ML18106A595

Forwards Correction to Licensee 971020 Response to RAI Re GL 96-06.Attachment of Subject Ltr Was Transmitted W/Portion of Paragraph on Page 5 Truncated.Replacement of Attachment, Requested

ML18106A595
Person / Time
Site:	Salem
Issue date:	05/08/1998
From:	Dawn Powell Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-96-06, GL-96-6, LR-N980204, NUDOCS 9805190011
Download: ML18106A595 (7)

Text

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• Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit MAY 0 81,998 LR-N980204 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Ladies and Gentlemen:

CORRETION TO REQUEST FOR ADDITIONAL INFORMATION GENERIC LETTER 96-06 SALEM GENERATING STATION UNIT NOS. 1AND2 FACILITY OPERA TING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 I

On October 20, 1997, Public Service Electric and Gas (PSE&G) submitted letter LR-N970664 in response to the NRC's Request for Additional Information (RAI) on Generic Letter 96-06 dated September 19, 1997. Due to an administrative error, the attachment of letter LR-N970664 was transmitted to the NRC with a portion of a paragraph on page 5 of the original attachment being truncated. This letter is re-transmitting the attachment to letter LR-N970664 with the additional information marked by a revision bar in the margin. Please replace the attachment contained with the October 20, 1997, letter with the corrected attachment transmitted with this letter.

If you have any questions concerning the above information, please do not hesitate to contact us.

Sincerely, D.R. Powell Director - Licensing/Regulation and 980519-ooii 9065oif -- Fuels PDR ADOCK 05000272 p PDR The power is in your hands.

95-2168 REV. 6/94 I_

Document Control Desk

• 2 MAY 0 8'JS98 LR-N980204 C Mr. Hubert J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. P. Milano, Licensing Project Manager - Salem U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Morris (X24)

USNRC Senior Resident Inspector - Salem Mr. K. Tosch, Manager, IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625

LR-N970664 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GENERIC LETTER 96-06 SALEM GENERATING STATION UNITS 1AND2 DOCKET NOS. 50-272 AND 50-311 FACILITY OPERATING LICENSES DPR-70 AND DPR-75

1. Provide a list of the pipelines penetrating containment that are susceptible to thermally-induced pressurization. For each susceptible pipeline, identify which of the following approaches were selected for resolution:

a) Document piping system operability against the appropriate Code requirements; b) Analytically demonstrate that the isolation valve will act as a controlled leakage point; c) Where practical, affected sections of piping will be drained by procedure during system alignment; or d) Install relief valves.

PSE&G Response to Question #1:

The PSE&G response to GL 96-06 dated January 28, 1997 (Reference LR-N97072) described the screening criteria used to identify piping susceptible to thermally-induced pressurization. The following table lists those penetrations that were identified as susceptible to thermally-induced pressurization for Salem Unit 2:

Penetration # Line Description Approach for Resolution M25 Safety Injection test/drain line a,d M25A Accumulator sample line a,d E30C Containment Pressure capillary tubing a E54A RVLIS capillary tubing a E54D RVLIS capillary tubing a E54E RVLIS capillary tubing a E55A RVLIS capillary tubing a 1 of 5

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E55C Containment Pressure capillary tubing a E55E RVLIS capillary tubing a E55F RVLIS capillary tubing a M22B Containment Pressure capillary tubing a M23 Containment Pressure capillary tubing a M24B Containment Pressure capillary tubing a M25B Containment Pressure capillary tubing a E30F Sample return line c E36D Sample return line c M17 Dead weight tester line to pressurizer c M66 Refueling cavity to refueling water c purification pump M66A Refueling water purification pump to c refueling cavity M22 Demineralized water supply to d containment hose connections M22A Primary water to pressurizer relief tank d line M27 Reactor Coolant Drain Tank discharge d M45 Containment sump pump discharge to d waste liquid

2. Provide summaries of the evaluations of those pipelines where approach a or b was used for resolution of thermally-induced pressurization. These summaries should describe the method of analysis, assumptions used in analysis, and results. Also include piping fabrication drawings.

