ML20198T325
| ML20198T325 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/18/1997 |
| From: | Schnepp R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20198T286 | List: |
| References | |
| EMF-96-180, EMF-96-180-R01, EMF-96-180-R1, NUDOCS 9801270030 | |
| Download: ML20198T325 (107) | |
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SIEMENS EMF 96180 Revision 1 i
Quad Cities Unit 2 Cycle 15 Plant Transient Analysis
,rq April 1997 i'
I x..s I Siemens Power Corporation Nuclear Division
Siemens Power Corporation Nuclear Division i
EMF 96180 ,
Revision 1 t
issue Date: 4/18/97 Quad Cities Unit 2 Cycle 15 ,
Plant Transient Analysis Prepared by:
/ /
R. R. Schnepp, Engineer BWR Safety Analysis Nuclear Engineering Analyses Performed by:
D. M. Knee R. R. Schnepp A. W. Will April 1997 pa) b g
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Customer Disclaimer important Notice Regarding Contents and Use of This Document Messe Mead CereMiy Siemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the agreement between Siemens Power Corporation and the Customer pursuant to which this document is issued.
Accordingly,3xcept as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person actino on its behalf;
- s. makes any warranty or representation, express or imp'ied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, me+, hod or process disclosed in this document will not infringe privately owned rights; or
- b. assumes any liabilities with respect to the use of, or for dama0es resulting from the use of, any information, apparatus, method, or process disclosed in this document.
The information contained herein is for the sole use of the Customer.
In order to avoid impairment of ri0 hts of Siemens Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writmo by Siemens Power Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension therect, unless expressly provided in the Agrnment. No rights or licenses in or to any patents are implied M the furnishing of this document.
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EMF 96180 :
Revision 1 !
Pagei Nature of Changes j T,v e . Quad Cities Unit 2 Cycle 15 Plant Transient Analysis
< Superseded issue: 0 :
M PaneNo. Nature of Channe 1 21,26,45, Updated MCPR safety limit results consistent with Reference 23.
414,427, 428,429, 6 15 2 26 Updated MCPR operating limits consistent with the increase in MCPR safety limits.
3 66,67,69, Updated reduced flow MCPR limits consistent with the increase in 610,611, MCPR safety limits.
612,613, 614,615 4 82 Updated References 14,15,18 and 23.
The NRC identified a nonconformance related to the additive constant uncertainty for ATRIUM
- 9B fuel used in the 13CPR safety limit analysis. The approach and revised MCPR safety limit described in Reference 23 have been incorporated into this report.
NOTE: The changed items are denoted by a vertical bar ( l ) in the left hand margin.
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I EMF.06180 Ravision 1 Pageli i
Table of Contento i
Section PJtat
1.0 INTRODUCTION
...........................................11 2.0
SUMMARY
...............................................21 3.0 DISPOSITION OF EVENTS ....................................31 4.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN . . . . . . . . . . . . . . . . . . . . . 41 4.1 De sign B a sis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 4.2 C alculational Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 4.3 Anticipated Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 4.3.1 Load Rejection No Bypass / Turbine Trip No Bypass . . . . . . . . 42 4.3.2 Feed water Controller Failure . . . . . . . . . . . . . . . . . . . . . . . . 4 3 4.3.3 Loss of Feedwater Heating ........................44 4.4 M C PR $ a f ety Limit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 4.5 Nuclear Instrument Response . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 5.0 MAXIMUM OVERPRESSURIZATION ANALYSIS . . . . . . . . . . . . . . . . . . . , . 51 -
G.1 De sign Ba sis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 5.2 Pressurization Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 5.3 Closure of All Main Steam isolation Valves ...................52 6.0 AN ALYSIS AT OFF RATED CONDITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6.1 Reduc ed Core Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6.2 Reduc ed Core Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 ,
6.2.1 Automatic Flow Control . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 6.2.2 Manual Flpw Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 7.0 EVALUATION OF EOD/EOOS CONDITIONS . . ...................71 7.1 Final Feedwater Temperature Reduction .....................,71 7.2 C oa s t d o w n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1 7.3 Combined Final Feedwater Temperatute Reduction /Coastdown , . . . . 72 7.4 Feedwater Heater (s) Out of Service . . . . . . . . . . . . . . . . . . . . . . . . . 72 7.5 Combined Feedwater Heaters Out of Service /Coastdown .........72
8.0 REFERENCES
.. ...........................................81 APPENDIX A Margin to Unpiped Safety Valves . . . . . . . . . . . . . ! . . . . . . A 1
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EMF 96180 l Revision 1 Page iii :
List of Tables 11td.t .P.ng 2.1 Quad Cities Unit 2 Cycle 15 Base Case ACPRs at Rated Power With T S S S i n s e rt io n T!m e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 2.2 Quad Cities Unit 2 Cycle 15 Base Case ACPRs at Rated Power With NSS insertion Times .........................................25 2.3 Quad Cities Unit 2 Cycle 15 Thermal Margin Summary . . . . . . . . . . . . . . . . 2 6 2.4 Quad Cities Unit 2 Cycle 15 Results of Plant Transient Analysis With ,
TS S S inse rtion Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 2.5 Quad Cities Unit 2 Cycle 15 Results of Plant Transient Analysis With NSS insertion Yimes ........................................29 ,
2.6 EOD and EOOS Operating Conditions ...........................211 3.1 Quad Cities Unit 2 Cycle 15 Evaluation of Plant Parameter Changes on Disposition of Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.1 Quad Cities Unit 2 Design Reactor and Plant Conditions ...............47 4.2 Quad Cities Unit 2 Significant Parameter Values Used in Analysis ........48 4.3 Control Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 11 4.4 Quad Cities Unit 2 Cycle 15 Comparison of LRNB, TTNB and FWCF Results With TSSS Insertion Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 412 4.5 Quad Cities Unit 2 Cycle 15 Comparison of LRNB, TTNB, and FWCF Results With NSS Insertion Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 13 4.6 Input for MCPR Safety Lie Analysis ...........................414 5.1 Base Case Quad Cities Unit 2 Cycle 15 Results Summary of ASME Overpressurization Analyses With TSSS Insertion Times . . . . . . . . . . . . . . . 53 6.1 Automatic Flow Control Excursion Path . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 6,2 Reduced Flow MCPR Limits for Automatic Flow Control ' Base Case OLMCPR) ................................................66 6.3 Reduced Flow MCPR Limits for Automatic Flow Control (EOD/EOOS OLMCPR) ................................................67 6.4 Manual Flow Control Excursion Path . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 6.5 Reduced Flow MCPR Limits for Manual Flow Control .................69 6.6 Flow Dependent MCPR Results GE9 Fuel (Penalty Not included) . . . . . . . . 6 10 6.7 Flow Dependent MCPR Results GE10 Fuel (Penalty Not included) .......611 6.8 Flow Dependent MCPR Results ATRIUM 9B Offset Fuel ..............612
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EMF 96180 '
Revision 1 Page iv
- List of Tables (Continued) lihlt EAat 7.1 Quad Cities Unit 2 Cycle 15 Final Feedwater Temperature Reduction MCPR Results and Comparison to Base Case (TSSS Insertion Times) . . . . . . . 73 7.2 '
Quad Cities Unit 2 Cycle 15 Final Feedwater Temperature Reduction . ,
MCPR Results and Comparison to Base Case (NSS Insertion Times) . . . . . . . 7 4 7.3 Quad Cities Unit 2 Cycle 15 Final Feedwater Temperature Reduction i ASME Overpressurization Analysis Results and Comparison to Base Case
. (T S S S i n sertion Time s) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 5 7.4 Quad Cities Unit 2 Cycle 16 Coastdown Operation MCPR Results and Comparison to Base Case (TSSS Inserticn Times) . . . . . . . . . . . . . . . . . . . . 7 6 7.5 Quad Cities Unit 2 Cycle 15 Coastdown Operation MCPR Results and Comparison to Base Case INSS Insertion Tirnes) . . . . . . . . . . . . . . . . . . . . . 7 7 7.6 Quad Cities Unit 2 Cycle 15 Coastdown Operation ASME Overpressurization Analysis Results and Comparison to Base Case (T SS S insertion Times) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 8 7.7 Quad Cities Unit 2 Cycle 15 Combined FFTR/Coastdown MCPR Results and Comparison to Base Case (TSSS Insertion Times) . . . . . . . . . . . s . . . . . . 79 7.8 Quad Cities Unit 2 Cycle 15 Combined FFTR/Coastdown MCPR Results and Comparison to Base Case (NSS Insertior. Times) . . . . . . . . . . . . . . . . 7 10 7.9 Quad Cities Unit 2 Cycle 15 Combined FFTR/Coastdown ASME Overpressurization Anslysis Results and Comparison to Base Case (TS S S insertion Times) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 11 7.10 Quad Cities Unit 2 Cycle 15 Feedwater Heater Out of Service MCPR Results Comparison to Base Cote (TSSS Insertion Times) . . . . . . . . . . . . . . 7 12 7.11 Quad Cities Unit 2 Cycle 15 Feedwater Hester Out of Service MCPR Results and Comparison to Base Case (NSS Insertion Times) . . . . . . . . . . . . 7 13 7.12 Quad Cities Unit 2 Cycle 15 Feedwater Hester Out of Service ASME Overpressurization Analysis Results and Comparison to Base Case (TS S S insertion Times) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 14 7.13 Quad Cities Unit 2 Cycle 15 Combined FHOOS/Coastdown MCPR Results and Comparison to Base Case (TSSS Insertion Times) . . . . . . . . . . . . . . . . . 7 15 7.14 Quad Cities Unit 2 Cycle 15 Combined FHOOS/Coastdown MCPR Results and Comparison to Base Case (NSS Insertion Times) . . . . . . . . . . . . . . . . 7 16
