify the conditions that constitute operable individual Rod Position Indications (IRPI) system channels, provide for an allowed out of service time for inoperable IRPI indicator channels, and provide compensatory measures to be taken when any channel is determined to be inoperable.
i Summary of Safety Evaluation The License Amendments were issued October 30,1998.
License Amendments 140 and 131 - Cooling Water System Emergency intake Design Basis Description of Change The amendments change the design basis of the cooling water system emergency
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intake line flow capacity. The changes also reclassify the intake canal for use during a seismic event, which would be an additional source of cooling water available during a design-basis earthquake. The amendments also reflect the completion of license conditions that were implemented as part of interim amendments 128/120 to reflect compensatory measures taken by NSP until a seismically qualified emergency cooling water source could be provided.
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February 8,1999 Summary of Safety Evaluation.
The License Amendments were issued November 4i1998.
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Page 15 of 18 February 8,1999 CHANGES TO REGULATORY COMMITMENTS Regulatory Commitment Change 97-06 As part of the response to Generic Letter 83-28, Prairie Island committed to:
4.2.3 Perform life testing of the reactor trip breakers (including the trip attachmentn) on an acceptable sample size.
4.2.4 Develop periodic replacement of breakers or components consistent with demonstrated life cycles.
The commitments will be deleted pursuant to Supplement 1 to GL 83-28 dated October 7,1992, which states, "In light of this RTB operating experience, the staff has concluded that the actions already completed pursuant to GL 83-28 have been effective in improving RTB reliability tc open and that further actions to adJress the end-of-life degradation in breaker reliability are not justified. Therefore,'. tie staff concludes that licensee actions in response to items 4.2.3 and 4.2.4 of GL 83-28 are not necessary. To the extent that licensees may have made commitments to programs for periodically replacing RTBs or components in responses to GL 83-28, they may review and modify these programs taking into account their plant-specific operating experience, maintenance programs, and root cause determination programs for RTBs."
Regulatory Commitment Change 97-07 As part of the response to Generic Letter 96-04, Prairie Island committed to tagging out spent fuel pool dilution sources to eliminate their use without adherence to new administrative controls. This commitment is deleted since License Amendments 129 and 121 on SFP boron credit detail the requirements necessary for using boron credit for criticality control. A new commitment was made to determine appropriate administrative controls to control the SFP boron concentration and water inventory.
Regulatory Commitment Change 97-08 As part of the response to Generic Letter 96-04, Prairie Island committed to placing dry boric acid in the immediate vicinity of the Spent Fuel Pit in the event manual addition was required. The commitment is deleted because License Amendments 129 and 121 on SFP boron credit do not require boric acid in the immediate vicinity of the SFP.
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February 8,1999 J
Regulatory Commitment Change 97 09 la I
As part of the response to Generic Letter 96-04, Prairie Island committed to changes.
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to procedures that provide emergency makeup to the SFP to limit use of non 4
borated sources. Sampling following makeup was to be performed to' verify boron.
concentration. This commitment is deleted cince the steps provided in this q
commitment are now addressed in License Amendments 129 and 121 on SFP.
l boron credit. The weekly sampling requirement now in Tech Specs was approved-
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by the NRC as being sufficient to detect any dilution.
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Regulatory Commitment Change 97-11 In NSP-to-NRC [[letter::05000282/LER-1996-011, :on 960528,discovered Degraded Steam Generator Sleeves.Caused by Inadequate Cleaning of Parent Tube Prior to Insertion of Sleeves.Implemented Improved Cleaning & Addl Nondestructive Exam Re Installation of Sleeves|letter dated June 27,1996]], NSP committed to "..100% examination I
of all sleeves using current state-of-the-art Appendix H qualified eddy current l
techniques.... A reexamination of all sleeve welds with resulting volumetric and-
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crack-like ET indication will be done by UT." '
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l Three sleeves could not be examined by ET due to partial collapse of the sleeve.
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Four sleeves could not be examined by UT due to partial collapse of the sleeve.
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One sleeve could not be examined by UT due to a stuck in-situ pressure test tool.
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i In summary, physical constraints prevented fulfillment of the commitment. Sleeves i -
which could not be examined were removed from seivice by plugging.
1 Regulatory Commitment Change 98 03 J
l In NSP-to-NRC letter (entitled Clarification of Snubber Information) dated January 27,1988 NSP committed to "... monitor Nylatron backup rings via the present testing program and change to a material with an increased resistance at the next Viton j
seal change."
This commitment is revised to read, "NSP will continue to monitor Nylatron backup J
rings via the current testing program and will continue use of the Nylatron backup i
i rings at present service of 15 years.
