ML20134C870

Submits 120-day Response to NRC GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions

ML20134C870
Person / Time
Site:	Monticello
Issue date:	01/28/1997
From:	Hill W NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-96-06, GL-96-6, NUDOCS 9702040102
Download: ML20134C870 (15)

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4 Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello. Minnesota 55362-9637 January 28,1997 Generic Letter 96-06 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 120 Day Response to NRC Generic Letter 96-06 Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," issued September 30,1996 provided licensee notification of safety-significant issues that could affect containment integrity and equipment operability during accident conditions. The generic letter requests information concerning waterhammer and two phase flow conditions in cooling water systems serving containment air cooling systems, and thermally induced over pressurization of fluid filled piping systems.

Generic Letter 96-06 contains the following requested actions.

Addressees are requested to determine:

(1) if containment air cooler cooling water systems are susceptible to either water hammer or two phase flow conditions during postulated accident conditions; (2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so the overpressurization of piping could occur.

In addition to the individual addressee's postulated accident conditions, these

items should be reviewed with respect to the scenarios referenced in the genedc letter.

1128/97 SIS X\ TEAM \D04560tMONTILMhGENITR\GL96-06120 AO'7W >

l 9702040102 970128 i PDR ADOCK 05000263 P PDR

USNRC NORTHERN STATES POWER COMPANY January 28,1997

, Page 2 Requested information 1

Within 120 days of the date of this genen'c letter, addressees are requested to

, submit a written summary report stating actions takcn in response to the requested actions noted above, conclusions that wers reached relative to i susceptibility for waterhammer and two-phase flow in the containment air cooler ,

cooling water system and overpressurization of piping that penetrates l 4 containment, the basis for continued operability of affected systems and i

l components as applicable, and corrective actions that were implemented or are  ;

planned to be implemented. If systems were found to be susceptible to the )

conditions that are discussed in this generic letter, identify the systems affected and describe the specific circumstances involved.

Accordingly, in NSP's 30 day response to Generic Letter 96-06, Monticello committed to the following:

Monticello will complete the actions requested by Generic Letter 96-06 and will submit a summary report of requested information in accordance with the schedule specified by the generic letter.

Attachment A to this submittal fulfills the above commitment and the 120 day Generic j Letter 96-06 required response. i 4

, Within this submittal, Monticello commits to the following:

i Monticello will install pressure relieving devices or other equivalent means to resolve the overpressure conditions identified by this submittal prior to startup from the next refueling outage. The next refueling outage is currently scheduled

, for January 1998.

Please contact Sam Shirey, Sr Licensing Engineer at (612) 295-1449, if you require further information.

Y William J Hill Plant Manager Monticello Nuclear Generating Plant

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) c: Regional Administrator-Ill, NRC NRR Project Manager, NRC Sr Resident inspector, NRC State of Minnesota, Attn: Kris Sanda j Attachments: Affidavit to the US Nuclear Regulatory Commission 1

Attachment A - 120 Day Response to NRC Generic Letter 96-06 i

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UNITED STATES NUCLEAR REGULATORY COMMISSION I

NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 120 DAY RESPONSE TO NRC GENERIC LETTER 96-06 ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS I I

Northern States Power Company, a Minnesota corporation, by letter dated January 28, 1997, provides the required 120 day response to NRC Generic Letter 96-06,

" Assurance of Equipment Operability and Containment Integrity During Design-Basis  !

Accident Conditions." This letter contains no restricted or other defense information.

NORTHERN STATES POvVER COMPANY j

By Mb" h //

William TI-lili Plant Manager

Monticello Nuclear Generating Plant On this clB day of hu u M(T '1 before me a notary public in and for said County, personally appeared William J Hil, Plant Manager, Monticello Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.

