ML17229A947
| ML17229A947 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/10/1998 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17229A948 | List: |
| References | |
| PSL-ENG-SEFJ-97, PSL-ENG-SEFJ-97-037, PSL-ENG-SEFJ-97-37, NUDOCS 9812220077 | |
| Download: ML17229A947 (65) | |
Text
PSL-ENG-SEFJ-97-037 Rev 0 Page 1 of 34 FLORIDAPOWER 4 LIGHTCO.
ST. LUCIEUNIT2 No Significant Hazards Evaluation Proposed License Amendment for Implementing Bounding Fuel Cycle Safety Analysis Report PSL-ENG-SEFJ-97-037 Rev 0 (Tracking ¹ 9S146)
Safety Related Nuclear Fuel Nuclear Engineering Department
PSL-ENG-SEFJ-97-037 Rev 0 Page 2 of 34 REVIEW AND APPROVAL RECORD ST.
LUCIE UNIT 02 TITLE Pro pseud License Amendment for Im lementin Boundin Fuel C cle Safet Anal sis Re ort LEAD DISCIPLINE Nucl ar Fue P L Fuel En ineerin ENGINEERING ORGANIZATION Nuclear En ineerin REVIEW/APPROVALI GROUP INTERFACE TYPE PREPARED VERIFIED INPUT REVIEW NlA FPL APPROVED APPROVED*
MECH ELECT IRC CIVIL DB CSI NUC FUEL r
X N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A
~ IP/f)I5 p~ yP Sk-
~ For Contractor Evals As Determined By Pro)acts
'eview Interface As A Nin On All 10CFR50.59 Evals and PLAs FPL PROJECTS APPROVAL:
DATE:
OTHER ZNTERFACES
1 ~
J
PSL-ENG-SEF J-97-037 Rev 0 Page 3 of 34 TABLEOF CONTENTS DESCRIPTION
.~Pa e 1.0 Introduction/Background 2.0 Description of Proposed Changes 3.0 Basis for Proposed Changes and Analyses ofEffects 4.0 Determination ofNo Significant Hazards 5.0 Conclusions 6.0 List ofAffected Documents 7.0 Verification Summary 8.0 References 28 30
.... 3 1
.... 3 1 31 ATTACHMENTS 1.
Marked-Up Technical Specifications Pages 2.
Safety Analysis Report for Initial Application of PAC and NPAC 19 250
PSL-ENG-SEFJ-97-037 Rev 0 Page 4 of 34
1.0 INTRODUCTION
/BACKGROUND:
Cycle 12 reload for St. Lucie Unit 2 will involve changes to the reload evaluation process to be implemented as part of the Nuclear Fuel Fabrication and Related Services Contract between Florida Power and Light Company and ABB-Combustion Engineering, Inc.
These changes will require amendment to the current Technical Specifications (TS).
The proposed changes to the reload evaluation process involves using conservative physics inputs to safety analyses that are expected to encompass a wide range of fuel management strategies so as to allow flexibilityin the future reload core designs.
This process is referred to here as the Reload Process Improvement (RPI).
The safety analyses affected by the RPI process and the proposed changes to Technical Specifications have been reanalyzed, taking into account the bounding physics parameters corresponding to various fuel management schemes.
The fuel management schemes considered include up to 24 month fuel cycles and the use of Gadolinia and Erbia burnable poison in the fuel rods. It is concluded that the safety analyses would continue to meet the applicable acceptance
- criteria, when considering the effects of the bounding physics parameters.
In evaluating future core designs, cycle specific physics input values willbe compared to these bounding values to determine the continued applicability of the safety analyses.
These parameters will be verified to bound the cycle specific values for every future fuel cycle and deviations, if any, will be addressed in cycle specific fuel reload documents.
Approval of the proposed TS changes willfacilitate the implementation of future core reloads under the provisions of 10 CFR 50.59.
The TS changes proposed in this license amendment for St. Lucie Unit 2 include replacing the TID-14844 thyroid dose conversion factors in TS 1.10 definition of DOSE EQUIVALENTI-131 with the dose conversion factors from ICRP-30, Supplement to Part 1.
From physics considerations, the SHUTDOWN MARGIN requirement limits in Modes 1, 2, 3, and 4 (TS 3.1.1.1) and Mode 5 (TS 3.1.1.2) are proposed to be relocated from TS to the Core Operating Limits Report (COLR).
Bases Figure B2.1-1, axial power distribution for Thermal Margin Safety LimitLines (TMSLL)is replaced with a new Figure B2.1-1. TMSLLFigure 2.1-1 remains unchanged and consistent with the axial shapes in the revised Figure B2.1-1, and the safety limits used in the RPI safety analysis report. The peak linear heat rate, specified as a safety limitin TS 2.1.1.2, is proposed to be deleted from the TS, consistent with NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants, as applied to analog protection systems.
Additional topical reports are included in the TS administrative Section 6.9.1.11 consistent with the methodology used in the RPI safety analyses report.
These methodologies are listed in the attached TS marked-up pages, and include no clad lift-offmethodology (CEN-372-P-A), the CEA withdrawal methodology (CEN-121(B)-P) and the core design methodology with Erbium (CENPD-382-P-A).
PSL-ENG-SEFJ-97-037 Rev 0 Page 5 of 34 The thermal margin calculations and the safety analyses are performed by Combustion Engineering, Inc. using Nuclear Regulatory Commission (NRC) approved computer codes and one supplement which is currently under NRC review.
This supplement pertains to the fuel performance methodology with gadolinia-urania burnable absorber rods (CENPD-275-P, Revision 1-P, Supplement 1-P).
The ABB-CE Setpoint Methodology Topical Report Supplement (CENPD-199-P Rev 1-P-A Supplement 2-P-A) is used in setpoint analysis where the Free Oscillation Technique is replaced with the Xenon Swing Technique in the generation of axial power shapes, and represents a change to the methods currently in use for St. Lucie Unit 2.
The fuel performance
- analysis, used in support of this evaluation, is based on the no clad lift-offmethodology documented in CEN-372-P-A. Both of these methods are approved by the NRC for generic use.
An analytical method change is implemented in relation to the use of the HERMITE code (CENPD-188-A), which, in one dimension, has been used to simulate the four pump loss of flow with an explicit space-time loss-of-flow model.
This model provides more accurate data on the hot channel response in a time dependent fashion.
The methodology described in CEN-289(A)-P is applied to calculate the rod bow penalty effects.
This methodology was previously approved for Arkansas Power &Light, ANO2. In particular, the rod bowing effects use the L /I dependence for the extrapolation of the 14 x 14 channel closure data.
This approach is justified for St. Lucie Unit 2 based on a comparative analysis of design features with respect to factors that influence rod bow.
For the CEA withdrawal (CEAW) event, the treatment of the change in integrated radial peak is modified in that the impact of the change in radial peak is explicitly calculated in place of the additive adjustment to the final overpower margin.
Also, the delta-T power trip, conservatively omitted previously, is credited in determining the most adverse case.
The revised method thus removes unnecessary conservatisms present in the previous analysis method as described in CEN-121(B)-P.
Applicable TS action statements and surveillance requirements are changed to be consistent with the above proposed TS changes.
The appropriate TS bases are also revised to reflect the proposed changes.
The bases for SAFETY LIMITS and DNB PARAMETERS continue to meet the minimum Departure from Nucleate Boiling Ratio (DNBR) design limitof 1.28 as defined by the CE-1 CHF correlation using the previously approved Extended Statistical Combination of Uncertainties (ESCU) methodology.
The analyses presented in this evaluation support the proposed TS and methods changes, and were performed with bounding physics inputs to cover a range of fuel management strategies for flexibilityin future cycle core reloads.
The analyses are intended to provide increased available margin with respect to the input parameters, so as to reduce the analysis rework due to variations in cycle specific inputs to the safety analyses.
This process has also resulted in the consequences
PSL-ENG-SEFJ-97-037 Rev 0 Page 6 of 34'f several events being calculated closer to the limits of acceptance criteria than existing Analysis of Record (AOR).
Re-analysis of certain design basis events has resulted in changes to event classification.
The feedwater line break event is classified as a very low probability event with Standard Review Plan (SRP) acceptance criteria for reactor coolant system (RCS) pressure of 120% of design pressure.
This is consistent with the analysis presented for St. Lucie Unit 2 Cycle 1. The CEA withdrawal event is categorized as a Required Overpower Margin (ROPM) event meeting the fuel centerline melt acceptance criteria.
A summary of the methodology changes is provided below.
First A lication for St. Lucie Unit 2 of Genericall A
roved Methodolo CEN-372-P-A (Fuel Performance) 2.
CENPD-382-P-A (Erbia Core Design)
CENPD-199-P Revision 1-P-A Supplement 2-P-A (Setpoint Methodology)
New Methodolo or Methodolo A
lication Chan e for St. Lucie Unit 2 CEN-121(B)-P (CEAW Methodology Previously approved for other CE NSSS)
CENPD-188-A (HERMITEST-LOF Application Previously approved for other CE NSSS)
CENPD-275-P, Revision 1-P, Supplement 1-P (Gadolinia Core Design-Under NRC Review)
CEN-289(A)-P (Rod Bow Effects Previously approved for ANO2, Arkansas Power 8c Light)
It is concluded in this evaluation that the proposed TS changes do not involve a significant hazards consideration.
Since this evaluation affects the plant safety analyses as related to the proposed TS changes and the RPI process, the evaluation has been classified as Safety Related.
PSL-ENG-SEF J-97-037 Rev 0 Page 7of 34
2.0 DESCRIPTION
OF PROPOSED CHANGES:
The proposed Technical Specifications changes are described below.
2.1 TS 1.10: DOSEE UIVALENTI-131 In the Definition of DOSE EQUIVALENTI-131, the reference for thyroid dose conversion factors is changed from "Table IIIofTID-14844" to "ICRP-30, Supplement to Part 1". The ICRP-30 dose conversion factors are used in the dose calculations reported in this evaluation.
2.2 Bases 2.1.1 REACTOR CORE FIGURE B2.1-1 AXIALPOWER DISTRIBUTIONFOR THERMAL MARGIN SAFETY LIMITS AND BASES 2.2.1 VARIABLE POWER LEVEL-HIGH Figure B2.1-1 is changed to reflect the impact of the RPI assumptions on the Fr values.
TS Figure 2.1-1 that is associated with Figure B2.1-1 remains unchanged.
The wording in the text of Bases 2.1.1 and Bases 2.2.1, related to the high power level trip setpoint and reference to the reactor protection system trip setpoint Table 2.2-1, is changed/corrected to reflect the actual analysis assumptions.
2.3 TS 2.1.1.2: PEAK LINEARHEATRATE This TS, specifying the peak linear heat rate limit corresponding to the fuel centerline melt, is deleted.
The fuel centerline melt criteria, however, willcontinue to be met by the safety analysis per the Bases 2.1.1.
2.4 TS 3/4.1.1.1: SHUTDOWN MARGIN-T,GREATER THAN200 F and Bases 3/4.1.1.1 SHUTDOWN MARGIN The SHUTDOWN MARGINlimitin this TS is moved to the COLR. This TS is changed to delete reference to "5000 pcm," and refer to COLR for the SHUTDOWN MARGINlimits.
Bases 3/4.1.1.1 is changed to delete "5000 pcm" and refer to COLR for the limitingvalue.
2.5 TS 3/4.1.1.2: SHUTDOWN MARGIN-T,LESS THANOR E UALTO 200 F and Bases 3/4.1.1.2 SHUTDOWN MARGIN The SHUTDOWN MARGINlimitin this TS is moved to the COLR. This TS is changed to delete reference to "3000 pcm," and refer to COLR for the SHUTDOWN MARGINlimits.
