ML20205M155
| ML20205M155 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/04/1986 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8604150111 | |
| Download: ML20205M155 (7) | |
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Northern States Pcwer Company 414 Nicollet Mall Minneapoks.Minnesott 55401 Telephone (612) 330-5500 April 4, 1986 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PIANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Additional Information Related to Implementation of NUREG-0737 Item II.F.2, Inadequate Core Cooling Instrumentation The purpose of this letter is to provide additional information related to the Prairie Island core exit thermocouple (CET) backup display and the model core exit thermocouple Technical Specifications proposed in our June 18, 1985 submittal.
In our October 22, 1985 letter on inadequate core cooling instrumentation, j
we committed to further evaluate the feasibility of upgrading the CET l
backup display to include four CET signals from each core quadrant and to provide either a schedule for upgrading the display or further justi-fication for deviation from the requirements of NUREG-0737, Item II.F.2.
We have completed our evaluation and now plan to provide a qualified backup display which will include at least four core exit thermocouples per quadrant.
The backup CET display will be part of the Inadequate Core Cooling Monitor (ICCM) package being purchased from Westinghouse. When completed, the ICCM modification will upgrade the existing RVLIS display to provide a qualified control room display of reactor vessel level, subcooling margin and CET readings. The ICCM will be in the same location as the existing RVLIS display, mounted in the incore rack in the control room.
The existing subcooling margin monitor will be removed.
Present plans call for installation of the ICCM to be completed by December of 1987. The eight qualified CET signals presently supplied t.o the subcooling margin monitor will be transferred to the ICCM by that date.
The remaining core exit thermocouples, presently connected to the existing plant process computer, will not be connected to the ICCM until completion of the new plant process computer and transfer of the core exit thermo-couples to the class 1E remote multiplexer units.
The CET's not connected to the qualified display must remain connected to the existing plant process computer to support reactor monitoring functions and cannot be g
paralleled to both the qualified display and the existing plant process 8604150111 860404
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PDR ADOCK 05000282
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Director of NRR Northem States Power Company April 4, 1986 Page 2 Because of delays in the plant process computer replacement project, described in our November 27, 1985 letter requesting an extension of time for completing NUREG-0737 Supplement No. 1 requirements, work on the plant precess computer replacement project not related to the safety parameter display system (SPDS) has been postponed and will not resume until January 1987.
As a result, transfer of the CET'.s to the new plant process computer class lE remote Multiplexer units and connection of the remaining CET's to the qualified display will not be completed until December 1988.
During the interim period, a display of eight fully qualified CET's will be maintained. The use of this interim display until completion of the SPDS was found acceptable by the NRC Safety Evaluation Report dated May 13, 1985. We believe that continued operation with an interim control room display of eight qualified CET's until completion of the new plant process computer will provide an adequate margin of safety because:
1.
A sufficient number of CET's is provided on the interim qualified display to adequately monitor post-accident core exit temperature distribution in the event the plant process computer display is lost. The cross-sectional area of a Prairie Island two-loop core is roughly half that of a Westinghouse four loop core and thus eight CET's (two per quadrant) are adequate to monitor the core exit for any localized overheating.
2.
A reliable method exists for obtaining core exit temperatures, independent of the plant process computer and the interim qualified displays. The capability exists to manually read all individual CET temperatures using a portable instrument in the bus rooms.
3.
The Emergency Operating procedure (EOP) responses would not be different if the number of CET's on the qualified display were greater than eight, since eight readings of core exit temperature would be sufficient to indicate what actions are necessary.
The model CET Technical Specifications proposed in our June 18, 1985 submittal were found to be unacceptable by the NRC Staff, and as a result we have agreed to revise the model CET Technical Specifications originally submitted. The revised model CET Technical Specifications are attached along with the revised model reactor vessel level indication system (RVLIS) Technical Specifications previously submitted in our October 22, 1985 letter.
The revised model CET Technical Specifications are consistent with the guidance provided in the Standard Technical Specifications.
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. Director of NRR N
April 4, 1986 Page 3 i
Please contact us if you have any questions related to the information we have provided.
Nw David Musolf Manager - Nuclear Support Services DMM/EFE/efe c: Regional Administrator-III, NRC 3
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NRR Project Manager, NRC Resident Inspector, NRC MPCA Attn:
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g TS.3.15-2 REV C. Specification - Reactor Vessel Level Instrumentation
- 1. The reactor vessel level instrumentation channels specified in Table TS.J.15-3 shall be operable.
