ML20101B997
| ML20101B997 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/28/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20101B988 | List: |
| References | |
| NUDOCS 9206050289 | |
| Download: ML20101B997 (4) | |
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UNITED STATES -
i fr NUCLEAR REGULATORY COMMISSION-WAS WNoToth D.C. 20666
' SAFETY EVALUATION ~BY THE OFFICE OF NVCLEAR H ACTOR REGULATION
- RELATED TO AMENDMENT N0. 51 ~ TO FACILITY OPERATING LICENSE NPF-68 AND-AMENDMENT NO. 30 TO FACILITY OPERATING LICENSE NPF-81 GEORGIA POWER COMPANY. ET AL.
V0GTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425
1.0 INTRODUCTION
l By-"ELV-03196, Application for Amends to Licenses NPF-68 & NPF-81,revising Loop Design Flow in TS Table 2.2-1 from 95,700 Gpm to 93,600 Gpm & RCS Sys Flow in [[TS" contains a listed "[" character as part of the property label and has therefore been classified as invalid. from 393,136 Gpm to 384,509 Gpm|letter dated November 12, 1991]], as supplemented-April 21, 1992, Georgia Power Company, et"al.- (the licensee) proposed license amendments to change the minimum required thermal-design flow (TDF) specified in the Technical Specifications (TSs) _for Vogtle Electric Generating Plant (Vogtle or the facility), Units 1 and.2. Specifically, the footnote in TS Table 2.2.1 for--
" Loop Design Flow"'would be changed to reduce the specified flow from 95,700
-gpm to 93,600.gpm. Similarly, in _TS 3.2.5.c, the " Reactor Coolant System
_(RCS)- Flow" specified in the. limiting condition for operation (Lf0) and
_ associated TS-Bases 3/4'.2.5 would be revised from 393,136 gpm to 484,509 gpm (including flow uncertainty).. The licensee's application also included related changes which-would apply ~ only prior to completion of the third
= refueling outage for Vogtl_e Unit 2.
However, since that refueling outage has
-now been completed,- these proposed changes are no longer needed and are not included-in these-amendments.-.The April. 21, 1992 letter provided additional information which did not change the-initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
During the-third refueling outage for Vogtle Unit -1 in late 1991, the licensee
. removed the:' resistance temperature detector (RTD) bypass. system used to
. measure.the hot leg. temperature and replaced it with direct immersion RTDs.
During this outage, the licensee also. began-a transitio_n in fuel.typelby replacing one third.of Unit l's core with Westinghouse's VANTAGE-5 fuel using
~. low leakage fuel loading pattern. The-same changes were made to-Unit 2 a
during its' third refueling outage which was recently completed. These changes were accomplished in accordance-with Amendments'43 through 46 for Unit-1, and L
Amendments _23lthrough 25 for Unit 2. The low leakage fuel loading pattern has resulted.in increased hot leg streaming-which causes an erroneous reduction in the.RCS flow rate measured via the calorimetric. heat' balance. To compensate for this problem and ensure that the RCS flow rate-TG limit can be met, the
--licensee has proposed the above TS changes which' reduce the allowable loop and RCS flow rate.
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. 3.0 EVALUATION in support of these proposed TS amendments, the licensee and Westinghouse have examined ec1n s&fety analysis that uses TDF as an input parameter.
Each analysis was_ either reanalyzed with a reduced RCS TDF of 374,400 gpm to determine the effect of the flow reduction, or evaluated to determine that the impact of the flow reduction was insignificant.
These analyses and evaluations assumed a power level of 3565 MWt, which provides results that are conservative with respect to the power level of 3411 MWt authorized by the current operating licenses.
As noted in Section 2 of this evaluation, the licensee had previously submitted a program to NRC for implementation of VANTAGE-5 fuel which has been approved by the NRC.
The analyses in that program were for a reduced loop TDF of 93,600 gpm (374,400 gpm for four loops) and a power level of 3565 MWt.
This included accident analyses for large and small break LOCAs, steam generator tube rupture, and a large spectrum of non-LOCA events dependant on fuel-related parameters.
Also, as part of an earlier program to relocate the lower steam generator instrument taps that are used to determine narrow-range level, the licensee had previously analyzed several non-LOCA events that are dependent on steam generator level based upon the reduced TDF.
These analyses were submitted in the licensee's application of May 29, 1990, and were approved by the NRC upon issuance of Amendments 34 (Unit 1) and 14 (Unit 2) on August 30, 1990.
