ML20216G037

Proposed Tech Specs 3.3 Re Engineered Safety Features

ML20216G037
Person / Time
Site:	Prairie Island
Issue date:	04/13/1998
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20216G013	List: ML20216G008 ML20216G033 ML20216G037
References
NUDOCS 9804170315
Download: ML20216G037 (10)

Text

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EXHIBIT A PRAIRIE ISLAND NUCLEAR GENERATING STATION Supplement to the License Amendment Request dated March 6,1998 Appendix A, Technical Specification Pages Marked Up Pages (shaded material to ta added, strike through mateiial to be removed)

TS.3.3-1 TS.3.3-3 B.3.3-2 *

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TS.3.3-1 l

3.3 NNGINEERED SAFETY FEATURES Acolicability Applies to the operating status of the engineered safety features.

Obiective To define those limiting conditions that are necessary for operation of engineered safety features: (1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.

Soecifications A. Safety Iniection and Residual Heat Removal Systems

1. A reactor shall not be made or maintained critical nor shall ,

reactor coolant system average temperature exceed 200*F unless l the following conditions are satisfied (except as specified in 3.3. A.2 below) :

a. The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 2500 ppm.

'b. Each reactor coolant system accumulator shall be OPERABLE vhen reactor coolant system pressure is greater than 1000 psig,

'OdRABILITY requires:

(1) The isolation valve is open (2) Volume is 1270 120 cubic feet of borated water (3) A minimum boron concentr tion of 1900 ppm (4) A nitrogen cover pressure of 740 1 30 psig

c. Two safety injection pumps are OPERABLE except that pump Centr:1 0.citch: 1.- th: centeel reer.-eb:11 :::t th: :quir m:nte Of5hippfylfiWd[ih Sections 3.3.A.3 and 3.3.A.4.

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d. Two residual heat removal pumps are OPERABLE.

e. Two residual heat exchangers are OPERABLE.

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TS.3.3-3 3.3.A.2.g. The valve position monitor lights or alarms for motor-operated valves specified in 3.3.A.1.g above may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided the valve position is verified once each shift.

3.  ?.t 1:::t ::: ::fety inj:: tie ;" p centr:1 cuitch 1.- th: ntr:1

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hat.R.CS whenever RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR except that both SI pumps may be run for up to one hour while conducting the integrated SI test ** when either of the following conditions is mets (a) There is a steam or gas bubble in the pressurizer and an isolation valve between the SI pump and the RCS is shut, or (b) The reactor vessel head is removed.

4. Seth fety inj:: tic: pu p centr:1 zuitch: ir the Centr:1 Recr ch:11 5: ir pu11:ut!!6Tsafepf]irijkst(ibpyumps*j*] hall {be[capablei pfj(pjI @ hilij @ p[)hA{RCS whene;ar RCS temperature is less than the temperature specified in the PTL.1 for disabling both safety injection pumps (except one or both pumps may be run as specified in 3.3. A.3 und 3.1. A.1.d. (2) ) .

5. Both reactor coolant system accumulators shall be isolated

• v'aenever RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR.

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• This specification does not apply whenever the reactor coolant system accumulators are depressurized or the reactor vessel head is removed.

• • 0ther SI system tests and operations may also be conducted under these conditions.

• • • This specification does not apply whenever the reactor vessel head is removed.

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B.3.3-2 3.3 ENGINEERED SAFETY FEATURES

. Based continued l

(1) Assuring with high reliability that the safety system will function '

properly if required to do so.

(2) Allowance of sufficient time to complete required repairs and testing using safe and proper procedures.

Assuming the reactor has been operating at full RATED THERMAL POWER for I at least 100 days, the magnitude of the decay heat decreases as follows I after initiating HOT SHUTDOWN.

Time After Shutdown Decay Heat. % of RATED POWER 1 min. 4.5 30 min. 2.0 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.62 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.96 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.62 Thus, the requiremcot for core cooling in case of a postulated loss-of-coolant accident while in the shutdown condition is significantly reduced below the requirements for a postulated loss-of-coolant acci-dent during POWER OPERATION. Putting the reactor in the HOT SHUTDOWN condition significantly reduced the potential consequences of a loss-of-coolant accident, and also allows mcre free access to some of the engineered safeguards components in order to effect repairs.

