ML20137L046
| ML20137L046 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/09/1995 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137K821 | List:
|
| References | |
| FOIA-96-485 JPN-PSL-SEFJ-95, JPN-PSL-SEFJ-95-039, JPN-PSL-SEFJ-95-39, NUDOCS 9704070144 | |
| Download: ML20137L046 (84) | |
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sf No Significant Hazards Evaluation Technical Specification Amendment to Reduce the Reactor Coolant System Pressure JPN-PSL-SEFJ-95-039 Rev1 October 9,1995 Safety Related Nuclear Fuel Nuclear Technical Support (bD 9704070144 9704o2 BS$DE 485 PDR
l, ucc-e7-9s Two s i svi rmace mcm7 ;one 4ev466999: p.e4 , JPF. PSL-SEFJ 95 039 + Rev. 1 Page 2 of 16 REVIEW AND APPROVAL RECORD PLANT - St. Lucie UNIT 1 TITLE Technical Soecification Amendment to Reduce the Reactor Coolant System Pressure LEAD DISCIPLINE Nuclear Fuel / PSL Fuel Enaineerina ENGINEERING ORGANIZATION Nuclear Technical Suooort REVIEW / APPROVAL: INTERFACE TYPE FPL GROUP PREPARED VERIFIED _ _ APPROVED APPROVED *
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JPN-PSL-SEFJ-95-039 i Rev 1 { l Page 3 of 16 ; ! 1 TABLE OF CONTENTS ' 1 l pESCRIPTION_ . .............. Pane' I
- 1. Introduction . . .......... .... 4
- 11. Description Of Proposed Change ........... ...4 I
111. Basis for Proposed Change . .................. 5 IV. No Significant Hazards Determination . k ...... ..... 13 ! V. List Of Affected Documents . . ...... . ...... 14 VI. Marked Up Technical Specification Pages . ..... .......... 15 Vil. References . .......... ..... 16 1 ATTACHMENTS l
- 1. Pressurizer Code Safety Valve Operability . .... ...... . 1 Page
- 2. Siemens Letter TMH:95:049, October 6,1995 . ......... .
9 Pages l 6 l
DEC-97-95 THU 11832 IMACE MCMT CORP 4074669991 P.96^ i ^ JPN-PSL-BEFJ-95-039 Rev 1 Page 4 of 16 " 1-
- 1. Introduction During the primary system heatup and pressurization, while returning to operation :
following the recent unplanned outage, the St. Lucie Unit 1 pressurizer system safety , valves experienced simmering and leakage at a reactor coolant system (RCS) pressure of approximately 2230 psia. The setpoint for these valves is 2500 psia +1% , and the current Technical Specifications (TS) minimum required F!CS operating pressure is t 7 2225 psia. To avoid problems of safety valves baking, operation through the end.of Cycle 13 at a RCS pressure of 2115 psia has usen evaluated and found to be acceptable. It has been determined that the operation at reduced RCS, pressure of 2115 psia does not involve a significant hazards consideration. Florida Power and , Light Company (FPL), therefore, proposes modification to St. Lucie Unit 1 TS 3.2.5 to i allow operation of St. Lucie Unit 1 at a reduced RCS pressure (2 2115 psia),110 psi below the current TS limit of 3 2225 psia, for the remainder of Cycle 13 operation. The FSAR analyses currently support a minimum nominal RCS pressure of 2235 psia for both normal operation and accident conditions. The nominal RCS pressure during . steady state operation as described in various FSAR sections is 2250 psia. However, l due to the temporary nature of the proposed change to the operating pressure, no changes will be made to the FSAR. i This evaluation is revised (Revision 1) to include additional information related to the .1
, pressurizer pressure program and the Chapter 15 events discussion. Additionally, Attachment 2 has been replaced with a revised Siemens letter providing detailed results g
of the re-analyses performed. The conclusions of the evaluation remain unchanged. il. Descriollon Of Proposed Channe l Technical Specification 3.2.5, Table 3.2-1, specifies limits on .RCS pressure, l , temperature, flow and axial shape index to assure that each of these parameters is maintained within the normal steady state operating envelope assumed in the safety analyses. These limits have been demonstrated to be adequate to maintain a minimum DNBR of 21.22 throughout each analyzed transient. I The proposed ' change reduces the TS 3.2.5 specified minimum pressurizer pressure I from 2225 psia to 2115 psia. The effect of this change on the DNBR margin and safety analysis has been evaluated and found to be acceptable. The marked up page of the Technical Specifications is in Section VI. A review of the Technical Specifications has concluded that no other TS is affected by the proposed change. Operating within the LCO limits for DNB parameters assures power operation within the TS reactor core safety limits. 4
, DEC-G7-05 TMU 11:33 INRCE NCNT CORP 4074669991 P.07 3
JPN-PSL-SEFJ-95-o39 Rev 1 Page 5 of 16 Ill. Basis for Proposed Channe The major issues considered in the evaluation of the proposed change are 1) review of safety analysis events, 2) impact on setpoint analysis, and 3) effect on fuel ! performance. Other issues addressed are radiological consequences, pressurizer pressure control, operability of safety valves, and primary and secondary operating l conditions. I Review of Safety Analysis Events _ All Safety Analysis Event Categories were evaluated to assess the potential impact of operation at reduced RCS prescure. The results of this review for each type of event are documented. The results of the review resulted in the following three classification categories: 1) No impact,2) Potential Impact, but Bounded by another transient, or 3) ' Impact requiring Evaluat.on. l For each Event Category, a discussion of the impact on the event due to the RCS pressure reduction is provided. A conclusion is stated for each event. A. Increase in Heat Removal by the Secondary System The FSAR events in this categor/ are evaluated for an increase in steam cooling due to the event initiator. A.1 increase in Steam Flow This event is currently bounded by the Loss of Flow event for DNBR. The effect of the pressure reduction on the relative DNBR would be similar for the two events.
Conclusion:
Bounded by Loss of Flow Event A.2 Inadvertent Opening of a Steam Generator This eventis affected by coolant Relief or Safety Valve activity and primary-to-secondary leak rate. There will be no impact due to the proposed RCS pressure ! reduction RCS because the initial l[L\ pressure l plays an insignificant role in this t analysis.
Conclusion:
No impact
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l l l 4 JPN-PSL-SEFJ-9s-039 Rev 1 Page 6 of 16 i A.3 Steam System Piping Failures inside and Consequences depend on the Outside of Containment cooldown rate. The proposed
- i lower initial pressure would I
have no significant effect. d !
Conclusion:
Sounded by Current Analysis B. Decrease in Heat Removal by the Secondary System The events in this category are anelyzed for the primary and secondary side overpressurization, for Departure from Nucleate Boiling (DNB), and for radiological l releases. l B.1 Loss of External Load (LOEL) The reduced RCS pressure will result in a larger rate of change of pressure when the l l high pressure trip is reached. This event is reanalyzed for the primary and secondary overpressurization. *;
Conclusion:
Impacted Event B.1 (Loss Of Extemal Load): This event is analyzed for both the primary and the secondary side overpressurization. The event is initiated by a loss of external load where the turbine stop valve is assumed to close rapidly. The reactor trip is on high pressurizer pressure. Operating at reduced pressure will result in primary-s.econdary energy imbalance for a longer time. Reanalysis of this event, performed at 2115 psia initial RCS pressure, showed that the peak primary pressure remains below the limit of 2750 psia, and th's secondary pressure remains below the limit of 1100 psia (Reference 1). 4 B.2 Loss of Condenser Vacuum (LOCV) The- effect of the RCS i pressure change would result in a faster pressurization for the LOCV. However, this event wiR remain bounded by B.1 because the impact on overpressurization will be l similar for the two events.
Conclusion:
Bounded by LOEL l B.3 Closure of Main Steam isolation Valve The impact for this event will be similar to that of LOEL. The Event will continue to
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azc-ov-9s THu 21:36 IMAGE MGMT CORP \ lf 4074669991 P.o1 j a JPN-PSL-SEFJ-95-039 Rev 1 Page 7 of 16 !I remain bounded by B.1.
Conclusion:
Bounded by LOEL B.4 t.oss of Normal AC Power This event is currently not I f(miting with respect to ONBR. The Effect of the reduced pressure on DNBR will be
%bf9hNNh*hN09 cimilar to that of the limiting Loss of Flow event.
Radiological consequences for this event are unaffected by a reduced initial RCS pressure because the TS leak rate assumed in the reference analysis remains u.nchanged. b
Conclusion:
DNB Event Bounded by Loss of Flow Radiological Consequences Not impacted B.5 Loss of Normal Feedwater This event is currently ., bounded by B.1 with respect to overpressurization, and will continue to remain bounded because of the similar effect on the two events. It will remain unaffected by a reduced RCS pressure relative to steam generator inventory.
Conclusion:
Pressurization Event Bounded by LOEL No impact on Steam Generator inventory B.6 Feedwater System Pipe Breaks initial RCS pressure has a similar impact as that of the bounding event A.3. b
Conclusion:
Potential Impact, Bounded by A.3 C. Decrease in Reactor Coolant System Flow Rate Events in this category are analyzed for fuel failure and adherence to DNBR limits. C.1 Loss of Forced Reactor Coolant Flow Reduced RCS pressure l ! impacts DNB margins. This event is re-evaluated as a
i DEC-OT-95 THU 11337 IMAGC MGMT CORP 4074669991 P.02
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part of DNBR margin in the setpoint analysis.
Conclusion:
Impacted - C.2 Reactor Coolant Pump Rotor Selzure/ RCS pressure will impact the 2 Shaft Break number of fuel rods calculated ' to fail during this event. This event is re-analyzed. with ; reduced RCS pressure for fuel failure. -
Conclusion:
Impacted Event C.2 (Reactor Coolant Pumo Rotor Seizurek This event involves pressurization I of the primary and secondary systems for which it is bounded by loss of external l load event. From DNBR considerations, this event is analyzed for fuel failure. A I reduction in initial RCS pressure will result in an increase in the number of fuel rod l failures. A reanalysis performed at reduced RCS pressure of 2115 psia showed that the radiological consequences due to fuel failure are bounded by the FSAR analysis ! (Reference 1), which assumes a fuel rod failure of 2.5%. l D. Reactivity and Power Distribution Anomalies Events in this category are analyzed for fuel failure and DNBR limits. D.1 Uncontrolled CEA Withdrawal from A reduced initial RCS pressure I Suberitical or Low Power Condition has no significant impact on this event. Therefore, this i event will continue to remain 1 bounded by D.2.
Conclusion:
Bounded by D.2 D.2 Uncontrolled CEA Withdrawal at Power Variable High Power (VHP) and Thermal Margin / Low Pressure (TM/LP) trips provide adequate protection. The proposed initial RCS pressure is within the range of current TM/LP verification analyses. The effect will be g similar to that of the bounding Loss of Flow event with respect to DNBR.
Conclusion:
Potential Impact, Bounded by Loss of Flow
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DEC-67-95 THU 11237 IMAGE MGMT CORP 4074669991 P.03 h JPN-PSL-SEFJ-95-o39 Rev 1 Page 9 of 16 D.3 CEA Misoperation A reduced initial RCS pressure willimpact the DNB results for the CEA Drop ! Event. The impact is evaluated in the setpoint analysis.
Conclusion:
Impacted D.4 CVCS Malfunctions A reduced initial RCS pressure will have no impact on these events, as it plays an insignificant role in these analyses. i
Conclusion:
No impact D.5 CEA Ejection Fuel failure criteria is based c the centerline melt and not on the DNB criteria. Therefore, reduced RCS pressure has no significant , impact.
Conclusion:
No impact a D.6 Inadvertent Loading and Operation of a This event is unaffected bythe 1 Fuel Assembly in an improper Location proposed RCS pressure reduction as pressure is not a consideration in this event.
Conclusion:
No impact E. Decrease in Reactor Coolant inventory important events in this category are: l E.1 inadvertent Opening of a Pressurizer TM/LP trip provides adequate Pressure Relief Valve protection. A Reduced RCS pressure will have insignificant impact on this event.
Conclusion:
No impact E.2 Steam Generator Tube Rupture (SGTR) A reduction in the initial RCS pressure will have a rninor ! beneficial impact on the l
m DEC-07-95 THU A1:41 ZMAGE MGMT CORP 4074669991 P.02 l JPN.PSL-SEFJ-95-039 i Rev 1 Page 10 of 16 1 I 1 radiological consequences of l this event, due to a slight decrease in the primary to secondary flow.
Conclusion:
Potential Impact, Bounded by FSAR Analysis
. E.3 Lcss of Coolant Accidents initial RCS pressure reduction ; . (SBLOCA and LBLOCA) will affect the initial mass '
inventory very slightly. The impact of this will be balanced by the beneficial effect due to h the reduced pressure.
Conclusion:
No impact
; F. Radioactive Releases from a System or Component Events in this category are unaffected by a reduction in the initial RCS pressure as Q l they are extemal to the RCS. ,
*i Imoset on Setooint Analyses !
Operation at 2115 psia is outside the current Technical Specification requirements and , therefore the Setpoint Analyses have to be evaluated. Protection against exceeding ; the Linear Heat Rate limit (21 kw/ft) is provided by the Local Power Density (LPD) ! Limiting Safety System Settings (LSSS) and LPD Limiting Conditions for Operation (LCO), which are dependent on the nuclear powei peaking and the core power level, but independent of the operating system pressure, and will therefore be unaffected. i The thermal margin low pressure (TM/LP) trip protects against the occurrence of DNB ' ! during steady state operation and for many, but not all (e.g. Loss of Flow and CEA Drop Events) Anticipated Operational Occurrences (AOos). A minimum proposed TS - pressure of 2115 psia is bounded by the current TM/LP verification analysis. which has considered a range of operating pressures from 1857 to 2400 psia (Reference 1). The reduction in RCS operating pressure (110 psi) will provide a reduced operating margin to the TM/LP trip setpoint. However, the TM/LP will continue to perform as required to ' provide DNB protection. The DNB LCO provides protection against DNB for transients involving changes in the : reactor coolant flow or changes in the power peaking which do not significantly change , I the axial shape index (ASI). The available margin toine current LCO is dependent on the rango of r nro nparating paramotors ovaluated. CEA Drop and Locc of Flow i ' transients are evaluated for the DNB LCO. The proposed operating pressure affects the allowed core power as a function of ASI which is used to verify the adequacy of the I e. l
_ ,_ _a . - - - _ _ _ DEC-E?-95 THU 11:47 IMAGE MGMT CORP 4074669991 P.11 JPN PSL-SEFJ-95-039 Rev 1 Page 11 of 16 1 current LCO for the Cycle 13 setpoint evaluation. The irapact of the reduced pressure on the power margin to the DNB LCO limits is calculated in Reference 1. It is determined that the DNB LCO margin will continue to be greater than 4% with the lower proposed TS pressure of 2115 psia. It is therefore concluded that sufficient margin is available in the setpoint analysis for the operation of St. Lucie Unit 1 with the proposed TS minimum pressure of 2115 psia. Fuel Performance Evaluation The effect of the proposed reduction in RCS pressure to 2115 psia has been evaluated in Reference 1, to verify that there is no adverse effect on the performance of St. Lucie Unit 1 fuel rods and assemblies in Cycle 13 core. The evaluation included the impact of the reduced pressure on the cladding transient stresses and strains, and the effect of the increased pressure differential between the fuel rod internal pressure and the RCS pressure on the rod internal pressure criteria. The evaluation has concluded that the maximum hoop stress will remain below the criterion of 56 ksi, and the maximum strain will be far less than the 1% criterion. The worst case int'ernal rod pressure has been determined to remain well below the design acceptance criterion. Additionally, Q all the other fuel rod design acceptance criteria related to the licensing basis and the _ performance of fuel assemblies will continue to be satisfied. Radioloalcal Consecuences The operation of St. Lucie Unit 1 at the reduced RCS pressure of 2115 psia will not create different types of effluents, and will not lead to a significant increase in any of the previously considered individual or cumulative occupational radiation exposures. Reanalyses performed with the reduced RCS pressure of 2115 psia, have shown that the RCS pressure will be maintained below 110% of design pressure for accident conditions, and the radiological consequences for all events will be bounded by the FSAR analyses. Pressurizer Pressure Procram The present instrument loops for control of pressurizer pressure have been reviewed with respect to the operation with the minimum TS pressure of 2115 psia, a reduction of 110 psi with respect to the current TS minimum pressure. The nominal operating pressure, if also reduced by 110 psi, would be 2140 psia. The necessary modifications ' to the pressure control system will involve changes to the low pressurizer pressure alarm, the backup heater control and the pressurizer spray and proportional heater controllers. The settings for the pressure control will be revised following the approval of the proposed TS amendment. I
D E c -o 7- 9 ts THu 12 42 Inacc Mcnr comp .iov466999: p.os l JPN-PSL-SEFJ-95-039 Rev 1 Page 12 of 16 Doerability of Safety valves The pressurizer safety valves are operable in their present condition and the unit can l be safely operated at the proposed reduced RCS pressure of 2115 psia. The operability of the valves is described in Attachment 1. l* RCS Temperature and Pressure l* l The reduction in minimum RCS pressure from 2225 psia to 2115 psia would reduce the l l core exit subcooling by about 7 F (from ~51"F to ~44 F), and would have minimal effect on the average RCS temperature and the RCS mass inventory (Reference 2). The d ; effect of the changes in thermal properties, due to the reduced pressure, on the core average temperature has been eshmated to be less than 0.2 F (decrease) and that on l, the mass inventory less than 0.25% (decrease). These small variations will have l; insignificant impact on any plant analyses involving these parameters. Furthermore, ! with the minimal effect on the primary temperatures, the secondary conditions would l remain practically unchanged.- I l Conclusion
, This evaluation has determined that the operation of St. Lucie Unit 1 at a reduced l
!7 pressure up to 2115 psia will satisfy all the safety analysis acceptance criteria, and will
- not require changes to any of the Reactor F>rotection System (RPS) and Engineered
! Safeguards Features Actuatior) System (ESFAS) setpoints as defined in the TS. Additionally, the proposed ieduction in the operating pressure will maintain sufficient margin above the TMILP trip floor pressure of 1887 psia and the safety injection I actuation pressure of 1600 psia, to ensure that there will be no spurious actuations. 1 ; 4
osC-or-9s Twu i : 4 ::: IMAGE NCM7 CORP 4074669991 ,_. A Od JPN-PSL-SEFJ-95-039 Rev.1 Page 13 of 16 IV. No Slanificant Hazards Determlnation Based on the standards in the Commission's regulation,10 CFR 50.92, a final determination that a proposed amendment involves no significant hazards consideration is made, 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 r.ccident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
- 1. Operation of the facility in ace'ordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
Operation at the proposed reduced RCS pressure does not affect the frequency of initiating events for any of the accidents previously evaluated. Thus, the probability of any accident previously analyzed ~is not increased. Evaluation of effects of the proposed change on departure from nucleate boiling ratio (DNBR) has shown that sufficient DNBR margin is maintained for accidents that challenge DNBR limits and all the acceptance criteria are satisfied. The Loss of Load, which is the limiting event for RCS overpressurization, was reanalyzed with the r' educed RCS pressure, and it has been shown that the pressure limit of 2750 psia is not exceeded. Additionally, seized rotor re-evaluation has shown that the existing FSAR limits on i fuel failure for this event are not impacted by the reduction in RCS pressure. I Therefore, there is no increase in the consequences of any accident previously evaluated, due to the proposed change.
