ML20196H367

Forwards Response to NRC 981029 RAI Re Further Info to Facilitate Completion of NRC Review of Pilgrims 120-day Response to NRC GL 96-06

ML20196H367
Person / Time
Site:	Pilgrim
Issue date:	11/30/1998
From:	Alexander J BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BECO-2.98.154, GL-96-06, GL-96-6, TAC-M96851, NUDOCS 9812090053
Download: ML20196H367 (6)

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g GL96-06 v

Sosfore 5 Neor Pilgrim Nuclear Power Station Rocky Hill Road ,

Plymouth, Massachusetts 02360-5599 t

l November 30,1998 l BECo Ltr. 2.93.154 I U.S. Nuclear Regulatory Commission  ;

ATTN: Document Control Desk l Washington, D.C. 20555-0001 1 Docket No. 50-293 License No. DPR-35 Response to Reauest for Additional Information Dated October 29.1998 For Resolution of Generic Letter (GL) 96-06 issues at Pilarim Nuclear Power Station. Unit 1 (TAC No. M96851)

This letter responds to the NRC Request for Additional Information (RAI) dated October 29,1998. The RAI requested further information to facilitate completion of the NRC's review of Pilgrim's January 28, 1998, 120 day response to Generic Letter 96-06, " Assurance of Equipment Operation and Containment integrity During Design Basis Accident Conditions." The responses to the NRC's four questions are provided as an attachment to this letter.

This letter contains no commitments. Should the NRC require further information on this issue, please contact P.M. Kahler at (508) 830-7939.

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.FAlexander R'egulatory Relations

. Group Manager PMK/bjt

! 298154 l 3 1 \

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9812090053 981130 f3 PDR ADOCK 05000293 E3 P PDR u O L

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. *op: Mr. Alan B. Wang, Project Manager U.S. Nuclear Regulatory Commission

, -. .- Project Directorate 13 Region I

! Omos of Nuclear Reactor Regulation 475 Allendale Road -

i - Mall Stop: OWFN 14B20 King of Prussia, PA 19406

!. U. S. Nuclear Regulatory Commission j l 1 White Flirit North Sr. Resident inspector )

11555 Rociwille Pike Pilgrim Nuclear Power Station j Rockville, MD 20852- 1 i

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Attachment to BECo Letter 2.98.154 Response to GL96-06 RAI Dated October 29,1998 The GL96-06 RAI contained four specific requests for information. The following is Pilgrim's response to the request.

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References:

1. NRC Generic Letter 96-06," Assurance of Equipment Operability and l Containment Integrity During Design-Basis Accident Conditions", dated September 30,1996.

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2. Letter from L.J. Olivier, BECo, to US NRC Document Control Desk,"120 Day Response to Generic Letter 96-06, Assurance of Equipment Operability and j integrity During Design-Basis Accident Conditions", dated January 28,1997.

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3. Letter from Alan B. Wang, US NRC. to L.J. Olivier, BECo, " Request for Additional l Information for Resolution of Generic Letter (GL) 96-06 issues at Pilgrim Nuclear i

Power Station, Unit 1 (TAC No. M968' ;", dated October 29,1998.

Reauest 1 Has the relief valve for the two reactor building closed cooling water lines been added to the IST program?

Response to Reauest 1 l Yes. Relief Valve PSV-4033 has been added to the IST program (PNPS Procedure 8.l.1.1).

Reauest 2 Have pressure relief devices been added to the two sump pump discharge lines?

Response to Reauest 2 l

To ensure that pressure could not build up in the sump pump discharge lines, a small hole was drilled in a pump discharge check valve disk in each line during Refuel Outage 11. The hole allows the pressure to be relieved back to the sump. This approach was taken due to the unavailability of 1 pressure relieving devices to support implementation of the corrective action during RFO 11. This I approach is also better with respect to ALARA, since rupture disks or relief valves would require i periodic testing or replacement. '

Reauest 3 Have the valves inside containment on the core spray sample line been opened?

Response to Reauest 3 Yes. Valves 1400-63A and 1400-64A were opened during RFO 11.

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Reaupst 4 l For the residual heat r:movil (RHR) shutdown cooling (SDC) lins pI=ss provids the following:  !

