ML20198N938

Provides Response to RAI Dtd 970916,re GL 96-06, Assurance of Equipment Operability & Containment Integrity During DBA Conditions. W/40 Oversize Drawings

ML20198N938
Person / Time
Site:	Monticello
Issue date:	10/30/1997
From:	Schibonski C NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-96-06, GL-96-6, TAC-M96835, NUDOCS 9711060197
Download: ML20198N938 (52)

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Northem States Power Company Montice i uc ear Generating Plant Monticello. Minnesota 65362-9637 October 30,1997 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 Response to Request for AdditionalInformation Dated September 16,1997 Related to the Monticello Rosponso to Generic Letter 96-06 (TAC NO M96835)

On October 25,1996 and January 28,1997 Monticello provided the requuod responses to Generic Letter 96-06 "Assuranco of Equipment Operability and Containment Integrity During Design Basis Accident Conditions."

By letter dated September 16,1997, the NRC staff requested additional information in order to complete review of Northern States Powers response to the Generic Letter. Attachment A and B to this letter provide the information requested by the staff.

This submittal contains no new NRC commitments, nor does it modify any prior commitments.

Please contact Sam Shirey,JS _Lipensing Engineer, at (612) 295-1449 if you require additional i rniER n re ated to 4qu sl.

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Craig . Schibenski Actin Plant Manager l' Mo icollo Nuclear Generating Plant (D

c: Regional Administrator lit, NRC w/o Att. B g \

NRR Project Manager, NRC w/o Att. B Sr Resident Inspector, NRC w/o Att. B Stato of Minnesota - w/o Att. B Attn: Kris Sanda s , (j , L I i,b' J Silberg w/o Att. B D1 o '

Affidavit to the US Nuclear Regulatory Commission

Attachment:

A: Montive!!o Nuclear Generating Plant Response to NRC Request for Additional Information dated September 16,1997 Related to the Monticello Response to Genoric Letter 96-06.

Attachment B: NRC requested drawings.

9711060197 971030-PDR ADOCK 05000263 1 E,N *

Mi P pon WM91 l H 4REmuGAnbRAtW1C

UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50 263 RESPONSE TO NRC REQUEST FOR ADDITIONAL lt" 3RMATION ON GENERIC LETTER 96-06 Northern States Power Company, a Minnosota corporatk.), by lettor dated October 30, 1997, providos the requested response to NRC Request for Additional Information on Gonoric Letter 96 06 dated September 16,1997. This letter contains no restricted or other dofonso information.

NORTH RN ATES P WE - RY By _ e ~ / ~

Crals chibonski Acti Plant Manager nticello Nuclear Generating Plant On this 30 day of OCTDb W $997 before me a notary puolic in and for said County, personally appeared Craig Schibonski, Acting Plant Manager, Monticello Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to oxocuto this document on behalf of Northern States Power Company, that he knows the contonts thoroof, and that to the best of his knowledge,information, and bellof the statements mado in it are true and that it is not Interposed for delay.

M pr M

, Samuel I. Shirey Notary Public Minnesota Sherburne County { - - - ^^^ ^ '~^

SM.lVEL 1. SHIREY My Commission Expiros January 31,2000 "

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Attachment A ,

Monticello Nuclear Generating Plant Response to NRC Request for Additional Infamation Dated September 16,1997 Related to the Monticello 3esoonse to Generic letter 96-06.

Reauestod Information The NRC requests that NSP provide summaries of the evaluations performed for the pipes in the thirteen affected penetrations. The summaries should include fabrication drawings of the piping sections evaluated and describe the method of analysis, including the assumptions usedin the analyses, and the results. The response from NSP should also explain how the criteria used for the evaluations meet the licensing basis criteria for Monticello andprovide the schedule fcr the completion of any modifications that may be required.