- 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.
There are no new failure modes or mechanisms associated with the operation of St. Lucie Unit 1 at the proposed reduced RCS pressure. Furthermore, the proposed. change will provide increased margin between the Pressurizer Safety Valves Setpoints and the nomin61 operating pressure. Therefore, the proposed change will not create evaluated. previously the possibility of a new or different kind of accident from any accident i 1 l m_. . . . . - -
_ _ , .- . . _ , _ . ~ . -.-..- - -- -. DEC-07-95 7HU 11: 43 IMAGE MGMT CORP 4074669991 _. P 05 ;. JPN-PSL-SEFJ-95-039 Rev 1 Page 14 of 16
- 3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
Evaluation of DNBR analyses with the reduced RCS pressure of 2115 psia has shcwn that sufficient DNBR margin is maintained for accidents that challenge the 4 DNBR limits. Reanalyses of the limiting RCS overpressurization event with the reduced RCS pressure has shown that the acceptance critada will continue to be met. Seized rotor reanalysis has determined that the fuel failure limits and the ' consequential radiological releases will be bounded by the FSAR analysis. No safety limits will be violated for any other accident analysis due to the operation at 1 the proposed reduced RCS pressure, and the existing acceptance criteria will continue to be satisfied. Th'e proposed amendment, therefore, will not involve a reduction in the margin of safety. Based on the determination made above and the supporting documentation, it is concluded that the proposed amendment does not involve a significant hazards consideration. V. List Of Affected Documents The following document is affected by the proposed change: Technical Specifications (See Section VI)
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- Page 15 of 16 5
VI. Marked Un Technical Specifications Psae l TABLE 3.2-1 DNB MARGIN LIMITS 4 Four Reactor Coolant Ptsnps Pa rameter. Operatino Adj Cold Leg Temperature 1549'F Pressurizer Pressure 3.,2225 psia
- Reactor Coolant Flow Rate > 355,000 gpm l
AXIAL SHAPE INDEX Figure 3.2-4 Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.
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X _ __ _ Add ST. LUCIE - UNIT 1 3/4 2 14 Amendment No. - 1 l l . - _ - - - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _
DEC-O7-95 THU 11: 44' IMAGE MCMT CORP 4074669991 P.07 9 JPH-PSL-SEFJ-95-039 Rev 1 Page 16 of 16 Vll. References
- 1. Siemens Letter, T. M. Howe to E. J. ' Wunderlich (FPL), " Supplemental Documentation for Reduced Technical Specification Pressure at St. Lucie Unit 1", TMH:95:049, October 6,1995
- 2. FPL Calculation PSL-1FJF-95-226, Revision 0,"Effect of tne Reduction in RCS Pressure from 2250 psia to 2115 psia on Primary System Conditions" i
1 I 4 e a a 1 i e d
JPN PSL-SENP-94-026 Revision 0 Page 3 of 36 REVIEW AND APPROVAL RECORD PLANT St. L_pcie unzT 1 TzrLz Steam Generator Eauivalency Report (SGER) SGER Safety Evaluation JPN-PSL-SENP-94-026 Revision 0 tzAo ozsezPLrwe Nuclear / Licensing zwazwzzazuo eacAnzzATzon B & W Nuclear Technologies (BWNT) REVIEW / APPROVAL: IN1TRFACE TYPE GROUP PREPARED VERIFIED APPROVE 0 FPL APPROVED
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~2ee BWN1 Signoff See BWNT Signoff See SWNT Signoff Mj/gG ELECT X N/A IRC X N/A CIVIL X See BWNT Signoff See BWNT Signoff See BWNT Signoff ! 49
'NUC" X See BWNT Signoff See BWNT Signoff See BWWT Signoffg j" f ES! X // *' ,
N/A NUC FUEL X See BWNT Signoff See BWNT Signoff See BWNT Signoff h. dd id[7[f L
- For Contractor Evals As Determined By Projects
** Review Interf ace As A Min On All 10CFR50.59 Evals and PLA6
/0 ( 'I h i FPL PRoJzCTS APPROVAL: '
DATE: 14/4-efsr4 V orama zurzarActs Babcock and Wilcox International (BWI) _e i l ' Ab h& V" Form 24, Rev 6/94
FPL S:f;ty Evduation Ns. JPN PSL SEN? 94-026, l Rev. 0, page 4 cf 36 SGER, Rev. O, Paga 291 of 472 4.7 10CFR50.59 EVALUATION The Code of Federal Regulations (CFR) allows c),anges to the f acility described in the Safety Analysis Report (SAR) without prior NRC approval unless the proposed change involves a change in the technical specifications or an "unreviewed safety question". Section 4.6.1 of the SGER describes an evaluation that shows that use of the BWI replacement steam generators (RSGs) does not involve a change to the St. Lucie 1 Technical Specifications. The following paragraphs describe an evaluation that assesses whether any aspect of question as defined the use of the RSGs involves an unreviewed safety by 10CFR50.59. As defined in 10CFR50.59, an unreviev.ed safety question exists if (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report (SAR) is increased; (ii) if the possibility of an accident or malfunction of a different type than any previously evaluated in the SAR is created; or (iii) the margin of safety as defined in the basis for any technical specification is reduced. The three criteria identified in 10CFR50.59 have been factored into the seven questions addressed in Subsections 4.7.1 through 4.'i.7 below. Assessment of each question as it pertains to use of the RSGs demonstrates that use of the RSGs does not create any unreviewed safety question. 4.7.1 Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the Safety Analysis Report? This question is addressed in three steps: First, accidents previously evaluated in the SAR for which the probability of occurrence could be affected by use of the RSGs are identified. Second, the differences between the RSGs and the original steam generators (OSGs) that could affect the probability of occurrence of these accidents are identified. Third, the OSG and RSG differences are evaluated to determine if the probability of these accidents is increased by use of the RSGs. These steps are addressed below. Sten 1 - Identification of Accidents for Which the Probability of Occurrence is Potentially Affected by Use of the RSGs Most of the accidents evaluated in the SAR are initiated by failures or inadvertent actuations of equipment and systems that are not related to use of the RSGs. All accidents evaluated in the SAR are evaluated with the RSGs in place of the OSGs to identify the ones that could be affected by use of 4.7-1
1 FPL Saf;ty Ev:luation N2. JPN-PSL SENP 90 026, SGER, R2v. O, P:ge 292 of 472 Rev. O, page 5 cf 36 L the RSGs. This evaluation is described in Section 4.2 of the Steam Generator Equivalency Report (SGER). It identifies the steam system piping failure and steam generator tube rupture (SGTR) as the only accidents whose probability of occurrence is potentially affected by use of the RSGs. Section 4.2 of the SGER shows the probability of occurrence of all other accidents evaluated in the SAR to be unaffected by use of the RSGs. Sten 2 - Identification of RSG Differences That Could Affect Accident Probability
- a. The only RSG difference from the OSGs that could affect the probability of occurrence of a steam system piping failure is identified in Section 4.2 of the SGER as the steam moiscure content. This difference is evaluated in Step 3A below.
- b. RSG differences from the OSGs that could affect the probability of occurrence of an SGTR are identified in Section 2.2.2 of the SGER as the number of tubes, tube dimensions, tube material, tube bundle configuration, tube support system, and tube-to-tubesheet joint design. These differences are evaluated in Step 3B below.
Sten 3 - Evaluation of Differences
- a. Table 4.7-1 shows that the RSGs operate at the same secondary side pressure, temperature, and flow as the OSGs. The only dif ference that could af fect the probability of a steam system piping failure is shown in SGER Section 4.2 to be RSG steam moisture content. The RSG steam moisture content is lower than the OSG steam moisture content. This is shown in Section l 2.2.7.3 of the SGER. The same operating conditions and the
! lower steam moisture content do not increase the severity of l conditions in the steam piping. Therefore, use of the RSGs does not increase the probability of occurrence of'a steam system piping failure.
- b. There are 8523 tubes in the RSG and 8519 tubes in the OSG, a difference of less than 0.05 percent. Therefore, the number of RSG and OSG tubes is considered, equal for this evaluation, and the difference in the number of tubes does not increase I the probability of an SGTR.
The RSG tubes are nominally 0.003 inches thinner than the OSG tubes. The RSG tube material (Alloy 690) has a higher allowable stress than does the Alloy 600 tubing used in the OSGs. Section 2.8 of the SGER compares the RSG and OSG tubing dimensions and properties and shows that the RSG tube critical l rupture pressure exceeds that of the OSG tube. Because the l 4.7-2 l l l
.. .~. . . . -~ ~ . ~ . - - - . - . - , _ . . . . ~ _ _ - . . - . . . . . - . . . . - - - - . - . . ,
I FPL Saf:ty Ev:fustion No. JPN-PSL-SENP 94-026 Rev. O peGe 6 cf 36 SGER, R3v. O, P ge 293 of 472 l critical' rupture pressure of the RSG tubes exceeds that of the i-OSG tubes, the reduced RSG tube thickness does not increase .! the probability of an SGTR.
)
L
-The RSG _ tube material is more resistant to primary and secondary side. corrosion and cracking than is the OSG - tube j material. ~ Several factors . contribute. to this increased corrosion resistance. These include higher chromium content i
and higher grain boundary carbide decoration. Sections 2.3.2
~
and 2.5.1.1 of the SGER compare the OSG and, RSG tube l materials, and- show that the RSG tube material resists corrosion better than the OSG tube material. Because the RSG tube material resists primary and secondary side corrosion better than the OSG tube material, the differences in tube l material of an SGTR. corrosion resistance do not increase the probability The RSG and OSG tube bundle configurations differ in tube bundle shape and tube bend radius. The RSG upper tube bundle shape consists of tubes with continuous, smooth, long-radius 3 bends, whereas the OSG has horizontal straight runs. The U-bends of the innermost RSG tubes are skewed to provide a _ longer minimum bend radius than those present in the OSG design. The RSG tube bundle configuration is used to evaluate the structural and vibrational adequacy of the RSG tubes. ! These evaluations are discussed in SGER Sections 2.2.4, 2.6.1, and 2.8. Section 2.2.2 states that the RSG tubes are { structurally equivalent or superior to the OSG tubes. The same parameters are used to evaluate RSG vibration control as were .used in the ,. OSG.
- The tubes _ are _ also evaluated for ;
mechanical resonance and pressure variation, and shown.to meet 1 the same frequency, stress and strain criteria as' the OSG tubes. Therefore, the RSG tubes and' tube bundle configuration are structurally as capable of controlling harmful modes of i vibration as the OSG tubes and supports. Therefore, the RSG tube bundle configuration does not increase the probability of an SGTR. The RSG tube-s are supported by lattice grids in the straight tube runs and by flat U-bar restraints in the bend region. This differs from the OSG " egg crate" design. The RSG tube support system is described in Section 2.2.4 of the SGER. This system is used in the evaluation of RSG tube structural ' and vibrational adequacy discussed in SGER Sections 2.2.4, 2.6.1, and 2.8. . Section 2.2.2 states that the RSG tubes are structurally equivalent or superior to the OSG tubes. The same parameters are used to evaluate RSG vibration control as i were used in the OSG. The tubes are also evaluated for mechanical resonance and pressure variation, and shown to meet the-same frequency, stress and strain criteria as the OSG tubes. Therefore, ' the RSG tubes and tube bundle support , system are structurally as capable of controlling harmful 4.7-3
FPL S: tty Ev;luati:n N:. JPN-PSL SENP-94-026. SGER, R2v. O, Paga 294 of 472 Rev. O, page 7 (f 36 modes of vibration e.1 the OSG tubes and supports. Therefore, the RSG tube support system does not increase the probability of an SGTR. The RSG tubes are joined to the tubesheets by welding and expansion. The tube-to-tubesheet weld is described in SGER Section 2.2.3, and shcwn to be designed, analyzed, performed and examined in accordance with ASME Fection III criteria. This is the same criteria applied to the OSG tube-to-tubesheet weld. The RSG tubes are hydraulically e.xpanded through the full depth of the tubesheet, the OSG tubes are not. Elimination of the tubesheet crevice in the RSG protects the tubes from secondary side stress-corrosion cracking in this area. Full-depth expansion of the RSG tubes is described in SGER Section 2.2.3.2. Because the RSG tube-to-tubesheet weld meets the same ASME criteriu as the OSG design, and because the RSG tube-to-tubesheet joint design includes the benefits of full-depth expansion, the difference in OSG and RSG tubc-to-tubesheet joint desL Jn does not increase the probability of an SGTR. Because each difference in RSG and OSG design that could affect the probability of an SGTR has been identified and shown not to increase the probability of an SGTR, use of the RSGs does not increase the probability of an SGTR. The paragraphs above describe review of the accidents previously evaluated in the SAR and identification of the steam system piping failure and SGTR as the only accidents whose probability of occurrence could be affected by use of the RSGs. The RSG and OSG differences that could affect the probability of occurrence of these accidents are identified and evaluated to show that use of the RSGs does not increase the probability of occurr?nce of these events. Therefore, use of the RSGs does not increase the probability of occurrence of accidents previously evaluated in the SAR. 4.7.2 Does the proposed activity increase the consequences of an accident previously evaluated in the Safety Analysis Report? This question is addressed ift three steps: First, accidents previously evaluated in the SAR whose consequences could be affected by . use of the RSGs are identified. Second, the dif f erences between the RSGE, and OSGs that could affect the dose consequences of these accidents are identified. Third, the RSG and OSG differences are evaluated to determine if the consequences of these accidents are increased by use of the RSGs. These steps are addressed below. 4.7-4
I FPL 5:f:ty Evduation Na, JPN PSL-SENP 94-026, SG5R, R3v. O, Paga 295 of 472 l j - Rev, O, page 8 cf 36 l I l Step 1 - Identification of Accidents for Which the Consequences are Potentially Affected by Use of the RSGs The attached Table 4.7-2 summarizes accidents previously evaluated in the SAR, lists their critical parameters, and i compares the OSG and . RSG parameter values used in the
- evaluation of each event. Table 4.7-2 identifies the SAR events that have dose acceptance criteria and where differences between the RSGs and OSGs exist that could affect
. the dose as (a) the steam generator tube rupture (SGTR) , and
- (b) waste gas decay tank rupture. j
- Sten 2
~
Identification of RSG Differences That Could Affect Accident Dose Consequences l RSG differences from the OSGs that could affect SAR accident dose consequences are identified in Table 4.7-2. They are shown for the events identified in Step 1 when the RSG and OSG , values of important parameters differ. These differences are j tube inside diameter and primary system volume. The effects I of these differences on the SAR accidents identified in Step 1 are evaluated below. Steo 3 - Evaluation of Differences
- a. The total SGTR primary-to-secondary system leakage is greater ;
for the RSG than for the OSG because the nominal RSG tube ! inside diameter is 0.006 inches larger than the OSG tube inside diameter. Because the RSG flow area is two percent greater than the OSG value, the integrated leakage would be two percent greater with the RSG than with the OSG. This . means the of fsite doses with the RSGs would be two percent greater than the values calculated with the OSGs. Therefore, the whole-body dose of 0.103 rem reported in the St._Lucie 1 FSAR would be 0.105 rem with the RSGs. Similarly, the thyroid dose of 0.41 rem reported in the FSAR would be'O.42 rem with the RSGs. These values compare - with the NRC-approved SAR (10CFR100) acceptance criteria of 300 and ~ 25 REM, respectively. The potential doses with the RSGs remain a ; small fraction of the 10CFR100 limits. Therefore, the negligible increases in dose remain below the existing NRC-approved SAR acceptance criteria, and the consequences of the SGTR are not increased by use of the RSGs. The details of this evaluation are documented in Section 4.2 of the SGER.