Bases 3/4.1.1.2 is changed to delete "3000 pcm" and refer to COLR for the limitingvalue.
PSL-ENG-SEF J-97-037 Rev 0 Page S of 34 2.6 TS 3.1.2.2 3.1.2.4 3.1.2.6 3.1.2.8 ACTION: REACTIVITYCONTROL SYSTEMS and Bases 3/4.1.2 BORATIONSYSTEMS The wording "at least 3000 pcm" in the ACTION statements of these TS is changed to "its COLR limit."
Bases 3/4.1.2 is changed to delete "3000 pcm" and refer to COLR for the limitingvalue.
2.7 TS 6.9.1.11:
CORE OPERATING LIMITSREPORT COLR a)
Specification 3.1.1.1, SHUTDOWN MARGIN -
T,z Greater Than 200 F
and specification 3.1.1.2, SHUTDOWN MARGIN - T,~ Less Than or Equal To 200 F, are added to the list of COLR specification limits listed in specification 6.9.1.11.a.
b)
The list of analytical methods that can be used to determine the core operating limits is expanded in specification 6.9.1.11.b to include the methods shown in Section 3.2.6.
2.8 Safet Anal sis Methodolo Chan es In addition to the proposed TS changes, the safety analysis reported in this evaluation includes the following changes from existing AOR relative to the application of analytical methods not previously (or generically) approved for use at St. Lucie Unit 2:
a)
The HERMITE code in one dimension is used when additional spatial detail is needed to model the transient behavior of the reactor core. In particular, the HERMITE code is used to simulate the four pump loss of flow with an explicit space time loss-of-flow model (ST-LOF).
This model provides a more accurate time dependent data for the hot channel response calculations. (see Section 3.6.2.1 of this report) b)
The rod bow penalty effects are calculated as described in CEN-289(A)-P.
The methodology described in CEN-289(A)-P was previously approved for Arkansas Power &
Light, ANO2.
In particular, the rod bowing effects are extrapolated from the 14 x 14 channel closure data using the L/I dependence.
This approach is justified for St. Lucie Unit 2 based on a comparative analysis of design features with respect to factors that influence rod bow. (see Section 3.4.2 of this report) c)
The analysis of CEA withdrawal (CEAW) event is performed as described in CEN-121(B)-
P.
However, the treatment of the change in integrated radial peak is modified in that the impact of the change in radial peak is explicitly calculated in place of the additive adjustment to the final overpower margin.
Also, the delta-T power trip, conservatively omitted previously, is credited in determining the most adverse case. (see Section 3.6.2.2 of this report)
The following supplement to ABB-CE's previously approved methodology is used in support of the analyses presented in this evaluation.
PSL-ENG-SEF J-97-037 Rev 0 Page 9 of 34 1.
CENPD-275-P, Revision 1-P, Supplement 1-P, "C-E Methodology for PWR Core Designs Containing Gadolinia-Urania Burnable Absorbers," June 1997 This Topical Report was submitted in June 1997 to the NRC for approval and is currently under review.
3.0 BASIS FOR PROPOSED CHANGES AND ANALYSISOF EFFECTS:
3.1 Introduction li This section contains a description of the analyses performed in support of the proposed Technical Specifications changes and the bounding values of input parameters.
The details of each analysis are in the attached Safety Analysis Report (SAR) (Attachment 2).
This Safety Analysis Report (Reference
- 1) documents the safety and setpoints analyses performed for initial application of the Physics Assessment Checklist (PAC) and Non-Physics Assessment Checklist (NPAC) for St.
Lucie Unit 2, in connection with the RPI process being implemented as part of the Nuclear Fuel Fabrication and Related Services Contract between Florida Power and Light and ABB-Combustion Engineering, Inc.
In the application of PAC and NPAC as part of the reload safety analysis, the relevant safety analysis input, including the physics input to safety, will be verified every cycle to conform with the requirements of setpoint and safety analysis checklists documented in the PAC and NPAC.
The PAC and the NPAC are generated consistent with the bounding analysis and the proposed Technical Specifications changes.
The attached SAR documents the bounding analysis, and serves as the basis to update the existing Updated Final Safety Analysis Report (UFSAR). The report also provides the basis for events not analyzed because they were not necessary to support the bounding safety analysis input.
The bounding analyses for several events are performed so as to result in consequences which are close to the acceptance limits.
The basis for the proposed TS changes is provided in Section 3.2.
Sections 3.3, 3.4, 3.5, 3.6, 3.7 and 3.8 provide a summary of the safety and setpoint analysis, and the methodologies used.
3.2 Basis for Pro osed Technical S ecifications Chan es 3.2.1 TS 1.10: DOSE E UIVALENTI-131 In the Definition of DOSE EQUIVALENTI-131, the reference for thyroid dose conversion factors is changed from "Table IIIofTID-14844" to "ICRP-30, Supplement to Part 1."
The derived limits in the International Commission on Radiological Protection (ICRP),
Publication 30 (Reference 2), represent more recent values for the thyroid dose conversion
PSL-ENG-SEFJ-97-037 Rev 0 Page 10 of 34
- factors, and incorporate the considerable advances in the state of knowledge of radionuclide limits for intakes.
These proposed ICRP dose conversion factors are consistent with the Federal Guidance Report No.
11 (Reference 3),
and the recommendations ofEnvironmental Protection Agency (EPA) 1987 guidance.
This change is also consistent with the corresponding definition in NUREG-1432.
The thyroid dose conversion factors from ICRP-30, Supplement to Part 1 are used in the analyses presented in the attached safety analysis report.
The consequences of these analyses are in compliance with SRP acceptance criteria for meeting 10 CFR 100 dose limits. For dose events that are not reanalyzed and for the Steam Generator Tube Rupture event, the doses are based on TID-14844 dose conversion factors and remain acceptable for the proposed change as they yield conservative results relative to the use of ICRP-30. The followingUFSAR events are analyzed for dose consequences:
Anticipated Operational Occurrences 1.
Inadvertent Opening of a Steam Generator Atmospheric Dump Valve or Safety Valve 2.
Loss of Offsite Power to the Station Auxiliaries B.
Postulated Accidents 1.
Steam System Piping Failures 2.
Feedwater Line Break 3.
Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft 4.
Control Element Assembly Ejection 5.
Pressurizer Pressure Decrease Events 6.
Small Primary Line Break Outside Containment 7.
Steam Generator Tube Rupture 3.2.2 Bases 2.1.1 REACTOR CORE FIGURE B2.1-1 AXIALPOWER DISTRIBUTIONFOR THERMAL MARGIN SAFETY LIMITS AND BASES 2.2.1 VARIABLE POWER LEVEL-HIGH The Bases Figure B2.1-1 is revised to reflect the bounding RPI analysis.
The combination of Fr and axial power shapes shown in the proposed Figure B2.1-1 are expected to bound the core designs of future fuel cycles.
The associated TMSLLremain valid for the revised axial shapes and peaking factors, and Figure 2.1-1 ofTS 2.1.1 remains unchanged.
The wording in the bases is changed to clarify the actual high power level trip value as 107%, which does not include the calibration and measurement uncertainties.
This change is consistent with Figure 2.1-1 Notation and with the analysis assumptions, which use a conservative trip setpoint value of greater than 107% to include the overall applicable uncertainties as a function of the types of analyses performed.
Also, reference to Table 2.1-1 is corrected to Table 2.2-1 which is the correct table defining the Reactor Protection System (RPS) trip setpoints.
PSL-ENG-SEF 5-97-037 Rev 0 Page 11 of 34 3.2.3 TS 2.1.1.2: PEAK LINEARHEATRATE The deletion from TS of the peak linear heat rate value is consistent with the corresponding specifications for analog protection systems in NUREG-1432.
There is no change to the analysis of record due to this proposed change.
Although the proposed change will eliminate the peak linear heat rate limit from the TS, the centerline melt limit remains applicable as a Specified Acceptable Fuel Design Limit(SAFDL). The safety analysis will continue to comply with the fuel centerline melt acceptance criteria, which is the basis for the derived peak linear heat rate limit. The Bases 2.1.1 specifically address the centerline melt as a SAFDL.
The analyses in the attached safety analysis report are performed consistent with Bases 2.1.1 for the fuel centerline melt criteria.
Since the centerline melt limit is fuel type and burnup dependent, specifying one numerical value as a centerline melt SAFDL becomes technically not feasible.
The safety analysis has in the past used in certain cases an event specific centerline melt SAFDL to determine fuel failures.
For UFSAR events, such as CEAW and CEA Ejection, the centerline melt limit has always been different from 22 kW/ft specified in TS 2.1.1.2.
The event specific limit, which deviated from TS 2.1.1.2, has been explicitlyjustified in the analysis of the respective events, performed consistent with the methodology previously reviewed and approved by the NRC. These analyses willcontinue to meet the intent of TS 2.1.1.2 that is captured in the Bases 2.1.1.
Since the Bases 2.1.1 remains applicable for the centerline melt as a SAFDL, the specification of centerline melt limit in TS 2.1.1.2 becomes redundant and not necessary.
The deletion of TS 2.1.1.2 is also justified based on the requirements of 10 CFR 50.36(c)(1)(i)(A) for safety limits for the following reasons:
a)
This TS does not provide limits on any process variable necessary to protect the integrity of RCS pressure boundary or fission product barriers that guard against the uncontrolled release of radioactivity.
b)
This TS does not provide any function to mitigate a design basis accident or transient involving integrity of a fission product barrier that guard against the uncontrolled release of radioactivity.
In addition, LCO 3.2.1 limits the peak linear heat rate to a more limiting value specified in the COLR. This LCO value is input as an initial condition in safety and setpoint analyses to ensure compliance with centerline melt criteria.
3.2.4 TS3/4.1.1.1:
SHUTDOWN MARGIN-T GREATERTHAN200 F Ec TS 3/4.1.1.2: SHUTDOWN MARGIN-T LESS THANOR E UALTO 200 F Bases 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN
PSL-ENG-SEFJ-97-037 Rev 0 Page 12 of 34 The proposed changes relocate the SHUTDOWN MARGINlimits to the COLR. Moving the SHUTDOWN MARGIN limits to the COLR provides the flexibilityto optimize the requirements based on cycle specific fuel management and design considerations, such as scram worth, burnable absorber loadings, soluble boron level, etc.
This approach is consistent with the concept of RPI implementation, which is intended to allow fuel management flexibility. There is no change to the SHUTDOWN MARGINvalue in this proposed amendment.
The SHUTDOWN MARGINrequirement for T,s greater than 200 F (Modes 1 through 4) during times late in cycle is determined by the results of the Steam Line Break analysis.
The specific late in cycle SHUTDOWN MARGIN requirement may vary from cycle to cycle since the scram worth and the power distribution may vary substantially from cycle-to-cycle.
Therefore, the proposed change will allow flexibility to accommodate cycle specific time-in-life SHUTDOWN MARGINrequirements under the provisions of 10 CFR 50.59, and thereby obviate the need for license amendment.
Specification 6.9.1.11.a requires core operating limits, such as SHUTDOWN MARGIN, to meet applicable limits of the safety analysis.
The SHUTDOWN MARGIN requirement for T,s less than or equal to 200 F (Mode 5) may vary from cycle-to-cycle based on cycle specific fuel management and design considerations.
The proposed change willallow flexibilityto accommodate cycle-to-cycle variations in SHUTDOWN MARGIN requirements necessary to meet the design basis under the provisions of 10 CFR 50.59, and thereby obviate the need for license amendment.
The changes proposed to the Bases 3/4.1.1.1 and 3/4.1.1.2 provide consistency with the proposed changes to the respective specifications.
The technical basis of the specifications remains unchanged.