The accumulator and refueling water tank conditions specified are consistent with those assumed in the LOCA analysis (Reference 2).

Specification 3.3.A.3 allows use of an SI pump to perform operations required at low RCS temperatures; e.g. raising accumulator levels in order to meet the level requirement of Specification 3.3. A.1.b(2) or ASME Section XI tests of the SI system check valves.

Specification 3.3.A.3 also allows use of both SI pumps at low tempera-tures for conduct of the integrated SI test and other S,I system testo and operations providing the pumps run for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this case, pressurizer level is maintained at less than 50% and a positive means of isolation is provided between the SI pumps and the RCS to prevent fluid injection into the RCS. This isolation is accomplished by using either a closed manual valve or a closed motor operated valve with

• the power removed. This combinat ion of conditions under strict administrative control assure that overpressurisation cannot occur. The option of having the reactor vessel head removed is allowed since in this case RCS overpressurization cannot occur.

Maint ining both : fety injection pump Centrol Peer centr:1 0;;itcher ir pulleut, :: p: ified ir 2.2.* ', and ical ting the ::curulatere, ::

epteified ir 2.2.*.5, .:111 provide-assur:ne that 4h plant Operating rendition ;ill be bounded by the :: umptirn: 2pplied tc the determination of the OPPC ::tpcinte ir th: ::: injection tranzien n:1y ! . Tht:: :ttpcint: .111 ::tunt: tP POnV: upon :n Ecr pre: cure incre :: :: m: int ir RCE precrure :ithir the ::: ptchie Operating region cf the pr : cur / temper:tur: 'britti fracture) limit curver ir the PTLE. 5 The previcien: ef th::: crecific: tion ::: not pplicab1: ^-- the l

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crprcccuricction con net Occur. Maintaining the safety injection pumps incapable of injecting into the RCS, as specified in 3.3.A.3 and i 3.3.A.4, and isolating the accumulators, as specified in 3.3.A.5, will provide assurance that the plant operating conditions will be bounded by the assumptions applied to the determination of the OPPS setpoints in the mass injection transient analysis. These setpoints will actuate the PORVs upon an RCS pressure increase to maintain RCS pressure within the acceptable operating region of the pressure / temperature (brittle fracture) limit curves in the PTLR. The provisions of these specifications are not applicable when the reactor vessel head is remcVed since in that condition, RCS overpressurization can not occur.

The safety injection pumps are rendered incapable of injecting into the RCS by employing at least two independent means to prevent a pump start such that a single action will not result in an injection into the RCS. l This may be accomplished through the pump control switch being placed in pullout with a blocking device installed over the control switch that would prevent an unplanned pump start. -

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EXHIBIT B PRAIRIE ISLAND NUCLEAR GENERATING STATION j Supplement to the License Amendment Request dated March 6,1998 t

l Appendix A, Technical Specification Pages Revised Pages TS.3.3-1 TS.3.3-3 B.3.3-2 i B.3.3-2a l

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TS.3.3-1 3.3' ENGINEERED SAFETY FEATURES Apolicability Applies to the operating status of the engineered safety features.

Obiective To define those limiting conditions that are necessary for operation of engineered safety features: (1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.

Hoecifications A. Safety Iniection and Residual Heat Removal Systems l

1. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200*F unless the following conditions are satisfied (axcept as specified in i 3.3.A.2 below):

I I a. The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 2500 ppm.

b. Each reactor coolant system accumulator shall be OPERABLE l when reactor coolant system pressure is greater than 1000 psig. j I

l OPERABILITY requires:

(1) The isolation valve is open (2) Volume is 1270 i20 cubic feet of borated water (3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of 740 30 psig

c. Two safety injection pumps are OPIRABLE e : cept as specified in Sections 3.3.A.3 and 3.3.A.4.

d. Two residual heat removal pumps are OPEIABI.E.

e. Two residual heat exchangers are OPERABLE.