- b. All of the radioactive gases stored in the waste gas decay tank _are assumed to be released in the waste gas decay tank >
I rupture accident. Whole-body and thyroid doses are calculated assuming that the contents of the waste gas decay tank are released instantaneously at ground level. The parameters that 4.7-5 I l
. ~ _ - - . .. . ._. . . .-
l FPL Srfety Evaluation Ns. JPN-PSL SENP 94-026, SGER. Rsv. O. Page 296 of 472 Rev. O, page 9 cf 36 could affect the dess for this event are RCS activity level . and RCS volume. RnS activity level is not affected by use of I the RSGs, but RCS vd ume is approximately one percent lc.rger with the RSGs than with the OSGs. This results in an increase of the calculated offsite thyroid and whole-body doses from 0.000878 and 0.128 REM.with the OSGs to 0.000887 and 0.129 REM with the RSGs. These values compare with the NRC-approved SAR l (10CFR100) acceptance criteria of 300 and 25 REM, ' respectively. Control room whole-body dose would increase by the same proportion from 0.000157 REM to 0.000159 REM. This compares with the NRC-approved SAR (10CFR100) acceptance criteria of 5 REM. Therefore, the negligible increases in dose remain below the existing NRC-approved SAR- acceptance i criteria, and the consequences of waste gas decay tank rupture I are not increased by use of the RSGs. The details of this ' evaluation are documented in Section 4.2 of the SGER. In the paragraphs above, the releases for SAR accidents that could be af fected by use of the RSGs are shown to meet the existing NRC-approved SAR acceptance criteria with the RSGs in place of the OSGs. Therefore, use of the RSGs does not increase the consequences of an accident previously evaluated in the SAR. 4.7.3 Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report? This question is addressed in three steps: First, the differences between the RSGs and OSGs that could af fect malfunction of the equipment assumed to function in the accident analysis are identified. Second, the equipment that could be affected by these RSG differences is identified. Third, the effect of the differences identified on the probability of a malfunction of the equipment identified is evaluated. For the purpose of this evaluation, meeting the existing NRC-approved SAR acceptance criteria is considered to demonstrate that the probability of a structural malfunction is not increased. These steps are addressed below. Sten 1 - Identification of RSG Differences That Could Affect Equipment Malfunction The probability of a malfunction of equipment and protection features that are assumed to function in the SAR accident analyses could be affected if the equipment or protection features ce required to operate at dif ferent conditions. The only RSG differences identified that potentially change the operating conditions of equipment or protection features 4.7-6
FPL S;f ty Evaluation N3. JPN-PSL SENP-95-026 SGER, R2v. O, Pagi 297 of 472 Rev. O, page 10 cf 36 assumed to function in the SAR accident analyses are RSG l weight and center of gravity. The RSGs are approximately ; l three percent heavier than the OSGs and their centers of , ! gravity at operating conditions are approximately two inches higher. ' l
\
l l Sten 2 - Identification of Equipment Affected by RSG Differences 1 ! As described in SGER Section 4.4, the RSGs are approximately I three percent heavier than the OSGs and the RSG centers of gravity are higher. Therefore, the loads imposed on the RSG supports and attached piping are slightly higher. Evaluation l of the increased RSG loads on supports and attached piping is l l discussed in Step 3 below.
)
1 i Steo 3 - Evaluation of Differences The effects of a three percent increase in weight and higher center of gravity for the RSGs are evaluated in Section 4.4 of the SGER. The evaluation shows that the existing supports and attached piping loads meet the NRC-approved SAR acceptance criteria ("specified-for-design" stress levels and loads) with the RSGs in place of the OSGs. An example of the evaluation I results is illustrated in the attached Table 4.7-3 which shows I the relationship between the OSG and RSG seismic loads on primary system components. These are compared with the l "specified-for-design loadsc" Because the stress levels and loads remain below the existing i NRC-approved SAR acceptance criteria with the RSGs in place of { the OSGs, the probability of a malfunction is considered to be l equivalent. Therefore, use of the RSGs does not increase the l probability of a malfunction of component supports or attached
)
piping. The above paragraphs identify increased RSG weight and higher j center of gravity as the only aspects of use of the RSGs that could af fect equipment important to safety. Steam gen'erator supports and attached piping are identified as the only equipment that could be affected. The evaluation shows that the increase in weight and higher center of gravity do not affect the probability of malfunction of this equipment. Therefore, use of the RSGs does not increase the probability of a malfunction of equipment important to safety previously analyzed in the SAR. l 4.7.4 Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report? , 4.7-7 1
l FPL Saf:ty Evaluation Ns. JPN-PSL-SFNP 94 026. SGER, Rsv. O, Paga 298 of 472 Rev. O. p;gs 11 cf 36 j This question is addressed in three steps: First, the differences ., between:the RSGs and OSGs that could affect the consequences of a ! malfunction of the equipment assumed to function in the SAR I
. accident analysis are identified. Second, the equipment affected is identified. Third, the effect of the differences on the 1 consequences is evaluated. These steps are presented below. ,
I i Sten 1 - Identification of RSG Differences That Could Affect Equipment Malfunction Consequences The RSGs could af fect the consequences of a malfunction of equipment by (a) affecting the sequence of events or thermal-
- hydraulic response of an accident, by (b) affecting assumed operator actions, or by (c) causing operation of systems important to safety outside their tperating limits.
j Dif ferences between the RSG and OSG that potentially increase
- the consequences of a malfunction of equipment assumed to j function in the SAR are addressed in turn below.
! a. As described in SGER Section 4.2, the RSG and OSG thermal-l hydraulic response to, and sequence of events for, accidents are equivalent. In addition, with the RSGs in place of the i OSGs, the equipment assumed to function in the SAR accident l' analysis functions within its operating limits. Because use , of the RSGs does.not change the sequence of events or thermal-t hydraulic response of any accident, no RSG differences'are identified that affect the consequences of a malfunction of j equipment assumed to function in the SAR.
- b. The operator actions assumed in response to the accidents
, analyzed in the SAR are documented in the plant operating procedures. Evaluation of plant - of f-normal and emergency operating procedures, described in SGER Section 4.6.3, shows that the actions they prescribe remain appropriate for the RSGs. Because use of the RSGs does not change the required operator actions, no RSG differences are identified that , af fect the consequences of a malfunction of equipment assumed l to function in the SAR.
- c. Review of the plant Technical Specifications with respect to the accident analyses presented in the SAR, described in SGER Section 4.6.1, showed that the operating limits used for the OSGs are appropriate and adequate for the RSGs. Because use l of the RSGs does not change the operating limits of systems :
important to safety, no RSG dif ferences are identified that l af fect the consequences of a malfunction of equipment assumed ! to function in the SAR. The paragraphs above show that use of the RSGs does not (a) af fect the sequence of events or thermal-hydraulic response of an accident, (b) af fect assumed operator actions, or (c) cause 1 4.7-8 l 4
. , - - - - - - - - . + . - . m-- --
FPL Saf:ty Evilustion Ns. JPN-P5L-SENP-94-026* SGER, Ray,0, P;gs 299 of 472 Rev. O, p:ge 12 ef 36 f i operation of systems important to safety outside, their operating limits. As such, there are no RSG differences that ' af fect the consequences of a malfunction of equipment assumed to function in the SAR. Sten 2 - Equipment Affected by Differences Since there are no' differences identified that affect the consequences of a malfunction of equipment assumed to function in the SAR, no equipment is affected. Sten 3 - Evaluation of Differences Because there are no RSG differences identified that affect the consequences of equipment assumed to function in the SAR, and because no affected equipment is identified, there are no differences to evaluate. Because there are no differences between the'RSGs and OSGs that could affect the consequences of a malfunction of equipment assumed to' function in the SAR, use of the RSGs does not increase the consequences of a malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. 4.7.5 Does the proposed activity create the possibility of an , accident of a different type than any previously l evaluated in the Safety Analysis Report? This question is addressed in two steps: First,_the differences ; between the RSGs and-OSGs that could have the potential to create l an accident'of a:different type than any previously evaluated in l the SAR are identified. Second, the differences identified in the i first step are evaluated for their potential to create an accident ; of a different type-than any previously evaluated in'the SAR. Sten l'- Identification of RSG Differences That Could Create a Different Type of Accident The RSGs connect to the same piping, instrumentation, and supports as the OSGs. This is described in SGER Section 2.1.3. Because the RSGs connect to and use the.same piping, instrumentation, and supports as the OSGs, no differences are created that could result in the possibility of an accident of a different type than any previously evaluated in the SAR. (There are two blowdown nozzles in the RSG design and one in the OSG design. Also, there are 12 water level taps in the RSG design and 9 in the OSG design. Because component nozzles 4.7-9
~ . _ - _ - ._- ..-- - -. .. . - . . - . ~ . _ - . _ _ . . ~ ~ - - . - - . . -
FPLRf:ty Evtluation Ns. JPN-PSL SENP-94-026, SGER, Rsv. O, Page 300 of 472 Rev. O page 13 cf 36 3 are not assumed to f ail, the additional nozzles cannot create the possibility of an accident of a different type than could occur with the OSGs.) sten 2 - Evaluation of Differences !
. Because use of the RSGs does not create dif ferences that could i result in the possibility of a different type of accident, there are no differences to evaluate.
l Because use of the RSGs does not create any differences that could i result in the possibility of an accident of a dif ferent type, their l use does not result in the possibility of an accident of a ' different type than types already evaluated in the SAR. l l l 4.7.6 Does the proposed activity create the possibility of a malfunction of equipment important to safety of a l different type than any previously evaluated in the SAR? ! l This question is addressed in two steps: First, the differences i between the RSGs and OSGs that could have the potential to create I , a malfunction of a different type than any previously evaluated in ! the SAR are identified. Second, the differences identified in the first step are evaluated for their potential to create a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. ,
)
1 Sten 1 - Identification of RSG Dif f erences That Could Create a Different Type of Malfunction l The RSGs connect to the same piping, instrumentation, and supports as the OSGs. This is described in SGER ..lection 2.1.3. Because the RSGs connect to and use the same piping, instrumentation, and supports as the OSGs, no dif ferences are created that.could result in the possibility of a malfunction of a different type than any previously evaluated in the SAR. (There are two blowdown nozzles in the RSG design and one in the OSG design. Also, there are 12 water level taps in the , RSG design and 9 in the OSG design. Because component nozzles l are not assumed to fail, the additional nozzles cannot create the possibility of a different type of equipment malfuncticu than could occur with the OSGs.) l Steo 2 - Evaluation of Differences I r l 4.7-10 l
. _ _ _ _ _ _ _ _ _ . _ _ _ . _. __- _._._. _ _____ _ _ _. -_-.__m . . - _ . _ __..___._. ...._.__ __
i 1 FPL S fsty Ev luation No. JPN PSL-SENP-9 *eO26 Rev. O, page 14 cf 36 SGER, Rsv. O, Pagi 301 cf 472 l Because use cf the RSGs does not create any differences that could result in the . possibility of a different type of malfunction, there are no differences to evaluate. i Because use of the RSGs does not create any differences that culd I result in the possibility of a malfunction of a different type, their use does not result in the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. 4.7.7 Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification? This question is addressed in four steps: First, a. review of the SAR accident analyses and . structural analyses identifies those events and components that could affect the margin of safety. (Reduction in the margin of safety is defined as the value of any acceptance criteria parameter exceeding the NRC-approved value that is reported in the SAR or modified by the NRC's SER. ) Second, j differences between the RSGs in the and firstOSGs thatidentified. could affect the events and components found step are Third, each i difference.is safety. evaluated to determine its effect on the margin'of Fourth, the Technical Specifications and their bases are reviewed to identify'any changes that are required by use of the RSGs, and to assess the effect of these changes on'the margin of safety. These steps are addressed below. ften 1 - Identification of SAR Accident and Structural Analyses' That Could Affect the Margin'of' Safety
, Evaluation of the SAR accidents with the RSGs in place of the OSGs is described in SGER Sections 4.2, 4.3, and 4.5.
Evaluation of the St. Lucie 1 structural analyses with the RSGs in place of the OSGs is described in SGER Section 4.4. The results of these evaluations identify the following events and structural components where the parametric results could exceed those for the OSGs:
- 1. Loss of feedwater.
- 2. . Waste gas decay tank ' rupture.
3. Low temperature over-pressure (RC pump start-up) .
- 4. LBLOCA containment response.
- 5. Steam generator tube rupture (SGTR) .
t
- 6. MSLB containment response.
- 7. RSG supports and attached piping loads and stresses.
i RSG 2. differences related to these items are discussed in Step i 4.7-11
FPL Sif:ty Evaluation Ns. JPN-PSL SENP 94-026, Rev. O, page 15 ef 36 $GER, Rev. O, Ptge 302 of 472 ] Sten 2 Identificat.fon of RSG Differences 1 The RSG differences that could affect Items 1 through 7 in
- Step 1 are shown in corresponding Items 1 through 7 below.
These items are identified in SGER Sections 4. 2, 4. 3, 4. 4, and 4.5. , 4 1. Less RSG secondary side. water inventory.
- 2. Greater RSG primary side liquid volume. !
3. Greater RSG heat transfer surface area.
- 4. Greater RSG primary side liquid volume.
5. 6. Greater RSG tube inside diameter. RSG integral flow restrictor. ;
*7 . !
Greater RSG weight and higher center of gravity. 1
- j These differences are evaluated in Step 3. I i
- Steo 3 -
Evaluation of Differences i The effect of the RSG differences from the OSG on each of the seven items identified in Step 1 are evaluated below for their
- effect on margin of safety.
1 1. Steam generator dry-out can occur earlier with the RSG because the RSG secondary side water inventory is less than that of the OSG. The evaluation of the loss-of-feedwater accident presented in SGER Section 4.2 shows that, for a loss of main feedwater with auxiliary feedwater 'available, a heat sink would be assured. Also, the NRC-approved SAR acceptance i criteria would safety with the RSGs. be met, and there would be no reduction of ! The evaluation also shows that, for a i total available, loss of feedwater from full power, with off-site. power 10 minutes of the Auxiliary Feedwater System. would be available for manual actuation i l in Therefore, the evaluation SGER Section 4.2 shows that i the NRC-approved FSAR !' acceptance criteria are met for the LOFW event with the RSGs, and use of the RSGs does not reduce the margin of safety for the Loss of Feedwater accident. i A St. Lucie FSAR' revisforiTis pl'a nned2that~willfaligithe"FSAR I acceptance icriteriAhwithJthe SRP ; Acceptance a identify)TReferencej3111as ;they analysisilof(!icriteriat recordj 'and
..The i conclusions. presented !ini thi~s . evaluation ofithe. St Ll Lucie' LOPW
- event
- J areOpredicat'ed %on 9 the(planned .;FSARf ; changes i ; being j implemented. RsThe rabovel: discussionsof rf thei St. #Lucieh LOFW evaluationi is ? toi beWerifledlwhent the?SGBRb isdreconciledf to the afinal have; ORSG Ldesig'nXand been;implementedF~^" ~ 5the?
^ " planned
' ' " ~ St.