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4 CMF 96180 Revision 1 Page v List of Fioures Fioure P. ant t
1.1 Quad Cities Unit 2 Operating Power / Flow Map . . . . . . . . . . . . . . . . . . . . . . . 1 'J 4.1 Load Rejection No Bypass at 100/108- Key Parameters . . . . . . . . . . . . . . 4 15 4.2 Load Rejection No Bypass at 100/108 Vessel Water Level .... . . . . . . 4 16 4.3 Load Rejection No Bypass at 100/108 Vessel Pressure Response . . . . . . . 417 4.4 Load Rejection No Bypass at 100/108- Safety / Relief Valve Flows . . . . . . . 418 4.5 Turbine Trip No Bypass at 100/108- Key Parameters . . . . . . . . . . . . . . . . 4 19 4.6 Turbine Trip No Bypass at 100/108 Vessel Water Level .............420 4.7 Turbine Trip No Bypass at 100/108 Vessel Pressure Response . . . . . . . . 4 21 4.8 Turbine Trip No Bypass at 100/108- Safety / Relief Valve Flows . . . . . . . . . 4 22 4.9 Feedwater Controller Failure at 100/108- Key Parameters . . . . . . . . . . . . 4 23 4.10 Feedwater Controller Failure at 100/108- Vessel Water Level . . . . . . . . . . 4 24 4.11 Feedwater Controller Failure at 100/108 Vessel Pressure Response . . . . . 4 25 4.12 Feedwater Centroller Failure at 100/108 Safety / Relief Valve Flows . . . . . 4 26 4.13 Radial Power Distribution for SLMCPR Deterrnination . . . . . . . . . . . . . . . . 4 27 '
4.14 Quad Cities Unit 2 Cycle 15 Safety Limit Local Peaking Factors With Channel Bow at Assembly Exposure of 22,500 mwd /MTU (SPCA9 372811GZH.ADV) ..................................428 4.15 Quad Cities Unit 2 Cycle 15 Safety Limit Local Peaking Factors With Channel Bow at Assembly Exposure of 20,000 mwd /MTU (SPCA9 358B 11GZL.ADV) ..................................429 5.1 MSIV Closure at 100/10,8- Key Paramet ers . . . . . . . . . . . . . . . . . . . . . . . . 5 4 5.2 MSIV Closure at 100/108 Vessel Pressure Response ................55 5.3 MSIV Closure at 100/108- Vessel Water Level . . . . . . . . . . . . . . . . . . . . . 5 6 5.4 MSIV Closure at 100/108 - Safety Valve Flow Rates .................57 6.1 Reduced Flow MCPR Limit for Automatic Flow Control (Bas e Ca se O L M C PR) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 13 6.2 Reduced Flow MCPR Limit for Automatic Flow Control (EOD/EOOS OLMCPR) ......................................614 6.3 Reduced Flow MCPR Limit for Manual Flow Control
( S L M C PR = 1.10) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 15
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EMF.96180 ;
' Revision 1 Page vi l
Nomenclature
, AFC automatic flow control ATWS anticipated transient without scram ,
BOC beginning of cycle CPR critical power ratio CRD contro! rod drive EFPH effective full power hours
,- EOD extended operating domain
- - EOFP end of full power EOOS equipment out of service 4
FFTR final feedwater temperature reduction FHOOS feedwater heater out of service FWCF feedwater controller f ailure LFWH loss of feedwater heating LHGR linear heat generatit..e rate LPRM local power range monitor LRNB load rejection no bypass LRNB USM load rejection no bypass unpiped safe:y valve margin MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MCPR, flow dependent minimum critical power ratio MFC manual flow control ,
MSIV main steam isoletion valve MSIVC USM main steam isolation valve closure unpiped safety valve margin NRC Nuclear Regulatory Commission NSS nominal scram speed OLMCPR operating limit minimura critical power ratio OOS out of service RPT recirculation pump trip RV relief valve i
RVOOS relisi valve out of service -
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EMF 96180 " .
Revision 1 Page vli l
Nomenclature (Continued)
SLMCPR safety limit minimum critical power ratio SLO single loop operation >
SPC Siemens Power Corporation Nuclear Division SRV safety / relief valve SRVOOS safety / relief valve out of service SVOOS safety valve out of service TCV turbine control valve TIP _
traversing incore prob-TIPOOS TIP strings out of service TLO two loop operation ,
Technical Specification scram speed TSSS TSV turbine stop valve TTNB turbine trip no bypass :
ACPR change in critical power ratio w~
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EMF 40180 Revision 1 Page 1 1 Quad Cities Urdt 2 Cycle il Mont T ansient Analysis
1.0 INTRODUCTION
This report describes the plant transient analyses performed by Slemens Power Corporation -
Nuclest Division (SPC)in support of the reload for Quad Cities Unit 2 Cycle 15 (OCB 1). The Cycle 15 core contains 364 exposed GE9 assemblies.144 exposed GE10 assembIles and 216 fresh ATRIUM
- 9B* offset assemblies. The ATRIUM g8 offset fuel assemblies use the SPC advanced channel and an offset lower tie plate. The limiting change in critical power ratio (ACPR) which precludes fuel damage to these fuel types in the event of anticipated plant transients during Cycle 15 operation is presented in this report.
For Quad Cities Unit 2 Cycle 15 (QC2C15), Commonwealth Edison Company (Comed) has responsibility for portions of the reload safety analysis. This document describes only the Cycle 15 analyses performed' by SPC: Comed analyses are described elsewhere. This document alone does not necessarily identif y the limiting events or the appropriate operating
. limits f or Cycle 15. The limiting events and operating limits must be determined in conjunction with results from Comed analyses. The scope of the analyses performed by SPC is defined in Reference 1.
The analyses reported in this document are performed using the plant transient analysis methodology approved by the Nuclear Regulatory Commission (NRC) for generic application to BWRs (Reference 8). The methods employed for this analysis include the use of the COTRANSA2 system anal als methods (Reference 2), the use of safety limit methodology (Reference 3), the use of ANFB critical power correlation (Reference 4) in XCOBRA T and XCOBRA, and the use of the CASMO 3G/MICROBURN B cme package (Reference 5). The transient analyses for Quad Cities Unit 2 Cycle 15 were performed with the parameters documented in Reference 7. This analysis supports operation in accordance with the power / flow operating map shown in Figure 1.1. The NRC technical limitations as stated in the W
ATRIUM is a trademark of Siemens. 1 v . _ _ _ - _ _ _ - _ -
EMF 9618C !
Revision 1 Page 12 methodology (References 2-6) have been fully satisfied by this anelysis. SPC has performed time step size sensitivity studies to assure that the numerical solution in the COTRANSA2 ,
code converged. Sections 6.0 and 7.0 describe the results of the off rated analysis performed to demonstrate that the full power minimum critical power ratio (MCPR) operating limit, together with the reduced flow MCPR limits, protect operation throughout this map.
The ATRIUM 9B offset fuel assemblies introduced to Quad Cities Unit 2 Cycle 15 have been evaluated to be hydrau$cally compatible with GE9/GE10 fuel resident in the reactor.
f Within this report, several Quad Cities licensing reports are mentioned. In summary, the major reports are identified as:
- The generic EOD/E005 report (Reference 17). Issues addressing generic EODIEOOS
- documentation, penalties, trends and other generic EODIEOOS data are ref erring to this report.
- The cycle spec /fic reload report (Re/etence 18). Issues addressing Cycle 15 analyses performed by SPC are referring to this report. The reload report is a summary of licent,ing limits.
- The cycle specific plant transient report (this report). Issues addressing Cycle 15 thermallimits, pressure rnargins and transients are referring to this report.
The structure of this report is given as:
- Section 2.0 is tM summary of thermal limits and pressure margins for Cycle 15 operation.
- Section 3.0is the Cycle 15 evaluation of the Quad Cities disposition of events and the identification of cycle specific analyses.
- Section 4.0 is the Cycle 15 transient analyses for thermal margin.
- Section 5.0 is the Cycle 15 ASME overprsssurization analyses.
- Section 6.0 is the Cycle 15 evaluation of off rated power and flow operation.
- Section 7.0is the Cycle 15 evaluation of the generic EOD/EOOS penalties defined in
, Reference 17.
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EMF 96180 Revision i Page 13
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. Percent of Roted now Figure 1,1 Quad Cities Unit 2 Operating Power / Flow Map
EMF 96180 Revision 1 Page 21 2.0
SUMMARY
The determination of thermal margin requirements for Quad Cities Unit 2 Cycle 15 was based on the, consideration of various operational transients. The most limiting transients for determination of thermal margins in C ad Cities applications in each general category of events are identified in Reference 17. The limiting MCPR transients determined in Reference 17 and considered in this report are the generator load rejection no bypass to the condenser (LRNB), turbine trip no bypass (TTNB), and the f eedwater controller f ailure (maximum demand) event (FWCF). The loss of feedwater heating event (LFWH)is the responsibility of Comed for Quad Cities Unit 2 Cycle 15. Other potentially limiting MCPR transients (such as the rod withdrawal error) are either considered in the cycle reload report or are the responsibility of Comed.
The change in critical power ratio (ACPRI for the transients is presented in Table 2.1 for Technical Specification scram speed (TSSS) and Table 2.2 for nominal scram speed (NSS).The MCPR safety limit (SLMCPR) analysis for Quad Cities Unit 2 Cycle 15 supports a value of 1.10 l for two loop operation (TLO) and 1.11 for single loop operation (SLO). These values apply to I all fuel types (GE9, GE10, and ATRIUM 9B offset) in the core for Cycle 15 and includes the effects of channel bow and up 'to 40% TIP strings out of service (TIPOOS). Therefore, the current SLMCPRs of 1.10/1.11 given in the Technical Specifications for TLO/ SLO remain I applicable.
