ML20087K808

Application for Amend to License NPF-3,revising TS 3/4.7.5.1 Re Ultimate Heat Sink

ML20087K808
Person / Time
Site:	Davis Besse
Issue date:	08/18/1995
From:	Jain S CENTERIOR ENERGY
To:
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ML20087K804	List: ML20087K800 ML20087K808 ML20087K831
References
2319, NUDOCS 9508240158
Download: ML20087K808 (11)

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! Dock:t'Nu:ber 50-346 i

• Lic;nsa Nurb:r NPF '

5:rici Nu ber 2319

' Enclosure-APPLICATION FOR AMENDMENT ,

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TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 h

Attached is the requested change to the Davis-Besse Nuclear Power ,

Station, Unit Number 1 Facility Operating License Number NPF-3. Also included 1s;the Safety Assessment and Significant Hazards Consideration i and the Environmental Assessment.

The proposed change (submitted under cover letter Serial Number 2319) concern:

Appendix A, Technical Specifications (TS):

i 3/4.7.5.1 Plant Systems, Ultimate Heat Sink i

i Fort John P. Stetz, Vice President - Nuclear 4

By:

S. C. Jain, Director - Engineering and Services-Sworn to and subscribed before me this 18th Day of August 1995

.h N6tary Publ c, State of Ohio LORIJ. STRAUSS

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Notary Pubht. Steen of Ohio My Commission Eagnres 3/22/98 :

f 9508240158 95081e PDR P ADOCK 05000346 PDR

1

Dockst Nu:b:r 50-346 '

'Licensa Nulbar NPF-3 Siricl Nunb:r 2319 '

Enclosure-Page.2

~ The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station (DBNPS), -

Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specification (TS) 3/4.7.5.1, Plant Systems, Ultimate Heat Sink.

A. Time Required to Implement: This change is to be implemented within 7 days after NRC issuance.

B. Reason for Change (License Amendment Request Number 95-0016):

The proposed change vould increase ~the allovable ultimate heat sink average water temperature, as specified in Technical Specification Limiting Condition for Operation 3.7.5.1.b, from f 85'F to f 90*F for the period of August 18, 1995 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> to September 17, 1995, 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />.

C. Safety Assessment and Significant Hazards Consideration: See Attachment 1.

D. Environmental Assessment: See Attachment 2.

E. Figures: See Attachment 3 P

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'Dockat Nunb r'50-346 Liccns: Nurb:r NPF-3 Serial Number 2319' Attachment l' SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION ~

FOR LICENSE AMENDMEfC REQUEST NUMBER 95-0016 ,

LAR 95-0016 Page 1 i

SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDMENT REQUEST NUMBER 95-0016 TITLE:

Revision of Technical Specification (TS) 3/4.7.5.1, Plant Systems, Ultimate Heat Sink, i

DESCRIPTION:

k The purpose of the proposed changes is to modify the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A Technical Specifications (TS) and associated Bases. The proposed change would increase the allovable ultimate heat sink (UHS) average water temperature, as specified in Technical Specification (TS)

Limiting Condition for Operation (LCO) 3.7.5.1.b, from f 85'F to f 90'F, for a period from August 18, 1995, 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, to September 17, 1995, 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />.

This proposed change is shovn on the attached marked-up TS page.

SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED: .

The systems and components affected are the: Service Water, Component Cooling Vater and Component Cooling Vater Loads. The Technical Specification Limiting Condition for Operation 3.7.5.1.b of f 85'F is affected.

FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES:

The SV system provides cooling water from the Forebay to the following Safe Shutdown equipment after either a Safe Shutdown Earthquake (SSE) or a Large Break Loss of Coolant Accident (LOCA): The Containment Air Coolers (CACs), the Component Cooling Vater (CCV) heat exchangers, the Emergency Core Cooling Systems (ECCS) room heat exchangers, the Control Room Emergency Ventilation '

System (CREVS), the seal water to the Hydrogen Dilution System Blovers. 'The SV system also provides a backup source of water for the Auxiliary Feedvater Pumps (AFPs) in the event of an SSE which renders the condensate storage tanks vater supply unavailable. SV is also a backup source of water for the CCV system when all other CCV makeup vater sources are not available following an SSE.

