ML17229A922
| ML17229A922 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/22/1998 |
| From: | Stall J FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML17229A923 | List: |
| References | |
| L-98-283, NUDOCS 9812010024 | |
| Download: ML17229A922 (26) | |
Text
p i 'r j CATEGORY 1 t
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9812010024 DOC.DATE: 98/11/22 NOTARIZED: YES DOCKET FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power &. Light Co. 05000335 AUTH. NAlrg AUTHOR AFFILIATION STALL,J.A. Florida Power 6 Light Co.
RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)
SUBJECT:
Application for amend to license DPR-67,revising thermal margin SL lines of TS Figure 2.1-1 to reflect increase in value of design min RCS flow from 345,000 gpm to 365,000 gpm & change flow rates stated in Tables 2.2-1 &. 3.2-1.
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0 Florida Power 6t Light Company, 6351 S. Ocean Drive, Jensen Beach, FL 34957 November 22, 1998 L-98-283 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 1 Docket No. 50-335 CS
Proposed License Amendment u o a dTSU dae Pursuant to 10 CFR 50.90, Florida Power 8 Light Company (FPL) requests to amend Facility Operating License DPR-67 for St. Lucie Unit 1 by incorporating the attached Technical Specifications (TS) revisions. The amendment will revise the Thermal Margin Safety Limit Lines of TS Figure 2.1-1 to reflect an increase in the value of design minimum Reactor Coolant System (RCS) flow from 345,000 gpm to 365,000 gpm, and change the flow rates stated in Tables 2.2-1 and 3.2-1 accordingly. The reactor protective instrumentation Reactor Coolant Flow-Low trip setpoint limits will be changed from 93% to 95% of design RCS flow. These proposed changes are intended to recover the analysis margin lost as a result of changes to the flow and low flow trip setpoint limits that were made necessary by steam generator tube plugging over more than 20 years of plant operation. The steam generators at St. Lucie Unit 1 were replaced during the 1997 refueling outage for the current operating cycle and, due to the resultant increase in RCS flow, the proposed changes will benefit the safety analyses while maintaining sufficient operating margins related to these parameters.
Other changes proposed in this license amendment include: a revision to TS 1.10 to update the reference for dose conversion factors used in Dose Equivalent Iodine-131 calculations from those listed in Table III of TID-14844 to those listed in ICRP-30; and administrative changes that delete the numerical value ofcdgcaltty analysis uncertainty described in fuel storage design features TS 5,6.1.s.t, update the analytical methods used for determining core operating limits listed in administrative controls
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TS 6.9.1.11, and revise Limiting Safety System Settings Bases 2.2 for theSteam Generator Pressure-Low trip to delete the numerical value of steam pressure descnbed as the full load operating point.
Please note that two of the Topical Reports shown in the revised list of analytical methods for TS 6.9.1.11.b are currently being reviewed by the NRC staff. The reports are identified in Part 3.6 of Attachment 1 (page 6 of 11). Including these Topical Reports in TS 6.9.1.11.b, therefore, is contingent
. upon completion of the staff's review of both reports and final approval thereof.
It is requested that the proposed amendment, if approved, be issued by August 15, 1999.
Attachment 1 is an evaluation of the proposed TS changes. Attachment 2 is the "Determination of No Significant Hazards Consideration." Attachment 3 contains a copy of the affected TS pages marked-up to show the proposed changes.
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98i20i0024 'ii8ii222t 05000335 PDR ADQCK PDR I an FPL Group company
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St. Lucie Unit 1 L-98-283 Docket No. 50-335 Page 2 CS
Proposed License Amendment u o a dae The proposed amendment has been reviewed by the St. Lucie Facility Review Group and the Florida Power 8 Light Company Nuclear Review Board. In accordance with 10 CFR 50.91 (b)(1), a copy of the proposed amendment is being forwarded to the State Designee for the State of Florida.
Please contact us if there are any questions about this submittal.
Very truly yours, J. A. Stall Vice President St. Lucie Plant JAS/RLD Attachments cc: Regional Administrator, Region II, USNRC.
Senior Resident Inspector, USNRC, St. Lucie Plant.
Mr. W.A. Passetti, Florida Department of Health and Rehabilitative Services.
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St. Lucie Unit 1 L-98-283 Docket No. 50-335 Page 3 CS
Proposed License Amendment u owa d SU date STATE OF FLORIDA )
) ss.
COUNTY OF ST. LUCIE )
J. A. Stall being first duly sworn, deposes and says:
That he is Vice President, St. Lucie Plant, for the Nuclear Division of Florida Power 8 Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.
J. A. Stall STATE OF FLORIDA COUNTY OF ST C.Oct e
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Sworn to and subscribed before me this 2 > day of /4dA'~ I0~, 19~
by J. A. Stall, who is personally known to me.