Monticello Response 4

General Method Of Analyses The piping c.omponents in the thirteen affected penetrations are designed or have been reconciled to meet the ANSI B31.1,1977 Edition through Winter 1978 addendum, code. The piping components are also designed in accordance with the Monticello USAR, Section 12.2.1.4 for ECCS load combinations. These design requirements are consistent with Monticello's licensing basis, To inciude the affects of the Load Case described in GL 96-06, the increase in pipe hoop and longitudinal stresses were calculated based upon the increase in intemal pipe pressures and were added to the stresses from the current piping analyses. As identified in GL 96-06, the increased intemal pressures are due to the water or air between closed valves being heated up due to a Design Bases Accident, in cases where the ANSI B31.1 allowable stresses could not be met, the stresses were shown to met the higher allowables of ASME B&PV Code, Section lll Appendix F 1986 Edition (ASME Section lli Appendix F). This is allowed for determining operability of Nonconforming Conditions by NRC GL 91-18, " Resolution of Degraded and Nonconforming Conditions and ca Operability",

Paragraph 6.13.

Below are additional discussions on how the intemal water pressures were calculated for penetrations X-12, X-17 and X-18. Also includad below are the results of the analysis by penetration.

, ~ Atttchment A Calculatina Water Pressure in Plos For The RHR Veer! Head Sorav Ploina and The RHR Shut Down Coolina Ploinn X-12: RHR Shutdown Coolina and X 17: RHR liead Sorav

- Temperature of the water is calculated inalde the pipe between the valve inside the Drywell, and the valve outside the Drywell. Three separate temperatures are calculated as a function of time; inside the Drywell, outside the Drywell, and at the penetration.

Driving temperature inside the Drywell is from either the small break accident _,

(SBA) LOCA, or design basis accident (DBA) LOCA whichever results in the

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higher pipe temperature. Pipe insulation inside the Drywell is assumed to be l blown off the piping, associated with the penetration. 1 Driving temperatiire outside the Drywell is the secondary containment temperature resulting from a loss of HVAC during a DBA LOCA. The pipe outside the Dry 'ellis insulated and contains a large volume of water. The analysis shows that for the duratio.1 of the event, the water temperature responds very little to the elevated Drywell temperature.

Once the pipe water temperature inside and outside the Drywell are calculated as a (anction of time, they are used as the driving temperature at the' ends of the l

water / steel pipe combination through the penetration. The temperature of the water / steel pipe combination is found as a function of time and length due to conduction from its ends. The steel pipe and water are treated as being in thermal equilibrium. The conduction heat transfer characteristics of this model (density, specific heat, and thermal conductivity) are in the same proportion as the cross sectional area.

To obtain the conduction temperature response through the length of the

- penetration, the penetration is modeled as an infinite wall with one directional

- heat transfer whose width is the same as the length of the penetration and with the driving temperature forces on each side of the wall. This is conservative since heat is lost from the penetration assembly into the Drywell bloshield. At any_ point in time, the average temperature of all the nodes through the wall is used to report the temperature of the penetration.

Pipe water temperature is solved by thermal resis'ance and capacity formulation as described in Section 4-7 of HEAT TRANSFER by Holman,6th Edition. Heat transfer is by convection from surface to fluid, and by conduction through solids.

_ Equations are modeled using a software spreadsheet which gives the temperature of each component: 1.e., pipe inside the Drywell, pipe outside the Drywell, and the penetration, each as a function of time starting from the

- beginning of the accident.

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, Att:chment A For pipe outside the Drywell, convection heat transfer coefficients from the air to the pipe were conservatively calculated using dry air properties.

Intemal convection coefficients from pipe to water were calculated per Section 7- ]

11 HEA_T TRANSFER by Holman , Free Convection in Enclosed Spaces. The RHR pipes were above the GrPr number allowed, therefore, the pipe coefficients for Free Convection ftom Vertical Surfaces, Section 7-4, and Ho:izontal Surfaces, Section 7-5, of HEAT TRANSFER by Holman were used.