TS 3.1.2.2 3.1.2.4 3.1.2.6 3.1.2.8 ACTION: REACTIVITYCONTROL SYSTEMS and Bases 3/4.1.2 BORATIONSYSTEMS Deleting the numerical value of SHUTDOWN MARGIN from these TS is an administrative change to ascertain consistency of the identified "Action" statements with the proposed SHUTDOWN MARGIN specifications.
The SHUTDOWN MARGIN in these "Action"statements is the value corresponding to specification 3.1.1.2 at 200 F that is proposed for relocation to the COLR in Section 3.2.4 above.
The changes proposed to the Bases 3/4.1.2 provide consistency with the proposed changes to the respective specifications "Action" statements.
The technical bases of the specifications and "Action"statements remain unchanged.
TS 6.9.1.11: CORE'OPERATING LIMITSREPORT COLR
PSL-ENG-SEFJ-97-037 Rev 0 Page 13 of 34 a)
This TS section lists specifications whose limits are'elocated to the COLR.
The addition of the specifications 3.1.1.1 and 3.1.1.2 to this list is justified based on the changes proposed in this license amendment.
b)
The methodologies included in this section of TS have been approved by the NRC for the appropriate applications.
The addition of methodologies to the current list in this section ofTS is described below.
40)
CEN-121(B)-P, "CEAW, Method of Analyzing Sequential Control Element Assembly Group Withdrawal Event for Analog Protected Systems,"
November 1979 41)
This methodology supports reclassification of CEA Withdrawal as an event protected by LCOs.
CEN-133(B), "FIESTA, A One Dimensional, Two Group Space-Time Kinetics Code for Calculating PWR Scram Reactivities," November 1979 This methodology updates the space time SCRAM reactivity calculations for several events and the space time calculation of heat flux during SCRAM for Loss ofFlow.
42) 43) 44)
CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 This methodology is the generically approved Extended Statistical Combination of Uncertainties (ESCU) methodology CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990 This methodology supports no Departure from Nucleate Boiling (DNB) propagation in fuel failure calculations in the Pre-Trip Steam Line Break, Post-Trip Steam Line Break and Seized Rotor events.
CENPD-183-A, "C-E Methods for Loss ofFlow Analysis," June 1984 This methodology supports the Loss of Flow event that is evaluated to confirm the Regulating CEA Insertion and Axial Shape Index COLR Limits, and SCRAM worth.
45)
CENPD-190-A, "C-E Method for Control Element Assembly Ejection Analysis," July 1976 This methodology supports the CEA Ejection event that is evaluated to confirm the Regulating CEA Insertion, Axial Shape Index and Linear Heat Rate COLR Limits.
PSL-ENG-SEFJ-97-037 Rev 0 Page 14 of 34 46)
CENPD-199-P, Rev.
1-P-A, Supplement 2-P-A, "C-E Setpoint Methodology", June 1998 This methodology supports the revised axial shape generation methodology, the elimination of Fmonitoring, the use of 3-D physics and thermal hydraulics, and the application of ABB-CE DNB analysis methodology to mixed core applications.
This methodology has been applied in the past for 10CFR50.59 applications and for specific license submittals.
The listed reference generically documents the approach and supports analysis of similar fuel designs from other vendors.
47)
CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers," August 1993 This methodology supports the fuel assembly designs with erbia burnable absorbers.
48)
CEN-396(L)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/KGfor St. Lucie Unit 2," November 1989 (NRC SER dated October 18,1991, Letter J. A. Norris (NRC) to J. H. Goldberg (FPL), TAC No. 75947) 49)
CENPD-269-P, Rev.
1-P, "Extended Burnup Operation of Combustion Engineering PWR Fuel," July 1984 The methodologies in Items 48 and 49 support extended burnup for St.
Lucie Unit 2. Item 49 is the generic topical, whereas Item 48 addresses the burnup extension to rod average burnup of 60 GWd/MTU specifically for St. Lucie Unit 2.
50)
CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2," December 1984 This methodology supports revised Rod Bow Penalty calculations based on measurement data and using L /I dependence.
This report was approved for Arkansas Nuclear One Unit 2 in 1985.
This methodology is used to calculate the rod bow penalty for DNBR limit.
51)
CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998
PSL-ENG-SEF J-97-037 Rev 0 Page 15 of 34 This methodology supports the use of ABB's S2M methodology for Small Break LOCA analysis and the methodology to address reactor coolant pump loop seal refillphenomena.
52)
CENPD-140-A, "Description of the CONTRANS Digital Computer Code for Containment Pressure and Temperature Transient Analysis," June 1976 This methodology supports the calculation of Containment High Pressure Trip time for Steam Line Break analysis.
53)
CEN-365(L), "Boric Acid Concentration Reduction Effort, Technical Bases and Operational Analysis," June 1988 (NRC SER dated March 13, 1989, Letter J. A. Norris (NRC) to W. F. Conway (FPL), TAC No. 69325)
This methodology supports the recalculation of SHUTDOWN MARGIN.
54)
DP-456, F. M. Stern (CE) to E. Case (NRC), dated August 19,
STN 50-470, "NRC SER Standard Reference System, CESSAR System 80," December 1975)
This reference supports the use of the SGN-III computer code to recalculate the time of Containment High Pressure Trip for Steam Line Break.
This is an acceptable code for calculating mass and energy releases for Steam Line Breaks per SRP 6.2.1.4.
3.3 Fuel Desi n Summa 3.3.1 Fuel Assembl Desi n
The mechanical design criteria for the fuel assembly, fuel rods and poison rods, that are used to evaluate fuel assembly mechanical performance are unchanged from criteria documented in the UFSAR and are satisfied for all fuel region designs.
The loading pattern of the fuel and poison rod assemblies, and the U23q enrichment in the fuel rods are established by the fuel management program.
The use of gadolinia and erbia as a poison is supported by References 7, 8 and 9. Reference 7 is currently under NRC review for reference as a topical report.
3.3.2 Thermal Desi n and Fuel Performance The thermal performance was evaluated for composite fuel pins (including gadolinia-urania or erbia-urania poison rods) that envelope the various anticipated fuel types using the NRC approved
PSL-ENG-SEFJ-97-037 Rev 0 Page 16 of 34 FATES3B version of the fuel evaluation model (References 4, 5 and 6).
The analysis included urania fuel; gadolinia-urania fuel with absorber concentrations of up to and including 8.0 weight percent gadolinia per References 7, 8; and erbia-urania fuel with absorber concentrations of up to and including 2.5 weight percent erbia per Reference 9.
The thermal performance analysis utilized a power history that envelopes the power and burnup levels of the peak pin at each burnup interval, from beginning of cycle (BOC) to end of cycle (EOC) burnups.
The burnup range analyzed was 0 to 60 MWD/kgU (References 10 and 11), a range in excess of that expected for future cycles at St. Lucie Unit 2.
The maximum fuel rod internal pressure was verified to remain below the maximum allowable gas pressure (Reference 12).
The effective power-to-centerline melt limit was determined to be in excess of the 22.0 kW/ftvalue currently assumed as the basis for the LHR LSSS Setpoints.
3.4 Thermal H draulic Anal sis 3.4.1 Steady state DNBR analyses of the bounding cycle design at the rated power level of 2700 MWt have been performed using the TORC computer
- code, the CE-1 Critical Heat Flux (CHF) correlation, the simplified TORC modeling methods, and the CETOP code described respectively in References 13 through 16, and 21.
The bounding cycle design is based on a core with all GUARDIAN fuel. The thermal hydraulic design for transition cycles is discussed in Section 3.4.3. The Extended Statistical Combination of Uncertainties (ESCU) methodology presented in Reference 17 was applied with St. Lucie Unit 2 specific data using the calculational factors and other uncertainty factors at the 95/95 confidence/probability level to define a design limitof 1.28 on the CE-1 minimum DNBR.
The bounding cycle DNBR limitincludes the following allowances:
1.
NRC specified allowances for TORC code uncertainty and CE-1 CHF correlation cross validation uncertainty as discussed in Reference 18.,
2.
NRC specified allowances for uncertainty associated with the CE-1 CHF correlation applied to CE HID grid fuel, discussed in Reference 19.
3.
Rod bow penalty as discussed in Section 3.4.2 below.
Bounding cycle assumptions willbe confirmed on a cycle by cycle basis.
3.4.2 Effects of Fuel Rod Bowin on DNBR Mar in Effects of fuel rod bowing on DNBR margin have been incorporated in the safety and setpoint analyses in the manner discussed in References 17, 20 and 22.
The bounding cycle analysis rod bow effects are calculated using the methodology described in Reference 22.
This methodology was previously approved for Arkansas Power & Light, ANO2.
In particular, the rod bowing effects use the L/Idependence for the extrapolation of the 14 x 14 channel closure data.
This approach is justified for St. Lucie Unit 2 based on a comparative analysis of design features with
PSL-ENG-SEF J-97-037 Rev 0 Page 17 of 34 respect to factors that influence rod bow.
Additional details on the justification of L/I extrapolation methodology for St. Lucie Unit 2 are presented in the attached SAR, Section 6.2.
The penalty used for this analysis, 1.2% on minimum DNBR, is valid for bundle burnups up to 31,700 MWD/MTU. This penalty is included in the 1.28 DNBR limit. For assemblies with burnups greater than 31,700 MWD/MTUsufficient margin is available due to the reduction in radial power peaking to offset potential rod bow penalties for these higher burnup assemblies.
Hence, the rod bow penalty based upon Reference 20 and the methodology in Reference 22 for 31,700 MWD/MTUis bounding for all assembly burnups expected for the bounding cycle.
3.4.3 Thermal H draulic Desi n for Transition C cles GUARDIAN fuel was first introduced at St. Lucie Unit 2 for the currently operating Cycle 11.
During transition cycles, when GUARDIAN and non-GUARDIAN fuel regions are present
, in the core, more favorable power distributions than those used in the bounding thermal-hydraulic design willcompensate for the effect of a reduction in inlet flow due to the GUARDIAN fuel assembly design.
Therefore, the bounding cycle design is expected to remain applicable to transition cycles. If the more favorable power distributions do not entirely compensate for the reduction in inlet flow to the GUARDIAN fuel for a transition cycle, an additional penalty will be applied to offset the uncompensated adverse effect due to the inlet flow reduction.
3.5
~OC An Emergency Core Cooling System (ECCS) performance analysis was performed for St. Lucie Unit 2 to demonstrate conformance to the ECCS performance acceptance criteria for light water nuclear power reactors. Analyses were performed for the limiting large and small Loss-of-Coolant Accident (LOCA) break sizes. In addition, a post-LOCA boric acid precipitation analysis was performed.
The proposed Technical Specifications changes do not affect the analyses in this section.
However, the analyses are performed to include the bounding physics input consistent with the RPI plan.
These analyses meet all the 10 CFR 50.46 acceptance criteria and are included in this submittal for documentation purposes as the new analyses of record.
These analyses are not required to support the proposed TS changes.
3.5.1 Lar e Break LOCA 3.5.1.1 Method of Anal sis The large break LOCA (LBLOCA)ECCS performance analysis was performed using ABB-CE's NRC-approved June 1985 Evaluation Model (Supplement 3-P-A to Reference 24). This is the same methodology used in the LBLOCA ECCS performance analysis documented in Section 15.6.6.1 of the St. Lucie Unit 2 UFSAR (Reference 23). Calculations were performed to evaluate fuel assembly designs currently in use or planned to be used, including the evaluation of both erbia
PSL-ENG-SEFJ-97-037 Rev 0 Page 18 of 34 and gadolinia burnable absorber fuel rod designs.
The analysis results are applicable to both axial blanketed and non-axial blanketed core designs.