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Sam'uel I. Shirey Notary Public - Minnesota f )

SMUEU. SHIREY lNOWPutuC 6 Sherburne County ~~

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My Commission Expires January 31,2000

Attachment A 120 DAY RESPONSE TO NRC GENERIC LETTER 96-06 Waterhammer and Two-phase Flow in the Containment Air Cooler Cooling Water System:

Affected Systems:

At Monticello, the Containment Air Cooler Cooling Water function is provided by the Reactor Building Closed Cooling Water System (RBCCW) which supplies coo!!ng water to the primary containment coolers. The Reactor Building Closed Cooling Water System (RBCCW) was determined not to be susceptible to damaging waterhammer or two-phase flow conditions.

Specific Circumstances:

WATERHAMMER: The RBCCW system may be susceptible to waterhammer during a design basis loss-of-coolant accident (LOCA) coincidental with a loss of normal auxiliary power. Under these conditions, the RBCCW pumps and the Drywell Atmospheric Cooling (DAC) fans would be tripped by an ECCS load shed signal. Steam voids could form in the DAC cooling since the saturation temperature of the RBCCW water is 264 F with the peak primary containment air ,

l temperature at 282*F for a large break, and up to 335 F for a small break LOCA.

Since the RBCCW pumps are tripped under these conditions, the operator receives low RBCCW system pressure control room annunciators. Annunciator response procedures for this condition is to close the RBCCW containment isolation valves. If the piping system inside the drywell were not leak tight, the '

pressure head supplied by the surge tank could dissipate and steam voids could ,

exist at drywell temperatures less than 264 F. Under these conditions,  !

reestablishing RBCCW flow to the drywell portion of the system could potentially J create a waterhammer situation. ,

Under the above loss of power scenario, there is no automatic restart of the RBCCW pumps. Additionally, the operating procedure for manually starting the RBCCW pumps require the pump discharge valve to be closed prior to starting the pump. Once the pump has started, the discharge valve is manually opened.

The manual opening of the discharge valve will result in a gradual restoration of flow and pressure and minimize any potential waterhammer due to collapse of steam voids. The controlled restoration of flow and pressure would result in very minor waterhammer as compared to events described in NUREG/CR-5220.

Therefore, any waterhammer that may occur is not expected to have the l

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USNRC NORTHERN STATES POWER COMPANY January 28,1997 Page 2 i

intensity to challenge the integrity of the piping system or containment j penetrations.

i-This start-up procedure is required to be followed whenever restarting a pump l after both pumps have been idle. This includes restarts after accident situations.

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] -TWO-PHASE FLOW: The conditions most likely to result in two-phase flow {

would be a small break LOCA where the peak drywell temperature reaches .

335 F. The DAC fans are tripped for a small break LOCA due to high drywell {

pressure. With no DAC fans running, RBCCW temperature above saturation  !

i (264 F) is not anticipated due to the heat transfer rate being very small. If the trip signal for the DAC fans is bypassed as permitted by the Emergency ,

Operating Procedures (EOPs), there would be a potential for two phase flow to 1

occur if drywell temperatures were to be maintained above 264 F for an i extended period. However, the EOPs require that the drywell sprays be used l before the drywell temperature reaches 281 F. For the small break LOCA, the

! drywell reaches 281 F requiring EOP operator actions to place the drywell j

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sprays into operation. DAC fans would not be placed in operation since the l drywell fan motor windings would be wet and could cause electrical faults after j i

the drywell sprays are initiated. Additionally, the sprays would drop the drywell I temperature to less than the RBCCW saturation temperature of 264 F such that

two-phase flow would no longer be a concern.

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) Conclusions Reached:

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! WATERHAMMER: Any waterhammer that may occur is not expected to have j the intensity to challenge the integrity of the piping system or containment penetrations. Operating practices already in place further minimize any potential impact due to pump restart.  !

