ML20065S759

Proposed Changes to Tech Specs Re Containment Sys,Refueling & Fuel Handling,Snubbers,Event Monitoring Instrumentation & ESF

ML20065S759
Person / Time
Site:	Prairie Island
Issue date:	10/29/1982
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20065S748	List: ML20065S746 ML20065S757 ML20065S759
References
NUDOCS 8211020007
Download: ML20065S759 (19)

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EXHIBIT B I.icense Ame'n dment Requent Dated October 29, 1982 Exhibit B, attached, consists of the following revised pages of the Appendix A Technical Specifications which incorporate the proposed changes, j PAGES TS-iii -

TS 3.6-2 Table TS.4.1-1 (Pg 5 of 5) . , ,

Table TS.4.4-1 (Pg 2 of 5) i Table TS.4.4-1 (Pg 4 of 5)

' '\s Table TS.4.4-1 (Pg 5 of 5)

TS.3.8-1 TS.3.8-3 TS.3.8-4 TS.3.12-1 (Pg 1 of 8)

TS.3.12-1 (Pg 2 of 8)

TS.3.12-1 (Pg 7 of 8)

TS.3.12-1 (Pg 8 of 8) ,

TS.3.15-1 TS.3.15-2 (new) ~

Table TS.3.15-1 .

Table TS.3.15-2 (new) ,s i

TS.4.5-2 4 ',

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APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES 1

i. TS TABLE TITLE 3.1-1 Unit 1 Reactor Vessel Toughness Data 3.1-2 Unit 2 Reactor Vessel Toughness Data.

,- 3.5-1 Engineered Safety. Features Initiation Instrument Limiting .

Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 4 .) 3.5-5 Instrument Operating Conditions for Ventilation Systems

'A 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System

(? ' 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation

, -3.9-2 Radioactive Gaseous Effluent Monitoring Instrumentation 3.12-1 Safety Related Snubbers l, 3.14-1 Safety Related Fire Detection. Instruments

! 3.15-1 Event Monitoring Instrumentation - Process j 3.15-2 Event Monitoring Instrumentation - Radiation

,4.1-1 Minimum Frequencies for Checks, Calibrations and Test of.

Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencie,s for Sampling Tests 4.2-1 Special Inservice Inspection Requirements.

4.4-1

. Unit 1 and Unit 2 Penetration Designation for' Leakage Tests .

4.10-1 . Radiation Environmental Monitoring Program (REMP)

. Sample Collection and Analysis 4.10-2 REMP L Maximum Values for the Lower Limits of Detection 4.10-3 REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection -

4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements .

4.17-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Rquirements i t.

4.17-3 Radioactive Liquid Waste Sampling and Analysis Program

'4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material in lp Liquid Effluents From Prairie Island Nuclear Generating i Plant (Per Unit)

5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition I

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TS.3.6-2 REV

4. Positive reactivity changes shall not be made by boron dilution when containment system integrity is not intact unless the boron concentration in the reactor is maintained 22100 ppm for the initial refueling and 22000 ppm for subsequent refuelings.

5. The vacuum breaker system shall be considered operable for containment system integrity when both valves in each of two vacuum breakers, including actuating and power circuits, are operable or when one vacuum breaker is daily demonstrated as-operable and the other has been inoperable for no more than 7 days under conditions for which containment integrity is required.

6. Automatic containment isolation valves listed in Table TS.4.4-1 shall be considered operable for containment system integrity when all auto-matic isolation valves, including actuation circuits, for each pene-tration are operable or the inoperable valve is deactivated in the closed position, or at least-one valve in each penetration having an inoperable valve is locked closed.

7. a. The 36-inch containment purge system double gasketed blind flanges shall be installed whenever the reactor is above cold shutdown.

b. The 18-inch containment inservice purge system double gasketed blind flanges shall be installed whenever the reactor is above cold shutdown except as noted below.

c. The inservice purge system may be operated above cold shutdown when required for safe-plant operation if the following conditions are met:

1. The debris screens are installed on the supply and exhaust ducts in containment.

2. Both valves shall satisfactorily pass a local leak rate test prior to use.

3. The two automatic primary containment isolation valves and the automatic shield building ventilation damper in each duct that penetrates containment shall be operable, including instruments and controls associated with them.