The 0.6 DEG/PD (0.6 Double-Ended Guillotine in the reactor coolant Pump Discharge leg) break was analyzed. This was previously identified as the limiting break (i.e., the break that resulted in the highest calculated peak cladding temperature) of the LBLOCA break spectrum. The break spectrum included both guillotine and slot breaks ranging in size from a 100% double-ended break to a 40% double-ended break. The reactor coolant pump discharge leg was demonstrated to be the limiting break location in Reference
- 24. It is limiting because both the core flow rate during blowdown and the core reflood rate are minimized for this location. The'0.6 DEG/PD remains the limitingbreak for this analysis because there were no plant changes incorporated into the analysis that impact the hydraulic calculations sufficiently to change the limitingbreak size. The results of the analysis are applicable to an end-of-cycle coastdown to an indicated cold leg temperature of 535'F.
3.5.1.2 Results Conformance of the results to the ECCS performance acceptance criteria is summarized below:
Parameter Peak Cladding Temperature Maximum Cladding Oxidation Core-wide Cladding Oxidation Result 2150 'F 8.03 %
(0.99 %
Criterion 2200 'F 17%
1%
3.5.2 Small Break LOCA 3.5.2.1 Method ofAnal sis The small break LOCA (SBLOCA) ECCS performance analysis was performed using ABB's SBLOCA Evaluation Model (Reference 25). It is the same methodology used in the SBLOCA ECCS performance analysis documented in Section 15.6.6.2 of the St. Lucie Unit 2 UFSAR (Reference 23).
The analysis was performed for the 0.045 ft break in the reactor coolant pump discharge leg. The 0.045 ft break was.previously identified as the limiting break (i.e., the break that resulted in the highest calculated peak cladding temperature) of the SBLOCA break spectrum. It remains the limitingbreak for this analysis because there were no plant changes incorporated into the analysis that impact the hydraulic calculations sufficiently to change the limiting break size. The reactor coolant pump discharge leg is the limiting break location because it maximizes the amount of spillage from the safety injection system.
The analysis was performed with fuel design data that produce results applicable to fuel assembly designs currently in use or planned to be used, including erbia and gadolinia burnable absorber fuel rod designs.
The analysis results are applicable to both axial blanketed and non-axial blanketed core designs, and an end-of-cycle coastdown to an indicated cold leg temperature of535'F.
PSL-ENG-SEFJ-97-037 Rev 0 Page 19 of 34 ABB has demonstrated and the NRC has approved (Reference
- 26) the capability of the ABB SBLOCA evaluation model to simulate the phenomena associated with the loop seal clearing concern described in Reference 27.
3.5.2.2 Results Conformance of the results to the ECCS performance acceptance criteria is summarized below:
Parameter Peak Cladding Temperature Maximum Cladding Oxidation Core-wide Cladding Oxidation Result 2055'F 5.8%
<0.68%
Criterion 2200'F 17%
1%
3.5.3 Post-LOCA Boric Acid Preci itation 3.5.3.1 Method of Anal sis The post-LOCA boric acid precipitation analysis was performed using ABB's NRC-approved post-LOCA long term cooling (LTC) Evaluation Model (Reference 28). This is the same methodology used in the boric acid precipitation analysis documented in Section 15.6.6.3 of the St.
Lucie Unit 2 UFSAR (Reference 23). The objective of the analysis is to demonstrate that the boric acid concentration in the core is maintained below its solubility limit following the limiting large break LOCA. The analysis uses the BORON computer code (Appendix C of Reference 28).
A boric acid solubility limit of 27.6 wt. % is used in the analysis. This is the solubility limit of boric acid in saturated water at a pressure of 14.7 psia. The value of 14.7 psia for the core pressure following a large break LOCA, as compared to the current AOR value of 20 psia, introduces an additional conservatism for the boric acid solubility limit.
3.5.3.2 Results The results of the analysis demonstrate that the boric acid concentration in the core is maintained below the solubility limitwhen the simultaneous hot and cold leg safety injection flow is initiated between two and six hours after the start of the LOCA.
3.5.4 Conclusions An ECCS performance
- analysis, consisting of the reanalysis of the limiting LBLOCA and SBLOCA and the reanalysis of post-LOCA boric acid precipitation, was performed for St. Lucie Unit 2.
The calculations were performed up to the time when cladding temperatures
- decrease, after accounting for the effects of core geometry changes followingcladding rupture, and adequate core cooling was demonstrated.
The results of the analysis demonstrate conformance to the ECCS acceptance criteria at an initial peak linear heat generation rate of 13.0 kW/ft.
PSL-ENG-SEF J-97-037 Rev 0 Page 20 of 34 3.6 Non-LOCA Anal sis The Non-LOCA safety analysis was performed for Unit 2 utilizing core physics and plant parameters that are anticipated to be bounding values for future cycles.
The Design Bases Events (DBE) considered in the safety analyses (listed in Table 3.6-1) are categorized into two major groups similar to the Cycle 2 safety analysis: Moderate Frequency events, referred to herein as Anticipated Operational Occurrences (AOO), and Postulated Accidents.
The DBE's were evaluated with respect to one or more of the following criteria: Offsite Dose, Reactor Coolant System
- Pressure, Fuel Performance, and Loss of Shutdown Margin.
The DBE's chosen for analysis for each criterion are the limiting events with respect to that criterion.
The CVCS Malfunction event was analyzed to ensure that the operator has sufficient time to terminate the event before the pressurizer fills solid. Allevents were investigated to assure that they meet their respective criterion for a reactor thermal power rating of2700 MWt.
The feedline break analysis does not involve fuel failure, and has the same acceptance criteria for radiological consequences as the AOO's i. e., small fraction of 10CFR 100 limits. Since the assumptions used in the feedline break analysis conservatively apply to the AOO's, it is concluded that the radiological consequences of the feedline break analysis also bound the results of the AOO's.
For events that are predicted to have fuel failures, acceptable values for fuel failure were derived from the dose acceptance limits.
Cycle specific analyses will verify the acceptability of these derived fuel failures for a particular reload.
The iodine dose conversion factors used in the safety analyses are taken from ICRP-30.
Doses not recalculated in this safety analysis report, are based on TID-14844 and willcontinue to remain conservative relative to the use of ICRP-30.
3.6.1 Methods of Anal sis AOO's are analyzed to assure that Specified Acceptable Fuel Design Limits (SAFDL) for Departure from Nucleate Boiling (DNB) and fuel Centerline Melt (CTM) are not exceeded.
The methodology used by Combustion Engineering integrates transient analysis results with the calculation of Limiting Safety System Settings (LSSS) and Limiting Conditions for Operation (LCO). This methodology is described in References 29 and 30.
AOO's are divided into two categories for evaluation with respect to fuel performance.
1.
Design Basis Events Requiring Reactor Protection System (RPS) trips to assure that SAFDL's are not exceeded 2.
Design Basis Events for which RPS trips and/or Sufficient Initial Steady State Margin (preserved by the LCO's) are necessary to prevent exceeding the SAFDL's.
Transient analyses for the first category determine the values of parameters used as input to the LSSS calculation, e.g., the pressure bias and the transient power decalibration term are calculated for DBE's in this category and the limiting values of these parameters are used as input to the Thermal Margin/Low Pressure (TM/LP)trip calculator.
PSL-ENG-SE&'J-97-037 Rev 0 Page 21 of 34 Transient analyses for DBE's in the second category determine the values of parameters used as input to the LSSS and/or LCO calculations, e.g., Required Overpower Margin (ROPM).
3.6.2 Chan es to Methods of Anal sis 3.6.2.1 Use of HERMITE in S ace-Time One Dimensional Mode One change in analytical methods involves the use of the HERMITE code (Reference 35) in one dimension when additional spatial detail is needed to model the transient behavior of the reactor core. HERMITE has been previously approved for use on other plants to simulate the four pump loss of flow with an explicit space time loss of flow model (ST-LOF). There are no fundamental core or nuclear steam supply system (NSSS) differences between these units and St. Lucie Unit 2 that would require major changes in the application of the HERMITE code to St. Lucie Unit 2.
Minor differences are accounted for in the specific modeling of the St. Lucie Unit 2 loss of flow transient.
The primary benefit of the use of HERMITE is to provide more accurate data on the hot channel response in a time dependent fashion, versus a static approach.
The use of the HERMITE code, in conjunction with a thermal hydraulic code, provides the thermal margin requirements and time dependent DNBR response for various anticipated operational occurrences and postulated accidents.
A description of the ST-LOF neutronics method used for one-dimensional calculations in HERMITEcan be found in Reference 37.
The synthesis of the axial power distribution and the planar radial power peaking factors provides a conservative representation of the hottest fuel assembly during the LOF transient, including maximum three dimensional power peaking effects.
This technique yields a conservative prediction of the minimum DNBR that can occur as a result of the LOF transient.
Revised LOP Method In the Reference 31 methodology it is assumed that the hot channel normalized heat flux decay is equivalent to the core average normalized heat flux decay, and the axial heat flux distribution is constant in time. The minimum DNBR calculated in this methodology assumes no decay of the hot channel heat flux.
The ST-LOF method uses HERMITE to calculate the core power in one dimension (axially) directly from CEA position versus time. It is assumed that the hot bundle normalized power decay is equivalent to the core average normalized power decay prior to the insertion of the CEA's.
As the CEA's are inserted in the core, the planar radial peaking factors are increased so that the hot channel power decreases less rapidly than core average power for the rodded planes. The hot bundle and core average axial heat flux distributions are each time-dependent.
The minimum DNBR value calculated with the ST-LOF method is based on the decay heat flux calculated by HERMITE at the time of minimum DNBR.
Reference 31 describes both static and dynamic methods for computing the DNBR. The ST-LOF method uses the static thermal hydraulic method for calculating the DNBR as described in Reference 31, Appendix A, except that CETOP is used in place of COSMO.
I'
PSL-ENG-SEFJ-97-037 Rev 0 Page 22 of 34 InitialTl>ennal Margin St. Lucie Unit 2 has an analog protection system and is provided with a Departure from Nucleate Boiling Limiting Condition for Operation (DNB LCO), which is a defined acceptable operating space in terms of core power and axial power distribution represented by Axial Shape Index (ASI).
The resultant DNB LCO envelope is generated based upon an adverse set of postulated conditions, such as the value of RCS flow,,consistent with the limits in Technical Specifications.
The low RCS flow condition forms a basis for the operating space within the DNB LCO envelope and provides the minimum initial thermal margin, but is also benign in terms of the prediction of margin degradation when modeled with HERMITE due to the relative proximity of the core to saturation conditions.
In adapting the application of the ST-LOF methods for analog plants, the ST-LOF transient is performed at the 'base'onditions consistent with the DNB LCO envelope.
These conditions are defined as the most adverse indicated values allowed by the LCOs. The Required Overpower Margin (ROPM) is quantified at these conditions (as a function of ASI) and set aside in the DNB LCO itself.
The increase in thermal margin associated with being at conditions less adverse than those used in the generation of the DNB LCO envelope is quantified. This additional thermal margin is used as a credit against any increase in ROPM associated with the initial conditions being different than the 'base'onditions.
Ifthis thermal margin credit is not sufficient to compensate for the increase in ROPM, the ROPM values used in the DNB LCO analysis are appropriately increased.
In all cases sufficient thermal margin requirements are established as inputs to the setpoint analysis.
In each application of the HERMITE ST-LOF methodology, plant specific data (e. g., trip setpoints, RPS delay times, holding coil delay times and CEA motion characteristics) are utilized to determine the appropriate time dependent core response.
3.6.2.2 Reclassification of AOO's from Protection via RPS Tri s to RPS and/or LCO's The method for the analysis of the Uncontrolled CEA Withdrawal (CEAW) at Power has been modified.
The CEA withdrawal is analyzed such that a combination of RPS trips and initial thermal margin are used to assure that SAFDL's are not violated. The CEAW methodology used is consistent with that outlined in Reference 32 (previously approved for other CE NSSS), except for the following:
In the CEAW event, the treatment of the change in integrated radial peak was previously an additive adjustment to the final overpower margin. This is being modified in that the impact of the change in radial peak is explicitly calculated using a thermal hydraulics code.