These analyses included:
FSAR Section 15.2.6 Loss of Nonemergency AC Power I
to Plant Auxiliaries FSAR Section 15.2.7 Loss of Normal Feedwater Flow FSAR Section 15.2.8 Feedwater System Pipe Break In its application of November 12, 1991, the licensee identified four additional events or evaluations which had not been addressed with the lower TDF in previously approved steam generator level tap relocation and VANTAGE-5 programs. -The revised evaluations are based upon the lower TDF. The four issues are:
1)
Inadvertent Opening of a Steam Generator Relief.or Safety Valve Event (FSAR Section 15.1.4) - To address this issue, the licensee referenced a reanalysis of the event assuming the lower TDF included as part of a power uprate submittal dated February 28, 1992. Additional details of the reanalysis were provided in the licensee's letter of April 21 992.
The analysis was performed using approved Westinghouse computer c as LOFTRAN and THINC-IV, and the approved Westinghouse W-3 departure from nucleate boiling (DNB) correlation (The W-3 correlation is based upon a minimum DNB ratio (DNBR) limit of 1.30).
The licensee indicated that the reanalysis accounted for all DNBR penalties (e.g., mixed core penalty) and was performed for both types of fuel presently in the Vogtle cores (Westinghouse's 17 x 17 low parasitic and VANTAGE-5).
The licensee reported that the minimum DNBR remained above the 1.30 limit.
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.: The licensee's reanalysis used approved methodologies-and appropriate
- assumptions,-and provided acceptable results.
The staff finds the analysis acceptable.
- 2)
Main Steamline Break Event (FSAR Section 15.1.5) - To address this issue the licensee refersacea a rc:ralysk of the event assuming the lower TDF included in the February 28, 1992 pover uprate submittal, as sup)1emented by information in-the April 21, 1992, letter.
The metiodologies used in this reanalysis are the same as used for item,(1) above, ~except that a DNBR limit criterion of 1.45 was app 1H to account
-for the calculation of primary system pressure to drop belo. '.000 psia.
-The licensee reported that the minimum DNBR for this event, considering DNBR penalties, remained above the 1.45 criterion.
The licensee's reanalysis used approved methodologies and appropriate ~. assumptions, and provided acceptable results.
The
. staff finds.the analysis acceptable.
3)
Main Steamline Break Information used for Superheat Study for Vogtle Units 1 and 2 (WCAP-11285) - This issue relates to the environmental qualification envelope for. equipment located outside containment.
The licensee's studies of this-issue were originally performed for the currently' licensed power level and were subsequently updated in the licensee's power uprate submittal. The licensee finds that TDF has a
. negligible impact.onlthe environmental-consequences to equipment located outside containment. The. staff agrees with the licensee's conclusion
--and finds the lower-TDF acceptable with' respect to the environmental qualification of equipment, based upon the current authorized power level. The. staff has not reviewed the licensee's conclusion with respect to proposed power level increases.
4)
Containment Design-Evaluation (FSAR Section 6.2.1.1 and 6.2.1.4) -
TheLlicensee _ addressed this item involving Steamline Break-.and toss-
-of Coolant Accident containment conditions by referencing containment canalyses-reported in the February 28, 1992,_ power uprate submittal.
.The referenced analyses were performed-by Westinghouse using its C0C0 computer code.
COC0.is a containment' analysis code that has been previously approved by: the NRC.
lThe licensee's analyses u' sed approved methodologies and appropriate.
assumptions, and provided acceptable results.
The staff finds the
-analyses acceptable.
The licensee has examined the effect of the reduction.in allowable RCS flow on
- trip setpoints.and-has determinedJthat no changes are required to the current settings.
Based on the information presented above, the licensee has concluded, and the staff' agrees, that reduction in TDF and the LC0 RCS flow value do not involve a significant increase in the probability or consequences of an accident previnusly evaluated. Moreover, the proposed lower TDF value does not cause l
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.any1 acceptance-criteria'for safety analyses, or the environmental envelopment for; equipment qualification, to be exceeded. The staf' therefore finds the proposed change _to'be acceptable for operation at the current authorized power levetof-3411 MWt.-
4.0 STATE CON 3ULTATION:
-In accordance with the Comission's regulations, the Georgia State official was: notified of the proposad issuance of.the amendments.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
These-amendments ' change a requirement with ' respect to the installation or use
-of facility components 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
- increasesin: individual or cumulative occupational exposure.
The Commission has previously issued a proposed finding that the' amendments involve no significant hazards consideration, a'nd there has been no public comment on such! finding (56 FR 61263 dated December 2,c-1991). 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 ortenvironmental asses ~sment need be prepared in_ connection with the issuance 'of these amendments.:
6.0 CONCLUSION
The'Commi'ssion-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 wil1~be conducted-in compliance with the Commission's regulations,
-and (3)1the. 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:.-D. Hood, PDII-3/DPRE H. Balukjian, SRXB/ DST '
Dated:
May_?8, 1992 d
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