8. During maintenance, construction and testing activities, containment integrity is considered intact if the auxiliary building special vent zone boundary is opened intermittently, provided such openings are under direct administrative control and can be reduced to less than 10 square feet within 6 minutes following an accident.

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TABLE TS.4.1-1 '

(Page 5 of 5)

Channel Functional Response Description Check Calibrate Test Test R.e.m.a.r. k_s

35. Post-Accident !!onitoring M R NA NA Includes all those in FSAR Table Instruments , 7.7-2 and Tables TS.3.15-1 and TS.3.15-2 not included elsewhere in this Table

36. Steam Exclusion U R H NA See FSAR Appendix I, Section Actuation System I.14.6

37. Overpressure NA R R NA Instrument Channels for PORV Mitigation Control Including Overpressure System Hitigation System

38. Degraded Voltage NA R H MA 4KV Safeguard Busses -

39. Loss of Voltage ,

NA R M NA 4KV Safeguard Busses S -

Each Shift D -

Da ily U -

Weekly

!!onthly Q -

Quarterly r*

R -

Each refueling shutdown M P -

Prior to each startup if not done previous week b

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Prior to each startup following shutdown in excess of 2 days if not done in the previous 30 days ,L NA -

Not applicable y O

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TABLE TS.4.4-1 (Pg 2 of 5)

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UNIT 1 AND UNIT 2 PENETRATION DESIGNATION FOR LEAKAGE TESTS Penetration Type Penetration Penetration Designation of Test No. (Notes 1,2) Description (Note 3) Test Method 17 Loop B Hot Leg Sample ABSVZ C Pneumatic (5) 18 Fuel Transfer Tube (4) ABSVZ B Pneumatic 18 Bellows Annulus A OILT 19 Service Air (4) ABSVZ B Pneumatic 20 Instrument Air Exterior C Pneumatic 21 RC Drain Tank Gas ABSVZ C Pneumatic to Analyzer 22 Containment Air ABSVZ C Pneumatic Sample In 23 Containment Air ABSVZ C ' Pneumatic Sample Out 24 Spare None 25A Containment Purge ABSVZ B Pneumatic Exhaust (4) 25B Containment Purge ABSVZ B Pneumatic Supply (4) 26 Containment Sump "A" ABSVZ C Pneumatic Discharge 2 7A-1,- S team' Genera tor Sealed A OILT 27A-2 Blowdown Sample 27B Fire Protection (4) ABSVZ B Pneumatic

-(51 in Unit 2) 27-1, 27-2 Pressure Instrument ABSVZ B Pneumatic (27C-1 and 27C-2 in Unit 2)-

27D Spare None 28A,28B Safety Injection ABSVZ H Hydrostatic 29A,29B Containment Spray ABSVZ H ilydrostatic 30A,30B Containment Sump ABSVZ H . Hydrostatic Suction

? s TABLE TS.4.4'l (Pg 4 of 5)

REV UNIT 1 AND UNIT 2 PENETRATION DESIGNATION FOR LEAKAGE TESTS Penetration Type Penetration Penetration Designation of Test No. (Notes 1,2) Description (Note 3) Test Method 42B (53 in Inservice Purge ABSVZ C Pneumatic Unit 2) Supply Valves (6) 42B (53 in Inservice Purge Annulus B Pneumatic Unit 2) Supply Blind Flange (4) 42C (54 in Containment Heating ABSVZ B Pneumatic Unit 2) Steam (4) 42D, 42E Spare None 42F (42E in Heating Steam ABSVZ B Pneumatic Unit 2) Condensate Return (4) 42F (42E in Heating Steam ABSVZ B Pneumatic Unit 2) Return Vent (4) 42G Spare None 43A (52 in Inservice Purge ABSVZ C Pneumatic (5)