Also, the delta-T power trip, which was conservatively omitted in Reference 32, is now credited in some cases in determining the most adverse case.
The revised method provides for a better representation of the plant configuration by removing unnecessary analysis conservatisms.
Other DBEs, analyzed with respect to fuel performance as specified in the SAR (Attachment 2, Table 8.0-4), have also been analyzed such that a combination of RPS trips and initial thermal margin are used to assure that SAFDLs are not violated. This change from prior analysis was done
t'
PSL-ENG-SEF J-97-037 Rev 0 Page 23 of 34 to optimize the thermal margin operating
- space, and remains consistent with the previously approved methodology.
3.6.3 Mathematical Models Plant response for Non-LOCA Events was simulated using the CESEC computer code (Reference 36). The STRIKINIIcomputer code (Reference 34) was used in the analyses of the CEA Ejection Event. As described previously, the HERMITEcode (Reference 35) in one dimension was used in the analysis of four pump loss of flow and can be used in other events where additional spatial detail is required.
Consistent with the original Cycle 1 analysis, the post trip steam line break analysis assumed negative reactivity credit due to the local heat up of the inlet fluid in the hot channel, which occurs near the location of the stuck CEA. This credit is based on three-dimensional coupled neutronic-thermal-hydraulic calculations performed with the HERMITE/TORC code (References 13 and 35).
Simulation of the fluid conditions within the hot channel of the reactor core and prediction of DNB was performed using the CETOP-D and TORC computer codes (References 13, 15 and 16).
Determination ofDNBR for the post trip return to power portion of the steam piping failure events is based on the correlation developed by R. V. McBeth (Reference 33) with corrections developed by Lee, consistent with the methodology employed in the Cycle 1 analysis.
3.6.4 In ut Parameters and Anal sis Assum tions The key parameters assumed in the transient analysis, and the specific initial conditions for each event are tabulated and documented in the SAR (Attachment 2).
Events were reanalyzed to account for the effects of changes in core physics and plant parameter values that are intended to bound anticipated changes in future cycles. The reanalyses use the thyroid dose conversion factors from ICRP-30.
Other analysis input changes include higher core inlet temperature, increase in primary safety valve tolerance, more negative moderator temperature coefficient and lower analysis value for low RCS flow trip setpoint than that used in the existing AOR.
The Reactor Protection System (RPS) and Engineering Safety Features Actuation System (ESFAS) instrumentation trip setpoints and delay times used in the analysis are provided in Table 8.0-7 of. In most cases, the analysis setpoint values in the table include uncertainties which are larger than those calculated for the instrumentation.
3.6.5 Anal sis Results and Conclusions The Non-LOCA analysis performed in support of the proposed TS and methodology changes, and also the bounding analysis input, meet all the safety analysis acceptance criteria. The radiological dose consequences, using the proposed thyroid dose conversion factors from ICRP-30, remain in compliance with acceptance criteria for satisfying 10 CFR 100'dose limits.
PSL-ENG-SEFJ-97-037 Rev 0
~Page 24 of 34 Since the analyses described in Attachment 2 used more adverse
- inputs, as compared to the analyses of record, they produce, in general, results that more closely approach the acceptance limits.
In the case of radiological consequences, results of several analyses are pushed to the desired dose limits (below the acceptance criteria) to define the acceptable limits for fuel failure.
For the feedline break event, the acceptance criteria used for peak RCS pressure is 120% of design pressure.
This is consistent with the SRP requirements and has been previously used for St. Lucie Unit 2 Cycle 1 feedline break analysis.
The analysis of CEA withdrawal event shows that the calculated increase in linear heat rate does not lead to the violation of fuel centerline melt limit.
Table 3.6-1 presents the acceptance criteria for each design basis event, which is either reanalyzed or evaluated, as appropriate, to ensure compliance with the respective acceptance criteria.
The details of each analysis and the results are provided in the attached SAR (Attachment 2).
PSL-ENG-SEF J-97-037 Rev 0 Page 25 of 34 Table 3.6-1 St. Lucie Unit 2, Design Basis Events Considered In The Safety Analysis (Section Numbers refer to those in the SAR ofAttachment 2)
~Dis osition
~Acce tence Section Sub-Section
~Descn tion Status Criteria 8.1 Increase in Heat Removal by the Secondary System'.1.1 8.1.2 8.1.3 8.1.4 8.1.5*
8.1.5a*
8.1.5b*
8.1.5c*
Decrease in Feedwater Temperature Increase in Feedwater Flow Increased Main Steam Flow Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve Steam System Piping Failures Inside Containment Pre-Trip Power Excursions Outside Containment Pre-Trip Power Excursions Post-Trip Analysis E
E A
E 3a, 3b 3a, 3b 3a, 3b la, 4a A
lc A
lc A
lc,4a 8.2 Decrease in Heat Removal by the Secondary System 8.2.1 8.2.2 8.2.3 8.2.4 8.2.5 8.2.6*
8.2.6a*
8.2.6b*
Loss of External Load Turbine Trip Loss of Condenser Vacuum Loss of Offsite Power to the Station Auxiliaries (LOAC)
Loss ofNormal Feedwater Feedwater Line Break Event Small Feedwater Line Break Event Feedwater Line Break Event with a Loss ofAC Bounded by 8.2.3 Bounded by 8.2.3 A
2a,2c E
la,2a Bounded by 8.2.3, 8.2.6 A
la,2a A
la,2b 8.3 Decrease in Reactor Coolant Flowrate 8.3.1 8.3.2 8.3.3*
Partial Loss ofForced Reactor Coolant Flow-Total Loss ofForced Reactor Coolant Flow Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft Bounded by 8.3.2 A
3a A
lc 8.4 Reactivity and Power Distribution Anomalies 8.4.1 8.4.2 8.4.3 8.4.4 8.4.5 8.4.6*
Uncontrolled CEA Withdrawal from a Subcritical or Low Power Condition Uncontrolled CEA Withdrawal at Power CEA Drop Event CVCS Malfunction (Inadvertent Boron Dilution)
Startup of an Inactive Reactor Coolant System Pump Event Control Element Assembly Ejection A
3a,3b A
2a,3a,3b A
3a,3b A
4a N/A A
lb,3b
PSL-E<NG-SEFJ-97-037 Rev 0 Page 26 of34 Table 3.6-1 (Cont'd.)
St. Lucie Unit 2, Design Basis Events Considered In The Safety Analysis (Section Numbers refer to those in the SAR of Attachment 2)
~Dis osition
~Acce tnnce Section Sub-Section
~Descri tion Status Criteria 8.5 Increase in Reactor Coolant System Inventory 8.5.1 CVCS Malfunction A
PZR Not Filling 8.5.2 InadvertentOperationoftheECCSDuringPowerOperation E
PZRNotFilling 8.6 Decrease in Reactor Coolant System Inventory 8.6.1 Pressurizer Pressure Decrease Events 8.6.2*
Small Primary Line Break Outside Containment 8.6.3*
Steam Generator Tube Rupture with a Concurrent Loss of Offsite Power A
lb,3a E
la A
lb 8.7 Miscellaneous 8.'7.1 Asymmetric Steam Generator Events A
3a Dis osition Status A
Reanalyzed E
Evaluated Acce tance Criteria 3
4 Acceptance la:
lb:
lc:
Acceptance 2a:
2b:
2c:
Acceptance 3a:
3b:
Acceptance 4a Criteria of Offsite Dose Small Fraction of 10 CFR 100 Limits Well Within 10 CFR 100 Limits Within 10 CFR 100 Limits Criteria of Pressurization RCS Pressure < 2750 psia RCS Pressure < 3000 psia Secondary Pressure < 1100 psia Criteria ofFuel Performance MDNBR> 1.28
< Fuel Centerline Melt Criteria ofShutdown Margin No Loss of Shutdown Margin
- Postulated Accidents
PSL-ENG-SEF J-97-037 Rev 0 Page 27 of 34 3.7 S~iA The Setpoint Analysis provides or confirms the Limiting Conditions for Operation (LCO), the Limiting Safety System Settings (LSSS) and the equipment setpoint requirements for St. Lucie Unit 2.
3.7.1 Method ofAnal sis The LCO and the LSSS limits are evaluated using the Setpoint Methodology (References 29 and 30), and the Extended Statistical Combination of Uncertainties Methodology (Reference
- 17) for the:
a.
Local Power Density LSSS (TS 2.2.1, Figures 2.2-1 &2.2-2) b.
Ex-core Detector Linear Heat Rate (LHR) LCO (TS 4.2.1.3) c.
Thermal Margin/Low Pressure LSSS (TS 2.2.1, Figures 2.2-3 &2.2-4) d.
Departure from Nucleate Boiling (DNB) LCO (TS 3.2.5)
The uncertainties for each of these limits are statistically combined.
The allowances for calibration and instrument drift, not included in the evaluation of setpoint limits, are applied as a penalty on the Technical Specification LSSS and LCO values when the equipment potentiometer settings are calculated.
This process ensures that the safety analysis remains in compliance with the Technical Specifications limits.
3.7.2 Results The LCO, LSSS and equipment setpoint requirements are verified to be in compliance with the design limits when each cycle design is established.
3.8 Other Dose Related Anal sis The doses in the analysis of record for the following were evaluated with respect to the proposed changes:
1.
Fuel Handling Accidents 2.
Control Room Doses 3.
Equipment Qualification It was determined that the analysis of record for the above would remain bounding when considering the effects of the proposed
- changes, including the replacement of TID-14844 with ICRP-30 for thyroid dose conversion factors.
PSL-ENG-SE<FJ-97-037 Rev 0 Page 28 of 34 4.0 DETERMINATIONOF NO SIGNIFICANTHAZARDS:
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 facilityin accordance with the proposed amendment would not:
1).
Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2).
Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3).
Involve a significant reduction in a margin of safety.
1). Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed amendment involves changes to the dose conversion factors used, in the thyroid dose calculations and the relocation of the SHUTDOWN MARGIN requirements for Modes 1
through 5 from TS to the Core Operating Limits Report (COLR).
Additionally, the peak linear heat rate value corresponding to centerline melt is deleted from the TS. The deletion of this TS remains consistent with the requirements of 10 CFR 50.36.
Bases Figure B2.1-1 is replaced with a new figure, consistent with the input assumptions of the safety analysis report.
The proposed amendment addresses analytical methods changes such as the use of HERMITE code in one dimensional mode for spatial
- details, the rod bow penalty calculations using L/I dependence discussed in CEN-289(A)-P, CEAW methodology change for crediting the delta-T power trip, and the methodology for core designs containing Gadolinia-Urania burnable absorbers (CENPD-275-P, Revision 1-P, Supplement 1-P).
None of these changes is a contributor to the initiation of previously evaluated accidents.
The changes to TS bases and the COLR methodology changes have no impact on the accident initiators. Accordingly, the probability of an accident previously evaluated is not significantly increased.
The proposed changes have been evaluated by Florida Power 8c Light (FPL) and Asea Brown Boveri - Combustion Engineering (ABB-CE).
The safety analyses assumed bounding physics parameters, and satisfy all the applicable acceptance criteria. Although specification 2.1.1.2 is deleted from TS, the safety analyses continue to meet the same centerline melt acceptance criteria as before and from which the peak linear heat rate value is derived.
Additionally, the peak linear heat rate value (corresponding to the centerline melt) does not meet the criteria specified in 10 CFR 50.36 for safety limits.
PSL-ENG-SEF J-97-037 Rev 0 Page 29 of 34 The changes to TS bases do not affect safety analysis results.