" Signature of No Public-State of Florida G EORGE R. teOOEg
!'.r:a.= MY 00MMISl0H CC t 062026 EXPlRES: June 17, 2000
";p,~" BondedTtuuNotaryPuSc Undeneltsa Name of Notary Public (Print, Type, or Stamp
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St. Luciq Unit 1
'ocket No. 50-335 Proposed License Amendment CS a dae e e 83 EVALUATIONOF PROPOSED TS CHANGES 1.0 Introduction 2.0 Background 3.0 Proposed Changes: Description and Bases/Justification 3.1 TS 1.10: DOSE EQUIVALENTI-131 3.2 TS Figure 2.1-1: REACTOR CORE THERMAL MARGIN SAFETY LIMIT-FOUR REACTOR COOLING PUMPS OPERATING 3.3 TS Table 2.2-1: REACTOR PROTECTIVE INSTRUMENTATIONTRIP SETPOINT LIMITS and TS Table 3.2-1: DNB MARGIN LIMITS 3.4 BASES 2.2: LIMITINGSAFETY SYSTEM SETTINGS 3.5 TS 5.6: FUEL STORAGE 3.6 TS 6.9.1.11: CORE OPERATING LIMITS REPORT (COLR) 4.0 Summary of Evaluation of Impact from Proposed Minimum RCS Flow and Low-Flow Trip Setpoint on Plant Analyses 4.1 Evaluation of Design Basis Events 4.2 Setpoint Analyses 4.3 Impact of the License Amendment on Other Analyses 5.0 Environmental Consideration 6.0 Conclusion
~ 7.0 References Note: Information in this attachment, in part, has been reformatted and edited from FPL Nuclear Engineering Safety Evaluation PSL-ENG-SEFJ-98-009: 11/5/98.
St. Lucie Unit 1 L-98-283
'ocket No. 50-335 Attachment 1 Proposed License Amendment Page 1 of 11 C U dae EVALUATIONOF PROPOSED TS CHANGES The proposed amendment to Facility Operating License DPR-67 for St. Lucie Unit 1 (PSL1) will revise the Thermal Margin Safety Limit Lines of Technical Specifications (TS) Figure 2.1-1 to reflect an increase in the value of design minimum Reactor Coolant System (RCS) flow from 345,000 gpm to 365,000 gpm, and change the flow rates stated in Tables 2.2-1 and 3.2-1 accordingly. The reactor protective instrumentation Reactor Coolant Flow-Low trip setpoint limits will be changed from 93% to 95% of design RCS flow. These proposed changes are intended to recover the analysis margin lost as a result of changes to the flow and low-flow trip setpoint limits that were made necessary by steam generator tube plugging over more than 20 years of plant operation. The steam generators at St. Lucie Unit 1 were replaced during the 1997 refueling outage for the current operating cycle and, due to the resultant increase in RCS flow, the proposed changes will benefit the safety analyses while maintaining sufficient operating margins related to these parameters.
Other changes proposed in this license amendment include: a revision to TS 1.10 to update the reference for dose conversion factors used in Dose Equivalent Iodine-131 calculations from those listed in Table III of TID-14844 to those listed in ICRP-30; and administrative changes that delete the numerical value of criticality analysis uncertainty described in fuel storage design features TS 5.6.1.a.1, update the analytical methods used for determining core operating limits listed in administrative controls TS 6.9.1.11, and revise Limiting Safety System Settings Bases 2.2 for the Steam Generator Pressure-Low trip to delete the numerical value of steam pressure described as the full load operating point. FPL has determined that the proposed revisions would not adversely affect the safety analyses conclusions supporting the operation of St. Lucie Unit 1, and it is concluded in Attachment 2 that the proposed amendment does not involve a significant hazards consideration.
20 Bac ou TS 1.10: "DOSE EQUIVALENT I-131 shall be that concentration of I-131 (pCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, l-132, I-133, 1-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, 'Calculation of Distance Factors for Power and Test Reactor Sites'." RCS Limiting Condition for Operation TS 3.4.8, "Specific Activity," establishes limits and surveillance requirements for DOSE EQUIVALENTl-131 in the primary coolant.
The St. Lucie Unit 1 steam generators were replaced in 1997 during the Cycle-15 refueling outage after more than 20 years of plant operation. RCS flow had decreased considerably with the old steam generators as a result, in part, of having approximately 25% (average) of the total number of tubes plugged. Revisions to the specified design minimum RCS flow through the years culminated with the currently specified value of 345,000 gpm. RCS flow with the replacement steam generators is greater than 400,000 gpm.