The MCPR operating limits (OLMCPRs) based on transients considered in this report are contained in Table 2.3. These limits are obtained by adding the ACPR for each fuel type for the limiting transient (Tables 2.1 and 2.2) to the plant Technical Specification two loop SLMCPR safety limit of 1.10. OLMCPRs are provided for all fuel types in the core for Cycle l
- 15. Key parameters from the transient analyses are provided in Tables 2.4 and 2.5.
Maximum system pressure for the ASME overpressure evaluation was calculated for the postulated closure of all main steam isolation valves (MSIVs) without credit for activation of the MSIV position scram, without pressure relief from the relief valves (RV),*and without pressure relief from tha safety / relief valve (SRV). The ATWS recirculation pump trip (RPT) at
. . . . . - . . , _ . , _ . _ , . . _ . _ _ .~ . . _. . _ . _ ,
- EMF 9618C, Revision 1 Page 2 2 20 psig is modeled. The results of this analysis, as shown in Table 2.3, indicate that the requirements of the ASME code regarding overpressure protection are me.t for the Quad Cities Unit 2 Cycle 15 core, Specifically, the peak vessel pressure limit of 1375 psig and the steam dome pressure limit of 1345 psig are protected.
The discussions and analyses in Sections 6.0 and 7.0 confirm that the full power MCPR operating limits adequately protect the core for reduced power and EOD/EOOS operation.
Analyses and limits presented in this report support operation with various combinations of EOD and EOOS conditions. The EOD/EOOS conditions addressed in this report are identified in Table 2.6.
To support extended operating domain (EOD)/ equipment out of service (EOOS) operation, the base case MCPR operating limits (OLMCPR) are increased by 0.04'd and the maximum overpressurization is increased by 5 psi.*'
Base case analyses refer to analyses that do not fully support EOD/EOOS conditions ano are representativo of normal operation. The base case analyses do support some EOD/EOOS conditions. In particular the 'was case analyses support ICF and RVOOS (SRVOOS for ASME overpressurization analyses).
Sensitivity analyses requested by Comed were perf ormed to determine the impact of operation with up to a 5 psi reduction in steam dome pressure. The analyses determined that ACPR would increase no more than 0.004 for the limiting events it is therefore concluded that operation with a reduction in stean; dome pressure of 5 psi has an insignificant effect on OLMCPRs. For overpressurization events (ASME overpressurization, ATWS, and unpiped safety valve opening analyses), a reduction in steam dome pressure results in increased The 0.04 OLMCPR penalty is required for operation with any combination of FFTR, FHOOS, and coastdown conditions. Other EOD/EOOS conditions require no OLMCPR penalty. The impact of SLO is applied to the SLMCPR. '
The 5 psi pressure penalty is required for operation with coastdown conditions. Other EOD/EOOS conditions require no pressurization panalt'.'. I
EMF 9618(*
Revision i Page 2 3 pressure margins; therefore, the overpressurization analyses with the higher steam dome pressure are bounding. A reduction in steam dome pressure does not increase the SLMCPR.
Higher values of pressure are conservative in the SLMCPR analysis.
ATRIUM 98 offset linear heat generation rate (LHGR) and maximum average planar linear heat generation rate (MAPLHOR) limits for Cycle 15 are presenttd in Reference 18. Composite power history curves for the FWCF and the LRNB/TTNB analyses are provided in Reference
- 24. The composite power history curves provided in Reference 24 can be used by Comed to assess compliance with mechanical design limits (1% strain criteria) for the coresident GE9 and GE10 fuel.
4
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EMF 96180
- Revision 1 Page 2 4 I
Table 2.1 Quad Cities Unh 2 Cycle 15 i Base Case ACPRs at Rated Power With TSSS Insertion Times A C PR
Transient - GE9 GE10 ATRIUM 98 Load Rejection No Bypass 100% Power /108% Flow 0.45 0.40 C ,34 100% Power /100% Flow 0.44 0.38 0.32 100% Power / 87% Flow 0.37 0.32 0.28 Turbine Trip No Bypass 100% Power '108% Flow 0.45 0.40 0.34 100% Power /1M% Flow 0.43 0.38 0.32 100% Power / 87% Flow 0.37 0.32 0.27 Feedwater Flow Controller Failure 100% Power /108% Flow 0.47 0.42 0.37 100% Power /100% Flow 0.46 0.41 0.35 100% Power / 87% Flow 0.39 0.35 0.30 Loss of Feedwater Heating ** S' 'N ACPRs presented are for third cycle GE9 fuel, second cycle GE 10 fuel, snd first cycle ATRIUM 9B fuel.
Analysis of the LFWH is the responsibility of Comed for Quad Cities Unit 2 Cycle 15.
I EMF.96 180 -
Revision 1 Page 2 5 Table 2.2 Quad Cities Unit 2 Cycle 15 Best Case ACPRs et Reted Power With NSS Insertion Times A C ',3
ATRIUM BG Transient GE9 GE10 Offset Load Rejection No Bypass 100% Power /108% Flow 0.45 0.40 0.34 100% Power /100% Flow 0.43 0.38 0.32 100% Power / 87% Flow 0.36 0.32 0.27 Turbine Trip No Bypass 100% Power /108% Flow 0.45 0.40 0.34 100% Power /100% Flow 0.43 0.38 0.31 100% Power / 87% Flow 0.36 0.32 0.26 Feedwater Flow Controller Failure 100% Power /108% Flow. 0.47 0.42 0.37 100% Power /100% Flow 0.45 0.40 0.34 100% Power / 87% Flow 0.38 0.34 0.29 Loss of Feedwater Heating S' "' ml
'*' ~
ACPRs presented are for third cycle GE9 fuel, second-cycle GE10 fuel, ahd first-cycle ATRIUM 9B offset fuel.
Analysis of the LFWH is the responsibility of Comed for Quad Cities Unit 2 Cycle 15.
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. . EMF 9618C Revision i L
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Table 2.3 ,
Quad Cities Unit 2 Cycle 15 Thermal Margin Summary MCPR Operating Limit"8 L
OLMCPR for Base Case / (EOD/EOOS)'*
ATRIUM 98 Transient GE9 GE10 Offset
- Feedwater Controller Failure-
[ l (100%P /108%F - TSSS)_ 1.57 / 1.61 1.52 /1.56 1.47 / 1.51 i c l (100%P /108%F NSS) 1.57 /1.61 1.52 /1.56 1.47 / 1.51 Maximum Pressurization (psig)
Lower Transient Steam Dome Plenum Steam Lines MSIV Closure Without 1290 1323 1290 Pcsition Scram (ASME)
(100%P /108%F, Base Case)
MSIV Closure Without 1295"' 1328"' 1295"'
(100%P /108%F, EOD/EOOS)
!* Based on a plant Technical Specification two-loop SLMCPR of 1.10 and analysis of the limiting system transient analyzed in this report. The actual cycle operating limit may -
be higher if analyses within Comed's scope of responsibility result in a.ACPR higher than those in Tables 2.1 and 2.2. For single loop operation, the Technical Specification 4
i SLO SLMCPR of 1.11 increases the OLMCPR by 0.01. Refer to Section 6.2 for reduced flow MCPR limits.- i Generic OLMCPR penalty of 0.04 is added to support EOD/EOOS operation with any )
combination of FFTR, FHOOS and coastdown conditions. Other EOD/EOOS conditions require no OLMCPR penalty. The 0.04 OLMCPR penalty was confirmed for Cycle 15-analyses.'
Generic pressure penalty of 5.0 psi is added to support EOD/EOOS operation with
' coastdown conditions. Other EOD/EOOS concutions require no pressurization penalty.
The 5.0 psi pressure penalty was confirmed for Cycle 15 analyses. ASME maximum -
overpressurization analyses are performed with TSSS insertion times only.
I
EMF 96180 Revision 1 Page 2 7 Table 2,4 Qund Cities Unit 2 Cycle 15 Results of Plant Transient Analysis With TSSS Insertion Times Maximum Maximum Maximum Core Average Vessel */
Neutron Flux Heat Flux Dome Pressure Event (% of Rated) (% of Rated) losla)
Load Rejection 603 133 1305 /1270 No Bypass (100%P /108%F)
Load Rejection 574 133 1304 /1272 No Bypass (100%P /100%F)
Load Rejection 507 131 1303 /1275 No Bypass (100%P / 87%F)
Turbine Trip 596 133 1304 /1270 No Bypass (100%P /108%F)
Turbine Trip 568 132 1303 /1271 No. Bypass (100%P /100%F)
Turbine Trip 504 131 1301 /1273 No Bypass (100%P / 87%F) l l
- Lower plenum pressure ~ j l
l
- - & - s n - -...a .
