ML20199F375
| ML20199F375 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/22/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199F368 | List: |
| References | |
| NUDOCS 9802030154 | |
| Download: ML20199F375 (4) | |
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[p%qk UNITED STATES g
.g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.c. 2055 0 0001 g
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 97 TO FACILITY OPERATING LICENSE NO. NPF-37, AMENDMENT NO. 97 TO FACILITY OPERATING LICENSE NO. NPF-66, AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. NPF 72, AND AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. NPF-77 COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT NOS.1 AND 2 ERAIDWOOD STATION. UNIT NOS.1 AND 2 DOCKET NOS. STN 50-454. STN 50-465. STN 50-456 AND STN 50-457 1.0 INTRCDUCTION By application dated January 30,1997, as supplemented by letter dated December 9,1997, Commonwealth Edison Company (Comed) the licensee for Byron and Braidwood Stations, requested changes to the Technical Specifications (TS) for the Byron and Braidwood facilities.
Additional information that did not change the initial proposed no significant hazards consideration determination was provided by letters dated May 23, August 8,1997, and January 7,1998. The proposed changes would reflect the higher loss-of-coolant accident (LOCA) peak containment pressure calculation due to steam generator (SG) replacements and would also implement TS administrative improvements having no techr'ical significance, Comed will be replacing the original Westinghouse D4 SG in the Byron and Braidwood #1 units with Babcock & Wilcox intemational (BWI) SG. The SG replacements are a 10 CFR 50.59 modification that will increase the reactor coolant system (RCS) volume and secondary side pressure. This results in a greater steem mass and energy release in the event of a LOCA. The greater m; ss and energy release, in tum, results in an increase of the calculated peak containment intemal pressuro (Pa) related to the design basis LOCA. Since Pa is specified in the TS, changes are needed to increase the test pressure for the #1 units. The TS changes for the #2 units are necessary because common TS pages are used fcr Units 1 and 2.
2.0 AFFECTED TECHNICAL SPECIFICATIONS The following changes are proposed:
TS 1.20.a defines Pa and specifies the current value as 44.4 psig. This TS would be changed to indicate that Pa will be 47.8 psig for Unit i after Cycle 8.
TS 4.6.1.3 wouid be changed to specify air lock door seal maximum leakage limits solely in terms of percentage of maximum allowable primary containment leakage rate (La) rather than in terms of both percentage of La and in standard cubic feet per hour (SCFH).
9802030154 900122 PDR ADOCK 05000454 P
2 TS 4.6.1.7.3 & 4 would be rsvised to specify that the containment purge eupply and Chaust isolation valve rer_ilient seal test pressure is "Pa" rather than "Pa,44.4 psig."
TS BASES 3/4.6.1.4 and 3/4.6.1.6 would be revised in several places to substitute the term Pa for "44.4. psig."
The above changes are dus to the increase in calculated oeak accident pressure from 44.4 psig to 47.8 psig. By (a) stanng the exact value of Pa in one place only (l.c., the TS Definitions section) and using the term "Pa" alone elsewhere, and (b) stating specified leakage limits in terms of percentage of La, the TS are simplified so as to -
- ize the number of TS pages affected by differences among units.
Alsc:
TS 5.4.2 2 would be rav; o indicate, corrected value of 12,340 cubic feet for the RCS volume and that for a ch Unit 1, the RCS volume of 12,340 cubic feet is increased I
by en additional 1,280 cubic feet as a result of SG replacement.
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This change would reflect the effects of SG replacernent on the RCS volume.
3.0 EVALUATION To support the increased RCS volume, the licensee has performed an evaluation of all accident andyses documented in the Updated Final Safety Analysis Report (UFSAR) to determine the effect of an increase in RCS volume on the consequences of the accidents analyzed. The results of the licensee's evaluation concluded thrt the increased RCS volume will not result in a reduction of the safety margin.
In the licensee's evaluation performed to determine the effect of the increased RCS volume on the peak con'ainment pressure following a large breck loss of coolant accident (LBLOCA), the calculation irdicated that the increased RCS volume couid cause the peak containment pressure to increase to 47.8 psig. However, the licensee states that this increased oack pressure is still below the containment dosign pressure of 50.0 psig.
Pa is defined in Appendix J as:
Fa (p.s.l.g.) means the calculated peak containment intemal pressure related to the desita basis accident and specified either in the technical specification or associated base?
The de I!gn basis accident for calculation of Pa is the postulated LOCA case which produces the h!ghest containment peak intemal pressure considering the complete spectrum of primary
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coolant piping system break sizes, break locations, single-failures and initial operating conditions.
A new maximum peak accident pressure was calculated by applying the incremental effects of the larger SG to the Finci Safety Analysis Report (FSAR) mass and energy release analysis. The licensso increased the blowdown phase data by 54,256 lbm and 33.084E+07 BTU to account for the niass ad energy MMd by the additional RCS fluid inventory of the replacement SG. For the reflood and post-reflood phases of a LOCA, the enthalphy of the iluid d:3 charged from 9 break t
after paasing through a SG would be higher with the replacement SG due to the higher l
secondary side temperature. To account for this effect, the licensee increased the enthalphy of j
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3-the fluid discharge during reflood and post reflood by 3.0 BTU /lbm. The 3 BTU /lbm value corresponds to an increase in secondary side T(sat) of 5.6 cegrees Fahrenheit. Also, the SG depressurization time was increased to allow for dissipation of the additional stored energy in ile larger matal mass and secondary side water inventury of the replacement SG's. The revised mass and energy data were input into a new CONTEMPT B&W containment model which had been previously benchmarked against the original licensing model using the original mess and miergy data. The licensee's analytical methodology is suffdently accurate and conservative for the purpose of calculating a new containment leakage rate test pressure. The licensee calculated the new Pa to be 47.8 psig. With regsrd to the qualification of equipment, the licensee determined that the revised temperature profile for the large break LOCA with the replacement steam generators is bounded by the equipment qualification envelope currently used to qualify equipment at the Byron and Braided Stations.
4.0
SUMMARY
The staff reviewed the licensee's submittals - 7d methodology for calculation of the new containment,neak accident (LOCA) pressure and found it acceptable. The proposed amendment will revise the containment test pressure "Pa" specification to conform with the new value. This Is consistent with Appendix J to 10 CFR, Part 50 which requires that the LOCA peak accident l
pressare be identified in the TSs or TS Bases. The recalculated value of Pa remains bounded by the coltainment design pressure thersb maintaining containment structural margins and is
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acceptable. The other TS changes are of an editorial nature and are acceptable on the basis that the increased RCS volume will not result in a reduction of the margin of safety.
5.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Iliinois State official was notified of the t
oroposed issuance of the amendments. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect tc 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 amer'dm?nts involve no significant increase in the amcunts, 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. Tne Commission has previously issued a proposed finding that the amendments involve no sigtiificant hazards considoration, and there has been no puL:ic comment on such finding (62 FR 19826 and 62 FR 66699). Accordingly, the amendments meet the ehgibility criteria for categorical exc!nion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmentalimpact statement or environmental assessment need be prepared in connection with the iscuance of the amendments.
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4 7.0 QpNCLUSION 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 proposci manner, (2) such activitics 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: W. Long C. Y. Lian0 Date: January 22, 1998 l
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