Unit 2) Exhaust Valves (6) 43A (52 in Inservice Purge Annulus B Pneumatic Unit 2) Exhaust Blind Flange (4) 43B,C,D Spares None 44 Containment Vessel ABSVZ B Pneumatic Pressurization (4)

45 Reactor Makeup to ABSVZ C Pneumatic Pressurizer Relief Tank 46A,46B Auxiliary Feedwater Sealed A OILT (46C,46D in Unit 2) 47 Electrical Sealed A OILT Penetration 47 Nitrogen-to Elect Sealed A OILT Penetration 48 Low Head SI ABSVZ H Hydrostatic 49A Instrumentation ABSVZ A OILT

) 49B (55 in Demineralized ABSVZ B Pneumatic .

Unit 2) Water (4)

! o TABLE TS.4.4-1 (Pg 5 of 5)

REV UNIT 1 AND UNIT 2 PENETRATION DESIGNATION FOR LEAKAGE TESTS Penetration Type Penetration Penetration . Designation of Test No. (Notes 1,2) Description (Note 3) Test Method 50 Post-LOCA Hydro- Exterior C ' Pneumatic gen Control Air Supply 50 Post-LOCA Hydro- Annulus C Pneumatic gen Control Vent 50 Sample to Gas Exterior C Pneumatic Analyzer Equipment Door Annulus- B Pneumatic (5)

Personnel Airlock Annulus B Pneumatic (5)'

Maintenance Air- Annulus B Pneumatic-(5) lock Notes:

1. Penetration numbers and description' identify the penetration. Additional information regarding penetrations is listed in FSAR Table 5.2-2.

2. Additional description of penetration function is contained in FSAR Appendix G.

3. Penetration Designations ABSVZ pipes connected to systems that are located in the Auxiliary Building Special Ventilation Zone

Exterior pipes-connected to systems that are exterior to the Shield l Building and ABSVZ Sealed pipes that will be sealed by water in. space between ' isolation l ' barriers following LOCA Annulus penetration that would leak to the Shield Building annulus following LOCA

4. These penetrations have blind flanges. Penetrations 18, 25A and 25B have

, blind flanges on inside only. Penetration 42B(53) and 43A(52) have a blind.

flange in the' annulus only.

5. Test pressure is applied in the opposite direction to the pressure that would exist when the component is required to perform its safety function.

! ' 6. The leakage test for this penetration is only required prior to use of the inservice purge system.

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.6 3.8 REFUELING AND FUEL HANDLING i-l Applicability

-Applies to operating limitations during ' fuel-handling and refueling operations.

1 j - Objectives i

To ensure that no incident could occur during fuel handling and refueling operations that would affect public health and safety.

i Specification-A. During refueling operations the following . conditions shall be satisfied:

1. The equipment hatch and at least'one door in each personnel air lock

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shall be closed. In addition, at least one isolation valve shall be j

operable or locked closed in each line which penetrates'the contain-

' ment and provides a direct path from containment atmosphere to the outside.

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'2. Radiation levels in fuel handling areas, the containment and the spent fuel storage pool areas-shall be monitored continuously.

3. The core suberitical neutron flux shall be continuously monitored by I

at least tw'o neutron monitors, each with continuous visual indication i

in the control room and one with audible indication in the containment, l which are in service whenever core geometry is being change.d. When core geometry is not being changed, at least one neutron flux monitor shall be in service.

,i j 4. During reactor vessel head removal and while loading and unloading i

fuel f rom the reactor, the minimum boron concentration of 2000 ppm

shall be maintained in.the. reactor coolant system. The required boron concentration shall be verified by chemical analysis daily.

'5. During movement of fuel assemblies or control rods out of the reactor

• vessel, at least 23 feet of water shall be maintained above the reactor vessel flange. The required water level shall be verified prior to moving fuel assemblies or control rods and at least once

every day while the cavity is flooded.

