ML20092F339

Safety Evaluation Supporting Amends 45 & 34 to Licenses NPF-66 & NPF-77,respectively

ML20092F339
Person / Time
Site:	Byron, Braidwood
Issue date:	02/12/1992
From:	Office of Nuclear Reactor Regulation
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ML20092F334	List: ML20092F339
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NUDOCS 9202190265
Download: ML20092F339 (4)

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. SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 45 TO FACILITY OPERATING LICENSE NO. NPF-66 i

AND AMENDMENT NO. 34 TO FACILITY OPERATING LICENSE NO. NPF-77 COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT NO. 2 BRAIDWOOD STATION, UNIT NO. 2 DOCKET N05. STN 50-455 AND STN 50-45)

1.0 INTRODUCTION

in a submittal dated June 28, 1991, the Commonwealth Edison Company (CECO) described reactor protection system (RPS) and engineered safety features actuation system (ESFAS) trip setpoint changes resulting from lowering of the lower narrow range steam generator (SG) level instrument taps at Byron and Braidwood No. 2 Units from 438 inches above the top of the SG tubesheet to 333 inches.above the tubesheet. The upper instrument tap for the Model D-5 SGs in the Byron and.Braidwood No. 2 Units design remains unchanged at 566 inches above the tubeshett. With the changes, the narrow range SG 1evel instrument taps for the No. 2 Units will be-at the same levels as those in the No. I Units which have Model D-4 SGs.

The-submittal also provided an assessment of the-impact of the changes on FSAR Chapter 15 analyses, and proposed Technical Specification (TS) changes to reflect the modifications.

2.0 ; STAFF EVALUATION-2.1 Setpoint_ Changes The Byron and Braidwood TSs express the SG water level low-1cw and high-high

. trips _in terms of percent of_ narrow range SG water level instrument span (NRS).. The-increase in the narrow range instrument span alters the correlation

-of levd1 expressed.in inches versus level expressed in percent of span.

Also included:in-the consideration of revised setpoints is velocity head.

Velocity head effects result in indicated levels for any given--power less than or equal to the actual level, with the amount of discrepancy varying directly but not proportionally with power.

The high-high and.-low-low SG level trip setpoints for the Byron and Braidwood No. 2 Units TSs account for the abcVe considerations, and are based on con-sistency with safety analysis assumptions and with the setpoint methodology described in the Westinghouse Topical Reports WCAP-12583 and WCAP-12523.

9202190265 920212 PDR.ADOCK 05000455 P-PDR

' This methodology, incorporating the above considerations, has been used in previous Byron and Braidwood applications. Since the basic methodology has not been changed for this use, we also find it applicable to Byron _and Braidwood No. 2 Units for this determination of setpoints.

2.2 Chapter 15 Analyses 2.2.1 Non-LOCA Event Analyses The submittal provided an assessment of the impact of the changes on Final Safety Analysis Report (FSAR) Chapter 15 analyses.

For most Chapter 15 events, the licensee found that the calculated results for existing Byron and Braidwood Updated FSAR analyses, performed assuming Model D-4 SGs, either would be unaffected by the changes or would remain bounding versus analyses assuming the mcdified Model D-5 SGs and associated trip settings.

Three Chapter 15 events were reanalyzed because of the potential for adverse effect due to the changes. The events are:

a.

Feedwater System Malfunction Causing an increase in Feedwater Flow (UFSAR Section 15.1.2)

This event was reanalyzed for both zero power and full power case conditions.

The zero power case was found to be bounded by the Uncentrolled RCCA Bank Withdrawal from a Subtritical or Low Power Condition event addressed in UFSAR Section 15.4.1.

For the full power case, the licensee reports that the calculated minimum departure from nucleate boiling ratic (MDNBR) remains above the safety analysis limit value throughout the transient.

This assures that departure from nucleate boiling (DNB) would not be encountered and that calculated event consequences meet criteria of acceptance, b.

Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature (UFSAR Section 15.1.1)

The licensee's submittal indicates that this event is bounded by the Increase in Feedwater Flow event discussed above, and the DNB basis is

met, c.

Loss of Non-Emergency AC Power to the Plant Auxiliaries / Loss of Flow (UFSAR15.2.6/15.2.7)

The licensee's submittal indicates that these reanalyses calculated that pressurizer overfill would not occur for these events and verified the natural circulation capability of the plants to prevent fuel or cladding damage during reactor pump coastdown.

The analyses of the above events were performed using the same methods as those for the UFSAR Chapter 15 event analyses of record. We find that these methods continue to be applicable.

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2.2.2-Steam Generator Tube Rupture (SGTR)

The licensee's submittal' indicated that.the SGTR consequences reported in the Byron and Braidwood UFSAR will:not be increased by the D-5 SG modifications.

The staff: concludes that_the finding of' acceptability for the SGTR analysis of-record continues to apply.

Submittal Section 4.3.2 discusses additional SGTR. analyses performed by the licensee which were submitted to the NRC for review.

Because this analysis is

-still under review,-we have'not considered.that analysis in our findings.

2.2.3 ~LOCA Analyses-

.The licensee's submittal' indicates.that LOCA analyses are not adversely affected by the changes because analysis assumptions are not changed. We find this acceptable.

33.0; TECHNICAL SPECIFTCATION CHANGES The licensee's submittal proposed changes to three TS pages to be implerrented L

--in the operating cycle after SG modification for each unit (Byron Unit 2 and-

P aidwood Unit 2) tol reflect the setpoint modifications discussed in Section 2.1 of.this report. These are:

a.

TS page 2-5, Table-2.2-1, item 13.b,- SG Water Level Low-Low Reactor Coolant System-(RCS)~ trip - Values for Total ^ Allowance (TA), parameters not measured on-a periodic basis (2), and Sensor Error (SE) are identified _as not applicable (N.A.).

The new Trip Setpoint is 36.3% of 1NRSiand the new Allowable Value is 35.4% of NRS.,

b.

TS page 3/4 3-25,-Ta51e 3.3-4, Item 5.b.2, SG Water Level High-High-

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Turbine Trip and Feedwater.-Isolation.--Values for TA.-Z, and.SE are

.-increased to 18.9..12.02, and 3.2, respectively. 'The new-Trip Setpoint is 80.Br of NRS and-the new Allowable;Value is 82.8% of NRS.

- c.- - =TS page'3/4 3-26, Table 3.3-4, Item 6.c.2, SG Water Level Low-Low Start Motor-Driven Pump and Diesel-Driven Pump - The new values are the same as in.a. above.

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Because Braidwood.-Unit 2, fuel. cycle 3 began in November 1991, the proposed TS change and the corresponding modifications to relocate-the

1oder sensing tap of the Unit 2 steam generator will not be effective until the start of fuel cycle 4.

Therefore, for Braidwood, Unit 2,

--the above proposed TS changes will only be effective ~for cycle 4 and after..The existing lTS will remain through cycle 3..

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. The licensee's submittal based its justification of these modified setpoints on consistency with FSAR Chapter 15 analyses assumptions as discussed in Section 2.2 of this report.

We find the licensee's submittal, describing lowered lower SG 1evel instrumentation taps, associated trip setpoint changes, and analytical justifications, acceptable based on use of a setpoint methodology which had been previously used in an approved application, and on justifications citing applicable UFSAR analyses and reanalyses using approved methodologies.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

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. 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 (56 FR 57692). 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 nas concluded, based on the considerations discus:ed abovc, 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:

F. Orr, SRXB Date:

February 12, 1992