ML20127J975

Submits Addl Info Re Implementation of NUREG-0737,Item II.F.2 Concerning Inadequate Core Cooling Instrumentation,In Response to Jr Miller 850513 Request.Draft Model Tech Specs Encl

ML20127J975
Person / Time
Site:	Prairie Island
Issue date:	06/18/1985
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 8506270206
Download: ML20127J975 (9)

Text

e es Northern States Power Company 414 Nicollet Mall Minneapons. %nnesota 55401 Telephone (612) 330 5500 June 18, 1985 Director Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 Prairie Island Nuclear Generating Plant Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Supplemental Information Related to Implementation of NUREG-0737 Item II.F.2. Inadquate Core Cooling Instrumentation The purpose of this letter is to provide the additional informa-tion and model Technical Specifications related to inadequate core cooling instrumentation requested in a letter dated May 13, 1985 l from Mr James R Miller, Chief, Operating Reactors Branch #3, Division of Licensing, USNRC.

Reactor Coolant ' Inventory Tracking System In recent conversations with our Project Manager in the Division of Licensing, we have noted that the description of the reactor vessel level instrumentation system (RVLIS) in the Safety Evaluation attached to Mr Miller's May 13, 1985 letter does not reflect some additional actions we have planned to complete our implementation of RVLIS:

a. RTD terminations inside containment in Unit No. I were originally made using terminal blocks. These terminations will be upgraded using fully qualified splices during the next refueling outage or cold shutdown of sufficient duration to complete the work.

b. Hardware installation was completed by May 30, 1985.

Completion of Emergency Operating Procedures and operator training is scheduled for completion by August 1, 1985.

c. The RVLIS installation utilizes a "high-reliability power source" as defined in NUREG-0737. This source is being further upgraded to a Class IE power supply, i

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Northem States Power Company Director of NRR June 18, 1985 Page 2 With respect to the six items requested in Mr Miller's May 13, 1985 letter, the following information is provided:

a. The Unit No. 1 RVLIS installation, except as noted above, is installed, functionally tested, and cali-brated. Test results are available for inspection during the planned NRC audit on June 20, 1985.

b. System preoperational testing and calibration was performed by personnel from Westinghouse Electric Corporation. All work was reviewed by Northern .

States Power Company. The system was found to per-form in accordance with design expectations and within design error tolerances.

c. The RVLIS installed by Northern States Power Company was designed by Westinghouse Electric Corporation.

The design was reviewed and found acceptable by the NRC Staff and approved for installation. This eval-

) uation is described in NUREG/CR-2628, March, 1982, The requirements of NUREG-0737, Item Section IV.

II.F.2 were found to be met by this system.

d. Model Technical Specifications covering RVLIS and other instrumentation for detection of inadequate core c oling are attached.

e. It is requested that the NRC Staff review the Prairie Island plant specific RVLIS installation for acceptance,

f. Emergency Operating Procedures (EOP's) used for operator training will conform to the technical con-tent of the NRC approved E0P guidlines for Westing-house reactors.

Subcooling Martin Monitor (SMM)

The current subcooling monitor installation is described in the SER attached to Mr Miller's May 13, 1985 report. Techni-cal Specifications for this installation are currently in place.

Alternative approaches to subcooling margin monitoring, using the new plant process computer system, are being studied.

Core Exit Thernocouples (CET)

The schedule specified in the SER attached to Mr Miller's May 13, 1985 letter now appears to be impossible to meet due to difficulties in computer hardware and software procurement.

Completion of the core exit thermocouple modification is

Northem States Power Company Director of NRR June 18, 1985 Page 3 dependent on hardware and software associated with the safety parameter display system modification. Regulatory Guide 1.97, Revision 2, instrumentation upgrades are also dependent on this hardware and software. We will discuss this matter in detail and provide the basis for schedule relief in sepa-rate correspondence.

Please contact us if you have any questions related to the infor-nation we have provided.

Dd David Musolf Manager - Nuclear Su ort Services c: Regional Administrator III, NRC Resident Inspector, NRC G Charnoff Attachment

b i

Northom States Power Company Attachment

Director of NRR l June 18, 1985

Prairie Island Nuclear Generating Plant Units 1 and 2 I

Inadequate Core Cooling Instrumentation Model Technical Specifications i

Model technical specifications specific to Prairie Island covering the requirements of the subcooling margin monitors, core exit thermocouples,

, and RVLIS are attached. The model specification for the subcooling margin monitor is essentially the same as presently exists in the Prairie Island Technical Specifications. The note explaining how the subcooling margin function was to be satisfied until qualified inputs are completed has been deleted since it is no longer applicable. The model subcooling margin monitor specification conforms with the NRC Standard Technical Specification i requirements of NUREG-0452.

The reactor vessel level instrumentation technical specification has been

incorporated as Section 3.15.C. It is consistent with the post accident i

monitoring specifications in section 3.15.A with the exception of the times allowed for inope mbility of the system. The seven days allowed for the number of operable. channels to be less than the required total number of channels has been increased to thirty days for the RVLIS instrumentation.

i Likewise the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> allowed for the number of operable channels to be l less than the minimum channels operable requirement has been increased to seven days.

. These times have been increased based on the nature and complexity of the reactor vessel level instrumentation and the time required to perform corrective maintenance on the system. Very little of the corrective maintenance on the RVLIS system can be performed by on-site personnel, most corrective maintenance will require the services of outside vendors.