6. At least one residual heat removal ~ pump shall be operable and- running.

The pump may be shutdown for up to one , hour to facilitate movement of fuel or core components.

7. If the water level above the cop of the reactor vessel flange is less i

i than 20 feet, except for control rod latching and unlatching operations, both residual heat removal loops shall be operable.

! 8. If Specification 3.8. A.6 or 3.8. A. 7 cannot be satisfied, all fuel handling operations in containment shall be suspended, the contain-ment, integrity requirements of~ Specification 3.8.A.1 shall be satisfied, and no reduction in reactor coolant boron concentration shall be made.

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i - Basis The equipment and general procedures to be utilized during refueling are dis-cussed in-the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in inter-locks and safety features, provide assurance that no incident could occur during the and safety.(({ueling operations Whenever changes that are would result not being in in made a hazard to publicone core geometry, health flux l monitor.i.s sufficient. This permits maintenance of the instrumentation. Con-tinuous monitoring of radiation levels (B above) and neutron flux provides immediate indication of an unsafe condition. The residual heat ccmoval pump

. is used to maintain a uniform boron concentration.

The shutdown margin indicated in A.S. above will keep.the core subcritical, even if all contiol rods were . withdrawn from the core. During refueling,-the reactor refueling cavity is filled with approximately 275,000 gallons of' borated water. The boron concentra ion of this . water is sufficient to maintain the reactor auberitical by approximately 10% ak/k in the cold condition with all rods inserted, and will also maintal rodswereinserted'intothereactor.i2yhecoresuberiticalevenifnocontrol Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained. A.6.

above allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

No movement of fuel in the reactor is permitted until the reactor has been

, subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel.

l analysis.(3Jhedelaytimeisconsistentwith,thefuelhandlingaccident-The spent fuel assemblies will be loaded into the spent fuel cask for shipment toareprocessingplantaftersufficientdecayoffissionpropggts. In loading the cask into a carrier, there is a potential drop of 66 feet The cask will not be loaded onto the carrier for shipment prior to a 3-month storage period. At this time, the radioactivity has decayed so that a release of fission products from all fuel assemblies in the cask would result in off-site doses less than 10 CFR Part 100. It is assumed, for this dose' analysis that-12 assemblies rupture after storage for 90 days. Other assumptions are the same as those used in the dropped fuel assembly accident in the SER, Section 15.

The resultant doses at the site boundary are 94 Rems to the thyroid and 1

Rem whole body.

The Spent Fuel Pool Special Ventilation System is a safeguards system which maintains a negative pressure in theispent fuel enclosure upon detection of high area radiation. The Spent Fuel Pool Normal Ventilation system is auto-4 matica11y isolated'and exhaust air is drawn through filter nodules containing I

a roughing filter, particulate filter, and a charcoal filter before discharge ~

to the environment via.one of the Shield Building' exhaust stacks. Two completely redundant trains are provided. The exhaust fan and filter of each train are shared with the corresponding train of the Containment In-service Purge System.

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of ,the iodine adsorbers in each SFPSVS filter train. The charcoal adsorbers are installed to reduce the potential /

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e TS.3.8-4 REV release of radioiodine to the environment. The in place test results should indicate a HEPA filter leakage of less than 1% through DOP testing and a charcoal adsorber leakage of less than 1% through halogenated hydrocarbon testing. The laboratory carbon sample test results should indicate a radio-active methyl iodide removal efficiency of at least 90% under test conditions which are more severe than accident conditions. The satisfactory completion of these periodic tests combined with the qualification testing conducted on new filters and adsorber provide a high level of assurance that the emergency air treatment systems will perform as predicted in the accident analyses.

During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provide sufficient shielding.