TS 2.1.1 requires that "the combination of THERMAL POWER, pressurizer pressure, and maximum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1." The curves of Figure
St. Lucie Unit 1 L-98-283
'ocket No. 50-335 i Attachment 1 CS
Proposed License Amendment o ad SUdae Page 2 of 11 2.1-1 represent the loci of points for these parameters, with four reactor cooling pumps operating, for which the minimum Departure from Nucleate Boiling Ratio (DNBR) is no less than the DNBR limit for the axial (power distribution) shape shown in Figure B2.1-1. The present Figure 2.1-1 shows thermal margin safety limit lines that are valid for reactor vessel flow of 345,000 gpm and is annotated accordingly.
TS 2.2.1 requires that the reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. The table presently lists the trip setpoint and the allowable value for the Reactor Coolant Flow-Low trip as z 93% of design reactor coolant flow with 4 pumps operating*. The asterisk is footnoted with the statement, "*Design reactor coolant flow with 4 pumps operating is 345,000 gpm."
TS 3.2.5 requires that DNB related parameters shall be maintained within the limits shown on Table 3.2-
- 1. This table lists the reactor coolant flow rate as ~ 345,000 gpm.
"
BASES 2.2, LIMITINGSAFETY SYSTEM SETTINGS states, The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setting of 600 psia is sufficiently below the full-load operating point of 800 psig so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of a 22 psi in the accident analyses."
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3.0 o osed C a es Desc o a 8 ses Just catio The affected TS pages, marked-up to show the proposed changes, are included in Attachment 3.
3.1 S 0 a e1-3 DOSE E UV - 3 In the Definition of DOSE EQUIVALENT 1-131, 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)'."
Basis/Justificatio: The derived limits in the International Commission on Radiological Protection (ICRP), Publication 30 (Reference 1), represent more recent values for the thyroid dose conversion factors, and incorporate the considerable advances in the state of knowledge of limits for intakes of radionuclides. The ICRP dose conversion factors are consistent with the Federal Guidance Report No. 11 (Reference 2), and the recommendations of Environmental Protection Agency (EPA) 1987 guidance. The syntax used for the ICRP-30 reference in proposed TS 1.10 is consistent with the corresponding definition in the Standard TS for Combustion Engineering Plants (NUREG-1432).
The safety analyses of record dose consequences are based on the TID-14844 thyroid dose conversion factors. Since these dose conversion factors produce conservative dose consequences as compared to the proposed change to ICRP-30, the dose consequences of all the analyses of record remain bounding for the proposed changes. The analyses
St. Lucie Unit 1 L-98-283
'ocket No. 50-335 Attachment 1 Proposed License Amendment Page 3 of 11
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CS u lowand S U date consequences thus would continue to remain in compliance with the acceptance criteria of 10 CFR 100 limits, and no reanalysis is required. However, the proposed change would allow the use of the ICRP values for dose consequences of any future reanalysis, consistent with the current industiy standards.
3.2 a G U S The current Thermal Margin Safety Limit Lines and the notation of Figure 2.1-1 reflect a design minimum RCS flow of 345,000 gpm, and are revised to include the effects of the proposed design reactor coolant flow of 365,000 gpm with four pumps operating. Figure 2.1-1 is associated with the TS Basis Figure B2.1-1 that remains unchanged.
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replacement steam generators was approximately 407,000 gpm. The expected flow with an assumed value of 15% steam generator tube plugging is in excess of 380,000 gpm. The proposed increase in the minimum design RCS flow to 365,000 gpm would thus provide sufficient margin to TS flow compliance with 15% steam generator tube plugging, after accounting for the flow measurement uncertainties. The revised TS Figure 2.1-1 includes the effects of the proposed reactor coolant system design flow rate of 365,000 gpm with four reactor coolant pumps operating. The new limits on plant operating space shown on the updated figure were obtained utilizing current fuel vendor approved methodology, and is associated with the axial power distribution shown in the Technical Specification Basis Figure B2.1-1.
The Thermal Margin Safety Limit Lines shown in Figure 2.1-1 do not include allowances for instrument error or fluctuations. As such, this figure cannot be directly used to verify any trip setpoints. Implementation of these safety limits in the plant does require consideration of instrument error and other uncertainty effects which are included in the generation of the Thermal Margin/Low Pressure (TM/LP) trip and Variable High Power Trip (VHPT) functions and in the pressure values selected for the actuation of Main Steam Safety Valves. This setpoint verification is performed every cycle to confirm the adequacy of the trip functions to provide necessary plant protection.
FPL has evaluated the impact of the proposed change to RCS design minimum flow on applicable plant analyses, and determined that the proposed revisions would not adversely affect the safety analyses conclusions supporting the operation of St. Lucie Unit 1. The results of this evaluation are summarized in Part 4.0 of this attachment.