The nominal calculated convection heat transfer coefficients may vary by *25%

as stated in HEAT TRANSFER by Holman, page 360. These lines were installed almst 30 years ago and have likely built up a thermally restrictive film which would tend to lower the heat transfer coefficients. For conservatism in predicting the highest water temperature, the nominal coefficient was increased by 25%

For pipe inside the Drywell, HEAT TRANSFER by Holman gives an approximation for the outside wall convecifix heat transfer coefficient of 700-a 4,400 Btu /Hr ft .F for condensation withoui jet impingement. A value of 1,000 Btu /Hr ftF was used.

Pipe insulation thermal conductivity, calculated surface convection heat transfer coefficients, and water density are corrected fc. the changes in temperatures.

For water contained between valvet, a 1*F increase in water temperature results in a 33 psi increasein water pressu.; as discussed on pages 3C-43 & 3C-44 of NUREG/CP-0152, " Proceeding of the Fourth NRC/ASME Symposium on Valve and Pump Testing", July 15-18,1996. This reference is the result of tests with heat applied outside a valve where the valve contains water, but no air. The resulting plot of pressure and temperature of the fluid is a straight line. Since the thcrmal expansion of water, thermal expansion of steel, and elastic expansion of steel are linear, it is appropriate to extend the line of NUREG/CP-0152 to 300'F which are water temperatures calculated in this calculation. The 33 psi pressure increase per degree increase is also used in NRC information notice IEIN 95-14,

" Susceptibility of Containment Sump Recirculation Gate Valves To Pressure Locking".

To obtain water pressure, each pipe segment is addressed separately. The temperature and pressure in each segment is individually calculated. The final pressure determined by combining the volume and pressure of each segment.

The combined water pressure is determined for each time step. The analyses is continued until peak water pressure is obtained.

The isolation valves are assumed closed at the beginning of the accident and -

stay closed through out the accident. Initial pressure was assumed to be the head of water acting on the penetration when the isolation valves were closed.

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. Att: chm:nt A -

Maximum calculated water pressure was 2,743 psig for X-12 and 1,411 prig for X-17, Calculatina Water Pressure In The Drywell Floor Drain Sumo Pipina

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X 18: Drvwell Floor Drain Suma

(, The methodology for calculating water pressure in the Drywell Floor Drain Sump

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piping is the same as used above for the RHR Lines with the following differences.

Both isolation valves are outside the Drywell and a higher percent of the piping is

] located inside the Drywell. The pressure boundary being heated up by the LOCA is hiside the Drywell down stream of the pump discharge check valves, out to the inner of the two isolation valves outside the Drywell. Conta!nment isolation valve measured leakage rates were used to verify that intemal pipe pressure would not exceed ASME Section Ill Appendix F allowable pressure. An equivalent leakage orifice size was calculated based on the results of the last g performed Appendix J leak rate testing.

l To obtain a leakage rate for the drain line in order to keep the pressure under the given allowrble pressure, a mass balance over a finite time period was calculated.

The calculated leakage rate required to maintain the water pressure under an ,

allowable pressure of 2,700 psig is 0.041 gpm. The calculated as left leakage '" Ae rato is 0.052 gpm at 2,700 psig.

Penetrations Which Meet the ASME Section ill Appendix F Allowables X-12: RHR Shutdown Coolina Piping components were analyzed between inboard containment isolation valve MO-2029 and outboard containment isolation valve MO-2030 inclusive. The piping components were qualified to the ASME Section lli Appendix F operability criteria for a maximum expected internal pressure of 3,306 psig.

Since 3,306 psig is greater than the peak calculated pressure of 2,743 psig, the components of this penetration meet the ASME Section ill Appendix F operability criteria.

X-17: RHR Head Sorav '

Piping components were analyzed between inboard containment isolation valve MO-2027 and outboard containment laolation valve MO-2026 inclusive. The Page 4

, Attachm:nt A piping components were qualified to the ASME Section ill Appendix F operability criteria for a maximum expected intemal pressure of 3,450 psig.

