ML18101A949
| ML18101A949 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/15/1995 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18101A948 | List: |
| References | |
| IEB-88-008, IEB-88-8, NUDOCS 9509050136 | |
| Download: ML18101A949 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE RESPONSE TO NRC BULLETIN 88-08 "THERMAL STRESSES IN PIPING CONNECTED TO THE REACTOR COOLANT SYSTEM" FOR SALEM NUCLEAR GENERATING STATION. UNITS 1 AND 2 PUBLIC SERVICE ELECTRIC AND GAS COMPANY DOCKET NOS. 50-272 AND 50-311
1.0 BACKGROUND
Bulletin 88-08, "Thermal Stresses, in Piping Connected tc Reactor Coolant System," was issued June 22, 1988.
Action 3 of the Bulletin requests that licensees implement programs to provide continuing assurance that unisolable sections of all piping connected to the Reactor Coolant System (RCS) will not be subjected to thermal cyclic stresses that could cause fatigue failure during the remaining life of the plant.
In response to this request, Public Service Electric and Gas Company (PSE&G) provided in Reference 1 a supplemental response to the Bulletin.
As part of that response, PSE&G committed to perform thermal fatigue analyses of the Pressurizer Auxiliary Spray and Normal and Alternate Charging lines, and to provide double valve isolation on the Safety Injection lines.
2.0 EVALUATION 2.1 Fatigue Analysis of the Normal and Alternate Charging, and Auxiliary Spray Lines In Reference 2, PSE&G submitted the results of thermal fatigue analyses for the Normal and Alternate charging lines and the Auxiliary Spray line. These analyses were performed by the Westinghouse Electric Corporation (W) and reported in a proprietary report (Reference 3).
In References 4 and 5, PSE&G provided additional information regarding the W submittal, in response to requests for additional information during teleconferences held on March 1, 1994, and August 18, 1994.
The purpose of these teleconferences was to obtain clarification regarding the basis for the thermohydraulic and fatigue analyses described in the W report.
The thermal fatigue analyses were based on work performed by W under sponsorship by EPRI and provided to the NRC in Reference 6. W performed experimental studies on thermal stratification and cycling in piping within the scope of Bulletins 88-08 and 88-11, and developed a methodology for calculating the thermal cyclic stresses in piping subjected to inadvertent low leakage flows.
This methodology is based on standard heat transfer analysis and thermal stress analysis principles. These stress ~anges and cycles are incorporated in ASME Section III Class 1 plant fatigue analyses to determine the cumulative usage factor (CUF) and the expected life of the piping.
9509050136 950815 PDR *ADOCfa<.05000272 G
j. ~ performed fatigue analyses based on the requirement of ASME Section III, 1986 Edition, Subsection NB-3653 and the fatigue curves in Figures I-9.2.1 and I-9.2.2 for austenitic stainless steels. Based on these analyses, ~ concluded that the CUFs for the Normal and Alternate Charging lines for design transients and postulated isolation valve leakage transients were less than 1.0 for the life of the plant.
The CUF for the Auxiliary Spray line under the same transients was determined to be less than 1.0 for 24 calendar years since plant start-up. The calculation of these values were based on conservative assumptions regarding the leakage flow rate, the thermal cycling frequency and magnitude, and certain aspects of Reference 6. (This reference has not been evaluated nor accepted by the staff.) The staff has reviewed the bases for these analyses and finds that, as described in Reference 3, they appear reasonable and acceptable for application at Salem Units 1 and 2.
In Reference 2, PSE&G also committed to re-evaluate the Salem Unit 1 Auxiliary Spray line by December 31, 1998, and the Salem Unit 2 Auxiliary Spray line by December 31, 2002.
These dates were chosen 3 years prior to the calendar dates when the CUF in the respective lines will attain the ASME Section III criterion of 1.0.
The staff finds this acceptable.
However, PSE&G has not provided specific information regarding this re-evaluation.
Based on the staff review of this re-evaluation PSE&G may be required to check for leakage at the end of the 24-year time interval. Should leakage be detected, PSE&G will be required to repair or replace the affected unisolable piping and isolation valves.
If leakage is not detected, PSE&G may be required to implement a program to detect leakage in these lines, to provide the requested assurance of Action 3 of the Bulletin for the life of the plant.
2.2 Leakage Monitoring of the Safety Injection Lines In Reference 1, PSE&G also committed to address the requirements of Action 3 of the Bulletin for the safety injection lines by installing two parallel motor-operated isolation valves both upstream and downstream from the Boron Injection Tank (BIT).
The closure of both sets of valves required the installation on the BIT of a thermal relief valve to provide overpressure protection, which PSE&G had committed and was prepared to install in the Fall of 1994 in Unit 2, and in the Spring of 1995 in Unit 1.
