ML20059C466
| ML20059C466 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/29/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059C453 | List: |
| References | |
| NUDOCS 9401050230 | |
| Download: ML20059C466 (5) | |
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UNITED STATES l
- i NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20556-0001 0
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9 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 104 TO FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 97 TO FACILITY OPERATING LICENSE NO NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY. INCm JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated September 13, 1993, Southern Nuclear Operating Company, Inc.
(the licensee), submitted a request for changes to the Joseph M. Farley Nuclear Plant, Units 1 and 2 (Farley), Technical Specifications (TS).
The praposed changes eliminate the low feedwater flow reactor trip (i.e., steam fic /feedwater flow mismatch in coincidence with low steam generator level) from the TS following the installation of a median signal selector (MSS) in the steam generator water level control system. The proposed amendment also reduces the low-low steam generator level trip setpoint and allowable value.
Specifically, the TS changes for both units include the following:
(1)
The Trip Setpoint and Allowable Value in Item 13 of TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, would be
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reduced from 17 percent to 15 percent and from 16 percent to 14.1 percent, respectively.
(2)
Item 14 of TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, and the associated Bases would be deleted.
(3)
Item 14 of TS Table 3.3-1, Reactor Trip System Instrumentation, 1
would be deleted.
(4)
Item 14 of TS Table 3.3-2, Reactor Trip System Instrumentation Response Times, would be deleted.
i (5)
The trip setpoint and allowable value in Item 6.b. of TS Table 3.3-4, Engineered Safety Feature Actuation System Instrumentation Trip Setpoints, would be reduced from 17 percent to 15 percent and l
from 16 percent to 14.4 percent, respectively.
I (6)
Item 14 of TS Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirements, would be deleted.
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9401050230 931229 PDR ADOCK 05000348 P
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. In support of these amendments the licensee submitted, as enclosures to the September 13, 1993, letter, two proprietary reports prepared by Westinghouse Electric Corporation: WCAP-13807, " Elimination of the Low Feedwater Flow Reactor Trip via Implementation of the Median Signal Selector (MSS) at Farley Units I and 2," and WCAP-13751, " Westinghouse Setpoint Methodology for Protection Systems - Farley Nuclear Plant Units 1 and 2."
1 2.0 EVALUATION Each of the steam generators at Farley, Units I and 2, has three independent narrow range water level detection instrument channels which provide input to the reactor protection system (RPS). The low-low steam generator water level protection function is configured with two-out-of-three actuation logic derived directly from these three narrow range level channels for each steam generator.
The two-out-of-three actuation logic also provides a starting signal for the auxiliary feedwater pumps.
The low-low steam generator water level reactor trip function is designed to preserve the steam generator as a heat sink for removal of residuai heat if there is a loss of normal feedwater.
The low feedwater flow reactor trip is configured to initiate a reactor trip during a condition of steam and feedwater flow mismatch on one-out-of-two channels in coincidence with low steam generator water level on one-out-of-two channels. The Farley FSAR does not assume that the low feedwater flow reactor trip mitigates the consequences of any analyzed accident.
In events such as a loss of normal feedwater or loss of all AC power, credit is only taken for the low-low steam generator water level reactor trip to ensure safe shutdown of the reactor.
At Farley, one of the steam generator water level instrument channels also supplies an input to the steam generator water level control system. As a result, a common instrument channel is used for both the RPS and the steam generator water level control system, separated electrically with a qualified isolation device. The low feedwater flow reactor trip was installed to satisfy the requirements of the Institute of Electric and Electronics Engineers Standard 279,1971 (IEEE Std. 279), " Criterion for Protection Systems for Nuclear Power Generating Stations," which is endorsed by 10 CFR Part 50.55a.
Section 4.7.3, " Single Random FaH e e," of IEEE Std. 279states, in part, "where a single random failure can cause a control system action that results in a generating station condition requiring protective action and also prevent proper action of a protective system channel designed to protect against the condition, the remaining redundant protection channels shall be capable of providing the protective action even when degraded by a second random failure." The purpose of the low feedwater reactor trip was to satisfy this criterion.
The staff reviewed WCAP-13807, " Elimination of the Low Feedwater Flow Reactor Trip via Implementation of the Median Signal Selector at Farley Units I and 2," which provides a detailed description regarding the addition of the median signal selector (MSS) to the steam generator water level control system.
The MSS is a hardware device that is designed to select the median of the three-narrow range steam generator water level instrument input-signals.
By selecting the median signal, a single random failure will not cause a control system action that results in a condition requiring protective action.
This
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, is because the failure of a single protection channel will not result in adverse control system behavior because the median signal of the three level l
channels will be selected for control purposes.
Thus, the possibility of adverse interaction between the steam generator water level control system and the RPS due to a single failure is eliminated, thereby satisfying the requirements of IEEE Std. 279 without credit being taken for the low feedwater flow reactor trip.