EMF 96-18C
, Revisi:n 1 Page 2 8 Table 2.4 Quad Cities Unit 2 Cycle 15 Results of Plant Transient Analysis With TSSS Insertion Times (Continued)
Maximum Maximum Maximum Core Average Vess el/
Neutron Flux Heat Flux Dome Pressure Event (% of Rated) (% of Rated) (osia)
Feedwater Flow 587 138 1200 /1166 Controller Failure (100%P /108%F)
Feedwater Flow 556 137 1198 /1167 Controller Failure (100%P /100%F)
Feedwater clow 485 135 1195 /1167 Controller Failure (100%P / 87%F)
MSIV Closure 306 131 1323 /1290 ASME Analysis (100%P /108%F)
MSIV Closure 305 130 1322 /1292 ASME Analysis (100%P /100%F)
MSIV Closure 298 127 1321 /1294 ASME Analysis (100%P / 87%F)
'*' Lower plenum pressure
EMF 9618C Revision 1 Page 2 9 Table 2,5 Quad Cities Unit 2 Cycle 15 Results of Plant Transient Analysis With NSS insertion Times Maximum Maximum Maximum Core Average Vessel */
Neutron Flux Heat Flux Dome Pressure Event (% of Rated) ._(% of Rated) (osia)
Load Rejection 596 133 1273 /1239 No Bypass (100%P /108%F)
Load Rejection 566 132 1273 /1242 No Bypass (100%P /100%F)
Load Rejection 495- 131 1272 /1245 No Bypass (100%P / 87%F)
587 132 1271 /1237 No Bypass (100%P /108%F)
Turbine Trip 557 132 1271 /1240 No Bypass (100%P /100%F)
Turbine Trip 488 130 1271 /1244 No Bypass (100%P / 87%F) s
- Lower plenum pressure
4
,y EMF 96180 Revision :
Page 210 Table 2.5 Quad Cities Unit 2 Cycle 15 Results of Plant Transient Analysis With NSS Insertion Times (Continued)
Maximum Maximum Maximum Core Average Vessel/
Neutron Flux Heat Flux Dome Pressure Event (% of Rated) (% of Rated) (osia)
Feedwater Flow 586 137 1176 /1141 Controller Failure (100%P /108%F)
Feedwater Flow 546 136 1172 /1140 Controller Failure (100%P /100%F)
Feedwater Flow 471 134 1168 /1140 Controller Feilure (100%P / 87%F)
Lower plenum pressure
.,m
EMF 96180 Revision 1 Page 211 Table 2.6 EOD and EOOS Operating Conditions Extended OperatM Domsk Condtlons e it' creased Core Flow (ICF)
- Coastdown
- Combined ICFFFTR
- Combined ICF/Coastdown
- Combined FFTR/Coastdown
- Combined ICF/FFTR/Coastdown Equipment Out of Service Condtions**
- Feedwater Heater (s) Out of Service
- Single-loop Operation (SLO) Recirculation Loop Out of Service"'
Relief Valve Out of Service (RVOOS)
- Safety / Relief Valve Safety Function Out of Service (SRVOOS) for ASME Events'*
- Up to 40% TIP Strings Out of Service (TIPOOS)i*
Base case analyses are performed with this condition.
EOOS conditions are supported for both EOD conditions and standard operating domain conditions.
"8 SLO adds 0.01 to the TLO SLMCPR.
l- 40% TIPOOS with 100% TIP strings available at startup,50% of theIPRMs OOS (LPRM substitution model on or off), and 2000 EFPH LPRM calibration interval. TIPOOS is evaluated in the SLMCPR analyses.
I
EMF 9618C Revision 1 Page 31 3.0 DISPOSITION OF EVENTS The initial disposition of events for Quad Cities is documented in Section 3.0 of Reference 17.
The initial Quad Cities disposition of events is based on the plant parameters documented in Reference 19. Differences between the OC2C15 plant parameters (Reference 7) and the plant parameters used in the initial disposition of events are identified in Table 3.1. In addition, the impact of the differences on the initial disposition are also provided in Table 3.1. The differences do not change the conclusions of the initial disposition of events provided in Section 3.0 of Reference 17. The Cycle 15 analyses are identified in Reference 1.
e e
_.m__-..._.._ _ _ ___.
EtAF-96180 Revision 1 Page 3 2 Table 3.1 Quad Cities Unh 2 Cycle 15 Evaluation of Plant Parameter Changes on Disposition of Events Parameter Change (From/To) impact Resolution CRD Enthalpy (Btu /lbm) None (insignificant). -
71.0 to 68.05 Cleanup Flow (kibm/hr) None (insignificant). -
270.0 to 125.0 Cleanup Entnalpy (Btu /lbm) None (insignificant). -
113.0 to 70.3 Feedwater Flow Uncertainty (%) Safety Limit Safety Limit is calculated on a cycle-1.76 to 2.62 specific basis. The new feedwater flow uncertainty was included in the cycle-specific analysis.
SV Flow Capacity (Ibm /sec) The increased SV flow The ASME MSIV closure event is 171.88 to 179.04 capacity may increase the evaluated on a cycle specific basis. The ASME maximum pressure EOD/EOOS pressure penalty is margin, reconfirmed for QC2C15.
The SRV is reconfirmed to be the most limiting SRV/SV out of-service for Cycle 15 analyses.
Turbine Bypass Delay (msec) Turbine bypass actuation is The FWCF event is evaluated on a cycle-100 to 150 slower. The FWCF specific basis. The EOD/EOOS MCPR w/ bypass event will penalty is reconfirmed for OC2C15.
become more severe.
NSS Times: OC2C15 will use Transients performed with As stated in Reference 17, results of the slower times than the slower NSS times become transient analyses using TSSS and NSS EOD/EOOS analyses. - more severe. insertion times demonstrate that scram speed has a minor impact on the EOD
/EOOS MCPR penalty and that the cycle-specific NSS times must be between the TSSS and NSS times used in Reference 17 for the generic EOD/EOOS MCPR penalty to be applicable.
OC215 analyses include thermal margin NSS insertion time evaluations to confirm Reference 17 conclusions.
l EMF 96180 i
l Revision 1 l Page 3 3 Table 3.1 Quad Cities Unit 2 Cycle 15 Evaluation of Plant Parameter Changes on Disposition of Events (Continued)
Parameter Change (From/To) Impact Resolution TCV Position (% Open) Less time is needed for the The LRNB event is evaluated on a cycle-65 to 4B TCV to fully close. The specific basis. OC2C15 analyses demon-LRNB and the ASME TCV strate that the LRN8 event bounds the closure overpressurization TTNB event. The EOD/EOOS MCPR events may become more penalty is reconfirmed for OC2C15.
severe. OC2C15 analyses also demonstrate that for ASME overpressurization analyses, the MSIV closure event bounds the TCV closure event.
Water Level Sensor Delay (sec) Added delay for the turbine The FWCF event is evaluated on a cycle-0.25 to 1.05 trip for the FWCF event. specific basis.Since the core is at steady-state conditions at the time of trip, the new delay will not impact ACPR or the EOD/E00S penalties. Reference 11 demonstrated that the FW control
, system has an insignificant impact on f ast transients.
FW Valva Stroke Time (soci The FW valve can respond A change in FW response will not affect 12 to 3 more quickly to the FW fast transients since the response of the demand. FW is slow relative to the change in events (core power, heat flux, etc.) for a fast transient. The change will not impact ACPR or the EOD/EOOS penal-ties. Reference 11 demonstrated that the FW control system has an insignificant impact on f ast transients, b
T.
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EMF 96180 '
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r 4.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN This section describes the analyses which were performed to determine the full power MCPR ,
operating limits for Cycle 15 of Quad Cities Unit 2.
d
- 4.1 Desian Basis The plant transient analyses for Quad Cities Unit 2 Cycle 15 determined that the limiting transient initial conditions were at rated power and 108% rated core flow. Rated reactor plant parameters for the analyses are shown in Table 4.1. The most limiting point in the cycle is when the control rods are fully withdrawn from the core. The thermal margins established for the end of full power (EOFP) capability are conservative for cases where control rods are partially inserted. The transient ar.alyses were performed assuming the conservative conditions in Table 4.2. All transients were performed with the most limiting RV (lowest set point) out of service. In addition, the relief function of the SRV was conservatively modeled as an RV (i.e., slower response time and lower flow capacity). .
Observance of the OLMCPR shown in Table 2.3 willprovide adequate protection against the occurrence of boiling transition during all anticipated transients considered in this section for Quad Cities Unit 2 Cycle 15, 4.2 Calculational Model COTRANSA2 (Reference 2), XCOBRA T (Reference 9), XCOBRA (Reference 6), and CASMO-3GIMICROBURN B (Reference 5) are the major codes used in the thermallimits' analyses as described in SPC's THERMEX methodology report (Reference 6) and neutronics methodology report (Reference 5). COTRANSA2 is a system transient simulation code which includes an axial one-dimensional neutronics model used to model the axial power shifts associated with the system overpressurization in the LRNB, TTNB, FWCF, and MSIV closure transients.
t XCOBRA T is a transient thermal-hydraulic code used in the analysis of thermal margins of the limiting fuel assembly. XCOBRA is a steady state thermal-hydraulic code used in the analysis of slow flow excursion events. Fuel pellet to-cladding gap conductance values used in the
EMF g6.180 ,
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analyses were based on RODEX2 (Reference 10) calculations for the Quad Cities Unit 2 Cycle- )
15 core configuration. The thermal margins of the f uel assemblies are evaluated in XCOBRA T, .
. XCOBRA, and MICROBURN B using the ANFB critica power correlation (Reference 4). The applicability of the- ANFB critical power correlation to GE9/GE10 fuel at Quad Cities is demonstrated in References 20 and 21.
in accordance with SPC methodology, possible limiting transients are evaluated using a-consistent set of bounding input. From the results of these transients, the limiting transient is identified as the FWCF at 100% power /108% flow. Table 4.2 summarizes the values used f or important parameters in the analysis. Table 4.3 provides the feedwater flow, recirculating coolant flow, and pressure regulation system settings used in the analysis.
4.3 Anticiosted Transients For Quad Cities Unit 2 Cycle 15 operation, specific events have been evaluated for thermal margin as outlined in Reference 17. These events are the LRNB, the TTNB, and the FWCF.
Reference 17 demonstrated that other categories of transients are either inherently self-limiting, bounded by one of these or are part of Comed's analysis responsibility. Reference 17 provides descriptions of the trarisients that are considered for the cycle specific evaluation.