^ ' 4Lucie/FSAR
~~ ~
{ changes
' " ^ ^
4 4 4
- 2. All of the radioactive gases stored in the waste gas decay tank are assumed to be released in the waste gas decay tank i
4.7-12 J
, , ..m,
FPL Safwty Evolustion Ns. JPN PSL SENP 94-026, l Rev. O, page 16 cf 36 SGER, Rsv. O, P:ge 303 of 472 l l rupture accident. Whole-body and thyroid doses are calculated assuming that the contents of the waste gas decay tank are released instantaneously at ground level. The parameters that-could affect the dose'for this event are RCS activity level and RCS volume. RCS activity' level is not affected by use of l the RSGs, but RCS volume is approximately one percent larger with the RSGs than with the OSGs. This results in an increase of the calculated offsite thyroid and whole-body doses from 0.000878 and 0.128 REM with the OSGs to 0.000887 and 0.129 REM with the RSGs. These values compare with the NRC-approved SAR l (10CFR100) acceptance criteria of 300 and 25 REM, ! respectively. Control room whole-body dose would increase by the same proportion from 0.000157 REM to 0.000159 REM. This i l compares with the NRC-approved SAR (10CFR100) acceptance criteria of 5 REM. Theretore, the negligible increases in , I dose remain below the existing NRC-approved SAR acceptance criteria .and use of the RSGs does not reduce the margin of safety for the waste gas decay tank rupture. The details of this event are presented in SGER Section 4.2. 3. One low temperature over-pressure event could be affected'by i use of the RSGs.because the RSGs have greater heat transfer surface area than do the OSGs. The' event is reactor coolant (RC) pump start-up with the steam generators hotter (50 F by definition) than the. RC system. This event causes RC temperature and pressure to increase because heat is i transferred from the steam generator to the RC system. heat transfer surface More temperature increase. creates the potential for faster RC i The evaluation presented in SGER i Section 4.5.4 shows that the RCS pressure remains below the maximum value specified in the SAR. Therefore, use of the RSGs does not reduce the margin of safety for the low l temperature over-pressure event. 4. The LBLOCA containment response is potentially affected by use of the RSGs because the RSGs increase the primary side water volume by approximately one percent. This increases the mass i available in the blowdown phase, and the resulting containment pressure. The evaluation presented in SGER Section 4.3.2.1 shows that although containment building pressure is increased f rom approximately 38.4 psig to approximately 38.7 psig by use of the RSGs, the peak containment pressure is below the NRC-approved SAR acceptance criterion of 44 psig. In addition, the LBLOCA and found containment' conditions with the RSGs were reviewed to meet the existing St. Lucie 1 equipment qualification envelopes. Therefore, use of.the RSGs does not reduce the margin of safety for the LBLOCA containment response. i l l S. The predicted total SGTR primary-to-secondary system leakage is greater for the RSG than for the OSG because the RSG tube l . inside diameter is approximately 0.006 inches larger than the 4.7-13
. _ _ . , . - - - ~ .-
FPL Saf;ty Evaluation N3. JPN-PSL-SENP 94-026, SGER, Riv. O, Page 304 of 472 Rev. O, page 17 cf 36 j l OSG tube inside diameter. Because the RSG flow area is two i percent greater than the OSG value, the integrated leakage would be two percent greater with the RSG than with the OSG. l This means the offsite doses with the RSGs would be two percent greater than the values calculated with the OSGs. , Therefore, the whole-body dose of 0.103 rem reported in the St. Lucie 1 FSAR would be 0.105 rem with the RSGs. Similarly, the thyroid dose of 0.41 rem reported in the FSAR would be 0.42 rem with the RSGs. These values compare with the NRC- l approved SAR (10CFR100) acceptance criteria of 300 and 25 REM, 1 respectively. The potential doses with the RSGs remain a small fraction of the 10CFR100 limits. Therefore, the ; negligible increases in dose remain below the existing NRC- l approved SAR acceptance criteria, and use of the RSGs does not 4 reduce the margin of safety for the SGTR. The details of this evaluation are documented in Section 4.2 of the SGER. l
- 6. The MSLB containment response is potentially affected by use of the RSGs because the RSGs contain an internal flow restrictor at the steam outlet. The flow restrictor limits the effective Main Steam Line Break area to 3.69 square feet compared with 5.355 square feet used in the containment analysis f or the OSGs. This reduces the calculated break flow and the calculated peak containment pressure for the MSLB with the RSGs. In addition, the MSLB containment conditions with the RSGs were reviewed and found to meet the existing St.
Lucie 1 equipment qualification envelopes. Therefore, the existing SAR analysis bounds the containment pressure response l with the RSGs and use of the RSGs does not reduce the margin l of safety. Details of the evaluation of the MSLB are l presented in SGER Section 4.2. l l
- 7. The RSG support loads and stresses could be potentially af fected by use of the RSGs because the RSGs are approximately three percent heavier and their centers of gravity are approximately two inches higher (at operating conditions) than those of the OSGs. The evaluation presented in SGER Section 4.4 shows that, although the increases in weights and centers of gravity increase the loads on certain supports and attached piping, the resulting loads and stresses are within the NRC-approved acceptance criteria in the SAR. Therefore, use of the RSGs does not reduce the margin of safety for the support and attached piping loads and stresses.
Evaluation of the SAR accident analyses showc that, while some RSG analysis results exceeded the corresponding OSG values, all existing NRC-approved SAR acceptance criteria are met with I the RSGs for every accident. Therefore, use of the RSGs does l not reduce the margins of safety in the SAR accident analyses. l The results of these evaluations are documented in SGER Sections 4.2, 4.3 and 4.5. In addition, increased support loads and stresses are shown in SGER Section 4.4 to remain 4.7-14 1
m - - __. . _ ~ _ _ _ m . - - .. ,_.m.__.~.. _ . . _ _ . _ . _ . _ _ . ._m._..- . _ . _ - _ _ _ . FPL Saf:ty Evaluation Na, JPN PSL-SENP 94-026, Rev. O, pegs 18 af 36 SGER, Rav. O, Pags 305 of 472 below the existing NRC-approved SAR acceptance criteria. l Therefore, use of the RSGs does not reduce the margins of safety in the SAR structural evaluations. l i l t Steo 4 - Evaluation of Technical Specifications and Bases Review of the St. Lucie 1 Technical Specifications and their. bases' showed that there are no changes required to the Technical Specifications or to their bases by use of the RSGs. Therefore, the existing Technical Specifications . and their
-bases remain applicable for the RSGs. The results of this review are ' documented in SGER Section 4.6.1 and in the Technical Specification Checklist -in SGER Appendix -A.
Therefore, the margin of safety in the Technical Specifications is not reduced by use of the RSGs. l Because-the NRC-approved acceptance criteria in the SAR accident and structural analyses are met with the RSGs in place of the OSGs, and because changes to the Technical Specifications are not required, use of the RSGs does not reduce the margin of safety as defined in the basis for any Technical Specification. 4.7.8 Conclusion' Sections 4.7.1 through 4.7.7 address the seven questions to make 1 the unreviewed safety question determination required by 10CFR50.59. Sections 4.7.1 through 4.7.7 demonstrate that there are no differences unreviewed safety question. between the RSGs and OSGs that create an This is accomplished by showing that use of the RSGs does not (i) increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report (SAR) ; (ii) create the possibility of an accident or malfunction of a different type than any previously evaluated in the SAR; or (iii) reduce the margin of safety as defined in the basis for the Technical Specifications. Because use of the RSGs does not require changes to the St. Lucie 1 Technical Specifications and does not create an unreviewed safety question, it is concluded that the RSGs can be used without prior NRC approval as allowed by 10CFR50.59. ' 4.7-15
- l. , ,. _. , _
I FPL S:f;ty Evtluation Ns. JPN-PSL SENP-94-026, SGER, Rev. O, Ptge 306 of 472 R;v. O, pag) 19 cf 36 4 TABLE 4.7-1 STEAM GENERATOR COMPARISON (Same as SGER Table 2.1-1) PARAMETER ST. LUCIE MILLSTONE ST. LUCIE UNIT 1 UNIT 2 UNIT 1 RSG DATA DATA OSG DATA 3 3 1646 f,3 Frimary side volume: no ttbes plugged 1698 ft 1693 ft Secondary side mass a 0 % futt power 213455 lb 216300 lb 223102 tb , a futt power 130959 lb 132700 lb 137970 lb Futt power steam flow 5.9E + 06 lb/hr same same i Pressure drop across divider plate (total) 370,000;000 lb/hr G74,000,000 lb/hr 370,000,000 lb/hr j 4 pwp flow (no plugs or sleeves) 31.7 psi 36.3 psi 32.3 pai
- 4 pwp flow (18% tubes plugged) 47.2 psi N/A 48.0 pri i
1 Primary side design pressure 2500 osia same same Secondary side design pressure 1000 psia 1015 psia 1000 psia Primary side design temperature 650'F same same Secondary side design toeperature 550*F same same y eary side operating pressure 2250 psia same same Steam outlet conditions: pressure 885 psia 880 psia 886 psia maxinun carryover 0.1 % 0.2% 0.25% (Guarantee) Feedwater temperature a full power 435'F same same (pre-stretch') Meat transfer rate a full power 4.63E + 09 Stu/hr 4.63E + 09 Stu/hr 4.62E + 09 stu/hr Flow restrictor flow area 2 2 530.9 in 530.9 in no credit taken Steam outlet norrte 1.0. 34 in same same Primary side heat transfer surface area: 82424.3 ft 2 82223.6 ft 2 78517 ft 2 no ttbes plugged (based on avg. I .D.) Secondary side heat transfer surface area: 93706.7 ft 2 93500.0 ft 2 90109 ft 2 no tubes plugged (based on avg. 0.0 ) htseer of tubes 8523 8523 8519 Tube 0.0.: nominal 0.75 in same same upper tolerance +0.000 in same same tower tolerance 0.005 in 0.0075 in 0.0075 in tube wall thickness: nominat 0.045 in 0.045 in 0.048 in tolerance +/-0.004 in +/-0.004 in */ 0.005 in TLbe material: 58-163 Alloy 690 code Attoy 690 Code Attoy 600 Case N*20-3 Case N 20 fTLbethermalconductivity: a 400'F 8.92 Stu/ft hr *F 8.92 stu/ft hr *F 10.1 stu/ft hr 'F G 500'F 9.54 8tu/ft hr 'F 9.54 8tu/ft hr 'F ' 10.6 Stu/ft-hr 'F G 600*F 10.167 Stu/ft hr- 10.167 8tu/ft-hr- 11.1 stu/ft br 'F
'F 'F 4.7-16 1
FPL Saf;ty Evaluation Ns. JPN-PSL SENP-94-026, SGER, Rav. O, Pege 307 of 472 Rev. O, page 20 cf 36 TABLE 4.7-1 (continued) STEAM GENERATOR COMPARISON ST. LUCIE MILLSTONE ST. LUCIE PARAMETER UNIT 1 UNIT 2 UNIT 1 RSG DATA RSG DATA OSG DATA T @ pitch 1.00 in 1.00 in 1.00 in
; (triangular) f Tube minimun strength (per ASME Code and code 40 Ksl 40 Ksl 35 Ksi Cn e N20): yield tensile 80 Ksi same same stcan Drun Mead Inside Radius 116.25" same same Thickness (min.) 3.25" same same steam Drun Shelt inside Radius 115.0" same same thickness (min.) 4.8125" same same Lower secondary Inside Radius 79.25" 79.25 78.125" cylinder thickness (min.) 3.875" 3.875" 4.3 75" Primary Need Inside Radius 75.0" same same Thickness (min.) 7.0" same same Dry Weight (Lbs) 1053500 1070400 1024317 CC (in) W.R.T Base 359.2 367.6 353.323 Nirmal Operating Weight (lbs) 1263000 1283000 1244708 CG (in) W.R.T. Base 356.1 365.3 354.497 FLonded (Pri. + Sec.) Weight (lbs) 1644000 1666600 1618175
, CG (in) W.R.T. Base 391.7 401.0 389.218 I
$hitt Side Manways (No. Ola.(in)) 2
- 18 2 - 16 2 - 16 Primery side Manways (No. Dia.(in)) 2 18 '2 18 2 - 16 Primary intet Nortles (No. - Dia.(in)) 1
- 42 same same Prima y Outlet Norties (No.
- Dia.(in)) 2 30 same same fsedwater Nortte Diameter (Nom.) (in) 18 same same Tubesheet thickness (Nom.) (in) 21.875 21.75 21.5 (Min.) (in) 21.5 21.5 Stay cylinder thickness (in) 6.313 min. 6.313 min. 6.313 nom.
$wport skirt Thickness (Nom.) (in) 2.75 same same (Min.) (in) 2.687 2.687 -
M ndholes (No. - Dia. (in)) 6-8 48 2 - 5.69 Inspection Ports (No. - Dia. (in)) 10 - 2 N/A N/A I Divider Plate thickness (Nom.) (in) 2.25 2.25 2.00 4.7 17
_ - . . - . -. . . _ _ . . . . . - . . . - . - - . . . . . - - . - . _ _ - - - - . . - - - ~ . . . - -
. . . . - - ~ -.
FPL S;f ty Evtluation No. JPN.PSL.SENP.94-026. SGER. Rev. O, Page 308 of 472 R:v,0, prge 21 cf 36 TABLE 4.7-1 (continued)
- STEAM GENERATOR COMPARISON
. ST. LUCIE MILLSTONE ST. LUCIE l PARAfETER UNIT 1 ta IT 2 talli 1 i RSC DATA RSG DATA OSG DATA I Ins t rument Taps (ho. - Dia. (in)) 4-1 4-1 4-1 l Recirculation Notrie Diameter (in) 3 (capped) 2 N/A Orun Tap (No. Ola. (in.)) 1-1 N/A 11 tottom Blowdown Nozzles (No. Dio. (in.)) 2-4 24 1-2 W ter Level Nozzles (No. - Dia. (In.)) 12 1 (2 capped) 12 1 91 , I j J t J i 0 4.7-18
_. ,m___, - .m- _ . _.._______._.~._.__.-...._.~__...______.m___. i FPL Scfaty Ev:luation No. JPN-PSL-SENP-94-026, SGER, R2v. O, Prge 309 of 472 Rev. O, page 22 cf 36 i 6 ' 1 EE 4 5 DE i MS a E! E E E E E i i I a .. m I i n 1 i r< l 1 N- 1 I J " g .II.I.I .II.I i D I.5I I
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, A I
71 T T A A 1 A C E 15 I I N O 1 7 R T S l NM1 T E A 1 l E G I T 1 5 E S Nl I S E R UM 1 t 1 fE 5 T C N1 AW RL C 15 NE 1 E 7 M 51 UAN1W I 1 E S V P D1 N MLA SF E N1 S E RA m O 5 F V t O1 i O R mOL 1 O E I S T C1 N I R E 0 PT SF E F E F IR F - N O U R f U A l t t A L T AA OW O O R S U SC S OL O L 4 U E GME 4 SP S OC U T OA L T CI S T El Sf MTS LA 2 3 5 0 4 1 5 1 5 1 5 1 5 1 5 - 1 1 1 1 1
Table 4.7-2 ACCIDENT ANALYSIS
SUMMARY
(Continued) y;
.r EVENT ACCEPTANCE BOUNDED NP0ftTANT ORIGNIAL BMD WVT 'o?, ,
FSAR Sf"10NI CRITER!A BT "' PARAMETIPS SG REF MW SG MF BT ACCtrl y T3AR CRtiffta em MDNBR i
- uo 15.?.1 p{ -
u j. RCS PRESS 15 2.1
' LDSS OF NORMAL
- FitDWATER FLOW [1515l SG PfttSt 1511 f
HEAT 3at 84117 L MfSC 3 00298 LBedfSS 32 NO ' C R TIS m DOSF 15.1.5 N
,,1, stEoWATiR sTSTEM m in BREAKS 115181 CENTERLUE FUEL 15.1.5 2
wtT y 13.3.1 1055 0F FORCED RC FLOW INITIAL RCS FLOW 355000 GPM" 14 I 50UR PUMP C04STDOW4) 1355000 GPM" 15 6 E00P MSISTANCil 0.0100 FT* " 15 0.8196 FT' " 15 (15151 MNesR llc WILIT TEMPERATURE p ', SAME Yts RC PUMP MRTIA SAME RPS LOW FLOW DELAT SAM RCS PftESS 1511 SG PfttSS 1S11 1512 BWR RECIRC MALFUIICTI0et TMIS tytNT 15 NOT APPUC ASLE TO ST. LUCIE 1 1513 MACTOR C0011MT PUMP NTIAL RCS FLOW 355005 GPM" 14 1355000 GPM" 15 ROTOR SE!ZUHf 115141 RC INLIT TEMPERATURE SAM LDCEO ROTOR MSIST. SAME TES RPS TftlP D(14T SAME e
._ O RCS Pft!SS 15.2.1 ,5 SG PMSS 15 1 i ?
P 2 i M R D u 4.7-21
Table 4.7-2 ACCIDENT ANALYSIS
SUMMARY
(Continued) :n ,i
*?
c DNO wii
'o T EVINT ACCEPTANCE BOUNDED IMPORTANT ORic!NAt ,,
FSAR SECTIONI CRiffRIA BT "' PARAMETERS SG { Mk 15 3.4
, THr3 [Vf 4715 NOT PART OF THE ST. LUCIE 1 LEINSING B ASts. o, Rf ACT RSERT RATE SAME AllAt POMR SHAPE SAME P MONBR tmCONTRottfD CEA IM L I OM A KINETES C0tFS SAME y 15 4.1 RCS FLOW 355000 GPM" 14 ->3%000 GPW3 15 2 .
