ML20081B429

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Proposed Radiological Tech Specs Per NUREG-0737 (Generic Ltr 83-36) Re Fire Detection Instrumentation
ML20081B429
Person / Time
Site: Cooper Entergy icon.png
Issue date: 02/29/1984
From:
NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20081B427 List:
References
RTR-NUREG-0737, RTR-NUREG-737 GL-83-36, NUDOCS 8403090116
Download: ML20081B429 (16)
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.. ,- l RADIOLOGICAL TECHNICAL SPECIFICATIONS l

TABLE OF CONTENTS l Page No.

1.0~ DEFINITIONS 1-5 LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 6 - 22 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 23 - 26 SURVEILLANCE

. LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 27 - 46 3.2 PROTECTIVE INSTRUMENTATION 4.2 47 - 92

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A. Primary Containment Isolation Functions 47 B. Core and Containment Cooling Systems Initiation 47 and Control (CS, LPCI, HPCI, RCIC, ADS)

C. Control Rod Block Actuation 47 D. Radiation Monitoring Systems - Isolation and 48 Initiation Functions

1. Steam Jet Air Ejector Off-Gas System 48
2. Reactor Building Isolation and Standby Gas 48 Treatment Initiation
3. Lig'uid Radwaste Discharge Isolation 48
4. Main Control Room Ventilation 48
5. Mechanical Vacuum Pump Isolation ,

49 E. Drywell Leak Detection 49 F. Primary Containment Surveillance Information 49 Readouts G. Recirculation Pump Trip 49 H. Post-Accident Monitoring 49 l 3.3 REACTIVITY CONTROL 4.3 93 - 106 A. Reactivity Limitations A 93 B. Control Rods B 94 C. Scram Insertion Times C 97 D. Reactivity Anomalies D 98 E. Recirculation Pumps E 98 F. Restrictions F 98 C. Scram Discharge Volume G 98 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113 A. Normal Operation A 107 B. Operation with Inoperable Components B 108 C. Sodium Pentaborate Solution C 108 8403090116 840229 PDRADOCK05000g p _ _ - . . _ , -

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TABLE OF CONTENTS (Cont'd.)

Page No.

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SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS

-6.2.1.B NPPD Safety Review and Audit Board'(SRAB) 222 1 B.1 Membership 223 B.2 Meeting Frequency 223 B.3 Quorum 223 B.4 Review 223 B.5 Authority 224 B.6 Records 225

.B.7 Procedures 225 B.8 Audits 225 6.3 Procedures and Programs 226

  • 6.3.1 Introduction 226

,6.3.2 Procedures 226 6.3.3 Maintenance and Test Procedures 226 6.3.4 Radiation Control Procedures 226

.A -High Radiation Areas 226a 6.3.5 Temporary Changes 226a

, 6.3.6 Exercise of Procedures 226a 6.3.7 Programs 226a

.A Systems Integrity Monitoring Program 226a

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.B Iodine Monitoring Program 226a

.C* Environmental Qualification Program 226a

.D Post-Accident Sampling System (PASS) 227 l 6.4 Record Retention- 228 6.4.1 5 year retention ,

228

! , 6.4.2 Life retention i . 228  !

6.4.3 2 year retention 229 6.5 Station Reporting Requirements 230 6.5.1 Routine Reports 230

.A Introduction 230

.B Startup Report 230

.C Annual Reports 230

.D Monthly Operating Report 231 6.5.2 Reportable Occurrences 231

.A Prompt Notification with Written Followup 232

.B Thirty Day Written Reports 234 6.5.3 Unique Reporting Requirements 235 6.6 Station Operating Records 228 6.6.1 (5 year retention) 228 6.6.2 (life retention) 228 6.6.3 (2 year retention) 229

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS

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-3.2.D (cont'd.) 4. 2.D (cont ' d . )

5. Mechanical Vacuum Pump Isolation 5. Mechanical Vacuum Pump Isolation
a. The mechanical vacuum pump shall The instrument surveillance require-be capable of.being automatically ments are given on Tables 4.1.1, 4.1.2, isolated and secured by a signal of and 4.2.D.

high radiation in the main steam line tunnel whenever the main steam isolation valves are open.

b. .If the; limits of (3.2.D.S.a) are not met, the vacuum pump shall be isolated.

E. Drywell Leak Detection E.. Drywell Leak Detection The limiting conditions of Instrumentation shall be calibrated operation for the instrumentation and checked as indicated in Table that monitors'drywell l'eak 4.2.E.

detection are given in Table 3.2.E.