The relocation of SHUTDOWN MARGIN requirements to COLR does not affect analysis results or consequences as the limits remain unchanged.
Future changes to these limits will be controlled per Generic Letter 88-16 under the provisions of 10 CFR 50.59.
The use of HERMITE code in one dimension, for space-time loss-of-flow simulation, has been successfully applied for other ABB-CE plants.
The use of HERMITE code in this mode, for St. Lucie Unit 2, is acceptable since there are no fundamental core and nuclear steam supply system (NSSS) differences between St. Lucie Unit 2 and these plants.
The analyses presented in this submittal include the use of a supplement to the gadolinia-urania core design methodology topical report. The change in the rod bow penalty effects similar to that approved for another ABB-CE plant is justified for St. Lucie Unit 2 based on a comparative analysis of factors influencing the rod bow.
The change in the CEAW analysis method removes unnecessary conservatisms as compared to the previous analysis method.
The validity of results and conclusions of this evaluation are contingent upon NRC approval of these revised methods.
The radiological dose consequences for applicable safety
- analyses, using the dose conversion factors from ICRP-30, Supplement to Part 1, satisfy the acceptance criteria established to ensure compliance with the 10 CFR 100 dose limits.
The COLR methodology changes proposed to be listed in TS are those previously approved for CE plants with changes as described above.
The use of these methodologies remains consistent with their applicability for safety analyses.
Therefore, the proposed changes do not significantly increase the probability or consequences of an accident previously evaluated.
2). Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed amendment involves changes to the Technical Specifications for the dose conversion factors used in the thyroid dose calculations, the deletion of TS 2.1.1.2, the replacement of Bases Figure B2.1-1, and the relocation of SHUTDOWN MARGIN requirements to the COLR.
Additionally, there are methodology changes related to the safety analyses reported in this submittal.
The methodology changes include the use of HERMITE code in one dimensional mode for space-time loss-of-flow simulations, revised rod bow DNB penalty calculations, CEAW analysis methodology change including the use of delta-T power trip, and a supplement to the methodology for core designs containing Gadolinia-Urania burnable absorbers (CENPD-275-P, Revision 1-P, Supplement 1-P).
None of these changes, including those to the TS bases, willaffect the plant configuration and there willbe no impact on any system performance.
PSL-ENG-SEFJ-97-037 Rev 0 Page 30 of 34 Therefore, this amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3). Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
The proposed changes to the Technical Specifications have been evaluated with respect to the safety analyses using either previously approved methodology or methodology currently under NRC review (CENPD-275-P, Revision 1-P, Supplement 1-P). The use of HERMITE code in one-dimensional mode for spatial details, for space-time loss-of-flow simulation, provides more accurate data for thermal margin calculations and has been used for similar applications at other plants.
The calculations of rod bow DNB penalty using L /I dependence has been previously approved for another ABB-CE plant and is justified for St. Lucie Unit 2 based on an analysis of important factors influencing the rod bow. The CEAW methodology change showed acceptable analysis results after conservatively accounting for appropriate uncertainties.
The safety analyses performed with this methodology used bounding physics parameters to allow flexibility for future cycles core designs.
The revised Bases Figure B2.1-1 is consistent with the attached safety analysis report. Deleting TS 2.1.1.2 is justified since the specified limit does not meet any of the criteria of 10 CFR 50.36, and the fuel centerline melt criteria applied to the Specified Acceptable Fuel Design Limit (SAFDL) is not changed.
The setpoint analyses and safety analyses of all design basis accidents meet the applicable acceptance criteria with respect to the radiological consequences,
- SAFDLs, primary and secondary overpressurization, and 10 CFR 50.46 requirements. The proposed amendment, therefore, willnot involve a significant reduction in the margin of safety.
Based on the determination made above, it is concluded that the proposed amendment does not involve a significant hazards consideration.
5.0 CONCLUSION
S:
The proposed changes to the Technical Specifications have been evaluated by FPL and ABB-CE using bounding physics parameters to cover a wide range of fuel management schemes.
One supplement to the ABB-CE's current topical report (CENPD-275-P, Revision 1-P, Supplement 1-P), used in the safety analysis report, is currently under NRC review.
The results of the safety analysis are, therefore, contingent upon the NRC approval of this supplement.
The safety analyses have incorporated methodology changes related to the analysis of CEAW event, the use of HERMITE code in space-time one dimensional mode to simulate the four pump loss of flow transient behavior of the reactor core, and in the determination of rod bowing effects on the minimum DNBR.
l1 l\\
PSL-ENG-SEFJ-97-037 Rev 0 Page 31 of 34 The proposed TS changes have been evaluated for their impact on the design basis accidents.
Contingent upon approval of the revised methods (previously mentioned),
the safety analyses demonstrate that all the applicable acceptance criteria are met for all the accidents analyzed.
The proposed license amendment, therefore, does not result in a significant hazard.
6.0 LIST OF AFFECTED DOCUMENTS:
The followingdocuments are affected by the proposed change:
Technical Specifications FSAR DBD Plant Procedures The FSAR and the DBD change packages willbe prepared, as part of the Cycle 12 Reload PC/M process, subsequent to the NRC approval of the proposed TS changes.
The marked-up Technical Specifications pages are provided in Attachment 1.
7.0 VERIFICATION
SUMMARY
This engineering evaluation, prepared per ENG QIs 2.0 and 2.7 has been logically developed using the necessary design and reference materials to support the conclusions derived.
The proposed license amendment is based on the reanalyses of FSAR events to accommodate the proposed TS changes and the bounding physics parameters for flexibility in the future core reloads.
This engineering evaluation has been properly classified, per ENG QI 2.6, as Nuclear Safety Related.
8.0 REFERENCES
1 SL2-FE-0198, Rev 04, "St. Lucie Unit 2 Safety Analysis Report for Initial Application of PAC and NPAC, Revision 4," December 1998 (see Attachment 2)
ICRP Publication 30, Part I and Supplement to Part I, "Limitsfor Intakes of Radionuclides by Workers," Annals of the ICRP, Vol 2, No. 3/4 and Vol 3, No. 1-4, 1979, Pergamon Press Federal Guidance Report No.
11, "Limiting Values of Radionuclide Intake And Air Concentration and Dose Conversion Factors For Inhalation, Submersion, And Ingestion,"
Office of Radiation Programs, U.S. Environmental Protection Agency, Washington D.C.,
1988 4
CENPD-139-P-A, "Fuel Evaluation Model," July 1974 5
CEN-161(B)-P-A, "Improvements to Fuel Evaluation Model," August 1989
PSL-ENG-SEWJ-97-037 Rev 0 Page 32 of 34 10 12 13 14 15 16 17 18 19 20 21 22 CEN-161(B)-P, Supplement 1-P-A, "Improvements to Fuel Evaluation Model," January 1992 CENPD-275-P, Revision 1-P, Supplement 1-P, "C-E Methodology for PWR Core Designs Containing Gadolinia-Urania Burnable Absorbers," June 1997 (under review)
CENPD-275-P, Revision 1-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May 1988 CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers," August 1993 CEN-396(L)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/KGfor St. Lucie Unit 2," November 1989 (Approval SER dated October 18,1991, Letter J. A. Norris (NRC) to J. H. Goldberg (FPL), TAC No. 75947)
CENPD-384-P, "Report on the Continued Applicability of 60 MWD/kgU for ABB Combustion Engineering PWR Fuel," September 1995 CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990 CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986 CENPD-162-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1, Uniform Axial Power Distribution," April 1975 CENPD-206-P-A, "TORC Code, Verification and Simplified Modeling Methods," June 1981 CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981 CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 Letter, Robert A. Clark (NRC) to William Cavanaugh III(AP8cL), "Operation of ANO-2 During Cycle 2," July 21, 1981 (Safety Evaluation Report and License Amendment No. 26 for ANO-2)
NUREG-0787, Supplement 1, "Safety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3," Docket No. 50-382, October 1981 CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983 CENPD-207-P-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 2, Non-uniform Axial Power Distribution," December 1984 CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2,"
December 1984, Docket No. 50-368 Letter R. S. Lee (NRC) to J. M. Griffin (AP&L),Enclosure 2, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 66 to Facility Operating
PSL-ENG-SEF J-97-037 Rev 0 Page 33 of 34 23 24 25 26 28 29 30 31 32 33 34 License No. NPF-6, Arkansas Power & Light Company, Arkansas Nuclear One, Unit 2, Docket No. 50-368," May 7, 1985 St. Lucie Unit 2 Updated Final Safety Analysis Report, Amendment 11 CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model,"
August 1974 CENPD-132P, Supplement 1, "Calculational Methods for the C-E Large Break LOCA Evaluation Model," February 1975 CENPD-132P, Supplement 2-P, "Calculational Methods for the C-E Large Break LOCA Evaluation Model," July 1975 CENPD-132, Supplement 3-P-A, "Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," June 1985 CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model,"
August 1974 CENPD-137, Supplement 1-P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977 CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCAEvaluation Model," April 1998 Letter, L.A. Wiens (NRC) to T.F. Plunkett (FPL), "Request for Additional Information-Small Break Loss of Coolant Accident - St. Lucie, Unit 1 (TAC No. M97471)," April 4, 1997 CENPD-254-P-A, "Post-LOCA Long Term Cooling Evaluation Model," June 1980 CENPD-199-P, Rev. 1-P-A, "C-E Setpoint Methodology: C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems", January 1986 CENPD-199-P, Rev. 1-P-A, Supplement 2-P-A, "CE Setpoint Methodology", June 1998 CENPD-183-A, "C-E Methods for Loss ofFlow Analysis," June 1984 CEN-121(B)-P, "CEAW, Method of Analyzing Sequential Control Element Assembly Group Withdrawal Event for Analog Protected Systems," November 1979 McBeth, R. V., "An Appraisal of Forced Convection Burnout Data," Proc. Instn. Mech.
Engrs., Vol. 180, Pt 3C, PP 37-50, 1965-1966 CENPD-135P, "STRIKIN-II,A Cylindrical Geometry Fuel Rod Heat Transfer Program,"
August 1974 CENPD-135P, Supplement 2, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications)," February 1975 CENPD-135, Supplement 4-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," August 1976
PSL-ENG-SEFJ-97-037 Rev 0 Page 34 of 34 CENPD-135P, Supplement 5, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 35 CENPD-188-A, "HERMITE: A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," July 1976 36 37 Letter, A. E. Scherer, Enclosure 1-P to LD-82-001, "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," December 1981 CEN-133(B), "FIESTA, A One Dimensional, Two Group Space-Time Kinetics Code for Calculating PWR Scram Reactivities," November 1979
PSL-EN 6-SEF J-97-037 Rev 0, Page of 19 ATTACHMENT 1 Marked-Up Technical Specifications Pages
PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page ~
of 19 INDEX SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS SECTION
- 2. 1 SAFETY LIMITS PAGE 2.1.1 2.1.1.1 REACTOR CORE.........
0NBR..
~
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2"1 2-1
- 2. 1.2 REACTOR COOLANT SYSTEM PRESSURE----.--.-.-.-.............
2"1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETPOINTS......
~ ----
~
~ ~ ~ - ~
~
~
2 2 BASES SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE.....
- 2. 1.2 REACTOR COOLANT SYSTEM PRESSURE...----...--..
PAGE B 2-1 B 2-3 2.2 LIMITING SAFETY SYSTEM SE1TINGS
- 2. 2.1 REACTOR TRIP SETPOINTS.
ST.
LUCIE - UNIT 2
PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page 3 of 19 DEFINITIONS ICRP-30, Supplement to Part 1, pages 192-212, Tables entitled, "Committed Dose Equivalent in Target Organs or DOSE E UIVALEHT I-131 Tissues per Intake ofUnit Activity(Sv/Bq)."
