ML17241A442
| ML17241A442 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/18/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17241A441 | List: |
| References | |
| NUDOCS 9908240124 | |
| Download: ML17241A442 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATIONBY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENTNO. 163TO FACILITYOPERATING LiCENSE NO. DPR-67 FLORIDA POWER AND LIGHTCOMPANY ST. LUCIE PLANT UNIT NO. 1 DOCKET NO. 50-335
1.0 INTRODUCTION
In a letter dated November 22, 1998, Florida Power & Light Company (FPL) requested changes to the St. Lucie, Unit 1, Technical Specifications (TS) in order to recover the analysis margin lost as a result of changes to the Reactor Coolant System (RCS) flowand low-flowtrip setpoint limits that were previously made to reflect steam generator (SG) tube plugging over more than 20 years of plant operation.
The SGs were replaced during the 1997 refueling outage for the current operating cycle (Cycle 15) and resulted in an increase in RCS flow compared to previous values with SG tube plugging.
Specifically, the Thermal Margin Safety LimitLines of TS Figure 2.1-1 would be revised to reflect an increase in the value of design minimum RCS flowfrom 345,000 gpm to 365,000 gpm. The flowrates stated in Tables 2.2-1 and 3.2-1 would be changed accordingly.
The reactor protective instrumentation Reactor Coolant Flow-Lowtrip setpoint limits would be increased from 93 percent to 95 percent of design flow.
In addition, the following TS changes are also proposed.
TS 5.6.1.a.1, "Design Features, Fuel Storage Criticality,"would be revised to delete the numerical value of criticality analysis uncertainty and reference the value described in Section 9.1 of the Updated Final Safety Analysis Report (UFSAR). The acceptable analytical methods used for determining core operating limits listed in TS 6.9.1.11," Administrative Controls, Core Operating Limits Report (COLR)," would be updated.
The Limiting Safety System Settings in Bases Section 2.2 for the Steam Generator Pressure-Low trip would be revised to delete the numerical value of steam pressure described as the full load operating point.
The specific change that the Nuclear Regulatory Commission (NRC) staff addressed was a proposed revision to TS 1.10 to change the reference for thyroid dose conversion factors used in Dose Equivalent Iodine-131 calculations from those listed in Table III of TID-14844 (Reference
- 1) to those listed in the International Commission on Radiological Protection (ICRP)
Publication 30 (Reference 2).
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PDR 2.0 EVALUATION 2.1 TS Fi ure2.1-1: ReactorCore ThermalMar in Safet Limit-Four ReactorCoolantPum s
~Oeratin The current 'I hermal Margin Safety LimitLines and the notation of Figure 2.1-1 reflect a design minimum RCS flowof 345,000 gpm with 4 pumps operating based on tube plugging in the previous SGs which have been replaced.
FPL has proposed to revise these to include the effects of the proposed design reactor coolant flowof 365,000 gpm with four pumps operating.
The associated TS Bases Figure B2.1-1 remains unchanged.
Based on the RCS flowof 407,000 gpm measured during the current Cycle 15 with the replacement SGs, the expected flowwith an assumed value of 15 percent SG tube plugging would be in excess of 380,000 gpm. Therefore, the proposed increase in minimum design flowto 365,000 gpm would provide sufficient margin to TS flowcompliance with 15 percent SG tube plugging after accounting for flowmeasurement uncertainties.
The new limits were obtained using approved methodology of the current fuel vendor, Siemens Power Corporation (SPC).
FPL has evaluated the impact of the proposed change to RCS design flowon applicable plant analyses.
The NRC staff has reviewed these evaluations, as discussed below, and has determined that the proposed increase in RCS design flowto 365,000 gpm would not adversely affect the safety analyses conclusions supporting operation of St. Lucie, Unit 1. Therefore, the proposed increase in RCS design flow is acceptable.
2.2 TS Table 2.2-1: Reactor Protective Instrumentation Tri Set oint Limits and TS Table 3.2-1: DNB Mar in Limits FPL has proposed to revise Table 2.2-1 to change the Reactor Coolant Flow-Lowtrip setpoint and allowable value from 93 percent of design reactor coolant flowwith four pumps operating to 95 percent of design reactor coolant flowwith four pumps operating.
The footnote would be changed to reflect the new design flowwith four pumps operating of 365,000 gpm.
In addition, Table 3.2-1 of TS 3.2.5 would be revised to change the Reactor Coolant Flow Rate from
>345,000 gpm to >365,000 gpm.
When the RCS flowfaIls below the trip setpoint, an automatic reactor trip signal is initiated to ensure that, for a degradation of RCS flowresulting from expected transients, local power density or departure from nucleate boiling ratio (DNBR) safety limits are not violated. The increase in value of the Reactor Coolant Flow-Lowtrip setpoint has a beneficial impact for those transients which rely on this trip since the trip would occur at a higher core flowrate leading to an earlier reactor trip. Therefore, the proposed trip setpoint increase is acceptable.