PSE&G Response to Question #2:

Approach (a) was used to evaluate fourteen lines. Approach (b) was not utilized.

With regard to approach (a), two calculations were completed to evaluate these fourteen lines. The lines associated with penetrations M25 and M25A were the subject of .one calculation. The balance of lines are capillary tubing and have been evaluated under a second calculation. The following discusses these lines by the calculation methodology used.

Penetrations M25 and M25A:

Figure 1 provides a sketch and fabrication spool drawings of the M25 penetration piping between isolation valves 2SJ60, 2SJ123, 2SJ53 and drain valves 2SJ332 and 2SJ238. The piping is 3/4 inch diameter, schedule 160, ASTM A376, Grade TP316 seamless pipe. This piping was evaluated to Code requirements (approach (a)).

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J Figure 2 provides a sketch and fabrication drawings of the M25A penetration tubing between isolation valves 2SS103 and 2SS27. This line is 3/8 inch diameter, ASTM A213, Grade TP304 seamless tubing with a wall thickness of 0.065 inches. This tubing was evaluated to Code requirements (approach (a)).

The evaluation determined that the M25A containment isolation globe valves 2SS27 and 2SS103 are capable of maintaining their integrity when the overpressure condition is applied under the valve seat. The 2SS27 valve is acceptable "as is." The 2SS103 isolation globe valve required modification so that the overpressure condition would be applied under the valve seat. This modification has been completed.

The maximum temperature and pressure profiles during design basis accidents, i.e., main steam line break (MSL8) and loss of coolant accident (LOCA), were used to calculate the heat input into the subject piping. The resulting temperature distribution of piping contents in containment associated with penetration M25A was determined to bound the temperature distribution for piping associated with penetration M25; therefore, the temperature profile calculation was performed for penetration M25A piping only.

The increased temperature in each pipe segment inside containment and the pipe segment passing through the containment wall were determined through computer modeling. Piping outside containment was not modeled in the temperature profile calculation because temperatures are not expected to rise significantly in this piping before the peak pressure occurs. The increase in water volume due to thermal expansion and expansion of the tubing due to higher pipe temperatures was calculated as a function of pipe internal pressure and temperature increase. Utilizing the temperature profiles for each segment and assuming a constant mass of water in the piping, the peak internal pipe pressure was calculated using an iterative process.

For M25 penetration piping, the design code of record is ANSI 831.1.0 - 1967, "Power Piping" for design and ANSI 831. 7 - 1969 for pipe materials. The maximum pressure due to thermal overpressurization was calculated to be less than 3000 psia. This results in a pipe stress of approximately 7, 192 psi. Since the ANSI 831.7 maximum allowable stress is 16,000 psi, thermal pressurization of this piping would result in pipe stress well within the code allowable stress values.

For M25 penetration valves, the maximum working pressure was identified based on the valve pressure class and material group (Reference ANSI 816.5 and 816.34). The maximum allowable working pressure was determined to be 3415 psig. The actual calculated pressure was determined to be 3000 psia; therefore the valves associated with penetration M25 also maintain their integrity during thermal overpressurization conditions and the penetration is considered acceptable.

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I For M25A tubing, the design code of record is ASME Ill, "Rules for Construction of Nuclear Power Plant Components/' 1974 edition. The maximum pressure in penetration M25A tubing was calculated to be less than 7000 psia. This results in the M25A tubing stresses of approximately 21,000 psi, or 86% of its yield strength. Since ASME Ill Table 1-7.2 note 2 provides for allowable stress values up to 90% yield strength, the stress resulting from thermal pressurization of this tubing is within the code allowable stress value.

For penetration M25A valves, the maximum working pressure for the M25A isolation valves was identified based on the valve pressure class and material group (Reference ANSI 816.5 and 816.34). An evaluation of valve integrity concluded that the M25A containment isolation valves will maintain their integrity during thermal overpressurization conditions when the overpressure condition is applied under the seat. Since these globe valves are installed such that the overpressure condition will be applied under the valve seat, penetration M25A valves will maintain their integrity during thermal overpressurization conditions and the penetration is considered acceptable.

