ML20141B918

Proposed Tech Specs Re Update of Design Basis Accident Containment Temp & Pressure Response

ML20141B918
Person / Time
Site:	Monticello
Issue date:	06/19/1997
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20141B908	List: ML20141B902 ML20141B911 ML20141B918 ML20141B927 ML20141B934
References
NUDOCS 9706240205
Download: ML20141B918 (10)

Text

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Exhibit B Monticello Nuclear Generating Plant Revision No. 2 to License Amendment Reauest Dated January 23.1997 Technical Specification Pages Marked Up With Proposed Working Changes Exhibit B consists of the existing Technical Specification pages marked up with the proposed changes. Existing pages affected by this change are listed below:

Pages 112 113 176 l

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9706240205 970619 PDR ADOCK 05000263 P

PDR 1

Bases M /4.5 Continued:

automatically controls three selected safety-relief valves although the safety analysis only takes credit for tuo valves.

It is therefore appropriate to permit one valve to be out-of-service for up to 7 days witho.it materially reducing system reliability.

B. Rih' Intertie Line An intertie line is provided to connect the RilR suctior$ line with the two R11R loop return lines.

This four-inch line is equipped with three isolation valves. The purpose of this line is to reduce the potential for water hammer in the recirculation and RilR system when required to cooldown with an isolated or idle recirculation system.

The isolation valves are opened during a cooldown to ensure a uniform cooldown of the RilR inj ection piping.

If one recirculation loop is isolated or idle, these valves and associated piping allow the operating loop to cool the isolated or idle loop.

The RIIR loop return line isolation valves receive a closure signal on LPCI initiation.

In the event of an inoperable return line isolation valve, there is a potential for some of the LPCI flow to be diverted to the broken loop during a loss of coolant accident.

Surveillance requirements have been established to periodically cycle the Ri!R intertie line isolation valves.

In the event of an inoperable RilR loop return line isolation valve, either the inoperable valve is closed or the other two isolation valves are closed to prevent diversion of LPCI flow.

The RllR intertie line flow is not permitted in the Run Mode to eliminate 1) the need to compensate for the small change in jet pump drive flow or 2) a reduction in core flow during a loss of coolant accident.

C.

Containment Spray / Cooling Systems Two containment spray / cooling subsystems of 't$e PHR s'ysIem~afe id~odided to' remove heat' energy from the" M

le '

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~ " ' '

Ontitinme containment and control torus and drywell pressure in the event of a loss of coolant accidant.

A i

containment spray / cooling subsystem consists of 2 RHR service water pumps, a PHR heat exchanger. 2 PHR on M nment and valves and piping necessary for Torus cooling and Drywell Spray.

Torus Spray is not R Pumps i

pumps,

<ind valves 4 considered part of a containment spray / cooling subsystem.

Placing a containment spray /coolin9 ed art of subsys t em into operation follwing a loss of coolant accident is a mamal operation.

<t centainmes Limited to } The most degradad condition for long term containment heat removal following the design basis loss of g

coolant accident results from the loss of one diesel generator.

Under these conditions, only one RHR

apability.

pump and one RHH service water pump in the redundant division can be used for containment spray / cooling. The containment temperature and pressure have been analyzed under these conditions assuming service water and initial suppression pool temparature are both 96F.

Acceptable margins to

,ither subs) containment design conditions have been demonstrated. Tharefore the containment spray / cooling system I servi <:e s

Jater pump C is more than ample to provide the required heat removal capability.

Refer to USAR Sections 5.2.3.3.

• maininl r

i 6.2.3.2.3, and B.4.1.3.

three pumps repair period i During normal plant operation, the containment spray / cooling system provides cooling of the suppression pe r gom -

pool water to maintain temperature within the limits specified in Specification 3.7.A.I.

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3.5/4.5 Bases 112 REV r

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Bases 3.5/4.5 Continued:

Tim RI R.se s ie ----

=pprcraicr tintained

-a&-SpOGM The surveillance requirements provide adequate assurance that the containment spray / cooling system will -

be operable when required. The head and f low requirerrent s specified for the RHR service water pumps provide assurance that the minimum required service water flow can be supplied to the RHR heat exchangers for tha most degraded condition for long-term containment heat removal following the design D.

RCIC basis loss of coolant accident.

The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor isolated from the turbine and when the feedwater system is not available. The pumping capacity of the RCIC system is sufficient to maintain'the water level above the core without any other water system in operation.

If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts.

The system may also be manually initiated at any time.

The llPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefore, the specification calls for an operability check of the llPCI system should the RCIC system be found to be inoperable.

The surveillsace requirements provide adequate assurance that the RCIC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

E.

Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times.

It is during refueling outages that major maintenance is performed and during such time that all core and containment spray / cooling subsystems may be out of service. This specification allows all core and containment spray / cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.

Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may by necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7.A.l.

In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.

3.5/4.5 Bases 113

Bases Continued:

Vent system downcomer submergence is three feet below the minimum specified suppression pool water level This length has been shown to result in reduced postulated accident loading of the torus while at t% same time assuring the downcomers remain submegd under all seismic and accident conditions and possess adequate condensation effectiveness.

