NL-14-1821, Proposed Inservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0 - Supplemental Information

Proposed Inservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0 - Supplemental Information

ML14317A261
Person / Time
Site:	Hatch
Issue date:	11/13/2014
From:	Pierce C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-14-1821 HNP-ISI-ALT-HDPE-01, Ver. 2.0
Download: ML14317A261 (6)

Text

Charles R. Pierce Southern Nuclear Regulatory Affa irs Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7872 Fax 205.992.7601 SOUTHERN<<\

NOV 1 3 2014 COMPANY Docket Nos.: 50-366 NL-14-1821 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington , D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Proposed lnservice Inspection Alternative HNP-ISI-ALT-HDPE-01, Version 2.0 Supplemental Information Ladies and Gentlemen:

By letter dated September 19, 2014, Southern Nuclear Operating Company (SNC) submitted proposed inservice inspection alternative HNP-ISI-ALT -HDPE-01, Version 2.0 (SNC letter NL-14-1250). During a public meeting held on October 14, 2014, the Nuclear Regulatory Commission (NRC) inquired about the Maximum Operating Pressure value listed in Section 1100 of the Alternative Technical Requirements (ATR) (Enclosure 2 to NL-14-1250). Enclosure 1 of this letter provides the NRC question and the SNC response.

Additionally, it was noticed prior to this public meeting that Table 321 0-3 of the ATR (page E2-23 of NL-14-1250) was inadvertently missing the "10 yr" and "50 yr" rows. Enclosure 2 of this letter provides the corrected table, and is a direct replacement for page E2-23 sent in the NL-14-1250 letter. As such, the "header" and "footer" of Enclosure 2 of this letter are consistent with page E2-23 of NL 1250.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

Respectfully submitted, C. R. Pierce Regulatory Affairs Director CRP/RMJ

U. S. Nuclear Regulatory Commission NL-14-1821 Page 2

Enclosures:

1. SNC Response to NRC Question Regarding Maximum Operating Pressure

2. Corrected Page E2-23 of Alternative Technical Requirements (Enclosure 2 to SNC letter NL-14-1250) cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski , Chairman , President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M. D. Meier, Vice President- Regulatory Affairs Mr. D. R. Madison , Vice President- Fleet Operations Mr. B. J. Adams , Vice President- Engineering Mr. G. L. Johnson , Regulatory Affairs Manager- Hatch RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch

Edwin I. Hatch Nuclear Plant- Unit 2 Proposed lnservice Inspection Alternative HNP-ISI-ALT-HDPE-01 , Version 2.0 Supplemental Information Enclosure 1 SNC Response to NRC Question Regarding Maximum Operating Pressure to NL-14-1821 SNC Response to NRC Question Regarding Maximum Operating Pressure NRC Question from October 14, 2014 Public Meeting What condition causes the 190 psig maximum operating pressure when the design pressure is only 180 psig for the Plant Service Water (PSW) Piping?

SNC Response The maximum operating pressure of 190 psig occurs in a "dead head" situation where there is zero flow in the PSW piping system while the pumps are running.

Decreases in the PSW system flow rate results in an increase in the total developed head from the pumps. Using the average total developed head from the pumps of 440 feet at 0 gallons per minute with a normal water temperature of 95 degrees Fahrenheit results in a pressure of approximately 190 psig developed from the pumps. This response is based on a review of Plant Hatch design basis documentation and discussions with senior staff of the original plant design architect engineer.

This increase in pressure is acceptable based on the design code of record for the piping. The underground PSW piping is designed to USA Standard Code for Pressure Piping (USAS) 831.7-1969. USAS 831.7-1969,3-701 states that the design conditions shall be the same as in Division 101 of USAS 831.1.0. USAS 831.1.0-1967, 101.2.2 and 102.2.4 allow for internal pressure to exceed design pressure for occasional operation for short periods if the stress in the pipe wall does not exceed the allowable stress values for the piping by more than 15%

during 10% of the operating period or by more than 20% during 1% of the operating period. The maximum operating pressure of 190 psig is only approximately 5% above the design pressure of 180 psig and would not result in a stress greater than 15% above the allowable stress. Operating the system in "dead head" condition will occur much less than 10% of the operating life of the system and is therefore acceptable.

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Edwin I. Hatch Nuclear Plant- Unit 2 Proposed lnservice Inspection Alternative HNP-181-ALT-HDPE-01, Version 2.0 Supplemental Information Enclosure 2 Corrected Page E2-23 of Alternative Technical Requirements (Enclosure 2 to SNC letter NL-14-1250) to NL-1 4-1250 Proposed Alternative Techn ical Requirements to ASME Section XI Requirements for Replacement of Class 3 Buried Piping in Accordance with 10 CFR 50 .55a(a)(3)(i)

Table 3133-1 SA, Allowable Secondary Stress Limit (psi)

Number of Equivalent Full Design Temperature Range Temperature Cycles, N 100°F 110°f 125 °F*

N::; 1000 3440 3280 3020 1000 < N::; 10000 2300 2 190 2032 10000 < N::; 25000 1950 1870 1735 25000 < N ::; 50000 1730 1650 1540 50000 < N ::;75000 1630 1540 1435 N > 75000 1530 1470 1365

• Values interpolated from standard tab le.

Table 3210-1 Maximum Allowable Ring Deflection !!max DR Qmax (%)

7.3 3.0 7 2.8*

• Value extrapolated from standard tab le.

Table 3210-2 Soil Support Factor, Fs E'N!E' (12*Brl)/D, in/in 1.5 2.0 2.5 3.0 4.0 5.0

0. 1 0.15 0.30 0.60 0.80 0.90 1.00 0.2 0.30 0.45 0.70 0.85 0.92 1.00 0.4 0.50 0.60 0.80 0.90 0.95 1.00 0.6 0.70 0.80 0.90 0.95 1.00 1.00 0.8 0.85 0.90 0.95 0.98 1.00 1.00 1.0 1.00 1.00 1.00 1.00 1.00 1.00 1.5 1.30 1.15 1.10 1.05 1.00 1.00 2.0 1.50 1.30 1.15 1. 10 1.05 1.00 3.0 1.75 1.45 1.30 1.20 1.08 1.00 5.0 2.00 1.60 1.40 1.25 1.10 1.00 TABLE 3210-3 MODULUS OF ELASTICITY OF POLYETHYLENE PIPE Epipe (psi)

Load Temperature (°F)

Duratio n  ::; 73 95

• 100 11 0 125*

0.5 hr 82000 63550 59900 52500 44300 1 hr 78000 60450 56900 49900 42 100 10 hr 65000 50400 47500 4 1600 35100 24 hr 60000 46500 43800 38400 32400 100 hr 55000 42650 40200 35200 29700 1000 hr 46000 35650 33600 29400 24850 I yr 40000 31000 29200 25600 21600 10 yr 34000 26350 24800 2 1800 18350 50 yr 29000 22500 21200 18600 15650

• Values interpolated from standard table.

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