F. Primary Containment Surveillance F. Primary Containment Surveillance Information Information The limiting conditions of Instrumentation shall be calibrated operation for the instrumentation

  • and checked as indicated in Table 4.2.F.

that provides surveillance information readouts are given in Table 3.2.F.

G. Recirculation Pump Trip G. Recirculation Pump Trip

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The limiting conditions for Instrumentation shall be functionally operation for the instrumentation tested and celibrated as indicated on that trips the recirculation Table 4.2.G.

pumps as a means of limiting the

. consequences of a failure to scram during an anticipated transient are given in Table 3.2.G.

H. Post-Accident Monitoring H. Post-Accident Monitoring The limiting conditions for oper- Instrumentation shall ba functionally ation for the instrumentation that tested and calibrated as indicated monitors post-accident conditions on Table 4.2.H.

are given in Table 3.2.H.

1 COOPER NUCLEAR STATION TABLE 3.2.F.' '.

PRIMARY CONTAINMENT SURVEILLANCE INSTRUMENTATION

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fiinimum_ Number- Action Required When Instrument 'of Operable- _

Minimum Condition Instrument :I.D. No. Range' Instrument Channals Not ' Satisfied (1)

, Reactor Water Level NBI-LI-85A '-150" to +60" 2 A,B,C:

3 NBI-LI-85B -150" to +60" Reactor Pressure RFC-PI-90A 0 - 1200 psig 2 A,B.C

RFC-PI-90B 0 - 1200 psig_

Drywell Pressure PC-PI-512A 0 - 80 psia 2 A,B,C PC-PR-512B O'- 80 psia PC-PR-IA 0 - 250 psig 1 F PC-PR-1B 0 - 250 psig 50 - 170*F

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Drywell Temperature PC-TR-503' '2 A,B,C

PC-TI-505 50 - 350*F ,

Drywell Level PC-LR-1A 0 - 100' 1 F PC-LR-1B 0 - 100'-

4 i Suppression Chamber / Torus PC-TR-21A 0 - 300*F 2 A,B,C Ol Air Temperature PC-TR-23, Ch I & 2 0 - 400*F 1

Suppression Chamber / Torus PC-TR-21B 0 - 300*F 2 A,B,C l Water Temperature PC-TR-22, Ch 1 & 2 0 - 400*F Suppression Chamber / Torus l Water Level PC-LI-10 (-4' to +6') 2 A,B,C PC-LR-ll (-4' to +6')

PC-LI-12 -10 to + 10" 2 A,B,C,E PC-LI-13 -10" to +10"

! PC-LR-1A 0 - 30' -

1 F i PC-LR-1B 0 - 30'

, Suppression Chamber / Torus PC-PR-20 0 - 2 psig 1 B,C j Pressure Control Rod Position N.A. Indicating Lights 1 A,B,C,D l Neutron Monitoring N.A. S.R.M., I.R.M., 1 A,B,C,D 1

LPRM j 0 - 100% power

! Torus to Drywell PC-dPR-20 0 - 2 psid 1 A,B,C,E l Differential Pressure

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Suppression Chamber / PC-PR-20/513 (2)_ 0 - 2 psig 1 ,

Drywell Pressure (AP) _

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. NOTES FOR TABLE-3.2.F 1.. The following actions will be taken if the minimum number of operable instrument channels as required are not available.

A. 'From and after the date that one of these para' meters is reduced to one indication, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made operable.

B. From and after the date that one of these parameters is n -

indicated in the control room, continued operation is peic;.ssible during the succeeding seven days unless such instrumentation is sooner made operable.

C. If-the requirements of A and B above cannot be net,.an orderly shutdown shall be initia~ted.within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. These surveillance instruments are considered to be redundant to each other.

E. In the event that both channels are inoperable and. indication cannot be restored in six.(6) hours, an orderly shutdown shall be initiated and the reactor shall be in Hot Shutdown in six.(6) hours and in a Cold Shutdown condition in the following eighteen (18) hours.

F. From and after the date that one of these parameters is not indicated in the control room, either restore the inoperable component (s) to operable status within 7 days of the event, or prepare and submit a Special Report to the Commission within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to operable status.

  • i 2. These instrument' channels are utilized as inputtfor the performance of a manual calculation to provide the second Torus to Drywell differential pressure indication.