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. i TS.3.3-3 1

3.3:A.2.g'. The valve position monitor lights or alarms for motor-operated  !

valves specified in 3.3.A.l.g above may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l provided the valve positica is verified once each shift.

3. A maximum of one safety injection pump shall be capable of injecting into the RCC whenever RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR except that both SI pumps may be run for up to one hour while conducting the integrated SI test ** when either of the following conditions is met.

(a) There is a steam or gas bubble in the pressurizer and an i isolation valve between the SI pump and the RCS is shut, or (b) The reactor vessel head is removed.

4. No safety injection pumps *** shall be capable of injecting into the RCS whenever RCS temperature is less than the temperature I specified in the PTLR for disabling both safety injection pumps f (except one or both pumps may be run as specified in 3.3. A.3 and {

3.1.A.l.d.(2)). 1

5. Both reactor coolant system accamulators shall be isolated

• whenever RCS temperature is less than the Over Pressure Protection System Enable Temperature specified in the PTLR.

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• This specification does not apply whenever thn reactor coolant system accumulators are depressurized or the reactor vessel head is removed.

• • 0ther SI system tests and operations may also be conduc.ted under these conditions.

• • • This specification does not apply whenever the reactor vessel head is removed.

O B.3.3-2 3.3 ENGINEERED SAFETY FEATURES Bases continued (1) Assuring with high reliability that the safety system will function properly if required to do so.

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(2) Allowance of sufficient time to complete required repairs and testing using safe and proper procedures.

Assuming the reactor has been operating at full RATED THERMAL POWER for j at least 100 days, the magnitude of the decay heat decreases as follows after initiating HOT SHUTDOWN.

Time After Shutdown Decay Heat. % of RATED POWER 1 nin. 4.5 30 min. 2.0 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.62 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.96 1 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.62 I Thus, the requirement for core cooling in case of a postulated loss-of-coolant accident while in the shutdown condition is significantly reduced below the requirements for a postulated loss of-coolant acci-dent during POWER OPERATION. Putting the reactor in the HOT SHUTDOWN ,

condition significantly reduced the potential consequences of a loss-of-coolant accidcut, and also allows more free access to some of the engineered safeguards components in order to effect repairs.

l The accumulator and refueling water tank conditions specified are l consistent with those assumed in the LOCA analysis (Reference 2).

l Specification 3.3.A.3 allows use of an SI pump to perform operations l required at low RCS temperatures; e.g., raising accumulator levels in i order to meet the level requirement of Specifiestion 3.3.A.l.b(2) or i

ASME Section XI tests of the SI system check valves.

l i Specification 3.3.A 3 also allows use of both SI pumps at low tecpera-tures for conduct of the integrated SI test and other SI system tests t id operations providing the pumps run for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this case, pressurizer level is maintained at less than 50% and a positive means of isolation is provided between the SI pumps and the RCS to prevent fluid injection into the RCS. This isolation is accomplished by using either a closed manual valve or a closed motor operated valve with the power removed. This combination of conditions under strict administrative control assure that overpressurization cannot occur. The option of having the reactor vessel head removed is allowed since in this case RCS overpressurization cannot occur.

Maintaining the safety injection pumps incapable of injecting into the RCS, as specified in 3.3.A 3 and 3.3.A.4, and isolating the l accumulators, as specified in 3.3.A.5, will provide assurance that the i piant operating conditions will be bounded by the assumptions applied to 1 the determination of the OPPS setpoints in the mass injection transient I analysis. These setpoints will actuate the PORVs upon an RCS pressure increase to maintain RCS pressure within the acceptable operating region of the pressure / temperature (brittle fracture) limit curves in the PTLR.

The provisions of these specifications are not applicable when the J

B.3.3-2a

' reactor vessel head is removed since in that condition, RCS '

overpressurization can not occur,  ;

The safety injection pumps are rendered incapable of injecting into the  ;

RCS by employing at least two independent means to prevent a pump start l such that a single action will not result in an injection into the RCS, This may be accomnlished through the pump control switch being placed in pullout with a bl-eking device installed over the contrcl switch that would prevent an unplanned pump start.

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