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TWO-PHASE FLOW: System design and operating procedure evaluation has ,

shown that two-phase flow is not expected to occur in the RBCCW system, j i  !

in the unlikely event that two-phase flow did occur, no safety related functions j would be affected. The DAC system has no safety related function and there are 4

no safety related loads supplied by RBCCW, requiring operation under post l j LOCA conditions. ]

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USNRC NORTHERN STATES POWER COMPANY January 28,1997 Page 3 Actions Taken:

No actions required. The RBCCW system was evaluated to determine its susceptibility to both waterhammer and two phase flow. It has been determined that current system configuration, operating practices and procedures preclude detrimental waterhammer and two-phase flow conditions.

Basis for Continued Operability:

Neither waterhammer or two-phase flow are considered likely events due to system design and current operating procedures. If waterhammer or two-phase flow were to occur, the consequences have been determined to be minor.

Additionally, the RBCCW system provides no safety function, no credit is taken for its use during an accident, and therefore loss of RBCCW would not be a concern.

Overpressurization of Containment Penetration Piping Affected Systems:

To evaluate Monticello for overpressurization, all primary containment penetrations and associated systems were reviewed. Penetrations and associated systems were identified as having a potential for overpressurization if they had the attributes of being a closed volume containing fluid, (either air or water), were at a low initial temperature, most of the volume could be heated by the containment atmosphere after an accident, and no installed pressure relieving device was available to prevent the buildup of pressure. The affected systems and their respective containment penetration numbers are:

1. Drywell Equipment Drain Sump (X-19)

2. Drywell Floor Drain Sump (X-18)

3. RHR Shutdown Cooling (X-12)

4. RHR H.ead Spray (X-17)

5. Reactor Water Cleanup (RWCU) (X-14).

6. Main Steam Line Drain (X-8)

7. Recirculation Sample Line (X-41)

8. Service Air (X-21)

9. Instrument Air (X-22)

10. Alternate N2 Supply to SRV C, H, F and MSIV Inboard Actuators (X-34A)

11. Alternate N2 Supply to SRV A, B and E (X-105B-G)

USNRC NORTHERN STATES POWER COMPANY January 28,1997 Page 4

12. Vacuum Breaker Line (X-2298)

13. Demineralized Water Line (X-20)

Specific Circumstances and Actions Taken:

All containment penetrations and associated piping systems were reviewed. Of all containment penetrations,13 lines were initially evaluated as being susceptible to overpressurization. These lines are discussed below along with their final disposition.

1. Drywell Equipment Drain Sump (X-19)

The enclosed boundary is formed between the sump pump discharge check valve and the outside containment isolation valves. The enclosed boundary is postulated to pressurize. Included within this boundary are two ball valves which i are both located inside primary containment. These valves are postulated to I leak or experience pressure boundary failure first since their pressure rating is l approximately 21/2 times less than the containment isolation valves and piping l components outside the drywell. i Since the ball valves inside primary containment will be the first component to fail or relieve pressure, Primary Containment integrity is maintained. Primary containment integrity is the only safety related functiori associated with this i' system. Installation of a permanent pressure relieving device will be pursued.

2. Drywell Floor Drain Sump (X-18)

The enclosed boundary is formed between the sump pump discharge check valve and the outside containment isolation valves. The water leakage rate calculated from current Appendix J test results is greater than the calculated leakage rate required to relieve pressure at the faulted pressure of ASME Section Ill, Appendix F. Additionally, non-quantified boundary leakage, i.e.,

leakage past the drywell sump pump discharge check valves could also assist in relieving pressure. Therefore during pressurization, the valve leakage will relieve pressure and maintain Primary Containment integrity.

Primary containment integrity is maintained which is the only safety related function associated with this system. Installation of a permanent pressure relieving device will be pursued.

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USNRC NORTHERN STATES POWER COMPANY

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! 3. RHR Shutdown Cooling (X-12) j The enclosed boundary is formed by piping bounded by normally closed inboard  !

and outboard isolation valves. Calculations show the pipe is operable in a j faulted condition using conservative assumptions. One assumption postulated is 3 that all pipe insulation is removed from the pipe as a result of the accident. This i j results in a greater heat transfer rate so the water inside the pipe heats up faster. .