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_ COOPER NUCLEAR STATION TABLE 3.2.H

- POST-ACCIDENT MONITORING INSTRUMENTATION REQUIREMENTS

  • Minimum Number of Action' Required When Instrument. Operable Instrument Component Operability

-Instrument ID Number Range Channels Is Not Assured Elevated Release Point (ERP) RMP-RM-3B 1.00E-2 to 1 A Monitor (High Range Noble 1.00E+5 pc/cc Gas) (Xe-133-Equivalent)

Turbine Building Ventilation RMV-RM-20B 1.00E-2 to 1 A Exhaust Monitor 1.00E+5 pc/cc -

(High Range Noble Cas) (Xe-133 Equivalent)

Radwaste/ Augmented Radwaste RMV-RM-30B' l.00E-2 to 1 A Exhaust Monitor 1.00E+5 pc/cc 4 (High Range Noble Gas) (Xe-133 y Equivalent) 1 Primary Containment Gross RMA-RM-40A 1.0-1.0E+7 R/Hr. 2 A Radiation Monitor RMA-RM-40B 1.0-1.0E+7 R/Hr.

  • Note: Other Post-Accident Monitoring Instrumentation is located in Table 3.2.F-Drywell Pressure, PC-PR-lA and IB, Drywell Level PC-LR-IA and IB, Suppression Chamber / Torus Water Level PC-LR-1A and IB 9

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t  !-NOTES 1FOR TABLE 3.2.H

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' Action:.

A. With the number of operable componer.ts less than required by the minimum c< .

~ component operable requirements -initiate the preplanned alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

, 1) ' > either restore the ir.7perable component (s) to operable status ~within

.- H7. days of-the event, or

. 12); ' prepare and submit a Special Report to the Commission within 14 days following the event' outlining the action taken, the cause of the

. . inoperability and the plans and schedule for restoring the system to operable status.

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COOPER NUCLEAR STATION .

TABLE 4.2.F- -

PRIMARY CONTAINMENT. SURVEILLANCE INSTRUMENTATION TEST AND CALIBRATION FREQUENCIES Instrument Instrunent I.D. No. -

Calibration' Frequency Instrument Check Reactor. Water Level NBI-LI-85A '

Once/6 Months Each Shift NBI-LI-85B Once/6 Months .Each Shift Reactor Pressure RFC-PI-90A Once/6 Months Each Shift RFC-PI-90B Once/6 Months .Each Shift Drywell Pressure PC-PR-512A Once/6 Months Each Shift PC-PI-512B Once/6 Months Each Shift PC-PR-1A ~ Once/6 Months Each Shift PC-PR-1B Once/6 Months Each Shift

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Drywell Temperature PC-TR-503_ once/6 Months Each Shift PC-TI-505 Once/6 Months Each-Shift Drywell Level PC-LR-1A- -Once/6 Months Each Shift PC-LR-1B -

Once/6 Months Each Shift 1 Suppression Chamber / Torus PC-TR-21A' Once/6 Months Each Shift

? Air Temperature PC-TR-23, Ch. 1&2 Once/6 Months

  • Each Shift Suppression Chamber / Torus PC-TR-21B_ Once/6 Months Each Shift j Water Temperature PC-TR-22, Ch. 1&2 Once/6 Months Each Shift Suppression Chamber / Torus PC-LI-10 Once/6 Months Each Shift l

Water Level .

Once/6 Months Each Shift PC-LR-ll PC-LI-12 ,i Once/6 Months Each Shift PC-LI-13 Once/6 Months Each Shift PC-LR-1A Once/6 Months Each Shift PC-LR-1B Once/6 Months Each Shift-Suppression Chamber / Torus PC-PR-20 Once/6 Months Each Shift Pressure Control Rod Position N.A. N.A. Each Shift Neutron Monitoring (APRM) N.A. Once/ Week Each Shift Torus to Drywell PC-dPR-20 Once/6 Months Each Shift Differential Pressure Suppression Chamber / PC-PR-20/513 (2) Once/6 Months Each Shift Drywell Pressure (AP)

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COOPER NUCLEAR STATION .

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TABLE 4.2.H .

CALIBRATION. FREQUENCY FOR POST-ACCIPENT MONITORING INSTRUMENTATION

  • I i Item Function -

Calibration Item 'ID Number ' Pest Frequency Frequency Elevated Release Point (ERP) .