3.3 CO 0 C S U TO 0 S a e - andTSTABL 32- D B G I ITS Pa e3/42-14 Table 2.2-1 is revised to change the Reactor Coolant Flow-Low TRIP SETPOINT and ALLOWABLEVALUEfrom 93% of design reactor coolant flow with 4 pumps operating to 95%
of design reactor coolant flow with 4 pumps operating. In the footnote at the bottom of page 2-4, the design reactor coolant flow with 4 pumps operating is changed from 345,000 gpm to 365,000 gpm.
St. Lucie Unit 1 L-98-283 Docket No. 50-335 Attachment 1
'CS
Proposed License Amendment u a SUdae Page 4 of 11 Table 3.2-1 of TS 3.2.5 is revised to change the Reactor Coolant Flow Rate from >345,000 gpm to >365,000 gpm.
RCS flow from 345,000 gpm to 365,000 gpm, and an increase in the Reactor Coolant Flow-Low TRIP SETPOINT and ALLOWABLEVALUES from 93% to 95% of the design reactor coolant flow with 4 pumps operating. These changes will provide increased analysis margin to the Departure from Nucleate Boiling Ratio (DNBR), while maintaining sufficient operating margins to these parameters due to the higher RCS flow resulting from the replacement steam generators.
The RCS flow measured during the operation of*Cycle 15 with the replacement steam generators was approximately 407,000 gpm. The expected flow with an assumed value of 15%
steam generator tube plugging is in excess of 380,000 gpm. The proposed increase in the minimum design RCS flow to 365,000 gpm would thus provide sufficient margin to TS flow compliance with 15% steam generator tube plugging after accounting for the flow measurement uncertainties. Additionally, the increase in the low-flow trip setpoint value to 95% of the design RCS flow would maintain adequate operational margin to the trip setpoint. These proposed values are comparable to the values of 370,000 gpm for design RCS flow and 95% of design RCS flow for the trip setpoint that existed prior to the changes that were made by License Amendment No. 130 to accommodate steam generator tube plugging up to 25% (average).
The revision to TS Table 3.2-1, DNB MARGIN LIMITS, is consistent with the revised design reactor coolant flow with 4 pumps operating, annotated in proposed Figure 2.2-1 and in the proposed footnote of Table 2.2-1.
FPL has evaluated the impact on applicable plant analyses from the proposed change to the Reactor Coolant Flow-Low TRIP SETPOINT and ALLOWABLEVALUES in conjunction with the proposed RCS minimum flow value, and determined that the proposed revisions would not adversely affect the safety analyses conclusions supporting the operation of St. Lucie Unit 1.
The results of this evaluation are summarized in Part 4.0 of this attachment 3.4 S GS SS GS e -5 The Bases for "Steam Generator Pressure-Low" is revised to delete the specific numerical value of 800 psig presently described as the "full-load operating point."
Pressure-Low reactor protective system trip setpoint currently describes the steam pressure full-load operating point as 800 psig. The proposed revision will delete this numerical value from the Bases.
The present full-load operating point following replacement of the St. Lucie Unit 1 steam generators is approximately 850 psig and will change over the life of the new steam generators, in part, as a function of steam generator tube plugging. The numerical value of the full-load steam pressure is contained in the bases in a simple descriptive context and, as such, has no bearing on the design requirement to have this setpoint sufficiently below the full-load operating
~ ~ Q St. Luciy Unit 1 L-98-283
'ocket No. 50-335 Attachment 1 CS
Proposed License Amendment u o a d SU date Page 5 of 11 point so as not to interfere with normal plant operation. The proposed change will not alter the meaning of the affected Bases statement and will obviate the need for future license amendment(s) to maintain its accuracy.
3.5 S 56' LS 0 AGE Pa e 5-5 TS 5.6.1.a. states that the spent fuel storage racks are designed and shall be maintained with:
"
- 1. A k equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 0.0065 b,k for uncertainties."
The numerical value for uncertainty allowance is deleted and TS 5.6.1.a.1 is revised to read:
"
Q less than or equal to 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report."
allowance specified in this TS. The fuel storage criticality analysis, however, would continue to account for proper uncertainties and biases as appropriate for the analytical methods in use.
There is no change to the fuel storage criticality analysis or the calculated value of Q due to this proposed revision.