. a) Provide the applicable design criteria for the piping and valves including the required load -l cornbinations,  ;

i b) Provide a drawing of the piping run between the isolation valves induding the lengths and l thicknesses of the piping segments and the type and thickness of the insulation, j c) Provide the maximum-calculated temperature and pressure for the pipe run. Describe, in- l l detail the method used to calculate these pressure and temperature values. This should indude a discussion of the heat transfer model used in the analysis and the basis for the heat transfer coeffidents usedin the analyses.  ;

I Response to Reauest 4 -

l l c) The original piping design of the RHR SDC suction line was USAS B31.1 (1967 Ed.) as amended  ;

j - by the PNPS FSAR Appendices A and C. During the recirculation piping replacement project, the stainless steel portion of piping downstream of valve MO-1001-50 was replaced. The replacement was ,

l . performed in accordance with ASME Boller and Pressure Vessel (B&PV) Code Section Ill,1980 Edition I

with Winter 1980 Addenda. As such, the RHR SDC suction line within the primary containment was

I reanalyzed in accordance with ASME B&PV Code Section Ill.

i The design pressure and temperature ratings of the RHR SDC suction line between MO-1001-50 and  :

j MO-1001-47 are 1135 psig and 562*F, respectively. The piping materials are as follows: l i

20 inch diamete Stainless steel SA-358 Gr. 316NG 0.786 inch minimum wall  !

20 inch diameter Carbon steel A-106 Gr. B 1.031 inch nominal wall i 1 inch diameter Carbon steel A-333 Gr. 6 0.250 inch nominal wall 3/4 inch diameter Carbon steel A-333 Gr. 6 0.219 inch nominal wall l

Valve MO-1001-50 is a Pressure Class 600 stainless steel (A351 CF8M) valve.

1 l_ Valve MO-1001-47 is a Pressure Class 600 carbon steel (A216WCB) valve. )

! In accordance with PNPS UFSAR Appendix A.3 and Appendix C.3, the piping was evaluated using a load combination of :

i l Dead Weight + intemal Pressure + Safe Shutdown Earthquake.  !

b) See attached drawing.

L c) A new analysis was performed on Pilgrim's RHR SDC suction line since our 120 day response. The r:sults of this analysis are discossed below.

_ The maximum-calculated temperature and pressure for the portion of piping within the containment are 260*F and 2894 psig. For the piping outside containment, the maximum calculated temperature and pressure are 143*F and 2894 psig.

l Due to the difference in environmental temperatures in the drywell and the RHR 'A' valve room due to i

the LOCA, this isolated portion of piping was split into two segments to perform the analysis. To

( ctocount for axial heat transfer along the pipe axis, which would determine the length of each segment, j a pin-fin model was used to calculate the temperature distribution along the pipe axis. The calculation l

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j determined t the axial thermal decay length based on a 10% crit:rion b approximitaly 14.9 f=t.

Since the decay b an cxponential decay, the effective length of piping inside the drywell subject ts 260*F was increased by 5 feet to account for tha axtl heat transf:r.

The thermal transient on the fluid inside the piping was determined using an ANSYS (Version 5.4)

, thermal model. Since insulation on the piping inside the containment was assumed to have been

' dislodged due to LOCA forces, separate models were developed for each segment. For piping inside l the containment, the bounding Main Steam Line Break (MSLB) temperature profile was used in the cnalysis. The analysis assumed that the MSLB temperature p ofile was at saturated steam conditions 7 within the drywell which results in higher than actual heat transfer ( based on steam condensation on the piping outer surface) between the drywell air and the pipe outer surface. Since the steam is not at i saturated conditions following the MSLB, this is a conservative assumption. The heat transfer i coefficient between the pipe inner surface and the water was determined using formulas for heat transfer based on free convection in an enclosed space.

I ~A similar analysis for the piping outside containment was performed except that the pipe insulation was l not assumed to be ' dislodged' as a result of the transient.

I- The results of the thermal models were then used as input into an ANSYS structural model. The ANSYS model is axisymmetric and consisted of:

o PLANE 42 2-D structural solid elements to represent the pipe elements, o SHELL 51 structural elements to provide a pressure boundary at one end of the pipe, c' FLUID 79 2-D contained fluid elements to represent the water, and o The nodes.were fully restrained in the axial direction at the other end of the pipe to provide a pressure boundary.

The ANSYS structural model expands the water based on the thermal transient, in conjunction with Expansion of the piping due to thermal and pressure expansion, to determine the maximum-calculated intemal fluid pressure.

Using the loading combinations required by the PNPS UFSAR, the resulting intemal fluid pressure of 2894 psig was determined to result in piping stresses within the UFSAR limits for this type of event.

l Residual Heat Removal Shutdown Cooling Supply Penetration X-12 Piping i l

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i 2" Nukon Blanket Insulation (Assumed DCA =20" Dia.,0.786" min. wall, SA-358 Gr.316NG Dislogded by LOCA ) . EL =20" Dia.,1.031" nom. wall, A-106 Gr. B Valve MO-1001-50 i 1'11 1 2" Asbestos insulation In 5 '6"/ Penetration Area i l

&EL  ! d( EL 45' 1" l O'

<l 15'8" >< 15'4" >4'

--D' :l 1' 2" Insulation Outside Containment is Mix 17'8" of Approx.50% Asbestos and 50% Fiberglass Blanket (2" Thick) k 4 EL k - EL 27' 5" HB Valve v MO-1001-47 l

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