Since 3,450 psig is greater than the peak calculated pressure of 1,411 psig, the components of this penetration meet the ASME Section lli Appendix F 1 operability criteria.

X-18: Drywell Floor. Drain Sumo Piping was analyzed from penetiation X-18 to the outer most containment -

isolation valve, AO-25418. The piping components were qualified to the ASME Section lll Appendix F operability criteria for a maximum expected intei .al pressure of 2,864 psig.

Since 2,864 psig is greater than 2,700 psig and the leakage at 2,700 psig of 0.041 gpm is less than the as left leakage of 0.052 gpm, the components of this penetration meet the ASME Section Ill Appendix F operability criteria.

Additionalinformation On Remainder Of Penetrations X-20: Demineralized Wa_tgr Since this line has been drained and procedures are in place to assure the line remains drained during reactor operation, the post accident environment of the Drywell will not result in stress !evels exceeding design allowables.

X-19: Drvwell Eauioment Drain Sumo The containment valves located outside the Drywell are, AO-2561 A (2"), AO-2561B (2"), CRW-112 (3/4"), CRW-113 (3/4") and CRW-104 (1"). These valves are 600# valves with a pressure rating of 1432 psig at 120'F. The normally closed ball valve inside the Drywell, AO-2560B (2'), was hydro-tested to 750 psig. Since ANSI B31.1 piping code required hydro testing to 1.5 times the rated component capacity, the rating of the ball valve would be 750/1.5 = 500 psig.

Since the valve outside of containment is rated for 1432 psig which is over 2.5 times greater than the valve inside containm3nt rating of 500 psig, it is concluded that the valve inside the Drywell will leak before the valves outside the Drywell.

Thus relieving the prusure and maintaining primary containment integrity, in addition, since AO.2561 A, AO-2561B, and AO-2560B are 2" diameter valves, and the valve inside containment has four 1/2" diameter versus four 5/8" diameter bonnet bolts the valves inside containment will leak first since it is weaker, s

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, Att: chm:nt A X-14: RWCU The RWCU systern is normally in operation and when the system isolates during _

a LOCA the water temperature trapped between the isolation valves will be greater than the peak containment temperature. Therefore, no increase in pressure will occur due to thermal conduction. Procedures have been written to

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assure that if this line is isolated during operations, part of the line between the isolation valves would be drained.-

X-41: Recirculation Sample The Recirculation Sample system is normally in operation and when the system isolates during a LOCA the water temperature trapped between the isolation valves will be greater than the peak containment temperature. Therefore, no increase in pre sure will occur due to thermal conduction. Procedures have been written to assure that if this line is isolated during operations, part of the line between the isolation valves would be drained.

X-8 Steam Drain The line is kept hot during normal operation and will cool when the steam is isolated. Therefore, it is not susceptible to the concems of GL 96-06.

X-21: Service Air and X-229B: Vacuum Breaker Air Supolv These two penetrations supply plant service air, X-21 to the Drywell, and X 229B to the Torus. The perfect gas law is used to calculated a line pressure assuming an initial conservative operating containment temperature of 100*F for the Drywell and 70*F for the Torus. The maximum temperatures are conservatively assumed to be 335'F for the Drywell and 200 F for the Torus. The initial pressure used is the relief valve setpoint for the plant service air. The final calculated accident line air pressure is 162 psig and 141 psig for the Drywell and Torus respectively. These lines weie conservatively analyzed using a bounding pressure of 200 psig.

Stress analyses show the piping will meet ANSI B31.1 allowables. The pressure rating of the valves are 600#, well above the assumed 200 psig maximum.

X-22: Instrument Air X-34A: Attemate N2 Supolv to SRV's and X-105B-G:

Altemate N2 Supolv te SRV's These three penetrations contain instrument air and nitrogen supply pressure.