The fuel storage criticality analysis uncertainties are currently delineated in detail in Table 9.1-12 of the Updated Final Safety Analysis Report (UFSAR). Since the uncertainty value is dependent on the methods and assumptions used in the criticality analysis, it is appropriate for this value to be described in the UFSAR. The proposed change is also consistent with the Standard TS for Combustion Engineering Plants (NUREG-1432). Any changes to the fuel storage criticality analysis and the corresponding uncertainties will be controlled pursuant to 10 CFR 50.59.
a 3.6 S69 CO 0 E G S 0 CO aes6-96-9a The analytical methods listed in TS 6.9.1.11.b are updated to include in a consistent manner the current fuel vendor (Siemens Power Corporation) methodology that is used or could be used to determine the St. Lucie Unit 1 core operating limits that are documented in the Core Operating Limits Report (COLR). Existing Items 3 through 11.e) are revised with editorial corrections to improve clarity and accuracy of report identification. Items 6.b), 11.f), and 12 through 16 are additions to the existing list. The revisions are presented as INSERT - B in Attachment 3.
properly identify the approved'ethodologies used or to be used in the safety analyses performed for determining the St. Lucie Unit 1 core operating limits that are documented in the COLR. The updated list is made consistent with proper identification of the fuel vendor (Siemens Power Corporation) Topical Reports as recommended by the fuel vendor (Reference
- 3) and reviewed by FPL..The proposed revision to this specification is administrative in nature.
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St. Luciy Unit 1 L-98-283 Docket No. 50-335 Attachment 1 Proposed' License Amendment Page 6 of 11 CS d U dae
- Two of the Topical Reports shown in the revised list are currently under review by the NRC staff. Therefore, inclusion of these Topical Reports (speciflied below) in TS 6.9.1.11.b is contingent upon completion of the staff's review and final approval.
~ebb: EMF-84-093(P) Revision 1, "Steam Line Break Methodology for PWRs,"
Siemens Power Corporation, June 1998 (This document is a revision to ANF 93).
gee~: EMF-2087(P) Revision 0, "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation, August 1998.
The methodologies that have been added to the list (Item 6.b), Item 11.f), and Items 12 through
- 16) could directly or indirectly be used in determining the COLR limits and are included to make the list comprehensive. These methodologies include those used in performing large break LOCA analyses, non-LOCA transient analyses and setpoint verification analyses, and those that support the use of gadolinia fuel and extended burnups.
4.0 Su a o a uat'o o c o o osed u CS o a dLo - o Set o to a al ses The Departure from Nucleate Boiling Ratio (DNBR) parameter is sensitive to changes in core flow rate as well as the value of core flow at the time of reactor trip. The"changes included in this proposed license amendment affect the plant safety analyses in the following manner: (a) an increase in core flow rate tends to increase the calculated DNBR. The DNBR parameter is a direct indication of available thermal margin, and an increase in the calculated minimum DNBR indicates that thermal margin for the corresponding transient has been increased; and (b) an increase in the value of the Reactor Coolant Flow-Low trip setpoint would have a beneficial impact on the calculated DNBR for certain transients. The increased low-flow trip setpoint would result in the reactor trip occurring at a higher core flow rate leading to an earlier reactor trip.
The evaluation of the effects on safety was performed by Siemens Power Corporation (Reference 4) and FPL, and is summarized below:
4.1 a ua 'o of Des'asis vents The current analyses of record use the design reactor coolant flow of 345,000 gpm and the Reactor Coolant Flow-Low trip setpoint of 93% of design reactor coolant flow. These analyses also support the plant operating conditions associated with the replacement steam generators relative to the steam generator pressure, RCS flow, reactor outlet temperature, et al.
(References 5 and 6).
A review of events described in the St. Lucie Unit 1 UFSAR was performed to assess the impact of an increase in the Reactor Coolant Flow Rate to 365,000 gpm, and an increase in the Reactor Coolant Flow-Low trip setpoint value to 95% of design reactor coolant flow. Based on this evaluation of design basis events, reanalysis is not required.
St. Luciy Unit 1 L-98-283
'ocket No. 50-335 Attachment 1 Proposed License Amendment Page 7 of 11 RCS 'mum FlowandTS U date crease 'at e o al b he Seconda S s em The events in this category are cooldown events and are analyzed for DNBR, fuel centerline melt (CLM), shutdown margin and radiological consequences. An increase in the design RCS flow rate would have an insignificant impact on the CLM and shutdown margin analyses and would be beneficial from DNBR considerations. The proposed changes therefore would not adversely affect fuel failures and subsequent dose consequences for events that allow fuel failures. The limiting event for fuel failures and dose consequences is Steam Line Break. The radiological consequences for events with no fuel failures are governed by the TS activity levels and the primary to secondary leak rates. These parameters remain unaffected by the proposed changes. The limiting event for dose consequences in this category is Inadvertent Opening of a Steam Generator Relief or Safety valve. The current analyses of record, therefore, would remain valid.
ece se'a e o alb eSeconda S se These events are bounded by the Loss of Forced Reactor Coolant flow event for DNBR and are mainly analyzed for the over-pressurization of primary and secondary systems. The proposed RCS flow changes would have an insignificant impact on the pressure rise for these events. The limiting event for over-pressurization is the Loss of External Load. The radiological consequences are based on the TS activity levels and the primary to secondary leak rates.