LAR 95-0016 Page 2 ,

The CACs safety function is to remove one half of the post-LOCA heat removal requirement. The CREVS safety function is to maintain Control Room (CTRM) habitability and CTRM Environmental Qualification (EO) requirements ($110'F).

Each ECCS room has two coolers. The ECCS room heat exchangers safety function is to maintain the ECCS room temperature fl22'F during a LOCA with only one cooler in operation assuming a SV temperature of 72*F. One heat exchanger in each ECCS room may be isolated if SV is <72'F and still be considered OPERABLE.

The Hydrogea Dilution Blovers safety function is to provide Auxiliary Building air to CTMT post-LOCA to decrease the post LOCA hydrogen concentration.

During a Design Basis Accident, the CCV supplies cooling water to the following essential components: High Pressure Injection (HPI) pumps 1 and 2 bearing oil coolers, Decay Heat Pumps /Lov Pressure Injection (DH/LPI) 1 and 2 bearing housing coolers, Decay Heat (DH) Coolers 1 and 2, Containment (CTMT) Gas Analyzer heat exchangers 1 and 2, and Emergency Diesel Generator (EDG) jacket cooling vater heat exchangers 1 and 2.

The HPI system provides emergency core cooling for small break LOCAs. The HPI system provides makeup to the Reactor Coolant System (RCS) due to contraction during excessive overcooling of the RCS. The HPI system also provides a source of borated water from the BVST for Shutdown Margin (SDM) requirements. The LPI system provides emergency core cooling and refill of the reactor vessel following a Large Break LOCA, post-LOCA CTMT sump recirculation capabilities, long term decay heat removal, and post-LOCA boron dilution. The DH coolers safety function is to provide adequate post-LOCA CTMT sump cooling requirements and normal and emergency shutdown cooling requirements.

The CTMT gas analyzers safety function is to be capable of immediately analyzing CTHT atmosphere to detect the build up of hydrogen in CTMT following a LOCA.

This function is required to maintain hydrogen levels to acceptable limits.

The EDG jacket cooling vater safety function is to maintain the required temperature of the lube oil and diesel engine. The basis for the safety related cooling water (CCV) source is to ensure that the EDGs vill perform their intended safety function during and after an accident.

EFFECTS ON SAFETY:

Because Service Vater (SV) serves loads during normal operations as well as during accident conditions, a temperature increase in its water supply vill affect any accident analyses initial conditions as well as the post accident response. The discussion below considers each of these aspects.

A. Normal Operation During normal operation, SV supplies cooling to the Containment Air Coolers (CAC's), the Component Cooling Vater (CCV) heat exchangers and the Turbine Plant Cooling Vater (TPCV) heat exchangers. An increase in the Intake Canal temperature to 90'F from an assumed maximum of 85'F vill increase the normal operating temperatures of the components served by these systems. While TPCV-cooled components do not affect

.; s LAR 95-0016- .i g Page 3', l e

plant safety, they may limit plant operation. The TPCV system j temperature is normally maintained at 85'F at the outlet of the heat -

exchangers. There is no Technical Specification associated with_this  ;

F value. A 90'F SV temperature vill not permit this'setpoint to be '

maintained. Therefore TPCV-cooled components will_be closely monitored  ;

and appropriate action taken before any device is allowed to operate  ;

above its design temperature..

I During normal operation, there is sufficient cooling capacity in the CCV system to accommodate the changes in the SV supply temperature. l The automatic controls on the system maintain the CCW, heat exchanger j

!- outlet temperature at 95'F. There is no Technical Specification i' associated with this value. Evaluation of the heat exchangers has-f determined that the 95'F outlet temperature can be maintained with a #

90'F SV inlet temperature. Consequently, all loads served by CCW i i

during normal power operation vill be operated at normal conditions and there is no impact on the plant.

i The CACs limit the normal containment air temperature, which is a design input to the LOCA and EQ analyses. Technical Specification i 3.6.1.5 limits this temperature to 120*F. Vith SV at approximately  ;

82'F, the containment air temperature is approximately-110'F. Because ,

heat removal through the containment vessel shell supplements heat i removal through the CACs, containment temperature increases will be less than any-increase in SV inlet temperature. An increase of 8'F in SV temperature vould not cause the containment air temperature to exceed 120'F.