The maximum temperature at the end of blowdown tested during the liumboldt Bay (I)' and Bodega Bay (2) tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170*F.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160*F during any period of relief valve operation with sonic conditions at the discharge exit.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open.

This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall he separated from that of the stuck-open relief valve to assure mixing and unifonnity of energy insertion to the pool.

For an initini maximum suppression chamber water terrperature of 96 F and conditions which lead to

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plement of-cc minimum containnant pressure, adequate net positive suction head (NPSH) is maintained for the core 3pS)

Contalument [ spray, RHR, and HPCI pumps under loss of coolant accident conditions.

E the few core spray, I-o.

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ro r.

hours after a loss-of-coolant accide D $ __...

R loop be inoperable and should the contain-ry....____,

..__,r____...3 ment pressure be reduced t teospieric pressure thro Duginy-means adequate NPSil would not be avail able.

Since a e y degraded condition must exist, the period of vulnerability to this event i i

y Specification 3.7.A.1.b by limiting the suppression pool initia1 D perature_and the re Tretiud uf uvetui.luu -ith uun Inuyu uble-RHR-lvvy.

(1)

Robbins, C. II. " Tests of Full Scale 1/48 Segment of the ilumboldt Bay Pressure Suppression Containment," GEAP-3596, November 17, 1960.

(2)

Bodega Bay Preliminary llazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

(3) General Electric NEDE-21885-P, " Hark I Contaitunent Program bowncomer Reduced Submerg nce Functional Assessment Report", June, 1978.

3.7 BASES 176

4 Exhibit C Monticello Nuclear Generating Plant Revision No. 2 to License Amendment Reauest Dated January 23.1997 Revised Monticello Technical Specification Pages Exhibit C consists of revised Technical Specification pages that incorporate the proposed changes. The pages included in the exhibit are as listed below:

Paaes 112 113 176 l

C-1

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Bases 3.5/4.5 Continued:

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automatically controls three selected safety-relief valves although the safety analysis only takes credit for two valves.

It is therefore appropriate to permit one valve to be out-of-service for up to 7 days without materially reducing system reliability.

l B.

RHR Intertie Line An intertie line is provided to connect the RHR suction line with the two RHR loop return lines. This f our -inch line is equipped with three isolation valves.

The purpose of this line is to reduce the potential for water f

hammer in the recirculation and RHR system when required to cooldown with an isolated or idle recirculation system.

The isolation valves are opened during a cooldown to ensure a uniform cooldown of the RHR injection i

piping.

If one recirculation loop is isolated or idle, these valves and associated piping allow the operating l

loop to cool the isolated or idle loop.

The RHR loop return line isolation valves receive a closure signal on LPCI initiation.

In the event of an inoperable return line isolation valve, there is a potential for some of the LPCI flow to be diverted to the broken loop during a loss of coolant accident.

Surveillance requirements j

have been established to periodically cycle the RHR intertie line isolation valves.

In the event of an inoperable RHR loop return line isolation valve, either the inoperable valve is closed or the other two E

isolation valves are closed to prevent diversion of LPCI flow.

The RHR intertie line flow is not permitted in the Run Mode to eliminate 1) the need to compensate for the small change in jet pump drive flow or 2) a l

reduction in core flow during a loss of coolant accident.

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C.

Containment Spray / Cooling Systems

[

t Two containment spray / cooling subsystems of the RHR system are provided to remove heat energy from the f

containment and control torus and drywell pressure in the event of a loss of coolant accident. A containment

(

spray / cooling subsystem consists of 2 RHR service water pumps, a RHR heat exchanger, 2 RHR pumps, and valves and piping necessary for Torus Cooling and Drywell Spray.

Torus Spray is not considered part of a containment spray / cooling subsystem.

Placing a containment spray / cooling subsystem into operation following a loss of coolant accident is a manual operation.

e I

The most degraded condition for long term containment heat removal following the design basis loss of coolant f

accident results from the loss of one diesel generator. Under these conditions, only one RHR pump and one RHR service water pump in the redundant division can be used for containment spray / cooling. The containment temperature and pressure have been analyzed under these conditions assuming service water and initial suppression pocl temperature are both 90*F.

Acceptable margins to containment design conditions have been demonstrated.

Therefore the containment spray / cooling system is more than ample to provide the required heat i

removal capability. Refer to USAR Sections 5.2.3.3, 6.2.3.2.3, and 8.4.1.3.

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During normal plant operation, the containment spray / cooling system provides cooling of the suppression pool water to maintain temperature within the limits specified in Specification 3.7.A.l.

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3.5/4.5 112 REV t

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Bases 3.5/4.5 Continued:

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The surveillance requirements provide adequate assurance that the containment spray / cooling system will be operable when required. The head and flow requirements specified for the RHR service water pumps provide I

assurance that the minimum required service water flow can be supplied to the RHR heat exchangers for the most d

degraded condition for long-term containment heat removal following the design basis loss of coolant accident.