These parameters remain unaffected by the proposed changes. The limiting event for dose consequences in this category is Loss of Non-Emergency AC Power to Station Auxiliaries. The Loss of Normal Feedwater event, analyzed for secondary side liquid inventory, would be less severe for the case of increased RCS flow due to the lower initial primary system stored energy that has to be removed by boilirig off the secondary side inventory. The current analyses of record, therefore, would remain valid.
ecrease 'ac o Coo ant S s e Flow These events are mainly analyzed for Departure from Nucleate Boiling (DNB) considerations.
Since increased RCS flow and low flow trip setpoint tend to increase the calculated DNBR, the analysis of record would remain bounding for the proposed changes.
e d we 's ribu ' alies The only events in this category impacted by the proposed changes are those analyzed relative to DNBR. Since increased RCS flow and low flow trip setpoint tend to increase the calculated DNBR, the analysis of record would remain bounding for the proposed changes.
cease'aco C oa e o There are no events in this category which are impacted by the proposed changes.
St. Luciq Unit 1 L-98-283
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Proposed License Amendment u o a SU dae Page 8 of 11 Dec ease 'ac o Coola t ve o Inadvertent opening of a Pressurizer Pressure Relief valve is analyzed for DNBR. Since an increase in the RCS flow tends to increase the DNBR, the current analysis remains bounding for this event.
Steam Generator Tube Rupture is analyzed for radiological consequences. These consequences are based on the RCS and secondary system activity levels and the primary to secondary break flow. The activity levels used in the analysis of record are the Technical Specifications maximum allowed activity levels, which remain unchanged. The driving force for the break flow is the pressure difference between the primary side and the secondary side. The proposed design RCS flow rate corresponds to a steam generator tube plugging level of between 15% and 25%. The primary to secondary pressure difference at these conditions would remain lower than the conservative pressure difference used in the analysis of record, which supports a higher tube plugging level. A decrease in the primary to secondary pressure difference would tend to reduce the primary to secondary break flow and result in lower dose consequences. The analysis of record would thus remain bounding.
The increase in core flow rate would cause a small reduction in the core average temperature.
The decreased temperature would tend to increase the rate of depressurization in the small break LOCA transient, and eventually result in a decrease in the peak cladding temperature (PCT). The decreased temperature would cause a small change in the blowdown characteristics of a large break LOCA transient. The effects of these changes are estimated to be small with an insignificant impact on the PCT. The analyses of record for small break and large break LOCAs would thus remain bounding for the proposed changes.
ad'oac i e e eases from a Subs s e o Co o en The core flow rate has no direct impact on the consequences of events in this category, which therefore remain bounding for the proposed changes.
4.2 The setpoint analyses verify the following setpoints:
DNB-Limiting Condition for Operation (DNB-LCO)
DNB-I imiting Safety System Settings (DNB-LSSS)
Local Power Density-Limiting Condition for Operation (LPD-LCO)
Local Power Density-Limiting Safety System Settings (LPD-LSSS)
Increases in the design RCS flow and the low flow trip setpoint have no impact on the LPD-LCO and LPD-LSSS verification analyses. Since an increase in the RCS flow tends to increase the DNBR, the current DNB-LCO and DNB-LSSS analyses would remain bounding for the case with increased RCS flow and low flow trip setpoint values.
St. Lucie Unit 1 L-98-283 Docket No. 50-335 Attachment 1 Proposed
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License Amendment Page 9 of 11 CS owand SU dae 4.3 eLicese ed e o 0 e a ses While the preceding discussion focused on the impact of the proposed changes to the RCS flow and the low-flow trip setpoint on the results of the plant UFSAR Chapter 15 Safety Analysis, a complete evaluation of the impact of the proposed changes requires the evaluation of several additional issues. A detailed evaluation of all these issues was performed for Cycle 15 (Reference 6) to address the impact of the increased RCS flow due to the replacement steam generators. The proposed change affects only the minimum flow analyses. Since the proposed change is to increase the minimum RCS flow, the impact of the change remains bounded by the minimum and maximum flow analyses of record. A summary of other important issues is provided below:
a a u al Circula io Ca abili Natural circulation capability is determined mainly by the plant configuration parameters, which remain unchanged by the proposed amendment, and is not significantly impacted by the steady state design minimum RCS flow and the low-flow trip setpoint values. The proposed changes thus would not have any significant adverse impact on the natural circulation cooling capability of the plant.
e Co a' essuo o BOC o Sea Sse a' A change in the steady state design RCS flow would impact the containment analysis only through a change in the initial stored energy. The proposed increase in the design RCS flow would cause a reduction in the RCS average temperature of about 1.5'F, and a corresponding reduction in the initial system energy. A lower initial stored energy would be beneficial from the containment analysis viewpoint because a lower initial stored energy would result in a lower peak containment pressure. The peak pressure in the analysis of record would thus remain bounding for the proposed changes.
eed ae se e e ea The existing system evaluation for the AFW system identified this event in conjunction with loss of off-site power as the limiting condition for plant operators to be able to initiate auxiliary feedwater flow. This event is benefited by a decrease in the initial average RCS temperature.
