ML18039A323
| ML18039A323 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/22/1998 |
| From: | Raghavan L NRC (Affiliation Not Assigned) |
| To: | Zeringue O TENNESSEE VALLEY AUTHORITY |
| References | |
| TAC-M99711, TAC-M99712, NUDOCS 9804280056 | |
| Download: ML18039A323 (12) | |
Text
April 22, 1998 Mr. O. J. Zeringue Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
BROWNS FERRY PLANT UNITS 2, AND 3 - REQUEST FOR ADDITIONAL INFORMATIONREGARDING TECHNICALSPECIFICATION CHANGE TS-384 REQUEST FOR LICENSE AMENDMENTFOR POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)
Dear Mr. Zeringue:
By letter dated October 1, 1997, Tennessee Valley Authority (TVA)proposed revisions'to the Browns Ferry Nuclear Plant, Units 2 and 3 Technical Specifications to permit operation of the units at the unrate power level of 3458 MWt. Our review of your application is in progress. The staff has determined that additional information is required to complete these reviews.
A description of the information required is provided in the enclosure.
Please provide the
'requested information by May 15, 1998. Your prompt response willassist us in completing timely reviews.
Ifyou have any questions, please write or call me at (301) 415-1471.
Sincerely, Docket Nos. 50-260, 50-296 Serial No. BFN-98-002 L. Raghavan, Senior Project Manager Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: See next page DISTRIBUTION'rowns Ferry r/f
- JJaudon, Rll Docket Files OGC RG allo PUBLIC JZwolinski REckenrode ADeAgazio ACRS KManoly OFC NAME DATE PDII-3/PM LRaghavan:c 4/
/98 PDII-3/LA BClayto 4/3.0 /98 EMEB/D RWessman 4/>~/98 PDII-3/D FHebdon 4/2~/98 OFFICIAL RECORD COPY DOCUMENT NAME: G:LBPN$99711RAI.EMB 9804280056 980422 PDR ADOCK 05000260 P
~ 'pril 22, 1998 Mr. O. J. Zeringue Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
BROWNS FERRY PLANT UNITS 2, AND3-REQUEST FOR ADDITIONAL INFORMATIONREGARDING TECHNICALSPECIFICATION CHANGE TS-384 REQUEST FOR LICENSE AMENDMENTFOR POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)
Dear. Mr..Zeringue:
By letter dated October 1, 1997, Tennessee Valley Authority (TVA)proposed revisions'to the Browns Ferry Nuclear Plant, Units 2 and 3 Technical Specifications to permit operation of the units at the unrate power level of 3458 MWt. Our review of your application is in progress. The staff has determined that additional information is required to complete these reviews.
A description of the information required is provided in the enclosure.
Please provide the requested information by May 15, 1998. Your prompt response willassist us in completing timely reviews.
Ifyou have any questions, please write or call me at (301) 415-1471.
Sincerely, Docket Nos. 50-260, 50-296 Serial No. BFN-98-002 L. Raghavan, Senior Project Manager Project Directorate II-3 Division of Reacto'r Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: See next page Browns Ferry r/f
- JJaudon, Rll Docket Files OGC RGallo PUBLIC JZwolinski
'Eckenrode ADeAgazio ACRS KManoly OFC PDII-3/PM NAME LRaghavan:c DATE 4/
/98 PDII-3/LA BClayton'/3.0
/98 EMEB/D RWessman 4/>898 PDII-3/D FHebdon 4/"-~/98 OFFICIAL RECORD COPY DOCUINENT NAINE: G:LBFNL99711RAI.EMB
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+**++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 22, 1998 Mr. O. J. Zeringue Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
BROWNS FERRY PLANT UNITS 2, AND 3 - REQUEST FOR ADDITIONAL INFORMATIONREGARDING TECHNICALSPECIFICATION CHANGE TS-384.
REQUEST FOR LICENSE AMENDMENTFOR POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)
Dear Mr. Zeringue:
By letter dated October 1, 1997, Tennessee Valley Authority (TVA)proposed revisions to the Browns Ferry Nuclear Plant, Units 2 and 3 Technical Specifications to permit operation of the units at the unrate power level of 3458 MWt. Our review of your application is in progress. The staff has determined that additional information is required to complete these reviews.
A description of the information required is provided in the enclosure.
Please provide the requested information by May 15, 1998. Your prompt response willassist us in completing timely reviews.
Ifyou have any questions, please write or call me at (301) 415-1471.
Sincerely, L. Raghavan, enior Project Manager Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos. 50-260, 50-296 Serial No. BFN-98-002
Enclosure:
As stated, cc: See next page
Mr. O. J. Zeringue Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT CC:
Mr. J. A. Scalice, Senior Vice President Nuclear Operations Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr'. Jack A: Bailey, Vice President'"
Engineering 8 Technical Services Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. C. M. Crane, Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 General Counsel Tennessee Valley Authority ET 10H 400 West Summit Hill Drive Knoxville, TN 37902 Mr. Raul R. Baron, General Manager Nuclear Assurance Tennessee Valley Authority 4J Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. Karl W. Singer, Plant Manager Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Mr. Mark J. Burzynski, Managar Nuclear Licensing Tennessee Valley Authority 4J Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801
'r. Tim'othy'E. Abney, Manager Licensing and Industry Affairs Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 37402-2801 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 61 Forsyth Street, SW., Suite 23T85
, Atlanta, GA 30303-3415 Mr. Leonard D. Wert Senior Resident Inspector U.S. Nuclear Regulatory Commission Browns Ferry Nuclear Plant I0833 Shaw Road Athens, AL35611 State Health Officer Alabama Dept. of Public Health 434 Monroe Street Montgomery, AL35180-170l Chairman Limestone County Commission 310 West Washington Street Athens, AL 35611
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With respect to revised stress and fatigue analyses of mechanical components as a result of the power uprate, please:
The following questions are related to the General Electric (GE), Licensing Topical Report NEDC-32751P, "Power Uprate Safety Analysis for The Browns Ferry Nuclear Plant (BFN),
Units 2 and 3," dated September 1997 which is Enclosure 5 to your letter dated October 1, 1997.
