ML20134J963
| ML20134J963 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/06/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20134J952 | List: |
| References | |
| NUDOCS 9702130039 | |
| Download: ML20134J963 (5) | |
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S UNITED STATES f
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 4001
.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REriULATION l
l RELATED TO AMENDMENT NOS. 191 AND 174 TO FACILITY OPERATING l
LICENSE NOS. DPR-70 AND DPR-75 1
PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY 1
i SALEM NUCLEAR GENERATING STATION. UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311
1.0 INTRODUCTION
By letter dated May 31, 1996, as supplemented December 23, 1996, the Public Service Electric & Gas Company (the licensee) submitted a request for changes to the Salem Nuclear Generating Station, Unit Nos. I and 2, Technical Specifications (TSs). The requested changes would (1) revise the reactor vessel level indication system (RVLIS), (2) revise the channel calibration definition to better address temperature detector channel calibration methodology, and (3) delete a requirement to install a jumper in the auxiliary feedwater actuation logic. The December 23, 1996, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION 2.1 Proposed TS Change To Revise RVLIS Action Statements The licensee proposes to revise the action statements associated with RVLIS.
RVLIS is part of the safety related display information that the operator uses to perform manual functions and to determine the effect of the manual actions following a reactor trip and subsequent shutdown.
RVLIS monitors the reactor vessel during abnormal plant conditions by using differential pressure (d/p) transmitters to measure vessel level or relative void content of the fluid surrounding the core.
RVLIS was upgraded during the Unit I and Unit 2 refueling outages in 1992 and l
1991. During the timeframe that RVLIS was being upgraded, a temporary variance at each unit allowed both RVLIS channels to be inoperable as long as the i
l required channels for reactor coolant system subcooling margin monitor and the
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core exit thermocouples (T/C) were operable. This temporary variance, as I
9702130039 970206 PDR ADOCK 05000272 P
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. j discussed in detail in Amendment 117 for Unit I dated February 12, 1991, and i
Amendment 95 for Unit 2 dated September 10, 1990, has now expired on both units.
As indicated in the enclosures to the February 12, 1991 staff correspondence, the note at the bottom of Table 3.3-11 page 3/4 3-55, " Accident Monitoring Instrumentation," Amendment 117 for Unit 1 indicates that Action 8 remains in i
effect until startup from the 10th refueling outage at which time RVLIS will be upgraded, and upon expiration, Action Statements 1 and 2 will apply. As indicated in the enclosures to the September 10, 1990 staff correspondence, the note at the bottom of Table 3.3-11, page 3/4 3-51, " Accident Monitoring Instrumentation," Amendment 95 for Unit 2, indicates that Action 8 will remain I
in effect until startup from the 6th refueling outage at which time RVLIS will be upgraded, and upon expiration, Action Statements 1 and 2 will apply.
Action Statement 1 indicates that with one channel of RVLIS accident monitoring instrumentation inoperable, restore the inoperable channel to operable status within 7 days or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Action statement 2 indicates that with two channels of RVLIS accident monitoring instrumentation inoperable, restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within the next 12 4
hours.
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The licensee proposes to change the present RVLIS TS action statements from a TS requirement which requires shutdown in 7 days if one channel is inoperable and shutdown in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if two channels are inoperable to a TS requirement which requires the promulgation of a special report if one channel is inoperable for 30 days and a special report if two channels are inoperable for 7 days. Thus, in the event of RVLIS inoperability, unit shutdown is not required. This proposed action is consistent with the guidelines of NUREG 1431, Revision 1, Westinghouse Standard Technical Specifications.
The special report is appropriate in lieu of shutdown since (1) alternative actions are identified in the TS before the loss of RVLIS functional capability, and (2) there is a low probability of unit conditions that would require information provided by the RVLIS instrumentation in that its function is only required during reactor shutdown conditions.
The requirements for the special report will be delineated in a new Salem Units 1 and 2 administrative Technical Specification Special Reports Section 6.9.4.
As indicated, the requirements for the special report in the administrative TSs are consistent with the action statements and reporting requirements for RVLIS as indicated in NUREG-1431, Revision 1, however, whereas the Westinghouse Standard TS states, "The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule...," the licensee's proposed change states, "The report shall outline the preplanned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and
. the plans and schedule..."