Since the proposed change to the RCS flow would decrease the RCS average temperature, this event would not be adversely impacted by the proposed changes.
o e eaueO er ressuePoecio OP A al sis A change in the RCS flow would have no impact on the mass addition part of the LTOP analysis. An increase in the RCS flow would however impact the energy addition part of the analysis due to its effect on the heat transfer rate from the secondary side to the primary side of the steam generators. With the proposed change to the RCS flow, the RCS flow would remain within the range of currently allowed minimum and maximum RCS flows for the operation of the plant. The effect of the RCS flow on the energy addition part of the analysis thus would remain bounded by the analysis of record. Other parameters affecting this analysis
St. Luciy Unit 1 L-98-283 Docket No. 50-335 Attachment 1 Proposed' License Amendment Page 10 of 11 CS u owandTSU date are the RCP heat output and the initial primang to secondary temperature difference. These parameters are not impacted by the proposed changes. Therefore, the proposed changes would have no significant adverse impact on this event.
0 e essure P otection nal sis In Appendix 5A of the St. Lucie Unit 1 UFSAR, an analysis documenting the sizing of the primary and secondary safety valves is presented. The intent of this analysis, which is based on a "worst case" loss of load event, is to size the safety valves. The safety valve flow rates are chosen such that the peak system pressure is less than 110% of the design pressure.
The impact of the proposed changes on the analysis of record for the Loss of External Load event was evaluated and found to be insignificant (discussed in Part 4.1). This analysis thus would continue to remain in compliance with the over-pressurization criteria. Therefore, it is concluded that pressurizer safety valves, main steam safety valves and the reactor protective system would continue to provide overpressure protection for the steam generators and the reactor coolant system when considering the effects of the proposed changes.
o CS o o S ess a s's The proposed change to the minimum RCS flow would result in a decrease in the corresponding core outlet temperature. The RCS temperatures and flows would continue to remain within the range of temperatures corresponding to the currently allowed minimum and maximum RCS flows. Therefore, there is no adverse impact on any acceptance criteria, and sufficient margin to stress limits would continue to remain available for all RCS components.
5.0 v' e tal Co s de ato The proposed license amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The proposed amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and no significant increase in individual or cumulative occupational radiation exposure. FPL has concluded that the proposed amendment involves no significant hazards consideration and meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and that, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment need not be prepared in connection with issuance of the amendment.
6.0 ~Co c us'o The results of this safety evaluation demonstrate that the proposed amendment does not result in a violation of any safety limit or analysis acceptance criterion. Therefore, there is no safety concern associated with the proposed license amendment. As shown in Attachment 2, the proposed changes do not involve a significant hazards consideration. FPL concludes that this license amendment is, therefore, acceptable with respect to the operation of St. Lucie Unit 1.
St. Lucie Unit 1 L-98-283 Docket No. 50-335 Attachment 1 CS i' Proposed License Amendment a SU date Page 11 of 11 7.0 ~ee aces
- 1. ICRP Publication 30, Part I and Supplement to Part I, "Limits for Intakes of Radionuclides by Workers,"Annals ofthe ICRP, Vol 2, No. 3/4 and Vol 3, No. 1-4, 1979, Pergamon Press
- 2. Federal Guidance Report No. 11, "Limiting Values 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 D.C., 1988
- 3. *Letter RIW:98:269, R. I. Wescott (SPC) to R. J. Rodriguez (FPL), "List of SPC Methodology for St. Lucie Unit 1 Technical Specifications," August 4, 1998
- 4. *EMF-97-044, Revision 1, "St. Lucie Unit 1, Safety Evaluation of an Increase in Minimum Technical Specification RCS Flow Rate to 365,000 gpm," Siemens Power Corporation, December 1997
- 5. *EMF-97-058, Revision 0, "St. Lucie Unit 1 Cycle 15 Safety Analysis Report (FSAR Chapter 15 Review and Mechanical)," Siemens Power Corporation, July 1997
- 6. * "Steam Generator Equivalency Report (SGER), Revision 1, SASE - Volume 2,"
Prepared by Framatome Nuclear Technologies (FTI), June 1997 Note: References annotated with an asterisk are currently retained by the Engineering Nuclear Fuels Group as part of FPL's Nuclear Engineering Department records.