In accordance with the USAR, Table E 2-7, " Assumptions for the LOCA Containment Evaluation" initial Drywell temperature is assumed to be 135 F.

The initial pressure is assumed to be the N2 supply regulator setpoint. SBA Page 6

, Att:chment A LOCA Drywell temperature and pressure are 335'F and 21.9 psig respectively in accordance with the USAR figures 5.2-25 and 5.2-26.

Using the perfect gas law, the line pressure outside the Drywell is calculated to be 145.2 psig and the line pressure inside the Diywell is 145.2 psig - 21.9 psig =

123.3 psig.

Other than for the operation of the SRV solenold valves and the 125 psig rated accumulator tanks associated with these solenoids, the piping components are qualified to 200 psig. The valves are rated at 600#,

i The accumulator and SRV solenoid valves are rated at 125 psig which is > 123.3 psig.

The allowable stresses of ANSI B31.1 are mei.

Schedule For Completion Of Modifications

' To met the ANSI B31.1 code allowables, pressure relieving devices are to be installed on penetrations X-12, X-17, X-18 and X-19 at the next refueling outage currently scheduled for mid March,1998.

A method to increase margin between postulated pressure and design pressure on penetrations X-22, X 34A, and X-105B-G will also be implemented during the next refueling outage.

4 Fabrication Drawinos Submitted With This Response The fabrication drawings for the thirteen penetration are shcwn below. The P&lD drawing have not been included since they are included in the Monticello USAR Chapter 15.

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,, Att chm:nt A renetration P&lD Fabrication No. Drawings Drawings X-8 NH 36241 FSK 786 NH 36033 NX-13142-42 X-12 NH 36247 NX 13142-49 NF-97003 NF-36890 NX-9235 7 NX 9235-11 X 14 NH 36254 -

X-17 NH-36247 NX-13142-20 NF 36892 NX-9235-44 NX-9235-45 X 18 NH-36043 FSK-747 FSK-747-1 FSK-747 2 FSK-747-3 X 19 NH-36044 FSK 746 FSK-746-1 FSK-746 2 FSK 902 X 20 NH-36039 FSK-842 -

FSK-842-1 FSK 842-2 X-21 NH-36049-4 FSK-961 FSK-9612 X-22 N!i 36049-12 NF93216-1-1 NH-36241-1 NF93216-1-2 NF93216-5-1 NF93216-5-2 NF93216--5-4 X-34A NH-36049-10 NF-74295 NH-36241-1 NF-74295-1 NF-74295-3 NF-74295-4 NO-74226.

NF93216-5-1 NF93216-5-2 X-41 NH-36243 -

X-105B-G NH-36049-10 NF-74295 NH-36241-1 NF-93216-5-4  !

NO-74225 229B NH-36049-14 NX-13142-19-5 l

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Attachment B Monticello Nuclear Generating Plant Response to NRC Request for Additional Information Dated September 16,1997 Related to the Monticello Response to Generic letter 96-06.

Penetration No. Fabrication

, Drawings X-8 FSK 786 NX 13142 42 X 12 NX-13142-49 NF-97003 NF 36890 NX 9235-7 NX 923511 X 17 NX 13142-20 NF 36892 NX 9235-44 NX-9235-45 X 18 FSK 747 FSK 747-1 FSK 747 2 FSK 747 3 X 19 FSK 746 FSK 746-1 FSK 746-2 FSK 902 X 20 FSK-842 FSK-8421 FSK 842-2 X-21 FSK-%1 FSK-9612 X 22 NF93216-1 1 NF93216-12 1

NF93216-51 NF93216-5-2 NF93216--5-4 X-34A NF-74295 NF-74295-1 NF-74295-3 NF 74295-4 NO-74226 NF93216-51 NF93216-5 2 X-105B-G NF-74295 NF-93216-5-4 NO-74225 229B NX-13142-19-5

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