Please note that approach (d) is also applicable to M25 and M25A penetration piping. In addition to the evaluations summarized above for the potential thermal overpressurization of piping penetrating containment, PSE&G evaluated piping segments located inside containment between leak tight isolation valves for overpressurization vulnerabilities. These evaluations identified that certain isolation valves inside containment, including those for M25 and M25A piping, required overpressure protection. As a result PSE&G installed relief valves upstream of 2SJ123 (M25 piping) and 2SS103 (M25A piping) to provide the required overpressure protection. Therefore, the table refers to approach (d) in addition to approach (a) for penetrations M25 and M25A.

Penetrations E30C, E54A, E54D, E54E, ESSA, E55C, ES5E, E55F, M22B, M23, M248 and M2SB:

Approach (a) was used for the lines associated with these penetrations.

Penetrations E30C, E55C, M228, M23, M248, M258 are containment atmospheric pressure sensing lines. These lines are 1/8 inch diameter ASTM 213 Grade 304 stainless steel capillary tubes filled with silicon oil. The lines sense containment pressure through a sensor bellows, exit containment though a containment penetration, and terminate at a pressure transmitter. Figure 3 provides the fabrication spool drawings for these lines.

Penetrations E54A, E54D, E54E, E55A, E55E and E55F are reactor vessel level sensing lines. These lines are 3/16 inch outer diameter, 3/32 inch inner diameter, Grade 304 stainless steel tubing filled with distilled water. In containment the lines sense reactor level through a hydraulic high volume sensor 4 of 5

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bellows of Grade 316L stainless steel. Each line exits containment and terminates on the process side of a differential pressure switch through a Grade 316 stainless steel bellows. Figure 4 provides fabrication drawings for these lines.

The reactor vessel level sensing lines E54A, D, E and E55A, E and F were evaluated to bound the containment pressure sensing line overpressure condition based on significantly shorter tubing runs than the reactor vessel level sensing lines. Also, the silicon oil in the containment pressure sensing lines has a lower coefficient of expansion than the distilled water in the reactor vessel sensing lines.

To calculate the maximum pressure in the reactor vessel tubing, the temperature profile during a design basis accident (MSLB or LOCA) was identified. The containment temperature profile was segmented into time intervals, and the containment temperature increase identified for each time interval. Heat transfer by convection into the capillary tubing inventory was calculated to a maximum

~emperature of 270F. The limiting distilled water inventory was calculated to expand by 2.18 cubic inches. The sensor bellows allows for expansion of 7.5.

cubic inches. Therefore, the sensor bellows will expand to accommodate the thermal overpressure condition without challenging the tubing.

3. Description of evaluation criteria and Salem licensing basis criteria.

PSE&G Response to Question #3:

The applicable licensing basis for the M25 penetration piping is the design -

requirements of ANSI 831.1.0-1967, "Power Piping" and the material requirements of ANSI 831.7 -1969, "Nuclear Power Piping." The allowable stresses resulting from thermal over pressure conditions were determined to be within the allowable values provided by the design code.

The applicable licensing basis for the M25A penetration tubing is the design and material requirements of Section Ill of the ASME Boiler and Pressure Vessel Code, "Nuclear Power Plant Components," July 1974 through 1975 Winter Addenda. The allowable stresses resulting from thermal over pressure conditions were determined to be within the allowable values provided by the design code.

The applicable licensing basis for the tubing passing through penetrations E30C, E54A, E54D, E54E, E55A, E55C, E55E, E55F, M22B, M23, M24B and M25B is ANSI 831.1.0 - 1967 for design requirements and ANSI 831.7 - 1967 for material requirements. Since hydraulic sensor bellows are installed in these lines that are capable of expanding during a thermal overpressurization event, the lines remain within code allowable stress values.

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