4.3.1 Load Relection No Bvoass/ Turbine Trio No Bvoans The LRNB is a more limiting transient than the TTNB transient. This conclusion has been arrived at through comparison of the scram delays and valve closure times for the LRNB and TTNB transients for the Quad Cities units. The initial position of the turbine control valve (TCV) is such that it closes f aster than the turbine stop valve (TSV). Tables 4.4 and 4.5 show results from Cycle 15 LRNB and TTNB analyses. Cycle 15 analyses verify that the LRNB event bounds .
the TTNB event.
t in the load rejection transient, steam flow is interrupted by an abrupt closure of the TCV. The resulting pressure increase causes a decrease in the void volume in the core, w'hich in twu creates a power excursion. This excursion is mitigated in part by Doppler broadening and 4
- , . . - . . - ~ . ,...-m
l o
. Eh4F 96180 I Revisien 1 Page 4 3 pressure relief, but the primary mechanisms for termination of the event are control rod insertion and regeneration of voids. A turbine trip is similar to the load rejection transient, the difference is that steam flow is interrupted by an abrupt closure of the TSV.
The important parameters for these transients include the power transient (integral power) determined by the void reactivity, which affects the initial power excursion rate and is part of the intrinsic shutdown mechanism, and the control rod worth, which determines the value of the scram reactivity. Other important inputs include the control rod movement parameters (scram delay and insertion speed), which determine the event characteristics following the initial mitigation of the power excursion. From Tables 4.4 and 4.5, the largest calculated limiting ACPR for the LRNB event was at 100% power /108% flow conditions for both TSSS and NSS insertion times. The NSS insertion times provide less than a 0.01 improvement in the MCPR limits. I 1
Figures 4.1 -4.8 illustrate the behavior of major system variables during the LRNB and TTNB f events at 100% power and 108% flow for TSSS insertion times. MCPR occurs at i l
approximately 0.7 second for the ATRIUM 9B offset fuel. '
)
4.3.2 Feedwater Controller Failure i
The feedwater controller failure to maximum demand leads to an increase in feedwater flow into the reactor vessel. The excessive feedwater flow increases the subcooling in the recirculating water returning to the reactor core. This reduction in moderator temperature will j result in the core power increasing to a higher equilibrium power level if no other actions j occur. Eventually, the level of water in the downcomer region will rise until th'e high water level trip set point (LB) is reached. A turbine trip initiated on high water level results in the j rapid closure of the TSV to prevent the transmission of liquid water to the turbine. The rapid closure of the TSV produces a compression wave in the steam line which results in core void collapse and increased core reactivity. The stop valve closure initiates a scram signal at 10%
TSV closure bnodeled as a 0.01 second delay) and the resulting control rod insertion I
~
terminates the power increase.
EMF 96!180 Rsvisi:n 1 Page 4-4 i
In the analysis,- the bypass valves do not operate before the turbine trip signal due to l
conservative control sy, tem assumptions (maximum combined flow limiter and bypass valve opening bias settings prevent bypass valve operation). However, the bypass valves do open as a result of the closure of the TSV. The bypass valves are assumed in the model to start
, opening 0.15 second after the start of TSV motion. The start of bypass valve opening I
corresponds to the time when the stop valves become fully closed plus a delay, of 0.05 second. Although a longer TSV stroke time would result in a longer delay in bypass valve opening, a fast TSV closure results in a more severe event even though the bypass valve opens earlier. The reactor pressure increase produced by the rapid stop valve closure is l . mitigated by the opening of the bypass valves. The bypass valve opening time assumed in the analysis is given in Table 4.2.
Tables 4.4 and 4.5 show the results from the Cycle 15 FWCF analyses. Figures 4.9-4.12 illustrate the behavior'of major system variables during tho FWCF transient at 100%
power /108% flow for TSSS insertion times. The MCPR occurs 46.9 seconds for the ATRIUM- ,
- 98 offset fuel. The TSV becomes fully closed at 46.4 seconds.
- 4.3.3 Loss of Feedwater Heatino f
The loss of feedwater heating leads to a gradualincrease in the subcooling of the waterin the
, lower plenum. The gradual power change allows the fuel thermal response to maintain pace with the increase in neutron flux. For the Quad Cities Unit 2 Cycle 15 re!oad, the analysis of the loss of feedwater heating transient is the responsibility of Comed.
!' 4.4 MCPR Safetv Limit s
1 1
The MCPR safety limit (SLMCPR) for Quad Cities Cait 2 Cycle 15 operation was determined using the methodology described in Ref erence 3. The main input parameters and uncertainties used in the safety limit analysis are listed in Table 4.6. The radial power uncertainty includes the effects of up to 40% TIP strings out of service (TIPOOS) with 100% TIP strings available at startup, up to 50% of the local power range monitors (LPRM) out of service, and an LPRM calibration interval of 2000 effective full power hours (EFPH) as discussed in Reference 13.
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The determination of the SLMCPR' explicitly includes the effects of channel bow and relies on y Lthe following assumptions:
i j
- Cycle 15 will not 'uso channels for more than one fuel bundle lifetime' The GE9 .
i fuel uses CarTech channels, the GE10 fuel uses the GE advanced channel, and the ATRIUM 98 offset fuel uses the SPC advanced channel.
- = The channel exposure for the highly exposed GE9 assemblies will not exceed 50,000 mwd /MTU for Cycle 15 based on the maximum fuel bundle exposure. )
- The GE advanced channel bow data for the GE10 fuelis provided in Referenc5s
- 22 and 23 and is valid as long as Quad Cities is loaded as a control cell core,-
the fresh fuel loaded into Quad Cities is offset into the wide wide gap, and no new GE10 channels are inserted into the core. .
1
~
- The effects of channel bow were determined using a 2x2 array with a conservative exposure configuration, i
Analyses were performed with input parameters (including the radial power and local peaking
- factor distributions) for each exposure step in the design basis step-through including an EOFP + 1500 mwd /MTU extension to cover coastdown operation. The analysis that produced
- the highest number of rods in boiling transition corresponds te Cycle 15 near and of cycle I (EOC) exposure of EOFP + 1500 mwd /MTU. The radial power distribution corresponding to this l oxposure is shown in Figure 4.13.
The limiting local power distribution for the Cycle 15 SPC fuel types with channel bow are
- shown in Figures 4.14 and 4.15.
The results of the analysis support a TLO SLMCPR of 1.10 for all fuel types residing in the l core. Protection of this limit will assure that at least 99.9% of the fuel rods in the core are
- cxpected to avoid boiling transition during normal. operation and anticipated operational occurrences. In addition, analyses were explicitly performed to support the EOD conditions of ICF and SLO. The' TLO limit of 1.10 and a SLO limit of 1.11 supports all normal and l t EOD/EOOS conditions identified in Table 2.6. The current Quad Cities Technical Specificatior; SLMCPR safety limit of 1.10 for TLO and 1.11 for SLO are applicable.
~
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t EMF 96180 -
Revision 1 Page 4 6 4.5 Nuclear Instrument Resoonse The impact of loading ATRIUM 9B offset fuelinto the Quad Cities core will not affect the nuclear instrument response. The neutronic lifetime is an important parameter affecting the time response of the incore detectors. The neutron lifetime is a function of the nuclear and mechanical design of the fuel assembly, the in channel void fraction, and the fuel exposure.
The neutron lifetimes are similar for the SPC and GE Quad Cities fuel with typical values of 39(10-') to 40(10-') secondr orv the ATRIUM 98 offset lattices and 41(10-e) to 43(10-')
seconds for the GE9/GE10 laidces as calculated with the CASMO 3G code at core everage void exposure conditions. Therefore, the neutron lifetimes for a full core of ATRIUM 9B offset fuel, a mixed core of ATRIUM 9B offset and GEa'GE10 fuel, and a full core of GE9/GE10 fuel are essentially equivalent.
4
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Table 4.1 l- Quad Cities Unit 2 Design Reactor and Plant Conditions k
Reactor Thermal Power 2511 MWt Total Core Flow 98.0 Mlb/hr .
Core Active Flow 85.9 Mlb/hr Core Bypass Row 12.1 Mlb/hr Core Inlet Enthalpy" 522.7 Btullbm Vessel Pressures Steam Oome 1020 psis Core Exit (upper plenum) 1030 psia '
Lower Plenum" 1054 psia Turbine Pressure -
, 965 psia Feedwater/ Steam Flow 9.76 Mlb/hr
~
Feedwater Enthalpy" -
314.1 Btu /lbm Recirculating Pump Flow (per pump) 16.7 Mlb/hr Includes water rod / channel flow.
These parameters vary slightly due to cycle variations (core configuration and power distribution) and to minor riifferences in heat balance calculations between computer codes. Differences are not significant.
EMF 96;180 nevisis .:
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Table 4.2 Quad Cities Unit 2 Significant Parameter Values Used in Analysis 1 l
1 High Neutron Flur. Trip 3013.2 MWt Time to Deenergize Pilot Scram 200 msee Solenoid Valves Time to Sense Fast Turbine Control 80 msec
- 1 Valve Closure Time From High Neutron Flux Trip to 290 msec *'
Control Rod Motion .
Turbine Stop Valve Stroke Time 100 msec Turbine Stop Valve Position Trip 90% open I
Turbine Control Valve Stroke Time 150 msec (total)
Core Average Fuel / Cladding Gap
- 2177 Btu /hr ft 2.op
- Conductance (cycle specific value)
- Includes a 50-msec delay for RPS logic transfer and a 30-msec delay until signal is received by RPS logic.
"' includes a 90-msse delay for signal to reach solenoid valves and a 200-nisec delay for pilot scram solenoid valves to deenergize.
