ML20199C172
| ML20199C172 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/15/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199C170 | List: |
| References | |
| NUDOCS 9801290154 | |
| Download: ML20199C172 (4) | |
Text
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UNITED STATES p-
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NUCLEAR REGULATORY COMMISSION f
j wAsHiNotoN, D.C. 30en 0001
%,.....f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
,RELATED TO AMENDMENT No. % TO FACILITY OPERATING LICENSE No. NPF 37.
AMENDMENT NO.
% TO FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT No.
87 TO FACILITY OPERATING LICENSE NO. NPF 72, At4D AMENDMENT NO.
87 TO FACILITY OPERATING LICENSE NO. NPF 77 COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT NOS.1 AND 2 BRAIDWOOD STATION. UNIT NOS.1 AND 2 DOCKET NOS. STN 50 454. STN 50-455. 8TN 50-456 AND STN 50-457
1.0 INTRODUCTION
By letter dated February 18,1997, as supplemented by letter dated September 22,1997, Commonwealth Edison Company (CemEd, or the licensee) proposed Technical Specification (TS) changes for Byron Station, Units 1 and 2, and Braiowood Station, Units 1 and 2, to support steam generator (SG) replacements at Byron, Unit 1, and Braldwood, Unit 1. The September 22, 1997, submittal provided additional clarifying information that did not change the initial proposed no significant hazaids consideration determination.
The licensee will be replacing the original Westinghouse D4 SGs at Byron, Unit 1, and Braidwood, Unit 1, with Babcock & Wilcox Intemational (BWI) SGs. Due to changes in the l
location of the SG level taps, the installation of the BWI SGs requires an increase of the SG water level operating range (i.e., the difference between the low low and the high high SG level setpoints in percent of narrow range span). Consequently, changes are necessary to the TS setpoints for SG water level reactor trip and engineered safety features actuation. These setpoints are found in TS 2.2.1 Table 2.21 TS 3.3.2 Table 3.3-4, TSSR 4.4.1.2.2, TSSR 4.4.1.3.2 and TS 3.4.1,4.1.6.
The TS setpoints and operating ranges fer Byron, Unit 2, and Braidwood, Unit 2, which will continue to operak with the existing Westinghouse SGs, remain unchanged; however, due to the common Technical Specification pages being used for Byron, Units 1 and 2, and Braidwood, Units 1 and 2, these amendments will appear on the pages for both units.
l 2.0 EVALUATION 1
Comed proposes to change the TS for Reactor Trip System Steam Generator Water Level Low-Low setpoint, the Engineered Safety Features Actuation System (ESFAS) Steam Generator Water Level Low-Low Auxiliary Feedwater (AFW) setpoint, and the ESFAS Steam Generator i
Water Level High High Turbine Trip and Feedwater Isolation setpoint. Comed also proposes to change the TS surveillance requirements for minimum water level in Modes 3,4 and 5 (loops filled).
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4 The current requirement for the Unit 1 Iow-low level setpoints is 33.0 percent of narrow range span (NRS) with an allowable value of 31.0 percent NRS. The current requirement for the
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Unit i high-high level setpoint is 81.4 percent NRS with an allowable value of 83.4 percent NRS.
Comed proposes to change these setpoints for Byron Unit 1 and Braidwood Unit 1 to 18.0 percent NR8 with an allowable value of 16.1 percent NRS for the low-low setpcints and 88.0 percent NR8 with an allowable value of 8g.g percent NR8 for the high high setpoint. While the narrow range span for the RSG (180 inches) has decreased as compared to the original Westinghouse Model D4 steam generators (080) (232 inches), the operating range for the replacement steam generators ( RSG) (126 inches) is increased as compared to the 08G (112.3 inches). The licensee indicated that this increase in the operating range minimizes the possibility of inadverient plant trips following ioed changes and feedwater transients.
i The current surveillance requirement for the minimum SG water level in Modes 3,4 and 5 (loops filled) is 41 percent NRS for Unit 1. Comed proposes to change these requirements to l
18 percent NR8.
i The current SG level setpoints are based on the limiting accident analyses with Westinghouse steam generators. The limiting accidents for the low-low GG level reactor trip and AFW flow initiation setpoints are the Loss of Normal Feedwater and Feedwater Line Break. The limiting accident for the high high SG level setpoint is the Feedwater system Malfunction, which results i
in an increase in feedwater flow to one or more steam generators. The intent of the surveillance requirement for a minimum SG inventory in Modes 3,4 and 5 (loops filled) is to remove decay l
heat and is met by ensuring the SG tube bundle is completely covered.