Monitor (High Range Noble RMP-RM-3B Once/ Month Once/ Cycle Cas)

Turbine Building Ventilation RMV-RM-20B Once/ Month Once/ Cycle Exhaust Monitor ^'

(High Range Noble Cas)

Radwaste/ Augmented Radwaste RMV-RM-30B Once/ Month _Once/ Cycle Exhaust Monitor 1

(High Range Noble Cas)

)

l Primary Containment Cross RMA-RM-40A,B Once/ Month Once/ Cycle 4 Radiation Monitors **

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  • Note: Other Post-Accident Monitoring Instrumentation calibration requirements are in Table 4.2.F-

, Drywell Pressure, PC-PR-1A and IB, Drywell Level PC-LR-1A and IB, Suppression Chamber / Torus i Water Level PC-LR-1A and IB. .

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    • CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below l 10 R/hr with an inatalled or portable gamma source.

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23.2! BASES (Cont'd)

J F. , Primary Containment Surveillance Information (Forleach parameter monitored,Las listed in Table 3.2.F, there are two (2) l channels of; instrumentation. By comparing readings between the two (2)

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c hannels, Ja near continuous surveillance of instrument performance is available.

fAny deviation.in' readings will initiate an early recalibration, there-

.by maintaining _the' quality of:the instrument readings.-

The operability of the reactor water level instrumentation in Tables 3/4.2.F ensures that sufficient information is available to monitor and assess-accident-

. situations. .

LG. Recirculation Pump Trip The recirculation rnap trip has been added as a means.of limiting the con-sequences of the1 unlikely occurrence of a failure to scram during an antici-pated transient. The response of the plant to this postulated event. falls within the envelope of study events given in General Electric Company Topical Report, NEDO-10349, dated March, 1971.

H. Post-Accident Monitoring ,

The post-accident monitoring instrumentation supplements existing instru-meatation that was designed to monitor primarily tne normal operational ranges of these paremeters. 1 Post-accident monitoring instrumentation

. ~ provides information for the ranges that may exist during the extreme

, conditions postulated to occur during and after some accidents.

REFERENCES

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1. " Low-Low Set Relief Logic System and Lower NSIV Water Level Trip for Cooper Nuclear Station", NEDE 22197, December 1982, General Electric Company.
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. -87a-

.6.3 (cont'd)

Position on_ Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License DPR-46 dated October 24, 1980.

B. By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for'all safety-related electrical equipment'in sufficient detail to

, document the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

D. Post-Accident Sampling System (PASS)

A program shall be established to' ensure the capability to obtain and analyze reactor' coolant, radioactive iodines and particulates in plant

. gaseous effluents, and' containment atmosphere samples under accident conditions. This program shall include training of personnel, procedures for sampling and analysis and provisions for operability of sampling and analysis equipment.

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[ Attachment 2 Revised Technical Specifications for Fire Detection Instruments Revised Page: 216k Nebraska Public : Power District requests a revision to the Technical Specifications in order to add new fire detector instrument numbers, and to correct two existing ID numbers on Table 3.14, as shown on t,he attached pages.

Evaluation of this Revision with Respect to 10CFR50.92 ,

A. The enclosed .Techn,1 cal Specification . chcnge is judged to involve no significant hazards based on the following:

1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluation:

The addition of four fire. detectors to those presently existing in the area of the Service Water Pumps will enhance protection (against fires). It therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed license amendment create the possibility for a new or different kind of accident from any accident previously evaluated?

Evaluation:

This equipment provides alarms (they cannot by themselves activate the suppression system) as an aid to fire detection. Previous accident analyses remain bounding with no creation of a new or different kind of accident.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Evaluation:

The margin of safety is enhanced because this improved fire detection capability provides additional aids for the operator in mitigating the effects of a fire.

B. Additional basis for proposed no significant hazards consideration determination:

The commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48CFR14870). The examples include:

"(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications. . ." The proposed Technical Specification adds new fira protection detectors which

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were installed in order to meet . the 20-foot separation requirement

. between Division I'and Division II pumps, valves, and conduits. Since it

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- reflects the installation of improved monitoring capability this Technical Specification change is an example of the type discussed above.

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INSTRUMENT LOCATION INSTRUMENT ID NO.