With this assumption, the piping exceeds the code allowables. Even though the  !

pipe exceeds code allowables and the valves exceed their design rating, the - i j piping / components meet faulted condition acceptance criteria as established in j i ASME Section lil, Appendix F. Primary containment integrity is maintained  !

which is the only safety related function associated with this section of piping. l 4 Installation of a permanent pressure relieving device will be pursued.

) 4. RHR Head Spray (X-17) l i l l

The enclosed boundary is formed by piping bounded by normally closed inboard '

i and outboard isolation valves. Calculations show the pipe is operable in a i faulted condition using conservative assumptions. One conservative assumption l i

postulated is that all pipe insulation is removed from the pipe as a result of the i accident. This results in a greater heat transfer rate so the water inside the pipe heats up faster. With this assumption, the piping exceeds code allowables.

l Even though the pipe exceeds code allowables and the valves exceed their

! design rating, the piping / components meet the acceptance criteria as j established in ASME Section Ill, Appendix F. Primary Containment integrity is j maintained which is the only safety related function of this section of piping.

- Installation of a permanent pressure relieving device will be pursued.

I j 5. Reactor Water Cleanup (X-14)

I The enclosed boundary is formed by piping bounded by inboard and outboard

) isolation valves. Since the process stream water in the RWCU piping of concern l is approximately 512 F, isolation of the piping by a containment isolation signal j will not result in any increase in water temperature so overpressurization cannot

occur. For infrequent maintenance activities that require closing of both

! penetration isolation valves during normal operation, administrative controls have

{ been established to drain part of the pipe once isolated. Draining eliminates the

! possibility of water heating up to overpressurize the piping. Primary j Containment integrity is maintained which is the only safety related function i associated with this system. No modifications are required with the i

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) administrative controls. However, in lieu of administrative controls, installation of i

a permanent pressure relieving device is being evaluated. l

6. Main Steam Line Drain (X-8) l

. l The enclosed boundary is formed by piping bounded by normally closed inboard .

and outboard isolation valves. By procedure, these valves are closed during  !

( startup under hot conditions at approximately En0 PSIG. During operation, process steam remains on both ends of the isoMuon boundary keeping the line j hot. No modifications are required. I s 1

7. Recirculation Sample Line (X-41)  !

i y The enclosed boundary is formed by piping bounded by inboard and outboard isolation valves. Since the process stream water in the recirculation sample line piping is approximately 520 F, isolation of the piping by a containment isolation q signal will not result in any water temperature increase so overpressurization cannot occur. For infrequent maintenance activities that result in the closing of ,

both penetration isolation valves, administrative controls have been established I

] to drain part of the pipe once isolated. Draining eliminates the potential of water heating up to overpressurize the piping. Primary Containment integrity is

maintained which is the only safety related function associated with this system.

No modifications are required with the administrative controls. However, in lieu

of administrative controls, installation of a permanent pressure relieving device is being evaluated.

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8. Service Air (X-21)

The boundary analyzed for the penetration is between containment valves and

the penetration. The internal pressure in this section of piping reaches 162 PSIG

{ due to the temperature increase of the air contained in the attached piping inside l the drywell. Calculations show the piping outside the drywell is within code  !

i allowables at 162 PSIG. Primary Containment integrity is maintained which is l the only safety related function associated with this section of piping. No  !

4 modifications are required.

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The first boundary analyzed for the penetration is between the outboard i

containment valve and the penetration. The internal pressure in this section of piping reaches 145.2 PSIG due to the temperature increase in the air / nitrogen J

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USNRC NORTHERN STATES POWER COMPANY January 28,1997 Page 7 contained in the attached piping inside the drywell Calculations show the piping outside the drywellis within code allowables at 162 PSIG. Primary Containment integrity is maintained which is the only safety related function associated with this section of piping. No modifications are required.