The licensee determined the impact of the RSGs on the limiting low-low Mpoint transients, the Loss of Normal Feedwater and Feedwater Line Break. The transients were analyzed with RELAPS/ MOD 2 B&W using the methodology approved by the staff in BAW 1016g A. The RELAP5 analysis incorporated a low-low setpoint of 0 percent NRS for the Feedwater Line Break end 10 percent NRS for the Loss of Normal Feedwater and demonstrated that all acceptance 4
criteria (listed below) for each transient have been met.
Loss of Normal Feedwater Pressure in the reactor coolant and main steam systems did not exceed 110 percent of
- the design value; the minimum departure from nucleate bolling ratio (DNBR) remained above the g5/g5 DNBR limit ; and the ultimate heat sink for decay heat removal was assured.
Feedwater Line Break 1
- Pressure in the reactor coolant and main steam systems did not exceed 110 percent of the design value; 3 the ultimate heat bink for decay heat removal was assured; 4
the core remained intad for effective cooling; and
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3 radiation doses remain bounded by those predicted for the steamline break event and, therefore, did not exceed a small fraction of the 10 CFR Part 100 exposure guidelines.
The licensee also determined the impact of the RSG on the limiting high high setpoint transient, Feedwater System Malfunction resulting in increased feedwater flow. The transient was analyzed using the RELAPS/ MOD 2 B&W computer code. The licensee demonstrated that the acceptance criteria for the Feedwater Malfunction transient, listed below, were met with a high-high setpoint of 100 percent NRS. The licensee also demonstrated that the RSGs do not overfill.
Feedwater Malfunction Pressure in the reactor coolant and main steam systems did not exceed 110 percent of the design value; and the minimum DNBR remained above the 95/95 DNBR lienit.
The TS low low and high high SG level setpoints and associated alloivable values were calculated by the licensee using the approved methodology of WCAP 12583. Uncertainties in the setpoint value were determined based on this methodology to be approximately 15 percent NRS for low low level (approximately 5 percent NRS for the Loss of Normal Feedwater) and 9 percent NRS for high high level. The setpoint was conservatively chosen by the licensee as 18 percent NRS for low low level and 88 percent NRS for high-high level. The setpoint allowance determined per WCAP 12583 is 1.g percent, which yields a TS allowable value of 16.1 percent NRS for low low level and 89.9 percent NRS for high-high level. The staff finds the setpoints and associated allowable value changes to be acceptable based on the accident analysis described above and the use of the approved WCAP 12583 methodology.
The licensee determined the impact of the RSGs on the surveillance requirement for a minimum inventory to remove decay heat in Modes 3,4 and 5 (loops filled). The licensee stated that the intent can be met by assuring that the tube bundle is completely covered. The licensee determined that the Unit 1 SG tubes are covered when the SG water levelis within the span of the narrow range levelindication, which can be assured by specifying a surveillance requirement water level that is equal to or greater than the low low level setpoint (18 percent NRS). The staff finds the proposed surveillance acceptable.
3.0
SUMMARY
The licensee proposed changes to the TS for Byron, Units 1 and 2, and Braidwood, Units 1 and 2, to reflect necessary changes to the low low and high-high steam generator level setpoints.
These changes are necessitated by the replacement of the original Westinghouse D4 steam generators with BWI steam generators and the subsequent decrease in narrow range span.
The licensee analyzed the limiting transients for both the low-low and high high steam generator level setpoints using approved methodologies, The licensee demonstrated that the acceptanct, criteria are met for the Updated Final Safety Analysis Report, Chapter 15, transients that are impacted by the setpoint changes. Therefore, the staff finds the licensee's safety analysis to be an acceptable basis for setpoint determination.
Further, the staff concludes that the proposed TS low-low and high high SG level reactor trip and l
engineered safety feature actuation setpoints and associated allowable values are consistent i
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with the methodology approved by tne staff in WCAP 12583. The staff, therefore, finds the Yd
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changes supporting the SG replacement at Byron, Unit 1, and Braidwood, UnN 1, and the corres.ponding T8 changes for Byron, Unit 2, and Braidwood, Unit 2, to be acceptable.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the lilinois State official was notified of the proposed issuance of the amendments. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
l The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. 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 !s no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment 3 involve no significant hazards consideration, and there has been no public t
comment on such finding (62 FR 11491). Accordingly, the amendments meet 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 amendments.
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 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 Contributors:
S. Bailey S. Brewer S.Rhow Date.
January 15, 1998 i
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