B. Accidents ,

An increase in SV temperature affects the temperature of the CCW system' <I and the reactor containment building temperature following a loss of coolant accident or a main steam line break accident. A design basis Loss of Coolant Accident (LOCA) imposes the greatest performance

• requirement on the SV system. In this postulated accident, it is assumed that a loss of offsite power occurs with concurrent failure of an emergency diesel generator. Thus, only one train of' service water  :

and ECCS pumps are assumed to be available. In a' design basis LOCA, the critical period for refilling and cooling the reactor core occurs .

within the first few minutes of the accident. Following refill of the reactor vessel, the fuel vill be adequately cooled. Service water-  ;

temperature does not directly impact core cooling during this portion of the event, because the reactor is refilled via the low pressure i injection (LPI) pumps using water from the borated vater storage tank (BVST). In addition, the available containment spray (CS) pump injects water directly into containment from the borated water storage tank without any cooling supplied by service water. Until the BVST is exhausted, cooling is not required for the decay heat coolers. The  !

initial blovdown is sufficiently quick, that heat removal from containment air coolers is not effective'in reducing the initial temperature / pressure spike in containment.

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I 1 LAR 95-0016 P ge 4 ,

Following consumption of the BVST inventory, containment cooling for preservation of containment equipment qualification must be maintained.

At this point, more than 30 minutes into the transient (depending on pump combinations in service), the suction of both the LPI pumps and the CS pumps is transferred to the containment emergency sump, where effluent from the break and CS has accumulated. A CCV heat exchanger, cooled by SV, is required to provide cooling to the Decay Heat Removal (DHR) heat exchangers. An increase in design SV temperature from 85'F to 90'F vill cause a similar increase in CCV temperature. Upon making the transfer to the emergency sump (which is at a higher temperature than the BVST), both the LPI injection temperature and the CS injection temperature increase, giving rise to a temporary increase in contain-ment temperature and pressure. This increase peaks at approximately 10,000 seconds and is lover than the initial pressure / temperature peak.

The long term containment temperature following a LOCA is controlled by the heat removal'through the DHR system and the containment air coolers. The impact of an increase in SV temperature on the heat removal through these systems is discussed below.

The previous containment analysis, as described in references ~2a and 3a, used a DHR heat exchanger heat transfer coefficient of 300 BTU /hr-ft2 *F. This value was conservatively used to bound potential >

degradation from the design heat exchanger performance of 478 BTU /hr-ft2 *F. However, the most recent performance tests (reference

4) indicate that the actual pecformance of the vorst DHR cooler is 407 BTU /hr-ft2 'F. Sensitivity of containment temperatures to DHR cooler performance has been explored in references 2a and 2b for a range between 250 and 478 BTU /hr-ft2 *F. From a graph of calculated containment temperature increase vs. DHR cooler heat transfer coefficients, a coefficient of 400 BTU /hr-ft2 *F vould provide a long term containment temperature approximately 7'F below the analyzed temperature. Thus, the margin available in DHR heat exchanger performance vill more than compensate for a 5'F increase in service water temperature. This conclusion is confirmed by reference 2c, where a containment temperature response profile was re-run using a heat transfer coefficient of 400 BTU /hr-ft2 *F and a service water temperature of 90'F rather than the original 85'F. The new short term containment temperature profile is unaffected, while the long term profile is lover.

Containment air coolers (CAC) vill receive SV vater at a temperature of 90'F. In the previous containment analysis, the CACs received 85'F cooling vater at an analyzed flow rate of 1150 gpm. Reference 5 indicates that the current CAC flow is greater than 1350 gpm. This flow represents the current flov balance performed under simulated accident conditions. The CACs have a large water side temperature rise in LOCA service, therefore, performance is more affected by flow rate than minor inlet temperature changes. From reference 2c, the CAC heat removal vill be greater than that used in the containment analysis for an inlet temperature of 90'F vith 1300 gpm SV flow. The greater performance vill continue well into the transient, when containment

LAR 95-0016 Page'5 3 temperatures are no longer high enough to affect equipment qualifica-tion. For Main Steam Line breaks, since CAC duty is better with 90'F inlet temperature and.1300 gpm SW flow (than with 85'F inlet tempera-ture and 1150 gpm SV flov), the containment temperatures are also not increased from the previous analyses.