D.

RCIC l

The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor isolated'from

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the turbine and when the feedwater system is not available. The pumping capacity of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation.

If the wate" level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts.

The system I

may also be manually initiated at any time.

h i

The HPC_ system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefore, the specification calls for an operability check of the HPCI system should the RCIC system be found to be inoperable.

r The surveillance requirements provide adequate assurance that the RCIC system will be operable when required.

[

All active components are testable and full flow can be demonstrated by recirculation through a test loop during j

reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

1 i

E.

Cold Shutdown and Refueling Requirements T

i The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times.

It is during refueling outages that major maintenance is performed and during such time that all core and containment spray / cooling subsystems may be out of service. This specification allows all core and-containment spray / cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel.

Thus events requiring core cooling are precluded.

i Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may by necessary to drain the suppression chamber for maintenance or for the inspection required by

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Specification 4.7.A.1.

In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.

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3.5/4.5 113 REV t

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Bases Continued:

Vent system downcomer submergence is three feet below the minimum specified suppression pool water level. This length has been shown to result in reduced postulated accident loading of the torus while at the same time assuring the downcomers remain submerged under all seismic and accident conditions and possess adequate condensation effectiveness.*

The maximum temperature at the end of blowdown tested during the Humboldt Bay '" and Bodega Bay

• tests was 170oF and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170oF.

t Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the

. i j

suppression pool is maintained below 160oF during any period of relief valve operation with sonic conditions at the discharge exit.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open.

This action would include:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

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For an initial maximum suppression chamber water temperature of 90

• F and conditions which lead to minimum 1

containment pressure, adequate net positive suction head (NPSH) is maintained for the core spray, RHR, and HPCI pumps under loss of coolant accident conditions.

(1) Robbins, C.H.

" Tests of Full Scale 1/48 Segment of the Humboldt Bay Pressure

- i Suppression Containment," GEAP-3596, November 17, 1960.

i (2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50 -205, December 28,-1962.

(3) General Electric NEDE-21885-P,

• Mark I Containment Program Downcomer Reduced Submergence Functional-Assessment Report," June, 1978.

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t 3.7 BASES 176 f

REV f

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Exhibit D Monticello Nuclear Generating Plant Revision No. 2 to License Amendment Request Dated January 23,1997 General Electric Report GE-NE-T2300731-2, "Monticello Nuclear Generating Plant LOCA Containment Analysis For Use in Evaluation of NPSH for the RHR and Core Spray Pumps,"

June,1997 i

Notes:

No credit is taken for offsite (normal AC) power sources for the LOCA scenario at Monticello.

This is typically done since availability of offsite power would normally be expected to reduce the consequences of the postulated accident. Consideration was provided for the availability of offsite power for the evaluation of ECCS pump NPSH, as provided in this report, since it was recognized that availability of offsite power might be more limiting for the long-term containment response with respect to NPSH margins. This does not imply a change in the design basis for Monticello, but was assumed for the purposes of providing a complete consideration of all possible conditions for impact on ECCS pump NPSH margin.

The pump combinations chosen for evaluation of NPSH provide a broad evaluation of the impact of various possible containment cooling heat exchanger capacities and pump flow rates.

Long term scenarios use conservative values for operating core spray pump flow rates and rated containment cooling flow rates for RHR pumps. These combinations conservatively evaluate potential plant operating conditions and bound the worst case impact on NPSH (higher suppression pool temperatures and lower containment pressures).

The following long-term scenarios were evaluated for containment response and its expected impact on NPSH:

1. Loss of offsite power with failure of one emergency diesel generator

2. Loss of offsite power with a postulated failure of the LPCI loop select logic to select the unbroken reactor recirculation loop for injection.

3. Loss of offsite power with the use of containment sprays and the consideration of a single failure that makes the LPCI injection valve inoperable.

4. Normal offsite power available with a postulated failure of the LPCI loop select logic to select the unbroken reactor recirculation loop for injection.

5. Normal offsito power available with the use of containment sprays and the consideration of a single failure that makes the LPCI injection valve inoperable.

The limiting case for ECCS pump NPSH for the long-term containment response is the single failure of one emergency diesel generator with loss of off-site power. This case results in the maximum suppression pool temperature.

D-1

The limiting case for ECCS pump NPSH for the short-term containment response assumes decay heat for 1880 Mwt, failure of LPCI loop selection logic, and loss of off-site AC power.

Feedwater heat is conservatively assumed to be added to containment. Alllow pressure ECCS pumps are assumed to be operating at runout f!ow rates. The Core Spray pumps are assumed to be injecting to the reactor with all LPCI flow being lost out the break to the drywell due to the postulated single failure of the LPCI loop selection logic to select the correct loop.

Ru'iout flow rates were reevaluated and the impact on the short-term containment response was determined. GE letter GLN-97-024, June 18,1997 (attached) provides these results. The short-term case using 1880 Mwt in GE-NE-T2300731-2 provides conservative results when considering the effects of revised runout flow rates.

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