In the Fall of 1994, PSE&G proposed ~ leakage monitoring program as an alternate method to address the concerns for the safety injection lines and avoid the installation of the
- thermal relief valve on the BIT.
Leakage monitoring was previously accepted by the staff in Reference 7 as an alternative to temperature monitoring at the Byrrin and Braidwood Nuclear Stations.
PSE&G proposes to maintain valves (SJ4 and SJ5) upstream of the BIT normally open and valves (SJ12 and SJ13) downstream of the BIT normally closed.
PSE&G will perform leakage tests on a quarterly basis during Operational Modes 1 through 3, to determine that leakage past the downstream isolation valves is below the acceptance criterion of 0.05 gallons per minute (gpm).
This criterion represents the smallest measurable flow of the instrumentation.
This instrumentation consists of two in-line flow meters (rotameters) arranged in series, which are not normally in service but are used periodically to perform technical specification reactor coolant system pressure isolation valve leakage surveillance.
The low range flow meter has a range of 0-1 gpm; stated to be accurate to 2% full scale. These meters are located in the Auxiliary Building, require little maintenance other than periodic cleaning and visual inspection, and are replaced every 5 years.
During a teleconference on October 9, 1994, and a subsequent meeting on October 27, 1994, the staff expressed a concern that leakage past the closed isolation valves might not be detected by the proposed test method.
In References 8 and 9, PSE&G provided responses to the staff's concern.
To address this concern, PSE&G performed a test during the last Salem Unit 2 outage and demonstrated that the low range flow meter is capable of detecting leakage through the flow lines at flow rates as low as 0.05 gpm.
The staff finds this acceptable.
PSE&G also stated that to-date no leakage through the safety injection lines has been detected. Should leakage be detected, PSE&G stated that corrective actions will be taken to ensure that the leakage criterion is not. exceeded.
The staff finds this acceptable, subject to the condition that these corrective actions ensure that no measurable leakage is detected at the end of each refueling outage, for the life of the plant.
The staff has reviewed the proposed monitoring program by PSE&G, and finds it acceptable, subject to the condition stated above.
- 3. O CONCLUSION The staff finds that PSE&G has provided the required assurance for the life of the plant, as requested in Action 3 of Bulletin 88-08, for the Safety Injection, and the Normal and Alternate Charging lines.
For the Auxiliary Spray lines, PSE&G has provided this assurance for 24 calendar years.
To provide the required assurance of Action 3 for these lines, PSE&G has committed to re-evaluate this line in Units 1 and 2 approximately 3 years prior to exceeding the 24-year time interval for each unit. Depending on the staff review of this re-evaluation, PSE&G may be required to check for leakage at the end of the 24-year time interval. If leakage is detected, PSE&G will be required to replace or repair the affected unisolable piping and the isolation valves.
If no leakage is detected, PSE&G may be required to implement a program to detect leakage in these lines.
For the Safety Injection lines in Units 1 and 2, the staff finds that PSE&G will provide the required assurance for the life of the plant by implementing a leakage monitoring program, which will detect and preclude leakage through the isolation valves on thes.e lines.
Principal Contributor:
M. Hartzman Date: August 15, 1995
4.0 REFERENCES
- 1.
Letter of March 1, 1993, from S. LaBruna, PSE&G, to the U.S. NRC Document Control Desk.
- 2.
Letter of January 24, 1994, from S. LaBruna, PSE&G, to the U.S. NRC Document Control Desk.
- 3.
WCAP-13898, "Evaluation of Salem Units 1 and 2 Charging, Alternate Charging and Auxiliary Spray Piping Per NRC Bulletin 88-08," Westinghouse Electric Corporation, December 1993. (Proprietary.}
- 4.
Letter of July 19, 1994, from S. LaBruna, PSE&G, to the U.S. NRC Document Control Desk.
- 5.
Letter of October 21, 1994, from S. LaBruna, PSE&G, to the U.S. NRC Document Control Desk.
- 6.
Letter of September 30, 1994, from A. Kenny, EPRI, to T. L. Chan, NRC, with enclosed EPRI Report TR-103581, "Thermal Stratification, Cycling, and Striping (TASCS},
11 March 1994. (Proprietary, Confidential.}
- 7.
Letter of September 24, 1992, from R. Pulsifer, NRC, to T. J. Kovatch, Commonwealth Edison Company.
- 8.
Letter of December 22, 1994, from S. LaBruna, PSE&G, to the U.S. NRC Document Control Desk.
- 9.
Letter of January 26, 1995, from H. Onorato, PSE&G, to L. Olshan, U.S. NRC.