The second boundary analyzed is established by accumulator check valves and their respective SRV air accumulators downstream. Assuming a drywell temperature of 335 F, the internal pressure this section of piping and accumulators reaches is 145.2 PSIG. The containment pressure at this temperature is 21.9 PSIG. Thus the internal pressure of the pipe and .

accumulators is the difference between 145.2 PSIG and 21.9 PSIG which equals 123.3 PSIG. Since the postulated pressure is less than the 125 PSIG design pressure of the piping and accumulators and less than the maximum operating pressure for the SRV solenoid valves, operability of all the components is maintained.

For the small break LOCA accident, drywell sprays are initiated in accordance with Emergency Operating PE tedures. The small break LOCA analysis ,

assumed drywell sprays are placed in service at 10 minutes. This lowers '

containment pressure and temperature and the temperature inside the piping such that unacceptable pressures are not expected to occur. The saies function of this piping is to provide pneumatic supply to one ADS SRV and one low low set SRV which will be maintained. Due to the small difference between the postulated pressure and the design pressure, a method to improve the margin will be pursued.

10. Alternate N2 Supply to SRV C, H, F and MSIV Inboard Actuators (X-34A)

The first boundary analyzed for the penetration is between the outboard valve and the penetration. The internal pressure in this section of piping reaches 145.2 PSIG due to the temperature increase in the air / nitrogen contained in the attached piping inside the drywell. Calculations show the piping outside the drywell is within code allowables at 174 PSIG. Primary Containment integrity and pressure integrity of Train B of the Alternate N2 system are maintained which are the only safety related functions associated with this section of piping.

The second boundary analyzed for the penetration !s between the penetration and SRV C, H, F and the MSIV inboard actuators and their respective air accumulators. Assuming a drywell temperature of 335 F, the internal pressure j this section of piping and accumulators reaches is 145.2 PSIG. The containment  ;

pressure at this temperature is 21.9 PSIG. Thus the internal pressure of the pipe l

USNRC NORTHERN STATES POWER COMPANY i January 28,1997 Page 8 r and accumulators is the difference between 145.2 PSIG and 21.9 PSIG which equals 123.3 PSIG. Since the postulated pressure is less than the 125 PSIG design pressure of the piping and accumulators, less than the 125 PSIG  ;

maximum operating pressure for the SRV solenoid valves and less than the 150 i PSIG maximum operating pressure fo.r the MSIV actuators, operability of all the j components is maintained.  ;

i For the small break LOCA, drywell sprays are initiated in accordance with Emergency Operating Procedures. The small break LOCA analysis assumed l drywell sprays are placed in service at 10 minutes. This lowers containment pressure and temperature and the temperature inside the piping such that  !

unacceptable pressures are not expected to occur. The safety function of this l piping is to provide pneumatic supply to one ADS SRV, to one low low set SRV i and to the inboard MSIVs to maintain their leaktightness which will be maintained. Due to the small difference between the postulated pressure and the design pressure, a method to improve the margin will be pursued.

11. Alternate N2 Supply to SRV A, B and E (X-105B-G) l The first boundary analyzed for penetration X-105B-G is between the outboard check valve and the penetration. The internal pressure in this section of piping l reaches 145.2 PSIG due to the temperature increase in the air / nitrogen contained in the attached piping inside the drywell. Calculations show the piping outside the drywellis within code allowables at 174 PSIG. Primary Containment integrity and pressure integrity of Train A of the Alternate N2 system are .

maintained which are the only safety related functions associated with this section of piping.

The second boundary analyzed for penetration X-105B-G is between the ,

penetration and SRV A, B and E. Assuming a drywell temperature of 335 F, the i internal pressure this section of piping and accumulators reaches is 145.2 PSIG. l The containment pressure at this temperature is 21.9 PSIG. Thus the internal pressure of the pipe is the difference between 145.2 PSIG and 21.9 PSIG which j equals 123.3 PSIG. Since the postulated pressure is less than the 125 PSIG design pressure of the piping and less than 125 PSIG maximum operating pressure for the SRV solenoid valves, operability of all the components is maintained.

For the small break LOCA, drywell sprays are initiated in accordance with  !