1.
(a)
Describe the analytical methodology, assumptions, loading combinations and allowable limits used for evaluating piping stresses and cumulative fatigue usage factors (CUF), pipe supports, load limits at nozzles, penetrations, guides, valves,
~ pumps, heat exchangers and anchors.
Please address all mechanical components affected by the power uprateincluding but not limited to, reactor vessel and internal components, control rod drive mechanisms, balance-of-plant (BOP) piping systems.
(b)
Indicate whether the analytical computer codes used in the evaluation are different from those'used in the original design-basis.
Ifnot, please justify their use including their qualification for such applications, and discuss how code differences are reconciled.
(c)
Provide a comparison of the maximum calculated stress, and CUF for the components evaluated, at the design basis and power uprate conditions.
Please indicate whether the revised. analyses are performed to the Code of record; otherwise describe the Code and Code edition used for the revised analyses and justify their use.
Discuss how the calculated CUFs compare to those resulting from the actual loading cycles based on the data recorded during plant operation.
(d)
CUFs in Table 3-3 for the reactor vessel is provided at four locations; feedwater nozzle, recirculation outlet nozzle, main closure stud and the support skirt. Provide a comparison of CUFs for the limiting reactor vessel components between the design basis analysis and the power uprate condition reconciled. Explain why the CUF for the feedwater nozzle based on the pre-uprate power condition is greater than the calculated CUF that incorporates the power uprate condition.
(e)
Please discuss whether you expect modifications to piping, equipment or their supports for the power uprate.
Ifany, list the piping systems and pipe supports requiring modification and discuss the nature of these modifications.
ENCLOSURE
It is stated (page 3-1) that main steam relief valves (MSRVs) operate in the safety (spring) mode only and the over pressure analysis assumes MSRV opening tolerance to be at least 3% above the nominal setpoints with one MSRV out of service.
Table 5-1 indicates that the assumed setpoints include a +3% tolerance.
The MSRVs are the Target Rock 2-stage design, which has a history of upward setpoint drift significantly greater than +3% in the safety mode (including some found to be effectively stuck closed during testing).
Please demonstrate that the 3% tolerance is conservative or provide analysis results for more representative Target Rock SRV setpoint performance.
You stated that an assessment of flow-induced vibration of the reactor internal
. component due to power uprate was performed.to estimate. the vibration levels by extrapolating the recorded vibration data at BFN and by using the operating experience of similar plants (Section 3.3.2.1).
Please provide a detailed sample evaluation and the technical basis for using the operating experience of similar plants in determining that the flow induced vibration at BFN Units 2 and 3 willremain within the acceptance limits.
You stated (Section 3.5) that the effects of power uprate have been evaluated for the recirculation loop piping using the present code of record, B31.1 Power Piping Code, 1967 Edition. You also stated that the piping was evaluated for compliance with the American Society of Mechanical Engineers (ASME) code stress criteria. Provide an explanation of why both B31.1-1967 Power Piping Code and ASME Code were used in the evaluation for the reactor coolant piping. Identify the piping analysis aspects for which both design codes were employed.
You stated (Section 3.5) that, based upon evaluations performed by GE for similar plants, the BFN recirculation piping is judged to be acceptable for flow induced vibration, due to 105% power uprate conditions.
Provide the technical basis for determining that the flow induced vibration levels in the recirculation piping system at BFN are bounded by evaluations performed for other plants.
Please discuss the functional performance of safety-related mechanical components (i.e., valves and pumps) affected by the power uprate to ensure that the performance specification and technical specification requirements (e.g., flow rates, close and open times) will be met for the proposed power uprate'.
Confirm that safety-related power-operated valves will be capable of performing their intended function(s) following the power uprate, including such affected parameters as fluid flow, temperature, pressure and differential pressure, and ambient temperature conditions.
Regarding Sections 3.5 and 4.1.2, please provide your evaluation ofthe piping systems attached to the torus shell, vent penetrations, pumps, and valves, that may be affected by the loss-of-coolant accident dynamic loads (pool swell, condensation oscillation, and chugging) considered in the evaluation for the power uprate.
Regarding Section 3.7, provide a detailed discussion of the effects of the steam flow increase, identified in Table 1-2, on the design basis analysis of the main steam piping due to main steam isolation valve (MSIV) closure and turbine stop valve closure loads.
Also, provide an evaluation of MSIVdue to the increase in the hydraulic pressure for the higher flow rate following the power uprate, as discussed in Section 4.7 of GE's Generic Evaluation (GE, Licensing Topical Report NEDC-32523P, "Generic Evaluation of General Electric Boiling Water Reactor Power Uprate - Supplement 1, Volume 1," dated June 1996).
9.
Please provide the evaluation of the feedwater heater for the power uprate with regard to vibration, stress and fatigue usage (Section 7.4.1).
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