In addition, recalibration of the system following maintenance will re-quire that the calibration data be sent off-site for confirmation. The combination of having to utilize outside vendors for corrective 1 maintenance and confimation of calibration data will make corrective 4

maintenance of the RVLIS system very time consuming, and impossible to complete within the times specified for the other post accident monitoring instrumentation. There are presently no requirements for the RVLIS system specified in the NRC Standard Technical Specifications. ,

The core exit thermocouple instrumentation technical specification has been incorporated as Section 3.15.D. The specification was based loosely on the core exit thermocouple requirements of Section 3.3.3.6 of the Standard Technical Specifications in that four thermocouple channels are required per core quadrant.

i i But our model specification differs from the requirements in the NRC i Standard Technical Specifications in the actions required if less than four channels per quadrant are available. Because of the limited number of thermocouples in each quadrant and the long lead times for some of the system cables (approximately 30 weeks), we are reluctant to agree to specifications that would require shutdown after seven days without four 1

Northem States Power Company Attachment Director of NRR June 18, 1985 thermocouples per quadrant. Instead we are proposing requirements that would allow continued operation with less than four thermocouples per quadrant provided corrective action is initiated and at least 16 thermo-couples are maintained operable for post accident monitoring. These requirements will provide adequate thermocouple coverage for determination of inadequate core cooling conditions while still providing enough flexi-bility to prevent unnecessary and extended unit. shutdowns because of thermocouple failures.

l9 W . . TS.3.15-2 C. Soecification - Reactor Vessel Level Instrumentation ,

1. The reactor vessel level instrumentation channels specified in Table TS.3.15-3 shall be operable.

2. tvich 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 thirty 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 hini: sum Channels Operable requirements of Table TS.3.15-3, either restore the minimum number of channels to Operable status within 7 da.ys 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 />. ,

D. Specification core Exit Thermocouple Instnzmentation*

1. Except as spe-ified in 3.15.D.2 and 3.15.D.3, a minimum of four core exit thermocouple instrumentation channels per core quadrant shall be operable.

2. With less than four core exit thermocouple instrumentation channels operable per core quadrant, a total of 16 core exit thermocouple instrumentation channels shall be maintained operable and action shall be taken to restore the core exit themocouple instrumentation system to four channels per core quadrant.

3. With less than four core exit thermocouple instrumentation channels operable per core quadrant and less than 16 total core exit thermocouple channels operable, restore the inoperable channels necessary to meet the requirements of 3.15.D.1 or 3.15.D.2 to operable status within 7 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 />.

• Effective upon operability of the Safety Parameter Display System Basis The operability of the event monitoring instrumentation ensures that sufficient information is available on selected pl: ant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recomr.endations of 3UREG-0578, "'OiI-2 Lassons Learned Task Force Status Report and Shor: Term Recommendations."

9 9 Q j( *

'W , .g TARI.E TS.3.15-1 EVENT HONITORit:0 INSTRHHENTATION - PROCESS f. CONTAINHENT .

l l

Required Total No. Hinimum Channels instrussent of Channels Operable i

1. Pressurizer Wastr Level 2 1

2. Auxiliary Feedwater Flow to Steam Cenerators 2/ steam gen 1/ steam gen j (One Channel Flow and one Channel Wide Range l Level for Eacle Steam Generator)

3. Reactor Coolant System Subcooling Hargin 2 1 l

4. Pressurizer Power Operated Relief Valve Position 2/ valve 1/ valve (One Comenon Channel Temperature, One Channel l Limit Switch per Valve, and one Channel Acoustic Sensor per Valve *)

l -

5. Pressurizer Power Operated Relief Block Valve Position 2/ valve 1/ valve (One Common Channel Temperature, One Channel 1.imit Switch per Valve, and one Channel Acoustic Sensor per Valve *)

6. Pressurizer Safety Valve Position 2/ valve 1/ valve (One Channel Temperatura per Valve and Comenon Acoustic Sensor **)

i l

7. a. Contalument Water I.evel (wide range) 2 1

b. Containment Water Level (narrow range) 2 1 l

S. Containment flydrogen Honitor (2 sensors per Channel) 2 1

( 9. Containment Pressure (wide range) 2 'l

• - A common acoustic sensor provides backup position indication for each pressurizer power operated h relief valve and its associated block valve. N

• • - The acoustic sensor channel is common to both valves. When operable, the acoustic sensor may be d considered as an operable channel for each valve. ,

I i

Table TS 3.15v3 EVENT HO:IITORING INSTRUMENTATION - REACTOR VESSEL LEVEL i

Required Total No. Minimum Channels Instrument of Channels Operable I 1.. Full Range 2 1

2. Dynamic Head Range 2 1 1

c D

m 0 _

DUlHrD12 ILY 1

i n

! E i F 0;

9 9 L J D

Tablo TS.4.1-1 --

y (Pcgo 4 ef 5)

Channel Functional itemponse Descridion Check calibrate Test Test Remarks

27. Turbine overspeed HA R H HA Protection Trip Channel

28. Deleted .

29. Deleted -

3D. Deleted

31. Soismic Honitors R R HA HA

32. Coolant Flow - RTD 8 R H HA Hypass Flowmeter

33. CHDH Cooling Shroud S HA' R HA FSAR page 3.2-56

34. Henctor Cap Exhaust Air S HA R HA Temperature 35a. Post. Accident Honitoring H R HA HA 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

• HA Includes all those in Table Radiation Instruments TS.3.15-2

c. Post-Accident Honitoring H R NA NA Includes all those in Table Reactor Vessel Level TS.3.15-3 Instrinnentation

d. Post-Accident Monitoring H R NA NA Core Exit Thermocouple ,

Instrumentation

36. Steam Exclusion Actuation W Y H HA See FSAR Appendix I, Section System 1.14.6

17. Overpressure Hitigation HA R R HA Instrument Channels for PORV System ControlIncludingoverpressurahmga yK Hitigation System j*

38. Degraded Voltage NA R H HA 8' ,N 4 KV Safeguard Husses & f'

39. Lous of Voltage HA R H HA O ,8 4 KV Safeguard Husses