NSP has monitored the silicone fluid through sampling at each refueling outage for j-viscosity and particulates. An increase in viscosity would Indicate an accelerated l
degradation of the snubber fluid due to the combined effects of the high temperature and radiation environment. An increase ~in particulates could be an indication of L
backup ring failure. Also, a visual inspection of the steam generator snubbers for fluid leakage from around the seals is included as part of the visual examination performed during each refueling outage. There has been no evidence of Nylatron L.
backup ring failure nor signs of leakage due to a backup ring failure. As-found
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February 8,1999 >
testing on the steam generator snubbers during their removal / replacement and the subsequent visual inspection of the backup ring after disassembly will provide empirical evidence of seal life for service conditions at Prairie Island.
Regulatory Commitment Change 98-05 i
Corrective action stated in Unit 1 LER 92-005 was "...to provide instructions for the operator to remove the fuses...for No.12 Diesel-Driven Cooling Water Pump
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Control Panel"in case of a control room evacuation.
Due to concems with " hot shutdown repair" action with pulling fuses, operator action was changed to open the DC knife switch at the local control panel for No.12 Diesel-Driven Cooling Water Pump.
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The instruction to open the DC knife switch is an acceptable operator action to mitigate the consequences of fire damage to the pump's breaker control.
i Compliance with Appendix R, Section Ill.G.3 and Ill.L'is achieved without reliance
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on an operator action which may be considered to be a " hot shutdown repair,". for.
which an exemption may be required.
j Regulatory Commitment Change 98-06
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Corrective action stated in Unit 1 LER 92-006 was "...to provide instructions for the operator to remove the fuses for the reactor vessel head vent solenoid valves" in case of a control room evacuation.
t t
L Due to concems with " hot shutdown repair" action with pulling fuses, operator action was changed to de-energize the 125VDC Panel associated with the reactor vessel j
head vent solenoid valves.
f The instruction to de-energize the 125VDC Panel is an acceptable operator action to ll fall the reactor vessel head vents closed. Compliance with Appendix R, Section j
lil.G.3 and Ill.L is achieved without reliance on an operator action which may be considered to be a " hot shutdown repair," for which an exemption may be required.
Regulatory Commitment Change 9847 l
Corrective action stated in Unit 1 LER 98-005 was " Doors have been opened on Unit 1 to preclude flooding concerns with respect to Unit 1 MSIVs "
i The corrective action has been changed to " Doors will be under administrative control to preclude flooding concems with respect to Unit 1 MSIVs."
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i Page 18 of 18 February 8,1999 I
Engineering Calculation ENG-ME-372 determined that 1427. pounds of force will be l
applied to the doors by 50 inches of water. If the center latches are disengaged, this' enough force to open the doors with or without the ventilation system operating.
Regulatory Commitment Change 98-08 in responses dated January 29 and May 1,1992, to Generic Letter 91-11, NSP committed to implementation of surveillance procedures to verify operability and correct configuration of inverters and instrument buses. Surveillance procedure SP2313 satisfies this commitment for Unit 2.
l On October 29,1998 Unit 1 entered a forced outage. Unit 2 was scheduled to be l
shutdown for refueling late on November 6. SP2313 was to be performed in the Unit 2 work schedule on November 7. The Unit 2 shutdown was subsequently.
postponed 2 days, which delayed implementation of the scheduled work activities.
When SP2313 was identified to be performed, it was past its late date by about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SP2313 was implemented as a verification surveillance to serve as a redundant I
check on the alignment of safeguards inverters and vital instrument buses. Other l
checks done during the period of the missed surveillance showed that there were no problems with the equipment. SP2313 will continue to be run as committed.
l This is a one time commitment change due to extraordinary unit status and I
schedule.
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l
- Mfot Num:
1999 --
0089 Dato
- 02/24/99 FROM
- Bruce Loesch/ Linda Reinhart Loc
- Prairie Island TO
- US NRC DOC CONTROL DESK Copy Num:-486 Holder : US NRC DOC CONTROL DESK SUBJECT : Revisions to CONTROLLED DOCUMENTS o*oe**************************************************************************
Proc: dure #
Rev Title Rsvisions:
==
USAR 16 UNDATED SAFETY ANALYSIS REPORT i
UPDATING INSTRUCTIONS I
Place this material in your Prairie Island Controlled Manual or File. Remove revised or cancelled material and recycle it.
Sign and date this letter i
in the space provided below within ten working days and return to Bruce l
Loesch or Linda Reinhart, Prairie Island Nuclear Plant, 1717 Wakonade Drive E.,
j Welch, MN 55089.
?
Contact Bruce Loesch (ext 4664) or Linda Reinhart (ext 4659) if you have any i
questions.
Received the material stated above and complied with the updating instructions 1
1 Date J
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