1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/
gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I"133, I-134 and I-135 actually. present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table an II E - AVERAGE DISINTEGRATION ENERGY Z.H ~E.NT
- l. 11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for
- isotopes, other than iodines, with half lives greater than 15..minutes, making up at least 95K of the total non-iodine activity in the coolant.
ENGINEERED SAFETY FEATURES
RESPONSE
TIME 1.12 The ENGINEERED SAFETY FEATURES
RESPONSE
TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i. e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).
Times shall include diesel generator starting and sequence loading delays where applicable.
FRE UEHCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
. GASEOUS RADWASTE TREATMENT. SYSTEM
- 1. 14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:
a.
Leakage (except CONTROLLED LEAKAGE) into closed
- systems, such as pump seal or valve packing leaks that are captured, and conducted to sump or collecting tank, or b.
Leakag'e into the containment atmosphere from sources that are both specifically located and known either not to interfere with the ope'ration of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the secondary system.
ST.
LUCIE - UNIT 2 1-3
PSL-ENG-SEF3-97-037, Rev 0 Attachment 1
Page 4.
of 19 2.0 SAF=TY LIMITS AHO LrM TING SAFETY SYSTFM SETTINGS,
- 2. 1 SAFETY LIMITS
- 2. 1. 1 REACTOR CORE OHBR
- 2. 1. 1. 1 The combination of THERMAL POWER, pressurizer
- pressure, and maximum cold leg coolant temperature shall 'not exce d the limits shown on Figure 2. 1-1.
APPL:CABILITY:
MODES 1 and 2,'Ci'fOlt:
Whenever t e combinaticn of THERMAL POWER, pressurizer pressure and maximum cold leg coolant temperature has exceoded the limits shown on Figur 2.'=',
oe in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comoly with the requirements or Specification
- 5. 7. 1.
~ LET+.
P" 'EAR HEA; RATE 2.'.".2 The pea 'r heat rate of the
=uel shall be maintained le s
han or oual to 22 0 kwtiit corresponding o cente. line fuel melt).
APP'CABIL>TY:.
MODES and 2.
AC '.:Oil:
Whenever the peak linear hea-rate of the
=uel corresponding to contar!ine iuel melt),
be in c mply with the requirements or Speci ica icn nas axce dec 22.0
--'va'i e
HOT STANDBY within 1 nour,
- 6. i. l.
REACTOR COOLANT SYSTEM PRESSURE
.:.2 The Reac:or Coolant System pressure shal'i no- =xce d 27"0 ps',
AP.
r CABILrTYI'NODES 1, 2, 3, 4, and 5.
'Ci.Qg:
C ~
NOQES and 2
Whenever the Reactor Coolan; System oressure ras in HOT STANDBY with the Reactor Coolan vs em Gressu within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and c"molv with -i;e reouiromerits oi So axce4oec 2/
0 w I tn in ca- '.on os i", oe I lit I
~
I
~
NODES 3,
-" and
=.
Wnenever the reactor Ccolanit System
~ educe ne Reaction C oiarat
>ys am o ess" and comoly with -he requirements cr Soec-ro u
~
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F I
PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page 5 of 19 T
R The restrictions of this safety limitprevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting willoccur. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMALPOWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation. The CE-1 DNB correlation has been developed to predict the DNB heat fluxand the location of DNB for axially uniform and non-uniform heat fiux distributions. The local DNB heat fiux ratio, DNBR, defined as the ratio of the heat fluxthat would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to the DNB-SAFDLof 1.28 in conjunction with the Extended Statistical Combination of Uncertainties (ESCU). This value is derived through a statistical combination of the system parameter probability distribution functions with the CE-1 DNB correlation uncertainty.
This value. corresponds to a 95% probability at a 95% confidence level that DNB willnot occur and is chosen as ah appropriate margin to DNB for all operating conditions, The curves of Figure 2,1-1 show conservative loci of points of THERMALPOWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating forwhich the DNB-SAFDLis not violated for the familyof axial shapes and corresponding radial peaks shown in Figure B 2.1-1. The limits in Figure 2.1-1 were calculated for reactor coolant inlet temperatures less than or equal to 580'F. The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limitthe m 'm value of reactor inlet temperature.
Reactor operation at THERMALPOWER levels high r th n RATEDTHERMALPOWER is prohibitedby the high power level trip setpoint sp 'd in The area of safe operation is below and to the'eft of these lines.
l07 2-2-I O
The conditions forthe Thermal rgin Safety Limitcurves in Figure 2.1-1 to be valid are shown on the figure.
The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with LimitingConditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the SpecIed Acceptable Fuel Design Limits on DNB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operational Occurrences.
ST. LUCIE-UNlT2 B 2-1 Amendment No. 8, W 9
I Cn Dl I
C
-I
'.0 ale RL PLACE.
WITR gee P,CUgc 1.0 1.6 l.-
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1.01 2!i l'.)IL IJ0 I'l.'llCEtlf 0l'CTIVE COltE LEtlGTll fllOH 001TOH fiIj)0e 0 2.1-1 Axial )I<lwer Ilist.ri0ut.in0 for tl)coral u)argin safely lio)its
1.80 1.60 z0 1.40-m h
1.20 Cl I 1.00 0
0 0.80 g 0.60
~ 040.
0z 0.20 F~=1.77 F>= 1.60 Cfl tbRQ CO
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0.00 10 20 30 40 50 60 70 PERCENT OF ACTIVECORE HEIGHT FROM BOTTOM 80 90 100 Figure B 2.1-1 Axial Power Oistributlons for Themal Margin Safety Limits
0
F I
IT N
IMITIN F
Y PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page Q
of 19 V
I P
v I-Hi A Reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure - High or Thermal Margin/Low Pressure Trip.
The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.61% above the indicated THERMALPOWER level. Operator action is required to increase the trip setpoint as THERMALPOWER is increased.
The trip setpoint is automatically decreased as THERMALPOWER decreases.
The trip setpoint has a maximum value of 107.0% of RATED THERMALPOWER and a minimum setpoint of 15.0% of RATED THERMALPOWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMALPOWER level at which a trip would be actuated is
'o of RATED THERMALPOWER, which is the value used in the safety analysis.
Pre iiz rP r -Hi h The Pressurizer Pressure-High trip, in conjunction with the pressurizer safety valves and main steam line safety valves, provides Reactor Coolant System protection against ovetpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2375 psia which is below the nominal liftsetting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.
rrn IM i
wP The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than the DNB-SAFDLof 1.28, in conjunction with ESCU methodology.
The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of hT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIALSHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHALPOWER TILTand the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.
Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time, measurement uncertainties and processing error. The allowances include: a variable (power dependent) allowance to compensate for potential power measurement error; an allowance to compensate for potential temperature measurement uncertainity; an allowance to compensate for pressure measurement error, and an allowance to compensate forthe time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.
ST. LUCIE-UNIT2 B24 anondm8nt No.e, SB 89
141 R
Tl TY PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page C) of 19 NMA IN-T AN
'F 3.1.1.1 The SHUTDOWN MARGINshall be Xpsggg
~ $,2.
ggy~~
g4Csike Cz coLg. ItiM7p With the SHUTDOWNMARGIN,immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 1720 ppm boron or equivaient until the required SHUTDOWN MARGINis restored.
4.1.1.1.1 The SHUTDOWN MARGINshall be determined to be a.
Within one hour after detection of an inoperable CEA(s} and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.
Ifthe inoperable CEA is not fullyinserted, and is immovable as a result of excessive friction or mechanical interference or is known to be untrippable, the above required SHUTDOWN MARGINshall be verNed acceptable with an increased allowance forthe withdrawn worth of the immovable or untrippable CEA(s).
b.
When in MODE 1 or MODE 2 with Keffgreater than or equal to 1.0, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that CEA group withdrawal is within the Power Dependent Insertion Limits of Specification 3.1.3.6.
C.
When in MODE 2 with Keffless than 1.0, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> priorto achieving reactor criticalityby verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.
See Special Test Exception 3.10.1.
ST. LUCIE-UNIT2 3(4 1-1 Amendment No. 26,89
REACTIVITY CONTROL SYSTEMS PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page
)~
of 19 SHUTDOWN MARGIN - T LESS THAN OR E UAL TO 200 F
LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be APPLICABILITY:
MODE 5.
ACTION:
GU Si c
0 CQLg
[;YA,g xH~EPT I
With the SHUTDOWN MARGIN
, immediately initiate and continue boration at greater than or equa to 40 gpm ot: a solution containing greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.
SURVEILLANCE RE UIREMENTS 4.1.1.2 BR)+ Q b.
The SHUTDOWN MARGIN shall be determined to be
~a@i'n 54: coLg Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable CEA s and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.
If the inoperable.CEA 'is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:
l.
2.
3.
4.
5.
6.
Reacto~ coolant system boron concentration, CEA position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration,and Samarium concentration.
C.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, when the Reactor Coolant System is drained below the hot leg centerline, by consideration of the factors in 4. 1. 1.2b.
and by verifying at least two charging pumps are rendered inoperable by racking out theit motor circuit breakers'T, LUCIE - UNIT 2 3/4 1-3 Amendment No. p, 25
RHKTIYITYCGKOIL'- SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page
) )
of 19 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:
a.,
- One flow path from the boric acid tBakeup tank(s) with the tank meeting Specification 3.1.2.8 par't a) or b); via a boric acid makeup pump through a charging pump to the Reactor Coolant System.
b.
One flow. path from the boric acid. makeup tank(s) with the tank meeting Specification 3.1.2.8 part a) m" b)., via.a gravity feed valve.through, a charging pump.to,.the Reactor Coolant System.
c.
The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System.
'I OR At least two of the following three boron injection flow.paths shall be OPERABLE:
,a.
One flow path from each boric acid makeup tank with the
. combined tank contents meeting Specification 3.1.2.8 c),
via 4otti boric acid. makeup yuteps through a charging-pump to the Reactor Coolant System..
b.
One flow path from each boric acid makeup tank with the combined tank contents meeting Specification 3.1.2.8 c),
via both gravity feed valves through a charging pump to
~
the Reactor Coolant System; c.
The flow path from the refueling water storage tank, via a charging pump to the Reactor Coolant System.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
Amendment No. H,2$, 40 3/4 1-8 With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or. be in at least NOT STANOBY and berated to a SNUTOONN NARGIN equivalent to~Neo'.@ <'4+<
at 200.'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />;, restore't least two flow paths to OPERABLE status within the next 7 days 'or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
YB Colts Lien I.
ST.'UCIE - UNIT 2
REACTIVITYCONTROL SYSTEMS CHARGING PUMPS - OPERATING PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page
[2, of 19 3.1.2.4 At least two charging pumps shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBYand borated to a SHUTDOWN MARGINequivalent to t 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABL tus within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
tg CJD ~.. EMiI-4.12.4.1 4.1.2.4.2 At least two charging pumps shalt be demonstrated OPERABLE by verifying that each pump develops a flowrate of greater than or equal to 40 gpm when tested pursuant to the Inservice Testing Program.
At least once per 18 months verifythat each charging pump starts automatically on an SIAS test signal.
ST. LUCIE-UNIT2 3/41-10 Amendment No.e, &5~
REACTIVITYCONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - OPERATING PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page f3 of 19 3.1.2.6 At feast the boric acid makeup pump(s) in the bor n injection fiowpath(s) required OPERABLE pursuant to SpeciTication 3.1.2.2 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus ifthe flowpath through the boric acid pump(s) in Specification 3.1.2.2 is OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION'ith one boric acid makeup pump required forthe boron injection flowpath(s) pursuant to SpecTiication 3.1.2.2 inoperable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGINequivalent to t 200'F; restore the above required boric acid makeup pump(s) to OPERABL s atus within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
4.1.2.6 The above required boric acid makeup pump(s) shall be demonstrated OPERABLE by
~
~
~
verifying, that on recirculation flow, the pump(s) develop a discharge pressure of greater than or equal to 90 psig when tested pursuant to the Insewice Testing Program.