2.3 TS 1.10: Dose E uivalent l-131 In the definition of Dose Equivalent l-131, the reference for the thyroid dose conversion factors is changed from, "Table III of TID-14844, 'Calculation of Distance Factors for Power and Test Reactor Sites.'" to "ICRP-30, Supplement to Part 1, pages 192-212, Tables entitled,
'Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity(Sv/Bq)'."
The licensee's justification for making use of ICRP-30 thyroid dose conversion factors in place of those from TID-14844 is that the values in ICRP-30 are more recent and incorporate the considerable advances in the state of knowledge of limits for intakes of radionuclides.
- Also, the ICRP-30 dose conversion factors are consistent with Federal Guidance Report 11 (Reference 3) and the recommendations of Environmental Protection Agency 1987 guidance.
The staff has generally accepted the use of ICRP-30 dose conversion factors, and such use is consistent with current industry standards.
The licensee stated that the analyses of record for dose consequences use thyroid dose conversion factors from TID-14844. Since TID-14844 thyroid dose conversion factors give conservative dose results as compared to those in ICRP-30, the current dose analyses of record remain bounding and no reanalysis is necessary.
The licensee may and should use ICRP-30 thyroid dose conversion factors for future reanalyses.
The staff finds the licensee's proposed change to TS 1.10 and the use of the ICRP-30 thyroid dose conversion factors reflect current industry standards and account for advances in the state of knowledge of limits for intakes of radionuclides.
Therefore, the licensee's proposed change to TS 1.10 is acceptable.
2.4 TS5.6: Fuel Stora e
TS 5.6.1.a.1 states that the spent fuel storage racks are designed and shall be maintained with a Q equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 0.0065 ~k for uncertainties.
FPL has proposed to delete the numerical value of 0.0065 ~k, in TS 5.6.1.a.1, for the uncertainty allowance and reference Section 9.1 of the UFSAR for a description of these uncertainties.
The proposed TS 5.6 change affects a design feature.
The change does not affect a safety limit, limiting safety system setting, or limiting control setting as defined in 10 CFR 50.36.
Its removal from the TS does not affect a design feature, which, ifaltered or modified, would have a significant effect on safety.
The fuel storage criticality analysis uncertainties are dependent on the methods and assumptions used in the criticality analysis and are delineated in detail in Table 9.1-12 of the UFSAR. Any changes in these uncertainties due to revisions to the criticality analysis would be reflected in the UFSAR. This proposed change is also consistent with the Standard TS for Combustion Engineering Plants (NUREG-1432).
Based on the previous evaluation, the proposed change is acceptable.
2.5 TS6.9.1.11: Core 0 eratin Limits Re ort COLR FPL has proposed to update the analytical methods listed in TS 6.9.1.11.b.
The analytical methods listed in TS 6.9.1.11.b are approved methodologies used in the safety analyses performed for determining the St. Lucie, Unit 1, core operating limits that are documented in the COLR. The proposed modifications to the list would make it consistent with proper identification of the fuel vendor (SPC) topical reports and would improve the clarity and accuracy of report identification. Therefore, these proposed modifications are merely administrative in nature and are acceptable.
Two of the proposed references had not been approved by the NRC staff when these proposed modifications were submitt d with this amendment request.
One of the proposed references, EMF-84-093(P), Revision 1, subsequently has been approved by the NRC staff in a letter from F. Akstulewicz (NRC) to" J. F. Mallay (SPC), "Acceptance for Referencing of Topical Report EMF-84-093(P), Revision 1,
'Steam Line Break Methodology for PWRs'," dated February 16, 1999, and the accepted version, EMF-84-093(P)(A), Revision 1, which incorporates the acceptance letter, the staff safety evaluation, and an "A"(designating accepted) after the report identification symbol, was published by SPC in February 1999. Another proposed reference, EMF-2087(P), Revision 0, dated August 1998, has recently been approved by the NRC staff in a letter from C. Carpenter (NRC) to J. F. Mallay (SPC), "Acceptance for Referencing of Topical Report EMF-2087(P),
'SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCAApplications'," dated June 15, 1999. Therefore, these two reports are acceptable references for inclusion in TS 6.9.1.11.b.
2.6 Bases 2.2: Limitin Safet S stem Settin The Bases for "Steam Generator Pressure-Low" would be revised to delete the specific numerical value of 800 psig presently described as the "full-load operating point."
The present full-load operating point following replaceinent of the steam generators is approximately 850 psig.
However, the numerical value of the full-load steam pressure in the Bases is merely for descriptive purposes and has no bearing on the design requirement to have this setpoint sufficiently below the full-load operating point so as not to interfere with normal operation.
Therefore, the proposed change is editorial and therefore is acceptable.
2.7 Plant Safet Anal ses The current plant UFSAR Chapter 15 safety analyses of record use the design RCS flowof 345,000 gpm and the Reactor Coolant Flow-Low trip setpoint of 93 percent of design reactor coolant flow. SPC performed a safety evaluation of these analyses to assess the impact of an increase in the RCS flowto 365,000 gpm and an increase in the Reactor Coolant Flow-Low trip setpoint value to 95 percent of design reactor coolant flow.