2 Control Room FP-SD-17-1 FP-SD-17-2 FP-SD-17-3 3 Cable Spreading Room FP-SD-16-1 FP-SD-16-2 FP-SD-16-3 FP-SD-16-4 FP-SD-16-5 FP-SD-16-6 Cable Expansion Room FP-SD-16-7 FP-SD-16-8 4~ Switchgear Rooms DC Switchgear Rooms FP-SD-15-2, ,

FP-SD-15-3 Critical Switchgear Room FP-SD-22-1 FP-SD-22-2 5- Station Battery Rooms FP-SD-15-1 FP-SD-15-4

  • FP-SD-15-1A FP-SD-15-4A 6 Diesel Generator Rooms FP-SD-10-1 FP-SD-10-2

- FP-SD-10-3 FP-SD-10-4 CO2-SD-DG-1A CO2-SD-DG-1B CO2-SD-DG-1C CO2-SD-DG-1D CO2-SD-DG-2A CO2-SD-DG-2E CO2-SD-DG-2C CO2-SD-DG-2D 7 ~ Diesel Fuel Storage Rooms CO2-TD-DG-1A CO2-TD-DG-1B

'8. Safety Related Equipment not in Reactor Building RHR Service Water Booster Pumps FP-SD-14-3 Emergency Condensate Storage Tanks FP-SD-14-1 Service Water Pumps FP-SD-32-1 FP-SD-32-2 FP-SD-32-3 FP-SD-32-4 FP-FD-32-5 FP-FD-32-6 9 Auxiliary Relay Room & Reactor Protection System Rooms Auxiliary Relay Room FP-SD-15-9 Reactor Protection System Room 1A FP-SD-15-7 Reactor Protection System Room IB FP-SD-15-8

-216k-

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Attachment 3 m

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Revised Technical-Specifications for an Editorial Change Re01 sed Page: 58

' Nebraska . Public Power - District requests a change to the Technical i Specifications on page 58. This request changes "I-15" Hg "to "$15" Hg

- Vacuum".-

Evsluation of this Revision'with Respect to 10CFR50.92

- A. The enclosed; Technical Specification . change is judged to involve no ,

significant hazards' based on the following:

1.- Does the proposed license amendment involve a'significant increase in c the probability or consequences of an accident previously evaluated?

Evaluation:

Because'this change is of an editorial nature to clarify a_ number on

-a list it does not involve's significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed license amendment create the possibility for a new or different kind of- accident from any accident previously evaluated?

Evaluation:

Because this change is editorial in nature it does not create the possibility of a new or different kind of accident from any accident previously evaluated. , ,

3 .' Does the proposed amendment involve a significant reduction in a j . margin of safety?

. Evaluation:

Since' this - change involves the clarification in the current Technical Specifications it does not involve a significant reduction

. in'a margin of safety.

-B. Additional basis for proposed no significant hazards consideration

' determination:

The' commission has provided guidance concerning the application of the standards.for determining.whether a significant hazards consideration exists by providing certain examples (48CFR14870). The examples include:

'.'(i) A purely administrative change to Technical Specifications. . ."

- The change enclosed is clearly encompassed by this example.

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COOPER NUCLEAR STATION. 4.

TABLE 3.2.B (PACE 6) '

REACTOR CORE ISOLATION COOLING. SYSTEM _(RCIC) CIRCUITRY REQUIREMENTS Minimum Number of Action Require'd When

. Instrument Operable Components Component Operability:

Instrument I.D. No. Setting Limit Per Trip ' System (1) Is Not Assured RCIC High Turbine RCIC-PS-72, A & B <25 psig 1(2) A Exhaust Press.

RCIC Low Pump Suction RCIC-PS-67-1 ;115" Hg Vacuum 'l(2) ]

Press.

RCIC Steam Line Space RCIC-TS-79, A,B,C,6D .1200*F 2(4) A' Excess Temp. RCIC-TS-80, A,B,C,6D RCIC-TS-81, A,B,C,&D RCIC-TS-82, A,B,C,6D RCIC Steam Line High RCIC-dPIS-83 & 84 370" j,Sj,620" H90 1 A

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AP .

RCIC Steam Supply RCIC-PS-87, A,B.C,6D 3,50 psig 2(2) A di Press. Low

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RCIC Low Pump RCIC-FIS-57 -

2,40 gpm -

1(2) A l

Disch. Flow Pump Discharge Line CM-PS-269 3,10 psig (3) D Low Pressure i

RCIC Turbine Condition- RCIC-TDR-K9 13.5 3,T < 16.5 (3) E-al Supervisory Alarm Timer Reactor Low Water 10A-K80, A & B 2,-37" Indicated Level 2(2) A Level 10A-K79, A & 3 (NBI-LIS-72, A,B,C,

& D)

. Reactor liigh Water NBI-LIS-101, A & C #2 j,+58.5 Indicated Level 2(2) A Level ,

RCIC Steamline High RCIC-TDR-K12 2.7fr,13.3 see 1 A AP Actuation Timer RCIC-TDR-K32

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