SUBCRITIC AL OR LOW POWER i t ST ARTUP CONDITION 115111 RCS PMSS 1511 y SG PRESS 1511 !U 1542 tMCONTR0tttD CI A RE ACI MSERT RATE SAME in NTHORAWAL Al POMR AXIAL POWER SHAPE SAME D 115 1 11 MONBR KINETES COEFF'S SAME TES O u RC INLET TEMPERATURE SAME ? RCS FLOW 355000 GPM" 14 1355000 GPM" 15 RCS PRESS 1511 SG PRISS 1511 15.4.3 CEA MSOPERATION DRCCPED N00 RDRTH SAME 10ROPPED R001115.2.31 tifCTICS C0FTS SAME RADIAL POWER PEAK SAME RCS FLOW 35000 GPM" 14 13%800 GPM" 15 RCS PRESS 1511 SG PRESS 1511 i '
~
C ANT ST. LUCE-1 IS NOT LEENSED FOR PARiiAt PUMP OPERATIOft. THEREFCRE. THIS EVENT IS NOT APPLEA8tE TO ST. LUCIE-1 m 15 4.5 A AM THIS IVENT Q NOT APPLE A8tt TO ST. LUCIE-1 - P e
':?
S u k D u 4.7-22
Table 4.7-2 ACCIDENT ANALYSIS SUIPIARY (Continued) ,.,,
- .i ?,
pp IVINT FSAR SECil0N! ACCEPTANCE CRITERIA 80tMDED 8T "' IMPORTANT PARAMffERS ORIGMAL SG ESAR I CRffERIA
*f I <m
"" I"" "I Mk CVCS Mit!LMCil0NS THAT WIN SHUTDOWN MARGilt SAME o5 1148 REstti a t. DECRI ASE DI RCS PRISS DILUTION FLOW RAit SAME Y[S u g.
THE RCS BORON 3 3 2 CONCINTR AT10N 115 2'41 1 0 SG PRESS RCSVOLUME 1948 iT' " 4 1998 FT' " l e IIS t 8- t plADMRTENT (CADMG AND 3 114.7 Tlt$ ALCSENT IS NOT AFFEC'id Of SG REPLACEMINT.
, g, tDCAT!ON {15.3 38 h
1548 $7tCTRtw 0F Cf A ERCT104 DOSE EJECTED R00 WDRTH SAME Tis 1 ACC10tNTS 115 4.51 PEAT TUEL RWETICS COEfF1 gailAL FUEL ENTHALPT SAME y i YES INTHALPT SAME 6 M l g TMS (WNT IS NOT APPLEA8tf TC ST. LUCE 1 15.1 1 INADVERitNT OPERATPDff 0F (CCS DURWG POWER TRS IWNT IS IIGI PAR 7 0F fit ST. LUCE-1 LEENSWG Basis. OftRATION 1512 CHEMICAL AND VOLUME CONTROL SYSTIM 1985 fMNT IS NOT PART 08 THE ST. LUCE-1 ilCENSNIG BASIS , MAlfullCil0N THAT INCRI ASES RCS INVENTORY 115.3 BWR TRANS!TNTS THAT TIES IMNT IS NOT APPLC ABLE TO ST. lUCIE-1 MCRf ASE RCS RWENTORT 0 ' 9 6 s . 9 m b. u O 1 u 4.7-23
F t . l !! ' l7 f i; , 'i I t ! ! 't f !' , , . , h , at - # p f,(e. 2ytk_ 6 6*'O - ? O. ;" ? . y k & M mom? is . y$O( w ,= $ T . 3 WI t Y D S I 3 S S E E t O I Y M Y N N t Y I I t
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' W 1 1 O 4 O RT U 15 F 1 P C S A RR N T F O
, 8
, I 2 3 4 5
, 1 1 3 7 5
1 5 1 U 1 5 1 5 1
Table 4.7-2 ACCIDENT ANALYSIS
SUMMARY
(Continued) ,i g?
,, IVENT ACCEPTANCE ODUNDED - IMPORTANT 0 ORIGINAL IFSAR SfCil0NI CRITIRtA B T "' PARAMETERS SG TSAR CRiitRIA e<
em OVERPRESSt'RE PROTECfl0N Me 5M RCS PRISS 1511 15 2 2 APPtN0ft 5Al - c-LOW if MPER ATURE N DVERPRESSURE fliOP) Z PR0ftCil0N IAPPEN0ft 581 f
; y A.1053 Of RHR RCS PRISS MC AT Hf AI
- z
$8ME ifs i B. HPSt ACTUAil0N RCS PRESS HPSI FLOW SAME TES
?
6 C. P M HEATER RCS PRESS PZR HE ATER POW [R SAME ilt SG M.T. CAPACITT 132.1 1 1335 1 MBTURtR F M8 TUMR f C
- 0. RCP SIARTUP RCS PR(SS RCS FLOW R ATE 5=
SAMf NO its PORV StiP0 TNT SAME PORV arf A SAME FORV OPENNIG TIME SAMF NATUR Al CIRCUt ATION SG M.T. CAPActfi 132.1 1 133.8 1 COOLDOWN lAPPEN011 $C) RCS FLOW TES LOOP RESTSTANCE 0.019e Fi* " 15 0.0100 f f* " 15 OS.3 CONTAINMENT DESIGN fVALUAil0N 18.2.1.3! A. IBLOCA NT!Al RCS MASS 4.511 5 LBM g ppggg e 4fA 5 LBM 29 INITIAL .T ENERGi 2.5eE.8 STU No its 29 2.59E.8 BTU 29 O" E "N8 :n SG MASS stVENIORY '
"I *II 137970tRM 9 130959 19M D BRf AK ARE A ' '
5.355 FT' 30 3 89 FT' E i P 2 S on S ! O u 4.7-26
Table 4.7-2 ACCIDENT ANALYSIS SUNNARY (Continued) NOTES:
,'.o E.
- 1. Standard Review Plan section number. Im
$b
- 2. Minimum flow allowed by revised Technical Specifications. gk
- 3. Value for up to 18 percent plugged tubes. US g
- 4. Best-estimate loop resistance based on total core flow with no tube plugging. h
+
- 5. SG primary side liquid volume. !G l0
- 6. SG primary side flow resistance from tube sheet to tube sheet. $
?
6 9 2'
,0 D
Ld u k D m 4.7-27
i fable 4.7-3, Seismic Loads on Primary Syoum Component (s (came ma SGER Table 4.4-3) l Seismic Lead Esistin8 Reptocement
~
Seismic Component e wi Design g t Existing Calcuteted Specified for 8'I88 Percent 8etou {3
<8 Excitation Location Reptocement Deslen 8@
Manteam Calcuteted Nomiensa Desi8n Canbined Reactor vesset Fx (kips) 60.4 62.2 241.0 75 74 g ; X and Y Outlet Wortte , and Fy (kips) 15.2 15.7 253.0 94 94 ,{ Reactor vesset 10.0 90 90 Fr (kips) 1.0 1.0 Outlet Piping Mx (in-kips) 15.8 16.3 1042.0 98 98 g3 l Z 98.2 1041.0 91 91 o , ! My (in-kips) 95.3 ' c_ Ma (in-kips) 1267.5 1,305.5 43733.0 97 97 Reector vesset intet Fx (kips) 50.8 52.3 190.0 73 72 3 Nor2te Fy (kips) 17.9 18.4 146.0 88 87 g z ! Fr (kips) 47.4 48.8 57.0 17 14 ? , I I' Mx (In-kips) 2262.8 2,330.7 9638.0 TT 76 6 u My (in-kips) 1517.0 1,562.5 5981.0 75 74 ? . , Mr (in-kips) 1876.8 - 1,933.1 12770.0 85 85 16.9 17.4 30.0 44 42 Canbined Reactor vesset Fu (kips) 2 and Y outlet Nortte 99 99 and Fy (ktps) 2.4 2.5 172.0 Reettor vesset 40.0 84 84 f Outlet Piping Fr (kips) J.3 6.6 Mn (In-kips) .50.0 154.5 1170.0 87 37 My (In-kips) 800.0 824.0- 7521.0 89 89 , Mr (in-kips) 244.= 251.7 37646.0 99 99 l m peector vesset Inlet Fx (kips) 29.7 30.6 61.0 51 50 o Notste Fy (kips) 12.4 12.8 105.0 88 88 - 2 s 30.7 108.0 72 72 $ I Fr (kips) 29.8 Mx (in-kips) 1874.8 1,931.0 22392.0 92 91 Ny (in-kips) 731.1 753.0 10085.0 93 93 y f Mr (in-kips) 1548.9 1,595.4 14087.0 89 89 3 [ m i Note: The percent below design was calculated by o ! 100*(Specified for Design - Calculated)/Specified for Design. 2 l i' d l I 4.7-28 _ _ _ _ _ _ _ - - - - - - _ _ - _ - - _ _ _ _ _ - _- - - - - - - - . _ _ _ _ ._ - . _ __ _ - _ _ _ = _ . , _ , - - _ ____ _ - - - __ - _ ___- - ---_.-___ _ _-_- -___!
t Table 4.7-3, Seismic Loads on Prl...ary System Compontnte (cont'd) l Seismic Load Existing Seismic Conponent and Design g Component cet Percent getow Escitation Location Existing Calculated Reptscenent Specified for Below 3 "n Meninus Design Design $@ Calculated Maalaus Design 9 ,e Canbined Steam Generator Fa (kips) 60.9 62.7 149.0 N and Y 59 58 m* Intet Nortte end Fy (kips) 52.7 54.3 68.0 23 20 j <7 Steam Generator Intet Piping Fr (kips) 1.0 1.0 4.0 75 75
$ t.
o e$ Mn (in-kics) 73.8 76.0 ' 743.0 90 90 g3 My (in-kips) 85.8 88.4 708.0 88 88 ,k Mr (in-kips) 695.8 716.7 16828.0 96 % Steam Generator Fu (kips) 26.4 27.2 30.0 12 9 h Outlet Nozzle C Fy (kips) 31.7 32.7 134.0 76 76 g 26.4 z Fr (kips) 27.2 33.0 20 18 'P to Mx (in-ktps) 3018.3 3,108.8 12955.0 77 76 [ My (in-tips) 2708.2 2,789.4 3161.0 14 12 h Mz (in-kips) 3018.3 3,108.8 6686.0 55 54 Con 6ined Steam Generator Fn (kips) 17.0 17.5 27.0 37 35 Z and Y Intet Nortte and Fy (kips) 15.2 15.7 149.0 90 89 Steam Generator Inlet Piping F2 (kips) 6.4 6.6 7.0 9 6 Ma (in-kips) 356.5 367.2 5 75.0 38 36 , My (in-kips) 381.6 393.0 3793.0 90 90 Mz (in-kips) 63.4 65.3 11206.0 99 99 Steam Generator Outtet Na21te Fu (kips) 27.8 28.6 30.0 7 5 $ Fy (kips) 22.0 22.7 V3.0 76 76
-Q -
2 Fr (kips) 27.8 28.6 119.0 77 _ 76 [ i Mu (in-kips) 1866.0 1,922.0 6467.0 71 70 E m' My (in-kips) 1586.4 1,634.0 4251.0 63 62 g Mr (in-kips) 1866.0 1,922.0 7140.0 74 73 3 Note: The percent below design was calculatect by 7 100* (Specified for ' Design - Calculated) /Specified for Design. 2 U t 4.7-29
Table 4.7-3, Seismic Loads on Primary System Componento (cont'd) seismic Load Existine et t ou N Seismic Component and Design Below Excitation Location Component Existing Calcuteted Reptocement Specified for Dnip Design Desi p Maniaan Calculated Maniman -o ,e 10.7 11.0 '78.2 86 86 $ Canbined Reactor Cootent Pupp Fx (tips) w.7 N and Y Intet Nortte 31.7 32.7 137.0 77 76 Fy (kips) Fr (kips) 4.5 4.6 37.9 88 88 ," e(. Mu (in-kips) 1181.1 1,216.5 5713.2 79 79 g3 2 4976.0 86 85 o My (in-kips) 707.6 728.8 3,457.8 16461.0 80 79 Mr (in-kips) 3357.1 Fu (kips) 51.0 52.5 199.3 -74 74 3 8" Reactor Cootent Ptap Outlet Nottle 17.2 17.7 145.9 88 88 g Fy (kips) l z 38.9 40.0- 6 3 ? Fr (kips) 37.8 I Ma (in-kips) 1347.2 1,387.6 3999.0 66 65 6 2,245.7 8583.9 75 74 . My (in-kips) 2180.3 2,369.8 16658.0 86 86 Mr (In-kips) 2300.8 15.8 39.1 61 60 Combined Reactor Cootent Ps.sp Fu (kips) 15.3 _ Z and Y intet Nortte 21.9 22.6 94.9 77 76 Fy (tips) 11.9 109.9 89 89 Fr (kips) 11.6 966.8 11173.0 92 91 Mu (in-kips) 938.6 791.0 11862.0 94 93 My (in-kips) 768.0 2,524.8 8334.1 71 70 Mz (in-kips) 2451.3 - 29.8 30.7 '125.1 76 75 $ Reactor Cootent Pump Fa (kips) Outlet Norate 86 86 -Q 12.4 12.8 91.4 Fy (kips) 2 20.3 94.9 79 79 f Fr (kips) 19.7 864.4 17532.0 95 95 Mu (in-kips) 839.2 2 1,062.0 1100.0 6 3 ?, My (in-kips) 1031.1 1,299.9 6391.0 80 80 U Mr (in-kips) 1262.0 Note: The percent below design was calculatect by ~ 100* (Specified for Design - Calculated) /Specified for Design. n 4.7-30
k Table 4.7-3, Saicmic Loads on Pritnary System Componunt 9 (cont'd) , seisele Load Existing seismic Component and Design N"' Excitation Conponent m Location Entsting Calculated Replacement specified for Below Pe Maximaan Calculated Maximas. Design Design DW @ Canbined oe steem Generator M (in kips) 3323.6 3,&23.3 X and Y Duttet Piping 12000.0 72 71 bi
~
1 <m steam Generatt,r M (in-kips) 2235.1 2,302.2 Outlet Piping 12000.0 81 81 y $. r
- a. c5.
Ptsp Intet M (in-kips) 1794.0 1,847.8 5iping 12000.0 85 85 $3 2 P Ptsg Intet M (in-kips) 3628.1 3, 736.9 12000.0 70 69 c Piping
]
Ptop outlet Piping M (in-kips) 2955.1 3,043.8 12000.0 75 75 h m R.V. Intet M (in-klps) 2636.1 2,715.2 12000.0 78 77 2 Piping Conbined steam Generator M (in-kips) 1974.1 2,033.3 12000.0 14 83 6 2 and Y Outtet Piping steem Generator M (In-kips) 1537.8 1,583.9 12000.0 87 87 Outlet Piping j Ptsp Intet M (In-kips) 1193.5 1,229.3 12000.0 90 90 Piping _ Ptsp Intet M (in-kips) 2627.3 2,706.1 12000.0 78 77 Piping Ptsp outlet M (In-kips) 1318.2 1,357.7 12000.0 89 89 Piping R.V. Intet M (in-kips) 1952.8 2,011.4 12000.0 84 83 Piping Note: The percent below design was calculatecl by h 100*(Specified for Design - Calculated)/Specified for Design. ? ; if s P al 2 4 M 4.7-31 ,
I
?
i Table-4.7-3, Seismic Loads on Pru.ary System Componente-(cont'd) , Selsmic Load. Entsting I Seismic Component and Design * *"I m Component Encitetton Lorzion Existing Calcuteted Replacement Specified for- Bel *" Design Design Dnie @ Maniana Calcuteted itenlaum CE Combined React.,r vesset n (kips) 5.9 6.1 22.0 73 72 e t X and Y Outtet Support 4 : 214.7 221.1 473.0 55 53 V (kips)
=7 ),
Reactor Vesset M (kips) 261.3 269.1 1052.0 75 74 '
- Intet S y rt ((e.
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~
V (kips) _122.7 126.4 469.0 74 73 Steam Generator Fy (kips) 187.5 193.1 624.0 70 69 f f tower Support c Fr (kips) 44.8 46.1 54.0 17 15 g Ma (in-kips) 0.1 0.1 21.0 100 100 I .!
?
My (in-kips) 156.5 161.2 455.0 66 65 'g~ > 2 - ! Ma (in-kips) 2700.2 2,781.2- 24383.0 89 89 ? ; i o - steen Generator Fa (kips) 123.4 133.3 140.0 12 5
$ 'i Upper Stgiport w a
'~
Pressuriser, Fm (kipe) 22.2 22.9 82.5 73 72 Support Fy (kips) 24.6 25.3 80.7 70 69 i Mr (in-kips) 5681.8 5,852.3 17207.4 67 66 ! t Reactor Cootent Ptsip Fy (kips) 1.1 1.1 4.6 76 76 vertient Suport i Reactor Cootent Pump Fe (kips) 5.6 5.8 25.0 78 77 , Norizontet Steport ; Note: The percent below design was calculatect by . 100*(Specified for Design - Calculated)/Specified for Design. m :
- -?
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5 i o t t 9 1 . M.
=
Z 0. m 4.7-32 _ . . _ _ _ _ _ _ . - _ . _.m___._._______m.________._..___._____=._m_.__m _- _ _ _ _ _ _ _ _ _ _ _-- ______$ _ _ _ _ , , _ _ _ _ _ . - , . ..w..- .