1 The peak CCV temperature predicted by the containment analysis of reference 2c is less than 108'F at approximately 15000 seconds follow-

.ing.the accident. The peak CCW temperature in the existing analysis is slightly over 100'F. The increase is due to both the 5'F increase.in SV temperature and the greater credited heat removal by the DHR heat exchanger. A CCW temperature of 108'F is well within the design temp-erature of the essential components served by CCV. .From reference 6a' I and 6b, the design maximum cooling vater temperature supply to the ,

Emergency Diesel Generators and, the LPI bearings is 120'F. .The j maximum bearing temperature for HPI pump is 165'F. Plant surveillance  ;

testing data show that the bearing temperature tracks the CCW tempera- I ture; therefore, the HPI bearing temperature vill be below 165'F. l Therefore, safety related CCV loads vill be adequately served even  !

during the time of peak CCV thermal loading. _Non-safety related loads  !

(e.g. spent fuel pool cooling, containment loads, etc.) would have been automatically isolated by the safety features actuation system. How- a ever, these loads could be re-supplied at a later time ~, as required.

The ECCS room coolers are directly supplied by service' vater and are required to operate in design basis events since the normal ventilation 4 system is assumed to be unavailable. Two coolers are provided in each 2 of the two ECCS rooms. At lov SV temperatures (less than approximately 72*F), one cooler is sufficient to provide 100 percent of the required cooling capacity. At higher SV temperatures, existing administrative controls require that both ECCS room coolers ~in each.of the two ECCS rooms must be in service. From reference 2d, two ECCS room coolers are adequate up to a SV temperature in excess of 95'F, even with flow rates substantially degraded from normally accepted values.

The containmene hydrogen dilution blowers are "Nash" pumps which utilize less than 10 gpm SV to provide seal water. The.SV makeup connection to the CCW system vill be unaffected. The inlet temperature of SV for these uses is not critical.

The emergency suction for Auxiliary Feedvater is supplied by service I vater directly. An increase in SV temperature to 90'F represents a very small increase in initial liquid enthalpy when compared to the large increase in enthalpy encountered by.the feedvater as it is boiled in the steam generators. Due to the auple flow capacity of the auxil-iary feedvater pumps, the increase in service water inlet enthalpy will have negligible effect on system response. Similarly, 90' service water temperature vill provide adequate AFP bearing cooling when AFP's are taking suction from the SV system. Service water also provides cooling vater to the motor driven feedvater pump (MDFP) seal water and 1 bearing coolers. The vendor manual states that inlet cooling water is limited to 95'F. Therefore, increasing the SV to 90'F vill not ad-versely affect the HDFF.

~LAR 95-0016 P, age 6 ,

Control Room Emergency Ventilation units receive cooling vater directly from service water and are required during a LOCA. These units are also provided with an air cooled condensing coil which is designed for ambient temperatures of up to 95'F. The air cooled condensing unit is automatically selected if refrigerant pressure increases due to inade-quate water cooled condenser cooling. This is expected to occur at a SV temperature of approximately 110'F. Since the air cooled system backs up the Service Water cooled system, and increase in normal service water temperature does not impact the availability-of CREVS.

Thus, existing margin in equipment performance vill maintain contain-ment response within the existing profile for all times of importance following a LOCA. All equipment will operate as designed for all transients.

Under normal conditions, the ultimate heat sink is Lake Erie. Lake Erie is connected to the intake canal by a 96" diameter inlet pipe. In the unlikely event of a collapse of the non-seismic portion of the intake canal, the forebay contains sufficient vnter to provide conti-nuous cooling for more than 30 days. The analyses presently in the USAR assume that the intake structure forebay level is at least 562.0' International Great Lakes Datum (IGLD) and at an initial temperature of 85'F. Currently, the forebay level is approximately 569.5'. The low water datum of Lake Erie is 568.6 feet (I.G.L.D). The maximum varia-tions in the mean monthly level are 4.2 feet above and 1.2 feet below the datum for the 110-year period that the data has been collected. At 568.6' level, the' intake canal contains approximately twice the volume of water and 23 percent more initial forebay surface area (USAR figure 2.4-7) for heat transfer than assumed in the original analyses. There-fore, with the forebay at present levels, increasing the ultimate heat sink temperature from 85'F to 90'F does not impact the ability to provide continuous cooling for a period of 30 days. The connection between Lake Erie and the intake canal can be reestablished well within the 30 day period.