- . With the number of Operable reactor vessel level. instrumentation channels less than the Required Total Number of Channels shown on Table TS.3.15-3, either restore the inoperable channels to Operable status within. fourteen days, or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 3. With the number of Operable reactor vessel level instrumentation channels less than the Minimum Channels Operable requirements of
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Table TS.3.15-3, either restore the minimum number of channels to l
Operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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Basis The operability of the event =onitoring instrumentation ensures that sufficient infor=atica is available on selected pl' ant parameters to monitor acd assess these variables during and following an accident. This capability is consistent with the recommendations of NUREG-0578, "IfI-2 Lessons Learned Task Force Status Report and Short Ters Recommendations."
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Core exit thermocouple readings necessary to meet the requirements of Specification 3.15.A are available from the Plant Process Computer', the Control Room Core Exit Thermocouple Display or if no other readout is available, from test equipment readings from the Core Exit Thermocouple Junction Boxes.
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TABl.E TS.3.15-1 j
EVENT HONITORINC INSTRUHENTATION - PROCESS & CONTAINHENT Required Total No.
Hinimum Channels instrument of Channels Operable 1.
Pressurizer Wa ter Level 2
1 2.
Auxiliary Feedwater Flow to Steam Generators 2/ steam gen 1/ steam gen (One Channel Flow and One Channel Wide Range Level for Each Steam Cenerator) 3.
Reactor Coolant System Subcooling Margin 2
1 4.
Pressurizer Power Operated Relief Valve Position 2/ valve 1/ valve (One Common Channel Temperature, One Channel I.imit Switch per Valve, and One Channel Acoustic Sensor per Valve *)
5.
Pressurizer Power Operated Relief Block Valve Position 2/ valve 1/ valve (One Common Channel Temperature, One Channel Limit Switch per Valve, and one Channel Acoustic Sensor per Valve *)
6.
Pressurizer hafety Valve Position 2/ valve 1/ valve (One Channel Temperature per Valve and Common.
Acoustic Sensor **)
7.
a.
Containment Water I.evel (wide range) 2 1
b.
Containment Water Level (narrow range) 2 1
8.
Containment flydrogen Honitor (2 sensors per Channel),
2 1
9 Containment Pressure (wide range) 2
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- 10. Core Exit Thermocouples 4/ core quadrant 2/ core quadrant lg
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- A common acoustic sensor provides backup position indication for each pressurizer power operated U
relief valve and its associated block valve.
- The acoustic sensor channel is common to both valves. When operable, the acoustic sensor may be y,
considered as an operable channel for each valve.
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A DD1Hrall II4y Table TS.3.15-3 EVENT MONITORING INSTRUMENTATION - REACTOR VESSEL LEVEL Instrument Required Total No.
Minimum Channels of Channels Operable 1.
Reactor Vessel Level Instrumentation
- 2 1
1
- Includes the full range and dynamic head range n
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D' f Table TS.4.1-1
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Channel Functional
Response
Description Check Calibrate Test Test Remarks 27.
Turbine overspeed NA R
H NA Protection Trip Channel
.' 28.
Deleted 29.
Deleted 30.
Deleted 31.
Seismic Honitors R
R NA NA 32.
Coolant Flow - RTD S
R H
HA Bypass Flowmeter 33.
CHDH Cooling Shroud S
NA' R
HA FSAR page 3.2-56 34.
Reactor Cap Exhaust Air S
NA R
NA Temperature 35a. Post-Accident Honitoring H
R NA NA Includes all those in Table Instruments TS.3.15-1 (except for contain-ment hydrogen monitors which are separately specified in this table)
- b. Post-Accident Honitoring D
R H
NA Includes all those in Table Radiation Instruments TS.3.15-2
- c. Post-Accident Honitoring H
R NA NA Includes all those. in Table hmgy g.
Reactor Vessel Level TS.3.15-3 Instrumentation y7 I
Y 36.
Steam Exclusion Actuation W
Y H
NA See FSAR Appendix I, Section System I.14.6 w;
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37.
Overpressure H8tigation NA R-R NA Instrument Channels for PORV System Control Including Overpressure Hitigation System 38.
Degraded Voltage NA R
H HA 4 KV Safeguard Busses 39.
Luas of Voltage NA R
H NA 4 KV Safeguard Husses
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