The specifications require that at least one residual heat removal loop be in operation. This assures that sufficient cooling capacity is available to ccmove decay heat and maintain the water in the reactor below 140*F and that sufficient coolant circulation is maintained through the core to mini-mize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two residual heat removal loops operable when there is less than 20 feet of water above the vessel flange ensures that a single failure of the operating loop will not result in a complete loss of residual .

heat removal capability. With the reactor vessel head removed and 20 feet of water above the vessel flange, a large heat sink is available for core cooling. In the event of a failure of the operating RRR loop, adequate time is provided to initiate repairs or emergency procedures to cool the core.

The water level may be lowered to the top of the RCCA drive shaf ts for latching and unlatching. The basis for this allowance is (1) the refueling cavity pool has sufficient level to allow time to initiate repairs or emergency procedures to cool the core (2) during latching and unlatching the level is closely mon-itored because the activity uses this level as a reference point. (3) The time spent at this level is minimal.

References (1) FSAR Section 9.5.2

-(2) FSAR Table 3.2.1-1 (3) FSAR Section 14.2.1 (4) FSAR 'Section 9.6 (5) FSAR Page 9.5-2.0a

TABLE TS.3.12-1 (Page 1 of 8)

REV SAFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Radiation Snubber' Inaccessible Difficult -Area During No. Location Elevation (A or I) to Remove Shutdown UNIT I AFSH-22 A&B - Main and Aux- 773'-4k" A AFSH-36 iliary Steam 745'-7k" A AFSH-39 699'-10k" A AFSH-48 6 99 '-6 k" A MSDH-25 A&B 736'-6-7/16" A MSDH-26 A&B 756 '-7 k" A MSDH-29 756'-7k" A MSDH-30 736'-6-7/16" A MSH-48 A&B 739'-1-11/16" A MSH-62 A&B 735'-6" A '

MSH-63 - 756'-0" A MSH-64 743'-0" A MSH-65 748'-0" A MSH-66 753'-0" A MSH-67 743'-0" A MSH-68 A&B 755'-8" A MSH-69 A&B 748'-0" A MSH-101 729'-0" A MSH-102 735'-0" A

^MSH-103 A&B 737'-0" A UNIT II AFSH-2 Main and Aux- 749'-4" A AFSH-19 iliary Steam 745 '-7 k" A AFSH-20 745'-7k" A

AFSH-24 745'-6" A

AFSH-29 A&B 721'-1-9/16" A AFSH-33 707'-5" A

$ AFSH-39 6 96 '-6 k" A AFSH-40 696 '-6 k" A AFSH-44 750'-7 " A AFSH-46 750'-7" A i MSDH-17 739'-0" A MSDH-18 759'-0" A MSDH-19 739'-0" A MSDH-20 759'-0" A e

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REV SAFETY RELATED SNUBBERS Snubbers In High Accessible or Especially Radiation Snubber Inaccessible Difficult Area During No. Location Elevation - (A or I) to Remove Shutdown UNIT II MSH-23 A&B Main and Aux- 739'-1-3/16" A MSH-54 A&B iliary Steam 756'-0-1/16" I

!!SH-75 744'-0" A MSH-76 A&B 748'-0" A MSH-77 748'-0" A MSH-78 743'-0" A MSH-79 753'-0" A MSH-80 755'-0" A MSH-81 A&B 735'-9" A MSH-82 A&B 755'-8" A MSH-83 761'-13/16" I MSH-101 727'-0" A MSH-102 734'-0" A MSH-103A&B 736'-0" A UNIT I RHRRH-5 Safety Injection 723'-4k" I RHRRH-41 698'-11" I RHRRH-58 670'-0" A RHRRH-60 670'-0" A

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PPCH-160 718'-h" I RSIH-92 714'-11" I RSIH-93 714'-11" I RSIH-95 711'-2" I RSIH-96 711'-2" I RSIH-98 701'-2" I RSIH-163 717'-9" I RSIH-167 717'-9" I RSIH-413 A&B 722'-8" A RISH-414 716'-10" I RISH-442 717'-9 " I RSIH-469 707'-6 " I RSIH-476 707'-1-3/4" I SIRH-9 737'-0" I SIRH-11 718'-6" I SIRH-17 730'-0" 1 -

SIRH-18 730'-0" I SIRH-22 711'-4" I SIR!l-23 A&B 711'-4" I SIRH-26 705'-0" I

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TABLE TS.3.12-1 (Page 7 of 8)  !