Lov vater in the intake canal could also occur due to a maximum pro-bable meteorological event. This consists of a sustained WSV vind of 70 mph for a six hour duration. This could result in lake vater level decreasing below the level of the intake crib at 561.85' IGLD. The lov level condition from this meteorological event lasts for a maximum period of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. T.S. 3/4.7.5.1 requires a plant shutdovn if the forebay level reaches 562.0' IGLD. Since the low water condition due to a meteorological event is of a limited duration, increasing the ultimate heat sink temperature from 85'F to 90'F vill not affect the ability to safely shutdown the plant.

In summary, the proposed change to increase the allovable ultimate heat sink (UHS) average water temperature, as specified in Technical Specification (TS)

Limiting Condition for Operation (LCO) 3.7.5.1.b, from f 85'F to f 90'F, for a period from August 18, 1995, 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, to September 17, 1995, 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, vill not adversely affect plant safety.

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LAR 95-0016 Page 7 l

SIGNIFICANT HAZARDS CONSIDERATION:

The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for ,

determining whether a significant hazard exists due to a proposed amendment to l an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes vould: (1) Not involve a significant increase in the probability or consequence of an accident previously evaluated; (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. Toledo Edison has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes vould:

la. Not involve a significant increase in the probability of an accident previously evaluated because no accident initiators, conditions, or assumptions are significantly affected by the proposed change. The proposed change does not result in the operation of equipment important to safety outside their acceptable operating range.

Ib. Not involve a significant increase in the consequences of an accident previously evaluated because the proposed change does not change the source term, containment isolation, or allowable releases.

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2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because no new accident initiators or assumptions are introduced by the proposed change.

The proposed change does not result in installed equipment being operated in a manner outside its design operating range. No new or different equipment failure modes or mechanisms are introduced by the proposed change.

3. Not involve a significant reduction in a margin of safety because the proposed change is not a significant change to the initial conditions contributing to accident severity or consequences, consequently there are no significant reductions in a margin of safety.

CONCLUSION:

On the basis of the above, Toledo Edison has determined that the License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specifications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.

ATTACHMENT:

Attached is the proposed marked-up change to the Operating License.

LAR 95-0016 Page 8 ,

REFERENCES:

1. DBNPS Operating License NPF-3, Appendix A Technical Specifications through Amendment 199.

2. Calculations:

a. C-NSA-049.02-012, rev. O, Long Term Containment Response {ollowingaLOCAv/DHRCoolerU-300 BTU /hr-ft *F

b. C-NSA-049.02-011, rev. O, Effect of Degraded DHR Cooler 1-1 on Containment P/T Response Following a LOCA

c. C-NSA-60.05-006 Rev 0

d. C-NSA-032.02-003, rev. 3, Maximum Service Water Temperature

3. DBNPS Updated Safety Analysis Report through Revision 19.

USAR

a. Section 6.2, Containment System

b. Section 2.4.11, Lov Vater Considerations

c. Section 9.4.1, Control Room HVAC

d. Section 9.2.5, Ultimate Heat Sink

e. Section 9.2.1, Service Vater System I

4. 1991 DHR cooler performance tests DB-PF-04727 (09/06/91)

DB-PF-04703 (09/03/91)

5. SV Flow Balance Test DB-SP-401'9, DB-SP-4020 November 1994

6. System Descriptions:

a. SD-042, Revision 0, Decay Heat Removal System.

b. SD-003, Revision 3, Emergency Diesel Generators.

c. SD-029B, Revision 1, Control Room Emergency Ventilation System.

d. SD-18, Revision 1, Service Vater

e. SD-16, Revision 3, Component Cooling Vater

f. SD-23, Revision 2, Hydrogen Control

g. SD-22B, Revision 1, Containment Air Coolers

h. SD-38, Revision 2, High Pressure Injection

7. M480N-21, Vendor Manual for Motor Driven Feedvater Pump

8. Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants.  ;

9. NUREG-0136, Operating License NPF-3, Safety Evaluation Report, December 1976.