The steam generator water level control system MSS is functionally the same as the MSS design currently being used for the average temperature and delta-temperature control circuits in the RTD bypass elimination modification at Farley Unit 1.
The MSS used in the RTD bypass elimination circuitry was ipproved by the staff as documented in Amendment No. 87 to the Farley operating license dated March 8, 1993.
All three outputs from the steam generator narrow-range level channels are processed in the Westinghouse 7300 Process Protection Racks of the RPS, and are then input to the respective MSS. The single card MSS consists of operational amplifiers configured with input auctioneering (low and high),
feedback resistance and diode networks and adjustable input and output signal conditioning to produce an output signal to the steam generator water level control system that is equal (in voltage) to the median of the three narrow range steam generator level input signals. The MSS electronics are of a quality consistent with low failure rates and minimum maintenance requirements, and these components conform to protection system design requirements.
A separate relay card will be added to the steam generator water level control system to provide the capability for on-line testing of the MSS downstream of the protection system isolation devices and for calibration of the MSS. The relay card allows for testing of the MSS during plant operation without placing the low-low steam generator channels in trip.
The Farley operating and maintenance procedures will be revised to ensure that the procedures are consistent with, and support operation with an MSS, including administrative controls for operations with the MSS disabled or in test mode.
In addition, during testing of the steam generator narrow range level protection channels, the output of the MSS will be observed.
MSS calibration and functional testing will be included in the steam generator water level control system instrumentation calibration procedure which is currently conducted on a refueling basis and administratively controlled by the plant preventive maintenance program.
The isolation devices separating the low-low steam generator water level protection channels and the MSS of the steam generator water level control system are standard Westinghouse 7300 series equipment and were previously reviewed under WCAP-8892-A, " Westinghouse 7300 Series Process Control System Noise Tests," and accepted by the staff.
The licensee has stated that alarms will be added to the steam generator water l
level control system upon installing the MSS to indicate steam flow and feedwater finw mismatches and low steam generator level.
In addition, i
annunciators will be added to alert the operator in the event of a power I
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, supply failure on the MSS card and whenever the relay card test injection switches are enabled.
The licensee has also requested to revise the steam generator water level low-low reactor trip setpoint and the auxiliary feedwater actuation setpoint from 17 percent to 15 percent.
The revised low-low steam generator water level reactor trip setpoint was calculated using standard Westinghouse methodology as presented in WCAP-13751, " Westinghouse Setpoint Methodology for P6otection Systems - Farley Nuclear Plant Units 1 and 2."
This methodology is consistent with the Instrument Society of America Standard S67.04, 1987, "Setpoints for Nuclear Safety-Related Instrumentation used in Nuclear Power Plants.
1 The required setpoint is obtained by adding the specific channel statistical allowance (for channel inaccuracies) to the safety analysis limit.
The safety analysis limit for the low-low steam generator water level is O percent of narrow range level span. The channel statistical allowance was calculated to i
be 14.9 percent of the instrument span.
Thus, the results of the setpoint calculation indicate that sufficient margin exists to support a setpoint i
reduction from 17 percent to 15 percent without changing the safety analysis limit.
The revised setpoint of 15 percent still meets the safety analysis j
limit with the required channel accuracies included and a 0.1 percent j
additional margin.
In addition, calculations indicate that the allowable value can be reduced from 16 percent to 14.4 percent.
The licensee has stated that the revised trip setpoint would increase the margin between the steam generator water level low-low reactor trip and the normal operating band j
thereby providing additional margin to spurious reactor trips. Based on the i
above information, the staff finds the proposed changes acceptable.
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3.0 CONCLUSION
In summary, a review of the Farley safety analysis shows that no credit is taken for the reactor trip initiated by the low feedwater flow reactor trip in mitigating the consequences of any of the analyzed design basis accidents or transients. This trip was originally designed to satisfy the single random failure requirement specified in IEEE Std. 279, Section 4.7.3, for preventing adverse control and protective systems interaction. The MSS provides an acceptable alternative method of preventing interaction between the control and protection functions. As discussed above, the staff finds that utilization of the MSS is acceptable. Thus, with the addition of the MSS, the i
Farley steam generator water level control system meets the requirements of i
Section 4.7.3 of IEEE Std. 279 without the low feedwater flow reactor trip.
On this basis, the staff finds the proposed TS changes involving the elimination of the low feedwater flow reactor trip to be acceptable following the installation of the MSS.
The revised setpoints for the low-low steam generator water level reactor trip and auxiliary feedwater actuation are consistent with the safety limit assumed in the FSAR analysis and are consistent with approved setpoint methodology.
In addition, the revised setpoints remove an over conservatism which contributes to unnecessary reactor trips. On this basis, the staff finds the proposed change in the low-low steam generator water level reactor trip setpoint and allowable value to be acceptable.
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4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDE' RATION The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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-amendment involves no significant hazards consideration, and there has been no public comment on such finding-(58 FR 62157).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 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 j
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
J. Ganiere Date: December 29, 1993
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