REV SAFETY RELATED SNUBBERS Snubbers In High Snubber Accessible or Especially Radiation Inaccessible Difficult Area During No. Location Elevation (A or I) to Remove Shutdown UNIT II RCVCH-1396 Chemical & Vol 702'-10" I RCVCH-1505 Control 708'-6" I RCVCH-1513 710'-1" 1 RCVCH-1524 719'-1" I RCVCH-1574 721'-0" I RCVCH-1668 705'-5" I RCVCH-1373 722'-11" I RCVCH-1389 706'-1" I RRCH-253 704'-4" I RRCH-255 704'-8" I RRCH-261 707'-2" I RRCH-288 707'-2" I RRCH-291 704'-6" I RRCH-292 704'-7" I gg 7 CCH-304 Comp Cooling 717'-7" A CCH-373 712'-4" A CCH-376 A&B 700'-5" A CCH-377 703'-0" A CCH-378 708'-4" A CCH-380 670'-8" A CCH-381 A&B 671'-4" A CCH-397 699'-3" ~A CCH-398 A&B '~ "

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UNIT II -

CCH-161 Comp Cooling 717'-7" A CCH-166 719'-11" A CCH-167 720'-0" A CCH-172 720'-0" A CCH-173 708'-5" A CCH-176- 705'-3" A

! CCH-179 A&B 671'-4" A CCH-180 670'-8" A i CCH-181 708'-4" A l CCH-182 704'-2" A CCH-185 A&B 671'-4" A CCH-186 670'-10" gg 7 A RCSH-81 Containment Spray 760'-9" I

( RCSH-82 760'-8" I RCSH-83 A&B 732'-1" mnT II I CSH-75 A&B Containment Spray 731'-10" I CSH-76 752'-7" I CSH-79 751'-9" I CSH-82 A&B .731'-11" I CSH-83 767'-2" I

, CSH-84 767'-2" I i CSH-210 698'-0" I CSH-215 698'-0" A

CS H-224 710'-6" A

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o TABLE TS.3.12-1 (Page 8 of 8)

REV SAFETY RELATED SNUBBERS Snubbers In High Snubber Accessible or Especially- Radiation Inaccessible Difficult Area During No. 1.ocation Elevation (A or I) -to Remove Shutdown UNIT I RRHH-20 RHR 704'-3" A RRHH-62 705'-10" A MIIT II CVCRH-6 RHR 711'-0" I -

RRHH-21 704'-6" A MIIT II ZX-PSCH-127 ZX 707'-0" A O

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-l TS.3.15-1 REV 3.15 EVENT MONITORING INSTRUMENTATION Applicability Ipplies to plant instrumentation which does'not perform a protective function, but which provides information to monitor and assess important parameters during and following an accident. , l Objectives To ensure that sufficient information is available to operators to deter-mine the effects of and determine the course of an accient to the extent.

required to carry out required manual actions.

A. Specification - Process Monitors 1.TheeventmonitorihginstrumentationchannelsspecifiedinTable TS.3.15-1 shall be Operable.

2. With the number of Operable event monitoring instrumentation channels less than the Required Total Number.of Channels shown on  ;

Table TS.3.15-1, either restore the inop'erable channels to Operable status within seven 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 event monitoring instrumentation channels less than the Minimum Channels Operable requirements .of Table TS.3.15-1, either restore the minimum number of channels to

- 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 />.

B. Specification - Radiation Monitors

1. The event monitoring instrumentation channels specified in Table TS.3.15-2 shall be Operable.

2. With the number of Operable event monitoring instrumentation channels less than the Required Total Number of Channels shown on Table TS.3.15-2, either restore the inoperable channels to.

l Operable status within seven days, or prepare and submit a Special I

Report to the Commission pursuant to Technical Specification 6.7.B.2 within the next 30 days outlining the action taken, the cause of the inoperability, the plans and the schedule for restoring the system to Operable status.