The licensee indicates that the phrase "for inadequate core cooling" was added to clarify the monitoring function required since the proposed change only pertains to RVLIS whose sole function is to monitor for inadequate core cooling.
The staff agrees with the licensee's clarification and finds the above discussed TS change to revise the RVLIS action statements to be acceptable.
2.2 Proposed TS Change To Revise The Channel Calibration Definition To Better Address Temperature Detector Channel Calibration Methodology The licensee requested that Salem Units 1 and 2 TS Definition 1.4 pertaining to Channel Calibration be revised to replace the existing definition. The licensee proposes this revision to better account for temperature detector channel calibration methodology.
The licensee indicated that the proposed change ensures the required testing methodology aligns with standard industry methodology for instrument channels having a T/C or resistance temperature detector (RTD) as a sensor in order to prevent unnecessary removal of these sensors. Removal and reinsta11ation of RTDs solely for the purpose of calibration could introduce errors and increase personnel exposure. The licensee currently performs a cross calibration of the reactor coolant RTDs on a refueling outage frequency.
The licensee's current practice for completing channel calibrations for instrument channels having RTD or T/C sensors is to perform an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel, which is consistent with standard industry practice.
The issue of cross calibration was addressed in NUREG/CR-5560, " Aging of Nuclear Plant Resistance Temperature Detetors," which recognizes that on-line cross calibration can be a reasonable method for RTD calibration. However, as stated in NUREG/CR-5560, to perform in-situ calibration would normally require one or more newly calibrated RTDs to be used as a reference. Without a reference, the cross calibration will not account for common mode (systematic) drift and will only provide information on the consistency and not the accuracy of the installed RTDs. The cross calibration technique assumes that the average of the RTD measurements represents the true process temperature and that RTD drift is random and not systematic.
The results of studies referenced in NUREG/CR-5560 indicate that RTD drift is usually random.
However, the particular testing done to validate the cross calibration methodology in NUREG/CR-5560 utilized newly calibrated RTDs for the test.
The licensee's proposed TS change is consistent with the manner in which NUREG-1431, Westinghouse STS, defines channel calibration. At present, the licensee performs a cross calibration every refueling outage, but without a freshly calibrated RTD or T/C. The licensee proposes that whenever an RTD or T/C sens?ng element is replaced, the next required channel calibration will
9 include an inplace cross calibration which compares the other sensing elements with the recently installed sensing element. This cross calibration methodology provides for an on-line calibration testing capability.
The staff agrees with the licensee's proposed change to the definition of channel calibration and finds it acceptable.
2.3 Proposed TS Change To Delete A Requirement To Install A Jumper In The Auxiliary Feedwater Actuation Logic The licensee proposes to delete a TS requirement to install a jumper in the auxiliary feedwater actuation logic during certain required surveillance testing. This proposed TS change has been requested due to a design change in which the jumper function is now performed by an installed relay.
Previously, a requirement to install a jumper was added to Action Statement 21.b of Table 3.3-3 via Amendment 39 for Unit I and Amendment 116 for Unit 2.
The requirement states, "If the affected main feedwater pump is expected to be out of service for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the inoperable channel is jumpared so as to enable the start circuit of the auxiliary feedwater pump upon loss of the other main feedwater pump." This requirement ensured the actuation of auxiliary feedwater upon loss of the feedwater pumps.
The licensee is implementing a permanent design modification to ensure auxiliary feedwater pump actuation, therefore, a jumper would no longer be required. The requirement for the jumper, delineated in Action Statement 21.b of Table 3.3-3 would be deleted.
The modification, an installed relay actuation, performs the same function as that of the manual jumper.
The installation of the relay enables the I
auxiliary feedwater pump auto-start circuitry and precludes the need for manual installation of a jumper. Upon a steam generator feedwater pump trip, failsafe (de-energize to actuate) relays automatically perform the same function as the manual installation of the jumpers. The relay actuation condition is displayed to the operators by illumination of a control console pushbutton light for each steam generator feedwater pump, "SSFP TRIP-AFP AUTO ARMED."
The staff agrees with licensee's proposed TS change to delete Action Statement 21.b of Table 3.3-3 which requires the installation of a jumper based on the addition of a relay, and finds it acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments.
The State official, by letter dated June 18, 1996, had no comments.
e
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Fart 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (61 FR 30641). The amendments also relate to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
D. Spaulding Date: February 6, 1997