- Calculated by SPC for the Cycle 15 co's using RODEX2 at rated conditions,
EMF 96180 Revision 1 Page 4 9 Table 4.2 Quad Cities Unit 2 Significant Parameter Values Used in Analysis (Can?inued)
Safety / Relief Valve Performance Settings
Safety / Relief Valve (1 valve)
Capacity Per Valve (relief) 155.0 lbm/sec at 1120 psigib' Capacity Per Valve (safety) 166.1 lbm/sec at 1112,4 psig
Relief Valves Capacity (4 valves)'*
Capacity Per Valve 155.0 ILm/sec at 1120 psig Safety Valves Capacity (8 valves)
Capacity Per Valve 179.04 lbm/ soc at 1277.2 psig Safety / Relief Valve Delay / Stroke 1.85/250 msec'6' Relief Valves Delay / Stroke 1.85/250 msec MSIV Stroke Time 3.0 see MSIV Position Trip Set Point 90% open Condenser Bypass Valve Performance Total Capacity 1084 lbm/sec Delay to Openir.g (from the start of TSV motion) 150 msec Opening Time 0.11 sec (5% open), 0.25 see (80%
open),0.7 sec (100% open)
Fraction of Energy Generated in Fuel 0.965
Vessel Water Level (above separator skirt)
Normal 30 in Range of Operation (lower bound) 20 in High Level Trip 60 in Maximum Feedwater Runout Flow (2 pumps) 3307 lbm/sec Recirculating Pump Trip Set Point 1250 psig Steam Dome Pressure Valve set points are given in Reference 7.
'6' i The relief valve mode of the SRV is conservatively modeled with RV flow capacity and responsa time. ,
'*' I For ASME overpressurization event, SRV safety function is not credited.
]
One relie' valve at the lowest set point is not credited.
Reference 12. l l
EMF 96180
- Revision 1 Page 410 Table 4.2 Quad Cities Unit 2 Significant Perameter Values Used in Analysis (Continued)
Control Rod insertion Time TSSS Time NSS Time Pos' tion (Notch) (sec) (sec) 48 0.000 0.000 48 0.200 0.200 5% inserted 0.375 0.340 45 0.419 0.375 39 0.856 0.725 20% inserted 0.900 0.760 25 1.924 1.572 50% inserted 2.000 1.632 5 3.484 2.832 90% inserted 3.500 2.844 0 3.875 3.147 i
4 EMF 96180 Revision 1 Page _411 Table 4.3 Control Characteristics"'
Sensor Time Constants Pressure 500 msec Steam Flow /Feedwater Flow 250 msee Level 1.05 see Feedwater Control Mode Single Element *'
- Water Level Controller Proportional Gain 25%/ft Pressure Regulator Settings Lead 0.47 see Lag 7.2 see Gain 3.33 %/psid Bypass Flow Signal Bias 3.0%
Combined Steam Flow Limiter 1,05 Setting Turbine Maximum Steam Flow 2816.67 lbm/sec Recirculation Flow Control Mode Manust The transients, considered in cycle specific analyses are mitigated by reactor scram which has a response that is faster than the feedwater control system response. The inclusion of tile control system in the analysis model results in a more realistic
- calculated plant response. The representative parameters have an insignificant effect on pressure and thermal margins.
Quad Cities licensing analyses are insensitive to the feedwater control system e.
algorithms or settings. Single element mode provides slightly more conservative results compared to manual or three-element control mode for all events based on the Dresden study in Refererice 11.
EMF 96180 Revision 1 Page 412 Table 4.4 Quad Cities Unit 2 Cycle 15 Comparison of LRNB, TTNB and FWCF Results With TSSS Insertion Times Maximum Maximum ,.
Maximum Core Average Vessel */
Neutron Flux Heat Flux Dome Pressure State Point (% of Rated) (% of Rated) (psig) ACPR*'
- 100% Power /108% Flow LRNB 603 133 1305/1270 0.45 /0.40 /0.34 TTNB 596 133 1304 /1270 0.45 /0.40 / 0.34 FWCF 587 138 1200 /1166 0.47 /0.42 / 0.37 100% Power /100% Flow LRNB 574 133 1304 /1272 0.44 / 0.38 / 0.32 TTNB : 568 132 1303 /1271 0.43 / 0.38 /0.32 FWCF 556 137 1198 /1167 0.46 / 0.41 / 0.35 100% Power / 87% Flow LRNB 507 131 1303 /1275 0.37 /0.32 / 0.28
-TTNB 504 131 1301 /1273 0.37 /0.32 / 0.27' FWCF 485 135 1195 /1167 0.39 / 0.35 / 0.30 l
l i
- Lower plenum pressure ..
Values for third cycle GE9/second-cycle GE10/first-cycle ATRIUM 9B fuel.
ACPRs for the TTNB event are approximately 0.002-0.003 less than i.RNB event.
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Page 413 l t
Table 4.5
?
Quad Cities Unit 2 Cycle 15 Comparison of LRNB, TTNB, and FWCF Results -
With NSS Insertion Times S
Maximum Maximum Maximum Core Average Vessel'*/
Neutron Flux Heat Flux Dome Pressure State Point N
(% of Rated) (% of Reted) (psig) ACPR '"
1.)0% Power /108% Flow LRNB 596 133 1273 /1239 O / 5 / 0.40 / 0.34 TTNB 587 132 1271 /1237 0.45 / 0.40 /0.34 FWCF 586 137 1176 /1141 0.47 /0.42/0.37
-100% Power /100% Flow LRNB 566 132 1273 /1242 0.43 / 0.38 / 0.32 TTNB 557 132 1271 /1240 0.43 / 0.38 / 0.31 FWCF 546 136 1172 /1140 0.45 / 0.40 /0.34 100% Power / 87% Flow LRNB , 495 131 1272 /1245 0.36 / 0.32 / 0.27 TTNB -488 130 1271 /1244 0.36 / 0.32 / 0.26 FWCF ' 4 71 134 1168 /1140 0.38 / 0.34 / 0.29 ,
Lower p;enum pressure Values for third cycle GE9/second-cycle GE10/first cycle ATRIUM 98 fuel.
'd ACPRs for the TTNB event are approximately 0.002-0.003 less than LRNB event.
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EMF 96180 Revision 1 Page 414 Table 4.6 input for MCPR Safety Limit Analysis Fuel Related Uncertainties Statistical Parameter Uncertainty Source Do:ument Treatment I
l l
Plant Measurement Uncertainties
!. Uncertainty (%) Statistical Parameter Units Value (Reference 7) Treatment i
1 l
L m
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Revision 1 Page 416 404 ana
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Figure 4.2 Load Rejection No Bypass at 100/108-Vessel Water Level (Referenced to instrument Zero)
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EMF 9618C Revision 1 Page 417 isana 1soon.
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Vessel Pressure Response 6
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EMF 96180 Revision 1 Page 418 I
am RV, t Volve
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. I EMF 96180 t Revision 1 Page 419 i
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Key Parameters
e EMF 96180 t -
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Vessel Pressure Response
EMF 96180 V ,
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Safety / Relief Velve Flows
____......__.____.________q F
EMF 96180 i
. Gevision 1 ,
Page 4 27 i l
100 i i i i i i '
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j 60 -
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EMF 96180 -
Revision 1 Page 4 28 ,
1Contrel Rod Corner 1o lt Ir 1.007 1.015 1.028 1.026 1.088 1.074 0.999 1.010 0.973 io I1 l'.015 1.074 1.081 1.084 0.963 1.023 1.015 0.955 1.036 1R 1.028 1.081 1.076 1.103 1.118 1.075 1.014 0.989 1.023 l[
I IC l o 1.026 1.084 1.103 0.939 0.967 1.009
,I I Internal Ie 1.088 0.963 1.118 Water 1.059 0.841 0.988 1r I Channel l 1.074 1.023 1.075 1.009 0.924 0.971 I
l .
! 0.999 1.015 1.014 0.939 1.059 1.009 0.834 0.917 0.947 I
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l 1
- 0.973 1.036 1.023 1.009 0.988 0.971 0.947 0.915 0.826 I
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! Figure 4.14
- Quad Cities Urdt 2 Cycle 15 l Safety Limit Local Peaking Factors l Whh Channel Bow at Assembly Exposure of 22,500 mwd /MTU l (SPCA9 372811GZH ADV)
. , _ . _y.. _ . _ _ _ _ _ _ _ _ _ _ . - . , . _ _ - . _ _ , . . . - _ , _ _ , . _ _ . , _ _ , - , _ _ , - - , , _ . . - . _ , _ _ , , _ , _ . - - .
, EMF 96180 Revisi:n 1 Page 4 29 Contrel RodCorner :
o ;
1.011 1.022 1.032 1.028 1.093 1.079 1.000 1.015 0.977 l
r ;
o l 1 l'.022 1.085 1.084 1.085 0.963 1.021 1.013 0.958 1.042 i l
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y 1.032 1.084 1.075 1.105 1.121 1.075 1.008 0.985 .1.024 I
I C i o 1.028 1.085 1.105 0.937 0.960 1.007 I I
internal l
1.093 0.963 1.121 Water 1.057 0.837 0.985 l l Channel i 1.079 1.021 1.075 1.005 0.918 0.968 g I
i 1.000 1.013 1.008 0.937 1.057 1.005 0.829 0.910 0.945 I '
i I
I 1.015 0.958 0.985 0.960 0.837 0.918 0.910 0.828 0.915 l
\
l O.977 1.042 1.024 1.007 0.985 0.968 0.945 0.915 0.823 I I
I Figure 4.15 l Quod Cities Unh 2 Cycle 15 i Safety Limit Local Peeking Factors t With Channel Bow at Assembly Exposure of 20,000 mwd /MTUr l
(SPCAS 3588110ZL ADV) l
EMF 96180 R:visi:n 1 Page 51 5.0 MAXIMUM OVERPRESSURIZATION ANAL.YSIS This section describes the analysis of the maximum overpressurization event performed with COTRANSA2 (Reference 2)in compliance with the ASME code.