Table 4.7-3, Seismic Loads on Primary System Components (cont'd) Seismic Load Existint 2* - Seismic Congonent and Design Percent gg,p,g,n, SetoW Pe t Excitation Location Emlettra Calculated Replacement Specified for Desir p#* ' Maxinum Calculated Maniaun Design Design , L Combined Reactor Vessel N (kips) . 257.5 265.2 663.0 61 60 $ i 2 ard V Outlet support 38.8 392.0 90 90 u< m-* V (kips) 37.7 oe 134.4 138.4 304.0 56 54 Reactor Vesset Intet Support n (kips)
]a {-
V (kips) 177.3 182.6 692.0 74 74 172.9 178.1 405.0 57 56 t- I Steam Generator Fy (kips) Lower Support 2 Fr (kips) 81.6 84.0 397.0 79 79 -i, m Mu (in-kips) 5019.2 5,169.8 24422.0 79 79 [ , My (in-kips) 722.4 744.1 9772.0 93 92 Mr (in-kips) 0.3 0.3 4132.0 100 100 0 , steam Generator Fr (kips) 64.6 69.8 240.0 73 71 g Upper Support , Pressuriter Fx (kips) 24.6 25.3 80.6 69 69 Steport Fy (kips) 22.2 22.9 82.9 73 72 i Hu (in-kips) 5681.8 5,852.3 17101.5 67 66
- Reactor Coolant Ptsp Fy (kips) 0.7 0.7 9.2 92 92 ,
vertical styport Reacter Coolant Pe p Fe (kips) 18.9 19.5 25.0 24 22
's Morizontal support Note: The percent below design was calculatet by ,
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JPN-PSL tedP 84028 Revision 0 ' Page 3 et 38 ( REVIEN AND APPROVAL RECCItD i i l St. Lucia uwxr 1 , rust i rars.s Steam Generator Ech Recort (SGERI SGER Safety EMantion JPN-PSL-SENP-94-026. Revision O l t
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- 3 4 4.7 10CFR50.59 RVALukTION j the Code of Federal Regulations (CFR) allows Changes to the j facility described in the Safety Analysis Report (SAR) without
- prior MRC approval unless the proposed change involves a in j the technical specifications or an *unreviewed safety quest on'.
j Section 4.6.1 of the SGER describes an evaluation that shows that j use of the BWI replacement steam generators (RSGs) does not involve i a change to the St. Lucie 1 Technical Specifications. The J following paragraphs describe an evaluation that assesses whether j any aspect of the use of the RSGs involves an unreviewed enfety question as defined by 10CFR50.59. i As defined in 10CFR50.59, an unreviewed safety question exists if , . (i) the probability of occurrence or the consequences of am j l accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report (SAR) is increased; (ii) if the possibility of an accident or malfunctice of a different type than any previously evaluated in the SAR is created; or (iii) the margin of safety as defined in the basis for any technical specification is reduced.
. The three criteria identified in 10CFR50.59 have been factored into j the seven questions addressed in Subsections 4.7.1 through 4.7.7 l
below. Assessment of each question as it pertains to use of the i RSGs demonstrates that use of the, RSGs does not create Gay < unreviewed safety question. 4.7.1 Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the
- Satety Analysis Report?
..e . . . .
m This question is addressed f.a ' threerstapoph First#ac,cidentaip#: ! . I i :u-&;<: kpreviously evaluatedrin the;N occurrence could be affectedlby _ W
.s , # Second;%the; differences M M lM9 Wigeneratorsq0eGeb that could' h to determine 3 F
, 1' - these4 .tage. + j N '~ ~?addressed accidentsbelow. le increased ' by use of' the RSQa@W.Pf
.:. ,?
i j- S. tap,,.1 - Identification of Accidents for which the Probabili og Occurrence is Potentially Affected by Use of the y .m 2 4 -, m y; ,; n - Am- ' '-
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' that : are not related to use of"theitSGoa*/411
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J i evaluated in the SAR are evaluated with the RSGs is place og j the 00Gs to identuy the ones that could be affected by use og
! 4.7-1 a
} 1 i 9054 nov.O,Puen att of 472 FPL Seesty Ovehanten Ms. JPNML450P44438 . } nov.e, pass a se as , l l the EsGs. 'Ittie evaluation is described in Section 4.2 of the j steam Generator Equivalency Report (SGER) . It identifies the steam system piping failure and steam generator tube rupture (SGTR) as the only accidente whose probability of occurrence 1 is potentially affected by use of the R9Gs. Section 4.2 of i the SGER shows the probability of occurrence of all other i accidents evaluaged in the SAR to be unaffected by use of the j RSGs. Step 2 - Identification of RSG Differences ht Could Affect Accident Probability h a. The only RSG difference from the OSGs that could affect the probability of occurrence of a steam systesa piping failure is {c identified in Section 4.2 of the SGER as the steam moisture content. This difference is evaluated in Step 3A below. { 1 from the OSGs that could affect the l
- b. RSG differances probability of occurrence of an SGTR are identified in Section 2.2.2 of the SGER as the m=+ar of tubes, tube dimensions, l
tube material, tube bundle configuration, tube support system, and tube-to-tubesheet joint design. These differences are j evaluated in Step 33 below. ! S133L1 - Evaluation of Differences i
- a. Table 4.*. ..ows that the RSGs operate at the sans secondary l
i side pressure, tesperature, and flow as the 09Gs. The only 2 difference that could affect the probability of a steam system piping failure is shown in SGER Section 4.2 to be RSG steam moisture content. The RSG steam moisture content is lower i i than the 00G steam moisture content. This is shoest in Sectica
h 2.2.7.3 of the SGER.wtbessamagoperating conditions and tha.. '
! NWineresse,theaseverity l lower stema moisturm<E+E conditions in.the;steen pip "S.bNore,7 use;orthe i does not increase the pr**miey of.;. occurrence <of a stammL-.9 .yg j ', systempipingcamve.;,ww5w%m.,,p;ppep mw. m+ p qp cpug. e ;..;.v . g.g.,, a b. There are 8523 tubes.fa$the'l1tssland 8519 tubes in the 000. a ' " i ditterence of lesagtham 0.eszpercent.c.:ptherefore, the ausber.
} of Rs0 and 000 tubes is n=midaced equal for~this evalustiam,'"
and the difference in the musber of tubes does not increase the probability of an SGTR. ! The 33G tubes are ma=inalty 0.003 inches thianae than the OLG C tA1 oggt490 hea2a.inighero. j j w 6, uswaAtubesDv
~ ~
allosa etEse 7_"P M beefMoes f - ~W'th,)1_-Jy$CXhretheMis Osos. Sectica 2.0 of the sGut oosperes the RSS an$ C0G'tabing '
dimensions and properties and shoue that the RSG tube. critical rupture pressure eaceeds that of the 00G tube. Because the i j o 4.7-2 l I i ) i
l i 5 j i SOER. Re O. Pese 283 of 472
- PPL. .e.e no Seleer . Br.ahm.W.on
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l , j': ~ j $. , (f. ! critical rupture pressure of the RSG tubes exceeds that of the l OSG tubes, the reduced RSG tube thiczness does not increase ! i the probability of an SGTR. h l We RSG tube material is' more resistant to primary and i secondary side corrosion and cracking than is the OSG tube j material. Several factors contribute to this increased l corrosion resistance. These include higher chromium content i and higher grain boundary carbide decoration. Sections 2.3.2 and 2.5.1.1 of the SGRR compare the OSG and RSG tube materials, and show that the R8G tube material resists i corrosion better than the OSG tube material. Because the RSG i tube material resists primary and secondary side corrosion better than the OSG tube material, the differences in tube l material corrosion resistance do not increase the probability i of an SGn. i i The RSG and OSG tube bundle configurations differ in tube bundle shape and tube bend radius, he RSG upper tube bundle l i shape consists of tubes with continuous, smooth, long-radius i bends, whereas the OSG has horizontal straight runs. The U-bends of the innermost RSG tubes are skewed to provide a l longer minimum bend radius than those present in the OSG l design. The RSG tube bundle configuration is used to evaluate i the structural and vibrational adequacy of the RSG tubes. l These evaluations are discussed in SGRR Sections 2.2.4, 2.6.1, I and 2.8. Section 2.2.2 states that the RSG tubes are ! structurally equivalent or superior to the OGG tubes. Da j same parasc are used to evaluste RSG vibration control as 1 were used in ene 08G. Se tubes are also evaluated for l mechanical resonance and pressure variation, and shown to meet j the same frequency, stress and strain criteria as the 08G j tubes. Derefore, the RSG tubes and tube bundle configuratica i are structurally as capable of controlling harmful modes of vibration as the OSG tubes and . D erefore, the RSG l ! tube bundle centiguration does not the probability of an SGTE. . i . De RSG tubes are supported by lattice gride in the straightjz.;.,. J: ! tube runs and by flat U-bar restraints in the bood-:regionweg ! Mis differs from the 000
- egg crate
- design.' .19e<RSS tubeWh.W support j %is system is used in the evaluation of RSG tube ~ structural ggsyst 5 j and vibrational adequecy discussed in SGER Sectices 2.2.4 j 2. 6.1, and 2.8. Section 2.2.2 states that the RSG tubes are structurally equivalent or superior to the 000 tubes. Se j same parameters are used to evaluate RSG vibration control as were used in the 000. De - tubes .are .also. evaluated; fosh . e -
mechanical maaa and pressure'variatica7and'Shouahzl 52d l the same frequency, stress and strain criteria as the 000 i tubes. Merefore, the RSG tubes and tube bundle support system are structurally as capable of controlling harnaful } 4.7-3 e I l i i h
l
~
1 1 l
- ! ppt sessey eveheden No JPNML4SdP44428, 9054 Asv.0, Pass 2M of 473
- Asv.e,pase7etse ,
,. m ..- .
i i modes of vibration as the OSG tubes and supports. Therefore, } the RSG tube support system does not increase the probability j of an SGTR. I j The RSG tubes are joined to the tubesheets by welding and expansion. The tube-to-tubesheet weld is described in SGER
- Section 2.2.3, and shown to be designed, analyzed, performed i and examined in accordance with ASHI section III criteria.
t This is the same criteria applied to the OSG tube-to-tubesheet weld. The RSG tubes are hydraulically === dad through the l full depth of the tubesheet, the OSG tubes are not. l Rlimination of the tubesheet cre1,-ice in the RSG protects the j tubes from secondary side stress-corrosion cracking in this i area. Full-depth expansion of the RSG tubes is described in l SGER Section 2.2.3.2. Because the RSG tube-to-tubesheet weld ! meets the same ASME criteria as the OSG design, and because { the RSG tube-to-tubesheet joint design includes the benefits i of full-depth expansion, the difference in OSG and RSG tube-to-tubesheet joint design does not increase the probability of { an SGTR. ! ! Because each difference in RSG and OSG design that could
- , affact the probability of an SG11t has been identified and 5
) shown not to increase the probability of an SGTR, use of the l
i RSGs does not increase the probability of an SGTE. l The paragraphs above describe review of the accidents previously i evaluated in the SAR and identification of the steam system piping l failure and SGU the only accidents whose probability of i occurrence could be affected by use of the RSGs. The RSG and OSG differences that could affect the probability of occurrence of these accidents are identified and evaluated to show that use of the RSGs does not increase the probability of occurrence of'these Therefore, events. use of the RSGs does not increase the f probability of occurrence of accidents previously evaluated in the i SAR. i 4.7.2 Does the proposed activity increase the consequences of an accident previously evaluated in the Safety Analysis Report? This question is addressed in three steps: First, accidents ! previously evaluated in the SAR whose a nse p uces could be i affected by use of the RSGs are idr.ntified. Second, the j differences between the RSGs and OSGs that could affect the dose i consequences of these accidents are identified. Third, the RSG and 000 ditferences are evaluated,to deterninajiti.the conesquencestof % .- we*' these accidents are increased'by,insie'of'the~ RSGs. rThese steps are i addressed below. I i l l 4.7-4 i t A
- - - - - - - - - - - - - - _n-- ,v., - - , , - , , - - - - ..-
4 . _
- .s
- PFL Seesty 8vehanden 10s. #1WESSS44404, Sest Rev. O, Page 295 of 472 .
nov.e, pees e as se ;
- n. i I
i SlagL1 - Identification of Accidents for Which the Consequences j are Potentially Affected by Use of the RSGs 1 l The attached Table 4.7-2 susnarises accidents previously j evaluated in the SAR, lists their critical parameters, and j compares the OSG and RSG parameter values used in the
- evaluation of each event. Table 4.7-2 identifies the SAR 1 evente that have dose acce ptance criteria and where .
j differences between the Raga and, OSGs exist that could affect i the dose as (a) the steam generator tube rupture (SGTR), and { (b) waste gas decay tank rupture. 1 i ! Stasta - Identification of RSO Differences That could Affoct l Accident Dose Consequences .
~ - > .wy RSG ditferences from the ceGs that could affeet SAR accident ' '
1 l dose consequences are identified in Table.4.7-2. They are shown for the events identified in Step 1 when the RSG and CGG l values of important parameters differ. These differences are tube inside diameter and primary system volume. The effects i of these difforences on tha Gak accidents identified in Step 1 are evaluated below. Sten 3 - Evaluation of Differences
- a. The total SGTR primary-to-secondary system leakage is greater for the RSG than for the OGG because the naminal RsG tube inside diameter .006 inches larger than the OGG tube inside diameter. Because the RSG flow area is two percent greater than the OSG value, the integrated leakage would be two percent greater with the RSG than with the 00G. This ,
means the offsite doses with- the RSGs would be two percent greater than t.he values calculated with the ceGs. Therefore, the whole-body dose of 0.103 rea reported in the St. Emacie 1 FSAR would be 0.105 rem with the RSGs. Similarly, the thyroid dose of 0.41 rem reported in the FSAR mould be 0.42 rem with, the RsGs. These values compare with the Mac-approved SAR (10CFR100) acceptance criteria of 300 and 25 REN, . respectively. The.. y*==**=L doses with the RSGs ruumia a . d ..W i small fraction of(.the;s10CFR100 limits. Therefered the .!a"ef a ~~ l negligible lacreasonic,1aidose zamala:Delow:the existing . l approved SAR ' acceptance criteriapand the consequences of,thet ~ _ . .y i SGTR are not increased ~by une of the tsGs. The details of this evaluation are dae====*=d in section 4.2 of the SGER. . i
- b. All of the radioactive gases stored in the weste gas.
j
> M .: wsut jy & Q ,Z&a _" '
assuming that the ea=*==*= of the easte gas decay tank are released instan*===an=1y at ground level. The parameters that 4.7-5 1
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! could affect the dose-for this event are RCSRCS activity level is not and RCs volumna. l 1 the RSGs. but RCS volume is appi h tely This ces in results percent larger an increase with the RSGe than with the CeGs. 1 of the calculated offsite thyroid and whole-body doses frosa 0.000878 and These 0.128values REN compare with thewith 08Gs to 0.000887 SAR the NRC-approved and 0.129 RE i with the RSGs. criteria of 300 and 25 REN, (10CFR100) acceptance l respectively. Control room whole-body dose would increase This by ! the same proportion froen 0.000157 Rest to(10CFR100) 0.000159 REN. acceptance i i compares with the NBC-approved SAR Therefore, the negligible increases in i criteria of 5 REN. dose remain below the existing MRCMapproved SAR- acceptance l 0
- criteria, and the consequences oE.weste. gas decay tank rupture %-
are not increased Dyfuse of:ths'lRSGsAThesdect.11sfef . M: this.h..@:. j . l evaluation are documented in Section54.:2 ofttheisGER. In the paragraphs above, the releases for SAR =~idanea that could be affacted by use of the RSGe are shown to meet the existing place of MRC- the i approved SAR acceptance criteria the with the RSOs doesRSGs not inLacrease
' the l
! OGGs. Therefore, use of l consequences of an accident previously evaluated in the SAR. i ! 4.7.3 Does the propmed activity increase the probability of occurrence of a sanitunctica of equipment important to safety -wiously evaluated in the safety Analysis
- Report. :
i First, the differences . ! This questica is addressed in three steps:between the Rees analysis are and
' ceG l
equipment assumed to functica in the =~idane ! identified. Second, the equipment that could be affected by these is iAmmelfied. Third, the effect of the
'aso ditterances differencee identified.ca the pr**itity of Fora the am1purpose functionofofthis the j
j equipment identified is evaluated. evaluation, meeting the ~ existing NBC-approved SAE acceptance criteria is considered to . demonstrate that the pr**ility of a structural maifunctica is not:tacreased.- These steps age addressed I i 89"A8"'88....m E N M-e M fk.w$$[hNN ~ "" ~x+ 4;$I k; .da.;N}.d
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; - operating conditires of
- ' 4,.1-3 l
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t,.. asessmed to function liiN 7 ; meef Aase 336V.ESS"..s'.c. ' weight and center of gravity. The RSGs are apprestimately three percent heavier than the OSGs and their centers of gravity at operating condif0 me are apprestimately two inches higher. EtaIL2 - Identification of squipment Affected by RSG Differences
- As described in SGER Section 4.4, the RSGs are appreimmately three percent heavier than the OGGs and the RSG centers of gravity are higher. W erefore, the loads imposed on the RSG supports and attached piping are slightly higher. Evaluation
> of the increased Rae loads on supports and attached piping is '
discussed .in Step +.3 belawa k. :.a - ' 1. .m, gige ryg.. ! M e-3 m.engg . .... ~..a
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2133La - Rvaluation of Differences The effects of a three percent increase in weight and higher j r center of gravity for the RSGe are evaluated in Section 4.4 of ) the SGER. The evaluation shows that the existing supports and i attached piping loads meet the NRC-approved SAR acceptance criteria ("specified-for-design
- stress levels and loads) with the RSGs in place. of the OGGs. An example of the evaluation results is illustrated in the attached Table 4.7-3 which shows the relationship between the OSG and RSG seisatic loads on These are WM with the i primary system components.
j . *specified-for-da-ten loads." i Because the stress levela and loads remain below the existing ~ i NRC-approved SAR aW== criteria with the RSGs in place of i l j the OSGs, the pr*hdity of a am1tunctioniia; considered to be "2.q$s j equivalent. Therefore, use of the RSGs doeursot increase the i probability of a malfunction of component supports ce attache ( piping. j ' The above paragraphs identify increased RSS weight and higher i center of gravity as the only aspects of use of the RSGs that could af fact equipment important to safety. . Stensa generator supports and , attached piping are identified as the only'egnipsont that could be i ' affected. The evaluation shaus that tha incremos in weight anS 1% 4/1.; /. higher center of gravity xda?achaffectg,phe;. Wh"4*y.. of. _ .,.- this =? c --M'Thesats
-s i
malfunction increase theofprei =Hty of a malfuneHa#et. e Athe 'important 380s does toy not *c G 6,:p;,. l safety previously analysed in h"*[{-- ar @7$ 4- ip.% .gc jj.: ,21 - -
- g. -
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-m ' This question is addressed la three stopc4 'First; between the RSGs and CeGe that could affact the consequences of a salfunction of the equipment assumed to function in the SAR accident analysis are identified. Second, the equipment affected is identified. Third, the effoct of the ditierences on the consequences is evaluated. These steps are presented below.