St. Lucie Unit 1 Docket No. 50-335 CS 'i Proposed License Amendment u o a d SU date ee 3 DETERMINATIONOF NO SIGNIFICANT HAZARDS CONSIDERATION
St. Lucie Unit 1 L-98-283 Docket No. 50-335 Attachment 2 CS
Proposed License Amendment u o ad SUdate Page 1 of 2 DETERMINATIONOF NO SIGNIFICANT HAZARDS CONSIDERATION Descriptio of amendment request: The proposed amendment for St. Lucie Unit 1 will revise the Thermal Margin Safety Limit Lines of TS Figure 2;1-1 to reflect an increase in the value of design minimum Reactor Coolant System (RCS) flow from 345,000 gpm to 365,000 gpm, and change the flow rates stated in Tables 2.2-1 and 3.2-1 accordingly. The reactor protective instrumentation Reactor Coolant Flow-Low trip setpoint limits will be changed from 93% to 95% of design RCS flow. In addition, TS 1.10 will be revised to update the reference for dose conversion factors used in Dose Equivalent Iodine-131 calculations from those listed in Table III of TID-14844 to those listed in ICRP-30; and administrative changes are proposed that will delete the numerical value of criticality analysis uncertainty described in fuel storage design features TS 5.6.1.a.1, update the analytical methods used for determining core operating limits listed in administrative controls TS 6.9.1.11, and revise Limiting Safety System Settings Bases 2.2 for the Steam Generator Pressure-Low trip to delete the numerical value of steam pressure described as the full load operating point.
Pursuant to 10CFR50.92, a determination may be made that a proposed license amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows:
(1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
Replacement of the St. Lucie Unit 1 steam generators in 1997 resulted in an increase in RCS flow. The proposed amendment would increase the values of design minimum reactor coolant flow and the,low flow trip setpoint presently stated in the Technical Specifications (TS). These revisions are accompanied by a corresponding change to the Thermal Margin Safety Limit Lines of TS Figure 2.1-1.
The RCS flow related revisions do not change the probability of any previously evaluated accident, as they do not impact any plant component, structure or system affecting the accident initiators. The proposed changes would continue to maintain adequate operational margin to TS limits for RCS flow and the low-flow trip setpoint.
The proposed changes to the thyroid dose conversion factors from TID-14844 to ICRP-30, fuel storage TS 5.6.1.a.1, the list of analytical methods in TS 6.9.1.11, and the Bases for Steam Generator Pressure-Low trip setting have no relevance to, the accident initiators, and thus do not affect the frequency of occurrence of previously analyzed transients. Additionally, there are no changes to any active plant component due to these proposed changes.
The supporting evaluation of proposed TS changes demonstrates acceptable results for all the accidents previously analyzed, and it is concluded that the radiological consequences would remain within their established acceptance criteria when including the effects of increased RCS flow, increased low flow trip setpoint, and change to the thyroid dose conversion factors used in the determination of
St. Lucio Unit 1 L-98-283 Docket No. 50-335 Attachment 2 Proposed' License Amendment Page 2 of 2 CS u owand SU date dose consequences. Proposed changes to the Bases for the Steam Generator Pressure-Low trip setpoint, fuel storage design features, and the list of analytical methods in TS 6.9.1.11 are administrative in nature and do not impact current safety analyses.
Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.
This proposed amendment revises limiting flow parameters to derive analysis benefits from increased RCS flow due to the replacement steam generators, while assuring safe plant operation commensurate with the proposed RCS flow and low flow trip setpoint changes. These changes along with the proposed changes to the Bases for the Steam Generator Pressure-Low trip setpoint, dose conversion factors, the list of analytical methods in TS 6.9.1.11, and the fuel storage design features do not require modifications to the plant configuration, systems or components which would create new failure modes.,
There would be no change in the modes of operation of the plant. The design functions of all the safety systems remain unchanged. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
The proposed amendment revises limiting flow parameters to derive analysis benefits from increased RCS flow due to the replacement steam generators, while assuring safe plant operation commensurate with the proposed design minimum RCS flow and low-flow trip setpoint changes. FPL has evaluated the impact of the proposed changes on available margin to the acceptance criteria for Specified Acceptable Fuel Design Limits (SAFDL), 10CFR50.46(b) requirements, primary and secondary over-pressurization, peak containment pressure, potential radioactive releases, and existing limiting conditions for operation. With the proposed changes to the design minimum RCS flow, low-flow trip setpoint, and dose conversion factors, FPL has concluded that there would be no adverse impact to the existing safety analyses. The proposed changes to the Bases for the Steam Generator Pressure-Low trip setpoint, the list of analytical methods in TS 6.9.1.11, and the fuel storage design features are administrative in nature. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
Based on the discussion presented above and on the supporting Evaluation of Proposed TS Changes, FPL has concluded that this proposed license amendment involves no significant hazards consideration.