5.1 atsian Basis Rated reactor conditions are summarized in Table 4.1.These conditions are the same as those used in the transient analyses fc- *hermal margin. Conservatism was added to the transient by disallowing the operation of the four power actuated relief valves as required by the ASME code. Credit was not taken for either the relief or safety function of the SRV. The ATWS RPT trip was modeled at 1250 psig. Failure of the most critical active componer.t was assumed.
in this instance, the most critical active component is the direct scram on valve position. With the loading of ATRIUM 9B offset fuel assemblies into the Cycle 15 core, additional events were analyzed to verify that the closure of all MSlVs is the bounding pressurization event. The additional events include the closure of the TCVs and closure of the TSVs. All events are analyzed with direct scram on valve position disabled.
5.2 Pressurization Transients The position scram, which initiates reactor shutdown almost immediately upon MSIV movement, mitigates the effects of this event to the point that it does not contribute to the determination of thermal margins. Delaying the scram until the high flux trip set point is reached results in a substantially more severe transient.
Although the closure rate of the MSIVs is substantially slower than that of the TCVs or TSVs, the compressibility of the fluid in the steam lines provides significant damping of the compression wave associated with the TCV and TSV closure events to the point that the slower MSIV closure without direct scram results in nearly as severe a compression wave.
~
, Once the MSIVs are closed, the subsequent core pnwer production must be contained within a smaller system volume than that associated with the TCV or TSV closure events. Table 5.1
O f EMF.96180 !
Revision 1 -
Pags 5 2 l provides analysis resu'ts for the ASME events analyzed for Cycle 15. Cycle 15 analyses demonstrate that the MSIV cicsure event under these conservative assumptions results in a higher overpressure than either the TCV or TSV closure events. i i
5.3 Closure of All Main Steam isolation Valves This calculation assumed that all four steam lines were isolated at the containment boundary within 3 seconds. The valve characteristics and steam compressibility combine to delay the ;
arrival of the compression wave at the core until approximately 3 second:: from the initiation ;
of the MSIV stroke. Effective shutdown is delayed until approximately 5 seconds following initiation of the MSIV stroke because control rod performance is assumed to be at the Technical Specification limits. Only TSSS insertion times were used in the analyses.
, The limiting MSIV closure vessel pressure is attained at 100% power /108% flow. Results from i the limiting case are provided. The maximum vessel pressure (at the lower plenum) of 1323 psig was observed at 6.1 seconds. The maximum steam ! int. pressure of 1290 psig was observed at approximately 0.2 seconds. The maximum pressure in the steam dome of 1290 psig was also observed at 6.2 seconds. The relative values of maximum pressure during the MSIV closure transient indicate that the vessel and steam lines will be protected against overpressure limits defined in the ASME code when a pressure safety limit of 1375 psig in the lower plenum is protected. In addition, the Quad Cities Technical Specification steam dome pressure limit of 1345 psig is also protected (Reference 14).
Figures 5.1-5.4 litustrate the performance of major system variables during the MSIV closure overpressurization event at 100% power and 108% flow.
For EOD/EOOS operation, a generic pressure penalty of 5.0 psiis applied to the limiting ASME analysis (Reference 17). Even with this penalty, the MSIV closure transient has significant
-margin to the overpressure lirnits defined in the ASME code.
.-_,...,.s .,,...n. , ,. , , -,.pm-, m,,, , , _ _ _ ,. , - - , - . _ , , , . ,. , , .,. . . . , , - - , , . , . - , , - _ ,
. - . - . . -_-. _ - _ = - _ _ - - . _ _ - . . . -
Ef(F 96100 Revision i Page 5 3 l I
i Table 5.1 !
Base Case Quad Cities Unit 2 Cycle 15 I Pesults Summary of ASME Overpressurization Analyseh l With ~588 Insertion Times Maximum Pressurization (psig)
Transient Steam Dome Lower Plenum 4
MSIV Closure (100%P /108%Fn 1290 1323 (100%P /100%F) 1292 1322 (100%P / 87%F) 1294 1321 '
TCV Closure (100%P /108%F) 1288 1321 (100%P /100%F) 1289 1320 '
(100%P / 87%F) 1292 1318 TSV Closure (108%P /108%F) 1288 1321 (100%P /100%F) 1289 1320 (100%P / 87%F) 1292 1318 w --
-M er+- e- e-rt- ? ? e~~- -w.w- 9m- wiyr---r.w, --&w- pw-w -t-p er --- n
I EMF 96180
- i Revision 1 !
Page6 4 I
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M !
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a u a u u su Time, seconds Figure 5.1 j MSlV Closure at 100/108 l Key Parameters I.
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a u <a u u su Time, seconds Figure 5.2 M81V Closure et 100/108 Vessel Pressure Response f
4'
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Figure 5.3 MSIV Closure at 100/108 Vessel Water Level i (Referenced to Instrument Zero)
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- 1 EM! 96180 i Revision 1 i Page 5 7 l
t I
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Time, seconds Figure 5.4 !
MSIV Closure at 100/108 -
Safety Valve Flow Rates (SRV Assumed inoperable) >
a-h
- ee e.g
.~m .__,.t " ' - - - * -e-er (---- +-w-et =- +-W,t-<=- -m-**ev' +rw,m-ere w-mm+m=-- =+ -+- 4 e -+-ew '*--e -++- --
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EMF 90180 Revision 1 -
Page 61 r
6.0 ANALYSIS AT OFF RATED CONDITIONS
, Transient analysis of a BWR requires consideration of transients at off rated conditions. This section describes those evaluations performed in support of Cycle 15 that are not covered in Sections 4.0 and 5.0. This section specifically addresses reduced core power and core flow.
EOD/EOOS conditions are discussed in Section 7.0.
i 6.1 Reduced Core Power The base case cycle specific MCPR operating limits were determined using analyses perf ormed at full power and at end of cycle (EOC) exposure with all control rods fully withdrawn. Off-rated analyses are not used in setting the OLMCPR limit because there is sufficient MCPR margin at off rated conditions to ensure that the SLMCPR is not violated. The full power analysis will bound analyses at off rated conditions. At exposures earlierin the cycle, the core could potentially be at the OLMCPR at reduced power using control rods; however, the partially inserted control rods would result in a substantial increase in scram reactivity worth and in a ACPR less than the full power analysis.
Transient analyses were perfor'med with reduced power in Reference 17. The rer.ults of Reference 17 demonstrate that full power transients bound events at reduced power because of the increased instgin to thermal limits. The gain in steady state MCPR margin (the differerne between the steady state MCPR of the off rated power case and the steady state MCPR of the limiting full power ACPR case) is much greater than the increase, if any, in ACPR. Since the OC2C15 core design has similar characteristics of the Reference 17 core design (k eff,24 month cycle, etc.) and the power / flow map has not changed, thE conclusions of Reference 17 that full power transients bound events at reduced power are applicable for OC2C15.
6.2 Reduced Core Flow
~
LRNB, TTNB, FWCF, and MSIV closure transients were evaluated at 87% rated core flow and 100% rated power. For 100% power cases, analyses with 87% rated flow were bound by
EMF 96180 Revision .
Page 6 2 l analyses with 100% and 108% rated flow. The limiting event was FWCF. Reference 17 demonstrated that off rated core power and core flow transients were bound by rated power transients. Since the OC2C15 core design has similar characteristics of the Reference 17 core design (k ef f,24 month cycle, etc.) and the power / flow map has not changed, the conclusions of Reference 17 that full power transients bound events at reduced power and flow are applicable for OC2C15.
Analysis ior pump run up events from operation at less than rated recirculation pump capacity indicates the need for an augmentation of the full flow OLMCPR for lower flow conditions.
This is due to the potential for large reactor power increases should an uncontrolled pump flow increase occur.
The analysis establishes the reduced flow MCPR operating limits (MCPR,) necessary to protect the reactor fuel against boiling transition curing anticipated pump run up events from off rated core flow conditions for manual flow control (MFC). The analysis also establishes MCPR, limits to protect the OLMCPR for automatic flow control (AFC). The Quad Cities flow run up analyses use steep run up paths that bound GE9/GE10 and ATRIUM 9B offset equilibrium cores as well as transition cores from GE9/GE10 to ATRIUM 9B offset. Analyses are performed using XCOBRA (Reference 3) to calculate the change in critical power along a conservative flow run up path from 47 % power /30% flow to 120% power /110% flow for the MFC analysis. The flow run up path for the AFC analysis begins at 37% power /30% flow and ends at 100% power /108% flow. Linear extrapolation of the 40% and 30% core flow XCOBRA analysis results is used to obtain MCPR limits below 30% of rated core flow.
The MCPR, limits are shown in Figures 6.1 and 6.2 for the limiting fuelin Quad Cities Unit 2 Cycle 15 for the automatic flow control event. Figure 6.3 details MCPR, limits pertaining to the manual flow control event ior the limiting f uelin Quad Cities Unit 2 Cycle 15. The analysis results provide for operation up to EOFP and operation with EOD/EOOS. The cycle specific MCPR limit for Quad Cities Unit 2 shall be the maximum of the MCPR, limit depicted in these figures for the appropriate control mode and the full flow cycle specific OLMCPR. It is conservative to use the TLO MCPR, limit or full flow OLMCPR plus 0.01 (whicheUer is greater) f or SLO.This method is applied for operation up to EOFP and for EOD/EOOS conditions. Thesc
[
EMF 06180 Revision 1 Page 6 3 limits conservatively bound all transients from single loop conditions. The MCPR, limit is to i protect against boiling transition during flow excursions to maximum two pump flow; excursions to r,0ch high flows are not possible during single loop one pump operation. Thus, conservatively maintaining this two loop limit assures that there is even more thermal margin under single loop conditions than under two loop full power / full flow conditions.