Emergency Operating Procedures. The small break LOCA analysis assumed drywell sprays are placed in service at 10 minutes. This lowers containment l

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pressure and temperature and the temperature inside the piping such that )

unacceptable pressures are not expected to occur. The safety function of this  !

piping is to provide pneumatic supply to one ADS SRV and one low low set SRV, which will be maintained. Due to the small difference between the postulated pressure and the design pressures, a method to improve the margin will be pursued.

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12. Vacuum Breaker Line (X-229B) l The boundary analyzed for the penetration is between the outboard containment  !

isolation valve and the penetration.  !

j This line is designed for 125 PSIG. The internal pressure that this section of j piping reaches is 162 PSIG due to the temperature increase in the air contained i in the attached piping inside the suppression chamber. This determination conservatively used peak drywell temperature instead of peak suppression  ;

chamber temperature. Calculations show the piping is within code allowables at 162 PSIG. Primary Containment integrity is maintained which is the only safety related function associated with this section of piping. No modification is required.

13. Demineralized Water Line (X-20) l The boundary evaluated for the penetration is between ibe outboard i containment isolation valve and the penetration. l A portion of the boundary has been drained, i.e. from the inboard containment i isolation valve to hose station valves inside the drywell. With air initially at  !

ambient pressure and temperature in this section of piping, heating to accident l

condition temperatures will only raise the pressure to approximately 8 PSIG. i The piping is designed for 62 PSIG. Primary Containment integrity is maintained I which is the only safety related function associated with this section of piping. l This section of piping is only used during outages and administrative controls I have been established to ensure the line is always drained during normal I operation. The containment isolation valves associated with this penetration  !

have very low leakage rates when tested for Appendix J, such that the drained l section of piping is not expected to refill. However, changing the valve line-up upstream of the containment isolation valves would eliminate any possibility for refilling and would be a conservative action and is being evaluated.

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I USNRC NORTHERN STATES POWER COMPANY January 28,1997 Page 10 Conclusions Reached:

All penetrations and associated piping systems components have been determined to either remain within code allowables, or to meet operability criteria. Based on the above evaluations Primary Containment integrity is maintained and no safety functions are lost.

Basis for Continued Operability:

The following table summarizes the operability determinations for each of the affected systems along with the long term fixes if required.  :

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System Operability Long Term Fix 1 Drywell Ball valves inside primary containment installation of permanent .

Equipment will be the first component to fail or pressure relieving device will be I Drain Sump relieve pressure pursued 2 Drywell Floor With leakage piping / components meet

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Installation of permanent '

Drain Sump faulted condition acceptance criteria of pressure relieving device will be ASME Section Ill, Appendix F pursued

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3 RHR Piping / components meet faulted Installation of permanent l Shutdown condition acceptance criteria of ASME pressure relieving device will be l Cooling Section Ill, Appendix F pursued  !

4 RHR Head Piping / components meet faulted Installation of permanent Spray condition acceptance criteria of ASME pressure relieving device will be i Section Ill, Appendix F pursued

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5 Reactor Water Normally hot / administrative controls Installation of permanent Cleanup established to drain if line manualiy pressure relieving device being  !

isolated evaluated 6 Main Steam Normally hot None Line Drain 7 Recirculation Normally hot / administrative controls Installation of permanent Sample Line established to drain if line manually pressure relieving device being isolated evaluated 8 Service Air Within code allowables None 9 Instrument Air Within code allowables None Outside Containment 9 Instrument Air Within code allowables Method to improve margin will be inside pursued Containment 10 Alternate N2 Within code allowables None Outside 10 Alternate N2 Within code allowables Method to improve margin will be ,

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pursued I 11 Alternate N2 Within code allowables None Outside -

11 Alternate N2 Within code allowables Method to improve margin will be Inside pursued '

12 Vacuum Within code allowables None Breaker Line 13 Demineralized Administrative controls established to Evaluate enhancements to Water Line drain line administrative controls