ST. LUCIE-UNIT2 Gf4 1-12 Amendment No. 8,95,4e, 91
REACTIVITY CONTROL SYSTEMS BORATED MATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page /Q of 19 a.
Boric Acid Makeup Tank 2A in accordance with Figure 3.1-1.
b.
Boric Acid Make'up Tank 2B in accordance with Figure 3.1-1.
c.
Boric Acid Makeup Tanks 2A and 2B with a minimum combined contained borated water volume in accordance with Figure 3.1-1.
d.
The refueling water tank with:
1.
A minimum contained borated water volume of 417,100 gallons, 2.
A boron concentration of between 1720 and 2100 ppm of boron, and 3.
A'solution temperature of between 55'F and 100'F.
APPLICABILITY:
MODES 1, 2, 3 and 4.
COL@
(,mi ACTION:
3.1.2.8 At least two.of the following four borated water sources shall be OPERABLE:
'a 0 Mith the above required boric acid makeup tank(s) inoperable, restore the tank(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the nex 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOMN MARGIN equivalent to at 200'F; restore the above required boric acid makeup tan s
to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
Mith the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ot'e in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE RE UIREMENTS 4.1.2.8 At least two required borated water sources shall be demonstrated OPERABLE:
a.
At least once per 7 days by:
1.
Verifying the boron concentration in the water and b.
2.
Verifying the contained borated water volume of the water source.
I At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWT temperature when the outside air temperature is outside the range of 55'F and 100"F.
c.
At least'once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the Reactor Auxiliary Building air temperature is less', than 55'F, by verifying that the boric acid makeup tank solution is greater than 55'F.
ST.
LUC IE - UNIT 2 3/4 1-14 Amendment Ho. H,2$
> 40
PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page
( Q of 19 3/4. 1 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1. 1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions,
- 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T' The most restrictive avg'ondition occurs 'at EOL, with T at no load operating temperature, and is avg associated with a postulated steam line break accident and resulting uncon-trolled RCS ooldown.
In the analysis of this accident, a minimum SHUTDOWN MARGIN is required to control the reactivity transient.
Accordingly, the SHUTDOWN HARGIN requirement is based upon this limiting condition *and is consistent with FSAR safety analysis assumptions.
At earlier times in core life, the minimum SHUTDOWN HARGIN required for the most restric-tive conditions is less than Wsth T
Iess than or quaI to BOO F, '[
avg the reactivity transients resulting from any postulated accident re minimal and a ~SOQQ-pisa'HUTDOWN NARGIN rovtdes de nate protectton.
aS ~Cia;m i'n lCe C.OLic'/4.1.1.3 B0R0N BILUTIQN (oy +~ci;~~ 'ai 1
i ~
Ko A minimum flow rate of at least 3000 gpm provides adequate
- mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor: Coolant System.
A flow rate of at least 3000 gpm will circulate an equivalent Reactor Coolant System volume of 103931'ubic feet in approximately 26 minutes.
The reactivity change rate associated with boron concentration reductions will therefore be within the capability of operator recognition and control.
3/4. 1. 1. 4 HOOERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle.
The surveillance requirements for measurement of the HTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concei1tration associated with fuel burnup.
The confirmation that the measured HTC value is within its limit provides assur ances that the coef" ficient will be maintained within acceptable values throughout each fuel cycle.
ST.
LUCIE - UNIT 2 B 3/4 1-1 Amendment No. 5,
REACTIVITY CONTROL SYSTEMS BASES PSL-ENGCEFJ-97-037, Rev 0 Attachment 1
Page
) g of 19 n
3/4.1.1.5 HINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 515'F.
This limitation is required to ensure (1) the moderator temperature coefficient is within its
, analyzed temperature
- range, (2) the protective instrumentation is within its normal operating
- range, (3) the pressurizer is capable. of being in an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTNDT temperature.
3/4.1.2 BORATION SYSTEMS The boron injection. system ensures that negative reactivity control is available during each mode of 'facility operation.
The components required to perform this function include (1) borated water sources, (2) charging
- pumps, (3) separate flow paths, (4) boric acid makeup
- pumps, and (5) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 200'F, a minimum of two separate and redundant'oron injection systems are provided to ensure single functional capability in the event an assumed fai lure renders one of the systems inoperable.
Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue'isk to overall facility safet f injection system failures during the repair period..
e i7YliQ The boration capability of either'ystem is ficient to provide a
C.OL<
SHUTDOWN MARGIN from expected operating conditions o
after xenon 'decay and cooldown to 200'F.
The maximum expected boration capa lity requirement occurs at EOL from full power equilibrium xenon conditions.
This requirement can be met for a range of boric acid concentrations in the Boric Acid Makeup Tank (BAHT) and Refueling Water Tank (RWT).
This range is bounded by 5350 gallons of 3.5 weight percent (6119 ppm boron) from the BAHT and 16,000 gallons of 1720 ppm borated water from the RWT to 8650 gallons of 2.5 weight percent (4371 ppm boron) boric acid from BAHT and 12,000 gallons of 1720 ppm borated water from the RWT.
A minimum of 35,000 gallons of 1720 ppm boron is required from the RWT if it is to be used to borate the RCS alone.
With the RCS temperature below 200'F one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperab YIJ}iri 1o QcoLK IiYn t.
J Ife The boron cape ility required claw 00'p is based upon providinp a
SHUTDOWN MARGIN after xenon decay and cooldown from 200'F to 140'F.
ss condition requires either 6750 gallons of 1720 ppm - 2100 ppm borated water from the refueling water tank or 3550 gallons of 2.5 to 3.5 weight percent boric acid solution from the b'oric acid makeup tanks.
S'f.
LUCIE - UNIT 2 B 3/4 1-2 Amendment No. H, 2F, 40
PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page
) '7 of 19 NNUA RADIOLOGICALENVIRONMENTALOP RATIN REPORT (Continued) 6.9.1.9 At least once every 5 years, an estimate of the actual population within 10 miles of the plant shall be prepared and submitted to the NRC.
6.9.1.10 At least once every 10 years, an estimate of the actual population within 50 miles of the plant shall be prepared and submitted to the NRC.
6.9.'1.11 COR OPERATING LIMITSREPORT COLR Moderator Temperature Coefficient Movable Control Assemblies - CEA Position Regulating CEA Insertion Limits Linear Heat Rate Total Planar Radial Peaking Factors - F'~
Total Integrated Radial Peaking Factor-F,'NB Parameters - AxialShape Index Refueling Operations - Boron Concentration b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described ln the following documents or any approved Revisions and Supplements thereto:
e 1.
WCAP-11596-P-A, "QualiTication of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Westinghouse Proprietary).
2.. NF-TR-9541, "Nuclear Physics Methodology for Reload Design of Turkey Point &St. Lucie Nuclear Plants," Florida Power & LightCompany, Januaty 1995.
a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR forthe t.tdse@TQA Specification 3.1.1.4 Specification 3.1.3.1 Spec Tiication 3.1.3.6 Specification 3.2.1 Specification 3.2.2, SpecTiication 3.2.3 Spec Tiication 3.2.5 Specification 3.9.1 3.
5.
6.
CENPD-199-P, Rev. 1-P-A, "C-E Setpoint Methodology: CE Local Power Density and DNB LSSS and LCO Setpoint Methodology forAnalog Protection Systems," Januaty 1986.
CENPD-266-P-A, 'The ROCS and DITComputer Code for Nuclear Design,"
April 1983.
CENPD-275-P, Revision 1-P-A, "C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbeis," May 1988.
CENPD-188-A, "HERMITE: A Multi-Dimensional Space - Time Kinetics Code for PWR Transients," July 1976.
ST. LUCIE-UNIT2 6-20 Amendment No. 08;Bio-,H; 92
PSL-ENG-SEFJ-97-037, Rev 0 Attachment 1
Page
) g of 19 CO PERATING LIMIT EP RT COLR (Continued) b.
(Continued) 34.
Letter, J.A. Norris (NRC) to J.H. Goldberg (FPL), "St. Lucie Unit 2-issuance'f Amendment Re: Moderator Temperature Coefficient (TAC No.
M82517)," July 15, 1992.
35.
Letter, J.W. Williams, Jr. (FPL) to D.G. Eisenhut (NRC), "St. Lucie Unit No. 2, k kd..dd dd d
.2L d."
L-84-148, June 4, 1984.
36.
Letter, J.R. Miller(NRC) to J.W. Williams, Jr. (FPL), Docket No. 50-389, Regarding Unit 2 Cycle 2 License Approval (Amendment No. 8 to NPF-16 and SER), November 9, 1984 (Approval of Methodology contained in L-84-1 48).
37.
Letter, A.E. Scherer Enclosure 1-P to LD42401, "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System,"
December 1981.
38.
Safety Evaluation Report, "CESEC Digital Simulation of a Combustion Engineering Steam Supply System (TAC No.: 01142)," October 27, 1983.
39.
CENPD-282-P-A, Volumes 1, 2 and 3, and Supplement 1, "Technical Manual for the CENTS Code," Februaiy 1991, Februaiy1991, October 1991, and I
1 8 ~ ~ ~7 8
June 1993, respectively.
c.
The core operating limits shall be determined such that ail applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period speciTied for each report.
ST. LUCIE-UNIT2 Amendment No. 92
PSL-ENG-SEF J-97-037 Rev 0,
Page
$ 9 of 19 Insert A Specification 3.1.1.1 Specification 3.1.1.2 Insert B Shutdown Margin T,s Greater Than 200 F Shutdown Margin-T,sLess Than or Equal to 200 F 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 CEN-121(B)-P, "CEAW, Method of Analyzing Sequential Control Element Assembly Group Withdrawal Event for Analog Protected Systems," November 1979 CEN-133(B), "FIESTA, A One Dimensional, Two Group Space-Time Kinetics Code for Calculating PWR Scram Reactivities," November 1979 CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990 CENPD-183-A, "C-E Methods for Loss ofFlow Analysis," June 1984 CENPD-190-A, "C-E Method for Control Element Assembly Ejection Analysis," July 1976 CENPD-199-P, Rev. 1-P-A, Supplement 2-P-A, "CE Setpoint Methodology", June 1998 CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers," August 1993 CEN-396(L)-P, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/KG for St. Lucie Unit 2," November 1989 (NRC SER dated October 18,1991, Letter J. A. Norris (NRC) to J. H. Goldberg (FPL), TAC No. 75947)
CENPD-269-P, Rev. 1-P, "Extended Burnup Operation of Combustion Engineering PWR Fuel," July 1984 CEN-289(A)-P, "Revised Rod Bow Penalties for Arkansas Nuclear One Unit 2,"
December 1984 CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCAEvaluation Model," April 1998 CENPD-140-A, "Description of the CONTRANS Digital Computer Code for Containment Pressure and Temperature Transient Analysis," June 1976 CEN-365(L),
"Boric Acid Concentration Reduction Effort, Technical Bases and Operational Analysis," June 1988 (NRC SER dated March 13,1989, Letter J. A. Norris (NRC) to W. F. Conway (FPL), TAC No. 69325) ill DP-456, F. M. Stern (CE) to E. Case (NRC), dated August 19, 1974, Appendix 6B to CESSAR System 80 PSAR (NRC SER, NUREG-75/112, Docket No. STN 50-470, "NRC SER Standard Reference System, CESSAR System 80," December 1975)
PSL-ENG-SEF J-97-037 Rev 0, Page g
of 250 ATTACHMENT 2 St. Lucie Unit 2 Safety Analysis Report For InitialApplication ofPAC and NPAC