ML17353A618
| ML17353A618 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/26/1996 |
| From: | Croteau R NRC (Affiliation Not Assigned) |
| To: | Plunkett T FLORIDA POWER & LIGHT CO. |
| References | |
| TAC-M94314, TAC-M94315, NUDOCS 9603280122 | |
| Download: ML17353A618 (10) | |
Text
arch 26, 1996 Mr. T. F.'Plunkett President-NUclear Division Florida Power and Light'Company P.O.
Box 14000 Juno
- Beach, Florida 33408-0420
~TRIBUTION
.. attached sheet SUBJECT'URKEY POINT UNITS 3 AND 4 MEETING ON PROPOSED LICENSE AMENDMENTS THERMAL POWER UPRATE (TAC NOS H94314 AND H94315)
Dear Mr. Plunkett:
Members of your staff plan to meet with the NRC staff on April 4, 1996, to discuss information regarding technical specifications (TS) for the proposed thermal uprate submitted by [[letter::L-95-245, Application for Amends to Licenses DPR-31 & DPR-41,revising Definition of RTP from 2,200 Mwt to 2,300 Mwt.Proprietary TR WCAP-13719,Rev 2 & Nonproprietary TRs WCAP-13718,Rev 2 & WCAP-14276,Rev 1 Encl.Proprietary TR WCAP-13719 Withheld|letter dated December 18, 1995]].
In the meeting, please be prepared to address the issues described in the Enclosure to this letter.
If you have any questions, please call me at (301) 415-1475.
Sincerely, Richard P. Croteau, Project Manager Project Directorate II-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Docket Nos.
50-250 and 50-251
Enclosure:
As Stated cc w/enclosure:
See next page Document Name:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205554001 March 26, 1996 Mr. T. F. Plunkett President Nuclear Division Florida Power and Light Company P.O.
Box 14000 Juno Beach, Florida 33408-0420
SUBJECT:
TURKEY POINT UNITS 3 AND 4 MEETING ON PROPOSED LICENSE AMENDMENTS THERMAL POWER UPRATE (TAC NOS M94314 AND M94315)
Dear Mr. Plunkett:
Members of your staff plan to meet with the NRC staff on April 4, 1996, to discuss information regarding technical specifications (TS) for the proposed thermal upr ate submitted by [[letter::L-95-245, Application for Amends to Licenses DPR-31 & DPR-41,revising Definition of RTP from 2,200 Mwt to 2,300 Mwt.Proprietary TR WCAP-13719,Rev 2 & Nonproprietary TRs WCAP-13718,Rev 2 & WCAP-14276,Rev 1 Encl.Proprietary TR WCAP-13719 Withheld|letter dated December 18, 1995]].
In the meeting, please be prepared to address the issues described in the Enclosure to this letter.
If you have any questions, please call me at (301) 415-1475.
Sincer Docket Nos.
50-250 and 50-251
Enclosure:
As Stated cc w/enclosure:
See next page Richard P. Croteau, Project Manager Project Directorate II-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Mr. T. F'.;,Pl,upkett Florida Power".arid,:Light Company CC:
J.
R.
- Newman, Esquire
- Morgan, Lewis 5 Bockius 1800 M Street, N.W.
Washington, DC 20036 Jack Shreve, Public Counsel Office of the Public Counsel c/o The Florida Legislature ill West Hadison Avenue, Room 812 Tallahassee, Florida 32399-1400 John T. Butler, Esquire
- Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Mr. Robert J.
Hovey, Site Vice President Turkey Point Nuclear Plant Florida Power and Light Company P.O.
Box 029100 Miami, Florida 33102 Armando Vidal County Manager Hetropolitan Dade County ill NW 1 Street, 29th Floor Miami, Florida 33128 Senior Resident Inspector Turkey Point Nuclear Generating Station U.S. Nuclear Regulatory Commission P.O.
Box 1448.<<,-
Homestead:,-'~PIorida
-'33090 Hr. Bil.l PasCptti",;
Office of;:,Radial)on.'-Control Depart'ment",of.:,Heal th,.; and Rehabil itati've Services 1317 Winewood Blvd.
Tallahassee, Florida 32399-0700 Turkey Point Plant Mr. Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323 Attorney General Department of Legal Affairs;-; '.-'>, ">"
The Capitol Tallahassee, Florida 32304 Plant Manager Turkey Point Nuclear Plant
':;-".:"-.q9 ",,,", --',
Florida Power and Light Company P. 0.
Box 029100 Miami," Florida 33102 Hr. H.N. Paduano, Manager Licensing 8 Special Programs Florida Power and Light Company P.O.
Box 14000 Juno
- Beach, Florida 33408-0420 Mr. Gary E. Hollinger Licensing Manager Turkey Point Nuclear Plant P.O.
Box 4332 Princeton, Florida 33023-4332 Mr. Kerry Landis U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323-0199
I
THERNL UPRATE QUESTION TO BE DISCUSSED F
A.
Emergency'reparedness and Radiation Protection Branch 2.
3.
4.
5.
Pages 3 and 14 of 17 attachment 1: Reference is made to RG 1.52 (revision 2 issued in 1978).
Please address the use of more current standards or methods specified in more recent industry standards, such as ASIDE N510-1989 and AS'3803-1989, as opposed to the standards referenced in RG 1.52, revision 2 for filter testing.
Provide a detailed description of the fission product removal of the containment spray system and the extent to which credit is taken for the cleanup function in the analysis of the large break LOCA accident analysis (WCAP 14276, Rev.l, pg.3-148).
List the containment volumes not covered by the spray and the estimated forced or convective postaccident ventilation of these unsprayed volume.