Stag 1 - Identification of RSG Differences That could Affect Equipment Malfunction Consequences The RSGs could affect tLa consequences of a malfunction of equipment by (a) affecting the sequence of events or thexaml-by (hbatf=*8== assumed
? M.W W. hydraulic response of as accident,ingh & ** *
- 1. .
operator .- actionse seehr.dckf>caus MEsG%
. .? ~ 4 4 .important to. safetyw out.fdidedtheityWg ts#1T Differences between the R9E and 000 that potentially increase ~
' the consequences of a maltunction of equipment assumed to function in the SAR are addressed in turn below.
- a. As described in SGER Section 4.2, the RSG and 08G thermal-hydraulic response to, and sequence of events for, accidents I
are equivalent. In addition, with the RSGs in place of the ! OSGs, the equipment assumed to function in the SAR accident l analysis functions within its operating limits. Because use
- of the RSGs does not change the sequence of events or thermal-hydraulic response of any accident, no RSG differences are identified that affect the consequences of a malfunction of equipment e M to function in the SAR.
. .. : . . . .:. .a
- b. The operator actions ====ad in~ stos accidents --
E, analyzed in the SAR are dae===ated 'ther W :- n
. prae =AN. Evaluation. 00 plants off ~~== i . ,V.
operating prae-A wes, described in SGER eme*iaat " 31Eshous that tho actices they prescribe zemain s :-s-- .c i.a*% for the d,.'.- RSGs. Because use of the.ESGa does?notichange3theTregn;izedp@h... operator actions, no R8G - differem*ared lamtiftade that' - affect the consequences of a malfur*4an of'egeipment assumed to function in the SAR.
- c. Review of the plant Technical Specifications with respect to the accident analyses presente6 in the SAR, described in SGER Section 4.6.1, showed that tha operating limits used'for the OSGs are appropriate and adequate for the EsGs.- Becense use of the RSGs does not change the operating-limits of systems important to safety, no RfA differences affect the consequences ofL.smalf**aa*; aral*"=5 idaneif'la8 4 = T -
that E, h, .;m,;aw
~ ~
--t.._- -
;- s- ,
W pasgraphs abover show h w affect the sequence of events or thermal- ., L lic response of
=not k "'
an accident, (b) affect assumed operator actions, or (c) cause 4.7-8 i
-- _ _ . . _ . . - . - . - . - . - ~ - . - -- - -
--- - - - - - - - - - - 1 l l l
l ppL safety eveheelen No, JPOWOL4EMP44420, i hav. e, page 12 et as 30.'It, Asv. 8. Page 299 d 472
~
.Q'. s.4 operation of systems important to safety - outside Eheir i
k'.f Af'f w { operating limits. As such, there are no RSG diff8" faces thEf affect the consequences of a malfunction of equipment assumed to function in the SAR. Sten 2 - Equipment Affected by Differences i l Since there are no differences identified that affect the consequences of a malfunction of equipment assumed to function ! in the SAR, no equipment is affected. i E1SEL1 - Evaluation of Differences ! Because the conse there are no R9G differences identified that' affect A d em9 Wr quencesM na affected of equignment e assumed to function in the SAR,
- differences to evalums.s. quiyt is identified, there are no Because there are no diffarences between the RSGs and OSGa that 4
could affect the consequences of a malfunction of equipment assumed I to function in the SAR, use of the RSGs does not increase the i l i consequences of a malfunction of equipment important to safety i l previously evaluated it, the Safety Analysis Report. I
- 4.7.5 Does the proposed
- activity create the possibility of an )
! accider a ditferent type h ag pr M cwiety j evaluaten u the Safety Analysis Report? l ) . . This question is addressed in two steps: First, the differences , hetween the RSGs and ORGs that could have the ~~*4=1 to creata. { an accident of a different type than any Mously evaluated in the SAR are identified. Second, the differences identieted in the ,. first step are evaluated for their ,=*-i=1 to create an ma*44=* of a ditferent type than any previously evaluated in the SAR. Sten 1 - Identification of RSG Differences That Could Create a Different Type of Accident The RSGs connect to the same piping, instrumentation, and stupports as the OSGs. This is described in SGER Section j 2.1.3. Because the R9Gs connect to and use the same piping, instrumentation, and supports as the OSGs, no differences are created that could result in the possibility of an am-4 Anne of i
# a different type than any previously evaluatedria the.
GA(Theresarettwo blowdown nozzles in the'RSS designFand .SAR.oner in"t *,.~ the OSG design. Also, there are 12 water level tape in the RSG design and 9 in the OSG design. Because component nozzles 4.7-9
- _ _ _.._ _ _ , _ , _ _ _ _ "-*""*N444 a. e w-i,,ua,-,,.
i l Y' . j ) h. ik%meNynk^. - ..+. t At r= ' - nM are'act assusied to"faif,#the7 ? ~. A w } the possibility of as meidane of a different typethan ! occur with the 00Gs.) i I Step 2 - avaluation of Differences mecause use of the asGs does not create differences that could l result in the possibility of a different type of accident, there are no ditforences to evaluate. l 1 ' - ! IW Because use of the 390s does not create any differemose.that.& 4 zeeultiig,ther: ibi11ty, or..alg =~idanesozia.diffagent3 [EArf M r
. ..g
[h.$ fogy g>ekS5De
. ' " ' ' -~ .,u: .n g ~an%
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N.5 d,',2pg . . , Does the proposed activity create the poss'2llityiddN,.M8 3 , l 4.7.6 am1 function of equipment important to eafety of:a" A/ % ' ditferent type than any previously evaluated in the SARP This question is addressed in two steps: First, the differer.ces between the RSGs and OSGs that could have the potential to create 4 j a malfunction of a different type than any ptwviously evaluated la the SAR are identified. Second, the differences identified in the j ' first step are evaluated for their potential to create a . , am1 function of equipment '=mtad important la the to SAR.safety of a.:different.'
, ,7,: fg*
type , z.,% m-So @f than any~previously .
~
h;.M..N,h$@$ Mhdth@gfd@MM w w'
~
xdentif ~ ~ - istaas&intererenomes ' k n uitg . Mt~ ~
. 14s w :> 4.1 e m ,6 %_ %.3ev.%%p#1W t?'T!?#f@igdag the.
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P i.T.5Fh~tisWoosmic@hT'=Ogge,s instrumentatica, and as the M&" sg t in the ibility oria~mateam,+tmm 7 %f 1-created that could y evaluatseritthe(BREMR'9:i of a different type thea es in the RSG design'and etWia:c ' i-- 4 & (There
.the 000 are design.two blosdousAlso, chore are 12 unter level tape'in the . % 1 7
asG design and 91a the 000 desiga. Secomme component aossles . 1 are not assumed to fail, the mesitional aossles cannot-create 5 1. 1 the poseihility of a ditfarent type of equipannt mete ==+ei ..g m4f:. 6 than could occur with the 000s.)
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j
.e "'s.t.sess14c2.ae g g . 9.h[2---- -
J l g g,p 7 - h&~h( 7 sc%.W ::g,w m ... ! -: pc . ..x +. . %Mf .8
- Secause use of the'M~h'm,cre, ate' ._. , .
i could result in the possibility of a fiorent typk/of* . .
^,_t% ~t i
i malfunction, there are no ditierences to evaluate. l Because use of the RSGs does not create any differences that could l result in the possibility of a malfunction of a different type, ' their use does not result in the possibility of a malfunction of 3 equipment previously evaluated in the SAR. important to safety of a different type than any j a fe , 4.7.7 Does the proposed activity reduce the margin of safety as . i defined in the basis for any Stechnical Specificatientgvengsq
'." This questicacdmi N$doMsNydkribIah ~
j SAR accident analyses and structural analyses idsstifies? m A % evente and components that could affect the margia of safety. - l (Reduction in the margin of safety is defined as the.value of any j acceptance criteria parameter exceeding the NRC. approved value that l is reported in the SAR or modified by the NRC's SSR.) Second, i differences between the RSGs and OSGs that could affect the events i and components found in the first step are identified. Third, each I j difference is evaluated to determine its effect on the margin of safety. Fourth, the Technical Specifications and their bases are reviewed to identify any changes that are required by use of the l RSGs, and to assess the effect of these changes on the saggia of l safety. These steps are addressed below. - 1
- a .. ,, .'
+ :.%a. , m:--:: marc.. ,
2133 1J - Identification.of.SAkiAccident and Structural ~ ~ d CA^ l , ^"^- - - That could Atfactithe Margia.of." Safety;; $,. $$ ., a f.c i :t.:.l:X ? q s s.;.; . . . C avaluation of the$.SAR arw idanes witi,the asGs la ' j ! osQs Evaluation;of.the is described laa sGER sections 4.2J 4. M St.bmacie 'S sex =ce==1. analyses 6 ? i ~ RSQs in place ofytheF00Ggis@ihh!Am sGkR78ect . i 4 l The results of these evaluations identify the following e, vents % ~ # and structurni components where the parametric results could exceed thcts for the OGGs: l l 1. Loss of feedwater. ! 2. Maste gas decay tank rupture.
- 3. ' Low temperature over-pEassure (RC pump start-up) .
- 4. LBLOCh containment response.
j 5. Steam generator tube rupture (SGTR) . j 6. MSLB containment response. l 7. RSG supports and attached piping loads and stresses. RSGdiffN 2, e$beh kt kre# discussed ~1a Step" 4 4 d
) 4.7-11 i
4 i i t
'. _ _ - . . - . - -- . - - . - - - - - - - - - - - - - - - - - ^ - ~ ~ ^ ' " ' ~ ~ ~ ~ ~ ' i i PPL Safmy h No, M SENP44420, sogn, n ,. O. Page 302 of 472
- Asv. O, page 15 of 34 31SEL2 - Identification of RSG Differences
! no RSG differences that could affect Items 1 through 7 in ! Step 1 are shown in corresponding Items 1 through 7 below. ! These items are identified in SGER Sections 4.2, 4.3, 4.4, and j 4.5. I 1. Less RSG secondary side water inventory.
- 2. Greater RSG primary side liquid voltane.
- 3. Greater RSG heat transfer surface area.
j 4. Greater RSG primary side liquid volume. j 5. Greater RSG tube inside diameter. l 6. RSG integral flow restrictor. i 1
- 7. Greater RSG weight and higher center of gravity.
j i n ose ditferences are evaluated in Step 3. I ] \ l Sten 1 . Evaluation of Differences l
'!he ef fect of the RSG differences frcun the OSG on each of the I seven items identified in Step 1 are evaluated below for their j ef fect on margin of safety.
i 1. Steam generator dry out can occur earlier with the RSG because i the RSG secondary side water inventory is less than that of j the OSG. 'The evaluation of the loss-of-feedwater accident presented in SGER Section 4.2 shows that, for a loss of main j feedwater with auxiliary feedwater available, a heat sink i ' would be assurr ' Also, the NRC. approved SAR acceptance criteria would me r st, and there would be no reduction of safety with the RSGs. The evaluation also shows that, for a i total loss of feedwater from full power, with off-site power available 10 minutes would be availahle for mammat actuation i of the Auxiliary Feedwater System. 'therefore, the evaluation ! in SGER Section 4.2 shows that the NRC-approved FSAR acceptance criteria are met for the IDFW event withitheiRSGet. ,L i 8 and use of the RSGs does not reduce the margin of safety for the Loss of Feedwater accident. A St. Imcia FSAR revision'is w ==M1:ba
' acceptance criteria'f'with p- - if
*O isW i**W"N NERL event f ariiic ,
(( _ implemented ' a 1 f 4 ' evaluat Cha~ , g g i
.v.s. sw&g..n,m g . ,n -..v. >We% W h $&
- 2. . All of the . radioactive gases stored in the waste' gas decay tank are assumes to be released in the waste gas decay tank 4.7 12
I 1 l 1 ppt safety tvetussen No. JPMsWENP44434. 90tl. Asv. Da Pass 303 et 472 now.O page teof as l , i k - a .;ckfi:p .g ..
. 3 .,c rupture accident. Whole-body and thyroid doses are calculated assuming that the contents of the-waste gas decay tank are li released instantaneously at ground level, ne parameters that l
could affect the dose ior this event are Ecs activity level 1 P.nd RCS volune. RCS activity level is not affected by.use of ! the RSGs, but RCS volume is approximately one percent larger ! with the RSGs than with the OSGs. Sie results in an increase ' of the calculated offsite thyroid and whole body doses from l 0.000878 and 0.128 REN with the OSGs to 0.000887 and 0.129 REN with the RSGs. Rese values compare with the NRC approved SAR criteria of 300 and 25 REN, j (10CFR100) acceptance respectively. Control room whole body dose would increase by l the same p. W ien from 6.000199 REN to 0.000159 REN. His compares with the NRC-approved SAR (10CFR100) acceptance j criteria of 5 REN. Derefore, the negligible increases in l a dose remain below the existi. NRC approved SAR acceptance not reduce the margin of
- criteria and use of the RSCs safety for the waste gas decay task rupture. De details of l
g this event are presented in SGER Section 4.2. i j 3. One low temperature over pressure event could be affected by use of the RSGs because the RSGs have greater heat transfer curface area than do the OSGs. S e event is reactor coolant (RC) pumqp start-up with the steam generators hotter (50 F by definition) than the RC system. Mis event causes RC l and pressure to increase because heat is l i temperature transferred from the steam generator to the RC system. More l l heat transfer surface creates the potertial for faster RC tenqperature increase. De evaluation presented in SGER ' ! Section 4.5 aws that the RCS pressure remains below the maximum value specified in the SAR. Derefore, use of the RSGe does not reduce the margia of' safety for the low 3 temperature over-pressure event.
- 4. ne Latock containment response is potentially affacted by use of the RSGs because the RSGs increase the primary side watar i volume by .,y. :=nen1y'onageroest. dhis Sacreases the mass m available in the bloudous phase /aat the resulting enrainmane
! pressure. De evaluatica presented la SGER Sectica 4.3.3.1 shows that although eamtain= mat building pressure is increased i l from ayy. dmately 38.4 peig to syy. imately 38.7 peig by use
- of the RSGs, the peak containment pressure is below the NRC-i approved BAR acceptance criterion of 44 peig. Ya addition, the f.nfer*m contain==vit conditions with the R80s were reviewed ipment
- and found to meet the existing St. Incie 1 qualification envelopes. Derefore, use of the RSGs not j
reduce the margin of safety for the LBLock containment response. i irki- y. . .;c r b ge G ' j is greater for the BSG thE for the 000 becense the BSG tube j inside diameter is apprawi==tely 0.006 inches larger than the
! 4.7-13 i ,
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l sota, nov. o.rese aos er an m sesesy evesussen see. JmesumeP*ess. i Rev.9. pees 17 of 34 i 080 tube inside diameter. Because the R8G flow area.is two l percent greater than the 090 value, the integrated leakage would be two percent greater with the RSG than with the 090. l his means the of tsite doses with ths RSGs would be two j percent greater than the values calculated with the OSGs. Therefore, the whole body dose of 0.103 rem reported in the l l St. Lucie 1 FSAR would be 0.105 resa with the RSGs. Similarly, l j the thyroid dose of 0.41 rem These reported values in the FSAR compare with would the NRC-be i 0.42 rem with the RSGs. appecved SAR (10CFR100) acceptance criteria of 300 and 25 REN,a respectively. De potential doses with the RSGs remnainthe Derefore, i I small traction of the 10CFR100 limits. l
- negligible increases in dose remain below the existing MRC- l approved SAR acceptance critoria, and use of the RSGs does not l
' reduce the margin of safety for the SGM. D e details of this ) evaluation are documented in section 4.2 of the SGER. l i 5. D e MSLB containment response is potentially affected by use of the RSGs because the RSGs contain an internal flow restrictor at the steam outlet. The flow restrictor limits l l the effective Main Steam Line Break area to 3.69 square feet i j compared with 5.355 square feet used in the containment j analysis for the OSGs. This reduces the calculated break flow and the calculated pe~ak containment pressure for the MSLB with the RSGs. In addition, the MSLB containment conditions with l l the RSGs were reviewed and found to meet the existing St. ! Lucie 1 equipment qualification envelopes. M erefore, the l existing SAR analysis WM= the containment pressure response j
- with the RSGs and use of the RSGs does not reduce the margin - i of safety.
