ML20054K063
| ML20054K063 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/18/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054K061 | List: |
| References | |
| NUDOCS 8206300368 | |
| Download: ML20054K063 (5) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 DESIGUATED ORIGINAI, Certified By b/w Wi!
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SAFETY EVALUATION SY THE OFFICE CF MUCLEAR REACTOR. REGULATION SUPPORTIMG AMENDMENT NO. 65 TO FACILITY OPERATING LICENSE NO. DPR 40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285 Introduction By application dated November 17, 1981, the'0maha Public Power District (the District) requested an amendment to their Technical Specifications (TS) which adds setpoint, operability, and surveillance requirements for the safety-grade auxiliary feedwater automatic actuation system. By letter dated March 22, 1982, the District amended in part their November 17, 1981 application and ir.cluded revised setpoints for low steam generator water level, low steam generr. tor pressure, and steam generator differential pressure. The revised setpoints were the result.of a final optimization of auxiliary feedwater actuation system.(AFAS) setpoints based upon a final analysis of AFAS equipment and drift uncertainties performed by Combustion Engineering. By telephone call on May 4, 1982, the District was asked to describe the method used to calculate setpoint uncertainties, including the treatment of cali.bration errors, drift, instrument inaccuracies, and environmental effects (temperature, pressure, vibration, post accident effects, etc.). The District stated in the May 4,1982 telephone call that the setpoirit methodology used was that given in Combustion Engineering Report CEN-ll2(s) entitled " Plant Protection System Selection of Trip Setpoint Values" dated November 15, 1979.
The District's current TSs contain interim special TSs' which address control-grade au,xiliary feedwater automatic initiation. These control-grade TSs would be replaced by the District's proposed safety-grade TSs.' The following safety evaluation addresses the District's request.
Evaluation The Fort Calhoun Station's safety-grade auxiliary feedwater actuation system contains logic to automatically isolate a broken steam generator as well as providing auxiliary feedwater to the intact steam generator, thus providing for shutdown heat removal. The requirements for initiating auxiliary feedwater flow to a steam generator are a concurrent low steam generator level signal and a permissive signal based on the pressure in each steam generator. Figure 1 shows a simplified logic diagram for the system.
In the presence of a low steam generator level signal, (a signal equal to 28.2% water level is proposed),
a generator will be fed if its pressure exceeds 466.7 psia.
It will also be fed at.a pressure less than 466.7 psia'if it exhibits low water level and its pressure e'xceeds that of the second steam generator by a differential pressure of 119.7 psid, signifying that the second steam generator loop contains a break.
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t The 28.2% low steam generator water level TS limit is found by adding the low steam generator water level safety analysis setpoint (15%) to the low steam generator water level uncertainty value (13.2%). The low steam generator pressure TS limit and high steam generator delta pressure TS limit are determined in a similar manner. The staff has evaluated the proposed uncertainty values and safety analysis setpoints as described below.
The District used the setpoint methodology as contained in the Combustion Engineering report referenced above.
Section 2.3 of CEN-112(s) describes treatment of equipment errors. These errors include process instrumentation uncertainties (sensor, transmitter, and current loop) and protection system cabinet instrumentation uncertain-ties (bistable comparator and logic) as given by the equipment manufacturers.
The uncertainty values also account for exp'ected setpoint drift of the instrumentation and errors associated with' setting and checking the setpoints.
Accident environment errors are also considered. Examples of these are temperature effects, pressure effects, reference leg effects, seismic effects, and radiation effects. In addition, the. uncertainty value accounts for process error;(i.e., uncertainty in the value of the process parameter at the sensor).
The total instrument channel uncertainty represents the maximum uncertainty calculated that could occur.at any time during the periodic surveillance interval for the limiting event for which the function is required to operate.
CEN-112(s) states that this CE methodology used to calculate equipment set-points.is consistent ~with the guidance given in Regulatory Guide'1.105 (Instrument Setpoints).
The licensee has stated that their monthly test procedure requires the set-point to be returned to its normal value if the setpoint is found to be out of tolerance (i.e., less conservative than the Technical Specification set-point value). Typically, the setpoint is set slightly in the conservative direction from the TS setpoint value'. For example, the steam generator low water level TS setpoint is 28.2%. The setpoint is actually set between 29%
and 29.5%.
If the setpoint is found to be less than 28.2%, it is returned to within this range. The safety analysis setpoint for steam generator low water level is 15% which leaves a 13.2% margin allowed for ti.e total instru-ment channel uncertainty. A similar approach is used for low steam generator pressure and differential (delta) pressure.
The setpoints are located far enough away from the upper and lower limits of the instruments range such that the trip functions cannot be nullified due to drift beyond the instruments range. The steam generator low pressure setpoint is 466.7 psia. The pressure instrument range is 0 to 1200 psig.
3 The steam generator differential (delta) pressure setpoint is 119.7 psid.
. Based on our review of the licensee's submittal, we conclude that the pro-posed AFAS setpoints are consistent with the guidance given in Regulatory Guide 1.105 (Instrument Setpoints), and therefore, are acceptable.
It should be noted that although this Safety Evaluation has found the setpoint methodology described in CEN-112(s) adequate for determining the uncertainty values used in achieving Ft. Calhoun AFAS setpoints, it does not give staff acceptance of CEN-112(s) nor does it generically approve the use of'this methodology as applied to other plants.
The District provided a revised steamline break' analysis using the revised AFAS setpoint values to verify that adequate level will be maintained in the steam generators and that the broken steam generator will be properly isolated. The analysis methods that were used were the same as those used previously in the FSAR and in the analysis of the control-grade AFAS. We have reviewed the District's analyses and analysis results and conclude they are acceptable.
In addition to setpoint requirgments, the District proposed operability and surveillance requirements. We have reviewed the District's proposed auxiliary feedwater operability requirements and surveillance. requirements and find them acceptable.
A number of administrative changes were also proposed and.are acceptable.
Since auxiliary feedwater requirements are now part of Table 2-3, the auxiliary feedwater requirements that were contained in Table 2-5 and Table 2-6 and page 6-3 have been deleted.
The staff's review of the District'.s submittals related to NUREG-0737 item II.E.1.2 entitled " Auxiliary Feedwater Automatic Initiation and Flow Indica-tion".is continuing. The staff's review of the District's submittals related to " Main Steam Line Break with Continuous Feedwater Addition" is also con-tinuing. These' issues will be addressed at a later date.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amaunts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insigni-ficant from the standpoint of environmental impact and, pursuant to 10 CFR 151.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection'with the issuance of this amendment.
, Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be con-ducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
June 18, 1982 Principal Contributors:
E.G. Tourigny, Project Manager B. Hardin, Reactor Engineer i
R.A. Kendall, Reactor Engineer s
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