For the LOCA analysis, at least ten or more computer codes were used in your evaluations.
Discuss briefly why so many codes were used'~
and discuss the accuracy and veracity of the final results..'t Discuss the impact of power uprate on radiolysis.
Our experience "
indicates that, as a result of a power uprate, the production of oxygen by radiolysis after a
LOCA wil'l increase proportionally with the power level.
Does sufficient capacity exists in the licensee combustible gas control system to accommodate this increase in oxygen production.
Briefly discuss how the higher power level effects the source
- terms, onsite and offsite doses, and control room habitability during normal and accident conditions.
6.
Discuss how the radiation levels from both accident and normal operations are affected by the uprated power level.
B.
7.
Discuss the effects of the power uprate on coolant activation
- products, hcgyated corrosion products, and fission products.
Reactor Systems, Branch 7
I.
Please confirm'hat the methodology used in the transient and accident analyses documented in WCAP-14276, Rev 1, is consistent with that used in the UFSAR.
Identify any differences and discuss their acceptability.
2.
Please confirm that only safety grade systems and components are assumed in mitigating design basis events.
3.
4.
5.
6.
7.
8.
I s
Proc Vide thy".results of the analyses for Locked Rotor/Shaft Break accfdeilt aq44iting a loss of off-site power coincident with the event:"
Discuss the amount of fuel failure during the event and the calculated radiological consequences.
Are all fuel pins with DNBR below the NDNBR assumed to fail.
Provide the results of an analysis for a postulated main feed water line break.
Provide major transient curves for the reanalysis of the postulated main steam line break.
In your analysis of a large break LOCA, for the case of minimum ECCS
- case, the loss of the LHSI pump is assumed as the most limiting single failure.
Please discuss the potential loss of a diesel affecting ECCS.
Discuss the most limiting single failure assumed in the SGTR analysis in light of the maximum dose release.
Please confirm that the new proposed loop design flow rate of.85000 gpm is incorporated in the analyses documented in WCAP-14276>
R'ev -'1, including the loss of RCS flow and Locked Rotor/Shaft Break accidents.
9 Please confirm that the proposed setpoints for ESFAS are incorporated in the transient and accident analyses in. WCAP-14276, Rev 1.
- 10. Describe the steam generator tube plug level assumed in Non-LOCA event analyses.
Plant Systems Branch l.
2.
Please address the increase in the probability of turbine overspeed and associated turbine missile production due to plant operations at the proposed uprated power level.
In'page 5-33 of WCAP-14276, Rev.
1, Westinghouse indicated that for normal refuel)ng the maximum expected SFP heat load and temperatpe fo'r a 1/2 core offload at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown are 16.6 x 10 Btu/Hr and 147 F, respectively.
In page 14D-16 of the FSAR Appendix 14 D, Florida Power and Light Company stated that as the result of the expansion of spent fuel storage in the pool, the decay heat load for each pool increases to 16.98 x 10 Btu/Hr and the corresponding pool peak tr~nsient temperature after refueling increases to less than 141 F.
It is not clear why the pool with a higher heat load (16.96 x 10 Btu/ r vs 16.
x 10 Btu/Hr) would have a lower peak temperature (141 F vs 147 F).
Please provide detailed discussion for the above discrepancy.
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"). '6n.,abnormal'~operation without SFP cooling, the time to reach boi'l)ng iq 4.5..hours and the maximum boil off rate is 76.3 GPH.
Assumf'ng a loss of SFP cooling, provide the following information:
How long abnormal operation without SFP cooling is expected to be?
Mhat scenarios that lead to a loss of SFP cooling were considered?
Mhat actions would be required to restore SFP cooling2 Mill there be sufficient make-up water for the SFP?
Provided detail description of the make-up water sources.
How the pool boil-off will be collected/treated',
5.
It appears that Eg outside containment has not been addressed.
Please demonstrate what impact plant operations at the proposed uprated power level will have on Eg outside the containment.
It is stated on page 9.3-7 of the FSAR that the SFP cooling loop consists of a pump, heat exchanger and associated components (i.e.
filters, demineralizer, etc.),
and that in the event of a failure,:jf the SFP cooling pump, a
lOOX capacity spare pump is permanently piped into the SFP cooling system and is available as a standby pump.
However, in Figure 9.3-11 of the FSAR, three SFP cooling pumps are shown as part of the SFP cooling system.
Please provide clarification for this discrepancy.
In addition, please identify the safety class and capacity each of these pumps.
D.
Miscellaneous 2.
3.
Section 7.4, NON-RADIOLOGICAL EFFECTS, of WCAP-14276, Rev I, states that "Protection of the environment is assured by compliance with permits issued by federal, state, and local agencies."
Please confirm that none of these permits are affected by the proposed thermal uprate and no changes to the permits are necessary for the uprate.
Please discuss the maximum anticipated discharge temperature from the circulating water system during normal operation for the uprate condition and any limits which exist on the discharge temperature.
Table III-3, ANTICIPATED ANNUAL RELEASE OF RADIOACTIVE MATERIALS IN LIQUID EFFLUENTS FROM TURKEY POINT PLANT UNITS 3 AND 4 RECONCENTRATION FACTORS FOR COOLING CANAL SYSTEM, of the FES indicates that the table values were calculated for a power level of 2200 NMt.
Following a thermal uprate, will operation continue to be bounded by the values of FES Table III-3 and all other parameters in the FEST
Memorandum Dated
~03 26 96 ib Docket File PUBLIC PDII-I Reading S. Varga J. Zwolinski D. Hagan, T-4A-43 G. Hill, (4) T-SC-3 OGC ACRS G. Tracy
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