3. With the number of Operable event monitoring instrumentation l channels less than the Minimum Channels Operable requirement of l Table TS.3.15-2, initiate the preplanned alternate method of monitoring the appropriate parameters in addition to submitting the report required in (2) above.

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, r TS.3.15-2 REV ,

Basis The operability of 'the event monitoring instrumentation ensures that sufficient information is available on selected plant' parameters to monitor and assess these variables during and following an accident.

This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations."

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k TABLE TS.3.15-1 EVENT MONITORING INSTRUMENTATION - PROCESS Required Total No. Minimum Channels Instrument of Channels- Operable

1. Pressurizer Water 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 Generator)

3. Reactor Coolant System Subcooling Margin *** 2 1 I 4. Pressurizer Power Operated Relief Valve Position 2/ valve 1/ valve (One Common Channel Temperature, One Channel Limit Switch per Valve, and One Channel Acoustic Sensor per Valve *) ,

5. Pressurizer Power Operated Relief Block Valve 2/ valve 1/ valve .

Position i (One Common Channel Temperature, one Channel Limit Switch per Valve, and One Channel Acoustic Sensor per Valve *)

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

• - A common acoustic sensor provides backup position indication for each pressurizer E; 4g power operated relief valve and its associated block valve, p;

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- The acoustic sensor channel is common to both valves. When operable, the acoustic y>

sensor may be considered as an operable channel for each valve. y

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• • • - Fully qualified in'put instrumentation is being installed in accordance with the 8,

NRC's TitI Action Plan. Until installation is completed..this function will be satisfied using the plant process computer.

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TABLE TS.3.15 EVENT HONITORING INSTRUMENTATION - RADIATION ,

Required Total No. Minimum Channels Instrument of Channels Operable

1. Containment Radiation Monitors (Ili Range) 2 1 '

2. Steam Relief Activity Monitors 1/ steam line 1/ steam line

3. Illgh Range Shield Building Ventilation Monitors 1 1 e

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TS.4.5-2 REV

3. Containment Fan Coolers Each fan cooler unit shall be tested during each reactor refueling shutdown to verify proper operation of all essential features including low motor speed, cooling water valves, and normal ventilation system dampers. Individual unit performance will be monitored by observing the terminal temperaturea of the fan coil unit and by verifying a cooling water flow rate of greater than or equal to 900 gpm to each fan coil unit.

4. Component Cooling Water System

a. System tests shall be performed during each reactor refueling shutdown. Operation of the system will be initiated by tripping the actuation instrumentation.

b. The test will be considered satisfactory if control board indica-tion and visual observations indicate that all components have operated satisfactorily.

5. Cooling Water System

a. System tests shall be performed at each refueling shutdown. Tests shall consists of an automatic start of each diesel engine and automatic operation of valves required to mitigate accidents including those valves that isolate non-essential equipment from the system. Operation of the system will be initiated by a simulated accident signal to the actuation instrumentation. The tests will be considered satisfactory if control board indicatlon and visual observations indicate that all components have operated satisfactorily and if cooling water flow paths required for accident mitigation have been established.

b. At least once each 18 months, subject each diesel engine to a thorough inspection in accordance with procedures prepared in conjunction with the manufacturer's recommendations for this class of standby service.

B. Component Tests

'l. Pumps

a. The safety injection pumps, residual heat removal pumps and contain-ment spray pumps shall be started and operated at intervals of one month. Acceptable levels of performance shall be that the pumps start and reach their required developed heat on minimum recircula-tion flow and the control board indications and visual observations indicate that the pumps are operating properly for at least 15 minutes.

b. A test consisting of a manually" initiated start of each diesel engine, and assumption of load within,one minute, shall be conducted monthly. i 1

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