~
Mils of the evaluation of the MSLB are l l j presented in w ~ Section 4.'2. The RSG support loads and stresses could be potentially i l 7. i af facted by use of the 2SGs because the RSGs are apprenrimately 4 l three percent heavier and their centers of gravity are i cyprasrimately two inches higher (at operating conditionst than i those of the ceGs. S e evaluation presented in SGER Section 4.4 shows that, although the increases in weights and centere l of gravity increase the loads on certain supports and attached j piping, the resulting loads and stresses areThazefore, within the MRC- use of j app. d acceptanos criteria in the SAR. the RSGs does not reduce the margin of safety for the support and attached piping loads and stresses. - i' Evaluation of the SAR accident analyses whose that, while soms" n l RSG analysis results exceeded the cus=WyGeding OSG values. J all existing NRC approved SAR acceptance criteria are met with the RSGs for every accident. D erefore, use of the RSGs does not reduce;the marginsuot. safe in tha,.S&R -4 Aaa* anal m - - w e . . ,..sg 4 m y3 ~ i Sections 4.2, 4.3 and'4.5. In~addi . loads and stresses are shown in SGER Section'4.4 to remain 4.1-14 l l
j PPL sesw treasden Mt. JPW9b88MP44428, SOst, Rev.d. Page 306 et 472 l nov.e Psee is of as
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e below the existing Therefore, use of the MRCeapproved' RsGe does notSAR reduce acceptance the margins criteria'.7,ypM of- .E i l safety in the SAR structural evaluations. '._ } 812EL.i
- Evaluation of Technical Specifications and Bases Review of the Pt. Lucie 1 Technical ap gcications and their l bases showed that there are no changes regaired to the i
Technical specitications or to their bases by use of the RSGs. l nerefore, the existing Technical Specifications and their ! bases remain applicable for the R9Gs. The results of this l review are documented in SGEL SectionSGER 4.6.1Appendix and in the A. Technical Specification checklist in Serefore, the margin of safety in the Technical l l Specifications is not reduced by use of the RSGs. l Because the NRC. approved acceptance criteria in the SAR accident and structural analyses are met with the RSGs in place of the OSGs, ! and because changes to the Technical Specifications are not _ required, use of the RSGs does not reduce the margin of safety as defined in the basis for any Technical Specification. 4.7.8 Conclusion Sections 4.7.1 through 4.7.7 address the seven questions to make j the unreviewed safety questica deramination required by 10CFR50.59. Sections 4.7.1 through 4.7.7 d==anatrate that there .-~ l are no ditforences t 'n the RSGs and OSGs that create an l
- unreviewed safety question. This is ac m lished by showing that use of the RSGs does not (i) increase the probability of occurrence or the consequences of an accident or malfunction of equipment i important to safety previcrasly evaluated in the safety Analysis j Report (SAR); (ii) create the posf>ibility of an accident or j malfunction of a different type than any previously evaluated in -
the SAR: or (iii) reduce the, margin of safety as defined in the ~ basis for the Technical Specifirmeiamm. ) Because use of the R9Go doahnd-mequire changes to the St Imcie ., _ je ' 1 Technical Specifications and~ does not create an unreviewed safety - @@ggy question, it is concluded that the RSGs can be used uithout price *y. p-NRC approval as allowed by 10CFR50.59. :N u f, y. wS e . 1 k h?$;: ba)ihk,5,d & &
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- ACCIDEttr ANALYSIS St#GERY (Continued) 17 i Table 4.7-2 52 NOTES: m
- 1. Standard Review Plan section number. s 1
- 2. Minianun flow allowed by revised Technical Specifications. X l
{
- 3. Value for up to 18 percent plugged tubes.
total core flow with no tube plugging. I
- 4. Beet-estimate loop resistance based h t F. '
E
- 5. SG primary side liquid volume. %
SG primary side flow resistance"from tube sheet to tube sheet. I
- 6. 6
.I i
4 E 5 r f o E e w 0 u
-i 4.7-27
--- _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ -- - - - - ~ - - - - - . _ _ _ __ _
- _ _ ~ . . . . ~ . _ . . . _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
i i t I Table 4.7-3, Seismic Loads on Primary System Cotupon nts (came cc SGER Tc.ble 4.4 - 3) I, s.is-te t.se toi. tim
- Deecent _
Seteele @t crus Doefp Setes fy t ac e 'at t a6 L ocat ion taletig Calculated Septocement Spectflog for g,,g p nes t am Calcuteted seestana Desip Deele F4
- 60.6 42.2 241.8 75 2
Cas sined teactor veteel fa (ktps) I wul i k let mostle . 15 2 15.7 M3.8 94 96 fy (ktps) and Seecter veeset 3 gt. , 1.0 1.s 10.0 9e 9e g ennet cepig 's (k ts.) , n (In-klee) is.s 16.s ina2.e 9e 9e gk I er (In-kips) 95.3 98.2 1061.9 91 91 f 1,305.5 43T33.0 9F 97 h Nt (in-hips) 1267.* Roseter won et intet te (ktpe) 50. 52.3 190.0 F3 T2 h seente 144.0 88 87 tj Fy (htpe) 17 18.4 z 14
- fe (tips) 4F.4 44.8 57.0 17 Itt (le klpe) 2262.8 2,330.7 4 9634.0 FF 76 ny (in-ktpe) 151F.0 1,562.5 5961.0 M 76 .*
na (In-ktps) 1876.8 1,933.1 127F0.0 85 55 16.9 17.4 10 0 && 42 Coselned Reector veenet Fa (tipe) 2 erul T Outlet motste * *2.0 99 99 and Fy (ktpe) 2.4 2.5 . Seector vesset 6.4 6.6 &0.0 && 84 Outlet Pf pig ft (htpe) 150.0 154.5 1170.0 SF SF Iu (In-ktpe) sey (In-kips) 300.0 826.0 7521.0 39 39 at (In kips) 244.4 251.7 37644.0 99 99 teoster veneet latet fa (ktpe) 29.7 30.4 61.0 51 50 $ m Norste 12.9 105.0 at es ? Fy (ktpe) 12.4 :p fa (kles) 29.2 30.7 100.0 72 F2 $ 1876.8 1,931.0 22392.0 92 91 sen (In-klpe) i ny (In.kipe) 731.1 Ms.0 tooe5.0 93 93 { pe (in ktes) 15's.9 1,595.4 1soer.e o9 e, S Note: The percent below design was calculatec by ; 100*(Specified for Design - Calculated)/Specified for Design. 4.7-28
',%t% Table 4.7-3. Seismic Loads on Primary System Components (cont'd) g..s.._ - l raisiine seio-ic toed ,,%, percent setas Component end Dett e Below g,g g p fmy tr' w'c Cr w t taisting retruteted Septerement Specified for Location Design incisn'*en Centuteted Monteus Design f me = ian sa SS E 62.7 149.0 59 N ta (ktpe ) 60.9 Steen Co*+ ester ("*4" W Intet morale 54.3 68.0 23 20 7 j ' sa3 1 and fy (hips) 52.7 Ug j 4.0 F5 M ,, l Steen Ge+wrotor ta (ktpel 1.0 t.0 a Intet Pepine ft3.0 90 90 $ g. 5 73.8 16.0 Ma (in-ktpe) 108.0 83 38 f 85.8 88.4 4 My (In htps) F16.F 16824.0 96 96 g I Ita (in-ktpe) 695._8 _ 9 Fa (ktpe) 26.4 2F.2 10.0 12 h.- 9 team Generator -- 154.0 F6 Fe g I w i.t eeniis 31.7 32.7 z ty (tlpe) 9 g 27.2 33.0 20 tt fa (hlpe) 26 6 T 3,908.4 12951.0 FF F5 g mm (In ktpe) 3088.3 14 32 _
,5 2,F98.4 3541.0 per (la ktpe) 2FD8.2 u.se.o 55 se mm (ta stys) 3 cia.3 3,tos.s l
2' O 3F 31 tt 0 IF.1
$ teen Generator Fa (htpe)
Castened 90 30 ' Intet Donate 15.F 144.0 2 ard v fr (klpa) 15.2 and 4 F.0 9 __ Steme Generator Fa ${pe) 4.4 4.6 I letet Piptg 16 5FS.0 34 3%6.5 34F.2 as (In kSpe3 3773.0 90 90 385.4 393.0 soy (le.htye) 1120s.0 99 99 61.4 45.3 HR (In-htps) 24.6 33 0 7 g $ te (hlge) 2F.8 " j Steen Generator F6 76
= i.t .e.,is 22.F 9L0 -
fy (tige) 22.0
, , . ,. 6 ,,0.0 ,, ,6 5 fe (,,p., .O 6467 0 71 70 ,
1,922.0 i 1966.0
}
I nm (in kips) t5es.* i.434.0 c m .0 63 62 { my (In.ats.) 7150.0 76 F3 $ 1866.0 i,922.0 j ne (l kipe) 7
~ '
dJte. The percent below design was calculatect by 100*(Specified for Design - Calculated)/Specified for Design. d , i 4.7. <
susweeuner 3 Table 4.7-3, Seismic Loads on Primary System Components (cont'd) 1, 'c s. .e.
. . . . . , ,:,- =o.
, tetson or , _,0 t oc a t ier.
g.i. in,c.tcut.e.o
....u.
n.pi.c C.icui.i.e n 6 n, so.civ..a v., enie e.i onie mg, {,yy pu.y IC.F 11.0 78.2 86 06 C o=e' ned fa (klps) 3
. . -, heart.or
. ..Csot
.......nt 137.0 TF F6 8#
Fy (hipel 3: F 12.F " 4 g Fa (ktpe) 4.5 4.6 37.9 88 SS I j i Mu (in ktpe) 1181.1 1,216.1 5F13.2 T9 F9 g my (In-LIpe) 707 F28.3 4976.0 86 95 f 33' 3,457.8 16461.0 81 F9 Mr (in-ktps) teatter Coot.nt P g Fa (ktpe) 51 52.5 199.1 F4 F4 Outlet metale Fy (klps) I F.2 17.F 145.9 88 SS g Fe (ktpe) 31.8 34.9 40.0 6 1 mm (In.kipe) 1347.2 1,347.6 1989.0 6A 65 8583.9 F5 74
- sey (In-ktpe) 2180.3 2.245.F ,
i set (In. kip) 2300.8 2.360.8 16658.0 86 S4 Camblemt Beester Coet.nt Pump Fs (klp*3 15 3 I5 8 I *"d 1 I"I " IO j py (kly) 25.9 22.6 ' II pg (ggo.) 11.9 109.9 89 89
- f tt.6 g ,ggn.ggy .) 938.6 966.3 1 H T3.0 92 H my (In-ktps) 76s.0 F91.0 11s62.0 94 91 7451.3 2,$24.3 8334.1 71 70 sea (In klpe)
Reetter Cootent Puup Fa (ktpe) 29.8 30,7 125.1 76 75 M j Outlet monate 12.9 91.4 86 86 l Fy (hlpa) 12.4 l 19.7 20.3 94.9 79 79 98 (htps) O 839.2 864.6 1F5)2.0 95 95 l tea (in.kipe) - I my (in-kip.) tott.t 1.062.0 n00.0 6 3 { ng (en kip.) 1262.0 1,299.9 6391.0 so ao 0 [Ut e : The percent below design was calculatect by 7 l 100* (Specified for Design - Calculated) /Specified for Design. O . i l 4.7-i I
h Table 4.7-3, Seismic Loads on Primary System Cr=ponents (cont'd) Setente toed talettne Percoat , l toiselt Ceepenent ord Design g g tsletIng celeuteted Replacement Speelfled for Selow g ,g , ff , Dr*totton Lecetlen Wes* = m Cottulated lieslaue Seelen DeelP _ *g 12000.0 F2 71 is (In-ktpe) 31tl.6 3,423.3 , du ned Steen tenerator ! btlet Pipino 6 enr3 T 2235.1 2,302.2 12000.0 81 81 gh , l Stees Generator et (In klpe) g5 , tuttet Pipine 1794.0 1,84 F.S 12000.0 84 M g Pump Int et # (In-hlpe) g i Pipine ' i 12000.0 10 ee 3,T36.9 Pump tatet n (In kips) 36N P6 pine 12000.0 5 F5 3 Pump Outtot p (In klpe) 2f. 3.043.8 "f, ; PIPlr4 78 77 : 2636.1 2.215.2 12000.0
! R.V. Inlet n (In klp )
.ipi, '
1200s.0 aA al 6 C e ined Steam Cenerater n (In-klpo) 19F4.1 2.053.5 g Z and T guttet Pipleg ' 1,505.9 12000.0 SF Sr ' n (In-ktpel 111F.8 Steen Genereter ' guttet Pipteg
- m0s.e se se ow ,intet a tin-tip) i m .s t.220.3 [
P'el'4 2,Fes.t 12ess.e rs 77 a tin-alp ) 2s27.3 { s pu ,intet P8 pine t was.e se se 3 p , wetot a cla.mips) 131a.2 t.nr.F Paesev trees.e as as I G.v. Intet n (la-kles) its2.s 2.ett.s P8piae , I , Note: The percent below design was calculated by $ 100*(Specified for Design - Calculatedl/Specified for Design. 5 0
~
i 4.7 . i i
lI' f f l I< ),\l1 t e ,
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)
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@t o C f a
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- s. f y t ade c 1 9 6 3 6 1
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, a m e n
wo of 3 7 fp r r n it p
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u l ed bei s l t t ot et e - etee t e e r e t r t 4 D n t e s r e t r ae a ei e n n s s* e eS tf sw ep r g r s e s es e is t t e d o nt vuE vus p e n i r o ei ni eS oe ec e* l et t e t u ot e r r G G c ec ct b t c ne ot o t ne r e e r n e re os n rpce a eL t e ct c e t et eu s e r e F ri e t T no et eu en ee 0 W t r s e cr tI eS
. W.g' ov e e p(
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_ . . , _ . _ _ . . ._ .__.. ..._~. . m. - . _ _ _ . _ . . _.__ _.... _ _._ _ _ . _. . . . . m- . . . _ . . . . . .... _--.- . . . Table 4.7-3, Seismic 1.ca'Is on Primary SyGteta Cvg.pon2nto (cont'd) , teletire " Seteetc teed P m ent
~ ~- r e.or.t . . ,1, , e,
.e,- Per,,,
ent, , a te eenc ht erel 0eaien g ,,q,,,,,, ....,,,,.e,.....e., ,e ee.ie oesi.n
.....t .n m .. . . .....e c.icoietw m.nieu. oc ,
61 463.0 9 se (klpe') 217.5 265.2 90 90 $< ( toubined Beector vesset 34.4 392.0 1' c' 2 emi f twelet Sumart y (ktps) 11.1 16 54 y 7 138.4 304.0 e N (klpe) t14.4 Ft F4 g [ teetter vessel 142.4 692.0 Intet sumort W (ktpe) 117.3 SF 54 , 178.1 405.0 fy (hlpel 117.9 F9 79
' Steen Generator 44.0 397.0 Louse $wrt 8: thtps) 8' F9 79 g ,
pI S,149.8 24422.0 Itm (In htpel _5 9772.0 1 91 12
?&4.1 ser tin-ktpe) 7. . _
100 , 100 o 0.1 4132.0 " me (In-ktpel o.1 T3 71 g 69.8 240.0 ft (hlpe) 64 A Steam Gerwroter et 49 4 taper $ w it 30.6 ' 25.3 Ps (klpe) 24.6 72 T3 Presourtaer 82.9 22.9 t e st fy (ktpel 12.2 44 i 67 5,852.3 17101.$_ _ i l 5681.8 saa (In-ktps) 92 92 0.T 9.2 07 ' ty tklpe) heector Cootant Pump 22 26 1 verticai S qpert 19.5 .25.0 te.9 re ckspe) . seector cooient *w , morisertet s gport , g l The percent. below design was calculated by i
-2 Note: 100*(Specified for Design - Calculated) /Specified for Des gn. $
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ip I b, N. R. 4.7-33 S ._
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