For the design basis events, the increase in the design RCS flow rate has an insignificant effect on fuel centerline melt and shutdown margin and has a beneficial impact on DNBR considerations since it tends to increase the calculated DNBR. The proposed changes, therefore, would not adversely affect fuel failures and subsequent dose consequences for events that result in fuel failures. The proposed RCS flowchanges were also found to have an insignificant impact on the pressure rise for those events that are analyzed for the overpressurization of primary and secondary systems.
The decrease in core average temperature due to the increase in core flow rate would tend to increase the rate of depressurization in the small break loss of coolant accident (LOCA), resulting in a decrease in peak cladding temperature (PCT). Although the decreased temperature would cause a small change in the blowdown characteristics of a large break LOCA, the small effects of these changes would have an insignificant impact on the PCT. Therefore, the small break and large break LOCA analyses of record would remain bounding for the proposed changes.
The proposed flow increase would only impact the heat addition part of the Iow temperature overpressure protection (LTOP) analysis due to its effect on the heat transfer rate from the secondary side to the primary side of the steam generators.
However, the proposed flow would remain within the range of currently allowed minimum and maximum RCS flows for the operation of the plant. Therefore, the LTOP analysis would remain bounded by the analysis of record for the proposed RCS flowincrease.
In addition, since an increase in RCS flowtends to increase the resulting DNBR, the current DNB-LimitingCondition for Operation (LCO) and DNB-LimitingSafety System Settings (LSSS) setpoint analyses would remain bounding for the case with increased RCS flow and low flowtrip setpoint values.
The increases in design flow and low flowtrip setpoint have no impact on the Local Power Density (LPD)-LCO and the LPD-LSSS.
In summary, the staff has reviewed the evaluation of the St. Lucie, Unit 1, Chapter 15 safety analysis to assess the impact of the proposed increase in the RCS flow rate to 365,000 gpm and the proposed increase in the Reactor Coolant Flow-Lowtrip setpoint value to 95 percent of design reactor coolant flow. Based on our review, we conclude that the proposed changes are acceptable and that reanalysis of the current plant safety analyses is not required.
3.0 STATE CONSULTATION
By letter dated March 8, 1991, Mary E. Clark of the State of Florida, Department of Health and Rehabilitative Services, informed Deborah A. Miller, Licensing Assistant, U.S. NRC, that the State of Florida does not desire notification of issuance of license amendments.
Thus, the State official had no comments.
4.0 ENVIRONMENTALC NSIDERATION These amendments change a requirement with respect to 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 (64 FR 6696). Accordingly, these amendments meet the eligibilitycriteria 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 these amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public willnot 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 willnot be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
L. Kopp, SRXB M. Hart, SPSB Date:
August 18, 1999
References 1.
TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," Division of Licensing and Regulation, U.S. Atomic Energy Commission, Washington, DC, March 23,
'I962.
2.
ICRP Publication 30, Part 1 and Supplement to Part 1, "Limits for Intakes of Radionuclides by Workers," Annals of the ICRP, Vol. 2, No. 3/4 and Vol. 3, No. 1-4, 1979, Pergamon Press.
3.
Federal Guidance Report No. 11, "LimitingValues of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,"
Office of Radiation Programs, U.S. Environmental Protection Agency, Washington, DC, 1988.
Mr T. F. Plunkett Florida Power and Light Company ST. LUCIE PLANT CC:
Senior Resident Inspector St. Lucie Plant U.S. Nuclear Regulatory Commission P.O. Box 6090 Jensen Beach, Florida 34957 Joe Myers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 M. S. Ross, Attorney Florida Power L Light Company P.O. Box 14000 Juno Beach, FL 33408-0420 Mr. Douglas Anderson County Administrator St. Lucie County 2300 Virginia Avenue Fort Pierce, Florida 34982 Mr. WilliamA. Passetti, Chief Department of Health Bureau of Radiation Control 2020 Capital Circle, SE, Bin ¹C21 Tallahassee, Florida 32399-1741 J. A. Stall, Site Vice President St. Lucie Nuclear Plant 6351 South Ocean Drive Jensen Beach, Florida 34957 Mr. R. G. West Plant General Manager St. Lucie Nuclear Plant 6351 South Ocean Drive Jensen Beach, Florida 34957 E. J. Weinkam Licensing Manager St. Lucie Nuclear Plant 6351 South Ocean Drive Jensen Beach, Florida 34957 Mr. John Gianfrancesco Manager, Administrative Support and Special Projects P.O. Box 14000 Juno Beach, FL 33408-0420 Mr. Rajiv S. Kundalkar Vice President - Nuclear Engineering Florida Power & Light Company P.O. Box 14000 Juno Beach, FL 33408-0420 Mr. J. Kammel Radiological Emergency Planning Administrator Department of Public Safety 6000 SE. Tower Drive Stuart, Florida 34997