The automatic flow control analyses were performed to support the base case OLMCPRs as well as the EOD/EOOS OLMCPRs (refer to Table 2.3).
The MCPR, penalty described in Reference 20 has been applied to the GE9 and GE10 MCPR, limits. The penalty is a furection of core flow with a value of 0.0 at 100% rated and increasing linearly to 0.05 at 40% rated. The penalty is linearly extrapolated for flows less than 40% of rated. Calculated results that are bound by the MCPR, limits of Figures 6.1-6.3 are given in Tables 6.6-6.8.
6.2.1 AgEmatic Flow Control If the reactor is operated in the AFC mode, variations in core power should not result in CPRs less than the established OLMC'PR for rated conditions, if the rated condition MCPR limit is observed in a reduced flow condition, a subsequent increase in power to full power along the AFC control line may result in inadvertent degradation of fuel CPRs below this reference (full flow) OLMCPR limit. The probability of boilin0 transition conditions occurring during a subsequent anticipated event may increase beyond acceptable levelt 'f this were the case.
P SPC has determined the required MCPR, limit for off rated conditions to prevent the MCPR from degrading below the cycle full power OLMCPR limit during AFC operation. This was determined by evaluating the MCPR for a given reactor power distribution at varying total reactor power and flow conditions. The variations in total core power and flow were assumed to folicew the expected relationship fa' AFC operation (Table 6.1). The power distribution chosen was such that MCPR equaled the referenced OLMCPR at 100% rated power and 108% rated flow. The expected variation of core pressure and inlet coolant subcooling with reactor power level was also considered.
- - - - - _ - - - .- - - . - - -- _ _ =_.- . _ - - - . _ - - -
f EMF 96180 Revision 1 Page 6 4 The reduced flow MCPR limits for AFC are presented graphically in Figure 6.1 for the Cycle 15 fuel types, and in tabular form in Tables 6.2 and 6.3. The MCPR, limits provide the required protection during AFC operation for operation up to EOFP and operation with EOD/EOOS.
6.2.2 Manual Flow Control This section discusses pump excurtions when the plant is in MFC, i.e., not in AFC operation mode. Because the power / flow inctease due to a single pump excursion is bound by that of a two pump excursion, only a two pump excursion is evaluated for Cycle 15.The analysis of the two pump flow excursion indicates that the limiting event scenario is a gradual quasi-steady run up. These results indicate that MCPR would decrease below the SLMCPR if the full flow reference MCPR was observed at initial conditions. Thus, an augmented MCPR limit is needed for partial flow operation to protect the two pump excursion event.
The power / flow path used for the run up is shown in Table 6.4 and bounds that calculated for constant xenon.
The results of the two pump run up analyses for manual flow control are presented in Figure 6.3 and Table 6,5 for the' Cycle 15 fuel types. When in manual flow control, the cycle.
speelfic MCPR limit for Quad Cities Unit 2 shall be the maximum of the MFC MCPR, limit or the OLMCPR The MCPR, limits provide the required protection for operation up to EOFP and operation with EOD/EOOS.
e-
- - e - v - .-,s 4--,e - , - , , ,
. EMF 96180 novision 1 Pa9e 6 5 i i
i 1
Table 6.1
~
Automatic Flow Control i Excurolon Path !
i Recirculating Flow Power . ,
(% of Rated) (% of Rated) ,
108 100 .
100 94 l
90 86 1
80 78 .
70 69 i 60 61 50 53 40 45 30 37 3
1 h
e- -
i k
1
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- e<,e---,'c-. -
w---r +--rw-+v-- +% r y - - ' w--w<-;w.,e -w--. rr v,,wv yw.~,wr --w,-- -vnw -ww---i- ,,t~.v-v-w-=-ww-- ,y---
EMF 96180 >
Revision 1 Page 6 6
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I Table 6.2 Reduced Flow MCPR Limits for Automatic Flow Control i (Base Case OLMCPR) i L
Recirculation GE9 GE10 ATRIUM 9B Offset ;
Flow MCPR, Limit for MCPR, Limit for MCPR, Limit for l (% of Rated) OLMCPR = 1.57 0LMCPR = 1.52 OLMCPR = 1.47 l 108 1.57 1.52 1.47 ,.
l 30 2.92 2.82 2.77 1 0 3.82 3.70 3.59 i
ee
.3e 0-- -e Gt t OLEM = 157 .
i e-=a-9 CC t0 OLEM e 1.62 .
41
- 3. . .
58 > .
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} ss 2 ,.
8J
- i. .
16 - .
44 i .
0 to *e u ao eso iso fota Core rio. (E of Roted)
Nort: Larger view of the above graph is found in Figure 6.1.
h
EMF 90180 ,
Revision a l Page 6 7
{
i Table 6.3 Reduced Flow MCPR Limits for l Automatic Flow Control l (EOD/E008 OLMCPR) l l
Recirculation GE9 GE10 ATRIUM 98 Offset Flow MCPR, Limit for MCPR, Limit for MCPR, Limit for
(% of Rated) OLMCPR = 1.61 OLMCPR = 1.56 OLMCPR = 1.51 1 108 1,61 1.56 1.51 l-30 2.99 2.89 2.85 l r 0 3.91 3.79 3.72 l
,, l -..i., .
- aio m en . u
~ E37 Ii"'
i 80 .
I r '. .
SJ .
" to .
- i. .
it *
- i. i n
9 #0 .0 W to 100 120 leto' Core fio (4 of Roted)
NOTE: Larger view of the above graph is found in Figure 6.2. .
. l l
EMF 96180 l Revision 1 i Page 6 8 - l l
j i
Table S.4 Manual Mow Control l Excursion Path t
?
Recirculating Flow Power s
(% of Rated) (% of Rated) 7 110 _120 100 111 ,
90 102 80 93 ,
70 83 60 74 4
50 65 40 56 30 47 ;
- i Ek I
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EMF 90180 ,
Revision 1 Page 6 9 ,
l t
Table 6.5 Reduced Flow MCPR Limits for Manual Flow Control Recirculation Flow GE9 GE10 ATRIUM 9B Of(set
(% of Rated) MCPR Limit MCPR Limit MCPR Limit i
110 1.10 1.10 1.10 l 30 1.95 1.94 1.96 l 0 2.50 2.49 2.49 1 ss Gt9 WCPR(f) tmt
,, Ct10 WC8*(f) tat j
- ATRt>D6 Of f set p(f) (mt
.
- M9 peests
- K10 seeJts 5
{ I#
a
- AfDLW98 Offset n c. ;
I,, .
31
? is .
E * -
E i.
- g j s
is -
to a 0 20 40 to to too 120 Totot Core F6= (f. Roted)
NOTE: Lar9er view of the above graph is found in Figure 6.3. .
EMF 96180 Revision 1 .t Page 610 Table 6.6 Flow Dependent MCPR Rseults OE9 Fuel (Penalty Not included) i Automatic Flow Control MCPR l Core Flow Manual Flow 1.57 1.61
(% rated) Control OLMCPR OLMCPR 110 1.100 - -
l 108 - 1.570 1.610 1 100 . 1.175 1.663 1.705 l 90 1.254 1.786 1.832 l 80 1.337 1.920 1.970 1 70 1.420 2.062 2.117 l 60 1.506 2.212 2.272 1 50 1.599 2.375 2.439 l 40 1.706 2.561 2.630
'l 30 1M82 2.853 2.926 "e>'r- - -we-- i-m 1 c, w- 1'-g- , ,r-.w-r-w-er w-' .rhme- -w --
EMF 96180 Revision 1 Page 611 Table 6.7 Flow Dependent MCPR Results GE10 Fuel (Penalty Not included)
Automatic Flow Control MCPR i Core Flow Manual Flow 1.52 1.56 l
(% rated) Control OLMCPR OLMCPR 110 1.100 -- -
108 -- 1.520 1.560 l 100 1.175 1.609 1.652 l 90 1.253 1.727 1.774 l 80 1.336 1.855 1.906 l 70 1.419 1.991 2.047 l SO 1.504 2.135 2.196- l 50 1.596 2.291 2.356 l 40 1.703 2.472 2.541 l.
30 1.878 2.756 2.830 l 4
e 6
EMF 96180 Revision 1 i
Page 612 Table 6.8 Flow Dependent MCPR Results ATRIUM 9R offset Fuel Automatic Flow Control MCPR l Core Flow Manual Flow 1.47 1.51
(% rated) Control OLMCPR OLMCPR l 110 1.100 -- -
l 108 - 1.470 1.510 1 100 1.182 1.559 1.601 l 90 1.276 1.689 1.734 l 80 ' 1.372 1.829 1.880 1 70 1.470 1.979 2.035 1 60 1.568 2.136 2.198 1 50 1.671 2.303 2.370 i 40 1.785 2.490 2.561 1 30 1.960 2.765 2.850 I
l l
l
. EMF 96180 Revision 1 '
Page 613
=
4.0 , . . . . ,-
i . , . ,
c 3 GE9 OLMCPR = 1.57 $
, e o GE10 OLMCPR = 1.52 .
3.6 . r i ATmVM-98 Offset -
. OLMCPR = 1.47 -
3.4 -
.. s.2 -
-a .
[ s.0 Su . A n
{ 2.6 y 2.4 - -
u - *
.g 2.2 . -
a: -
2.0 -
1.s -
1.6 -
1 1 1 8 ! 1 0 1 E l t g g ) g a j g I j g i 1 E i a 3 0 20 40 60 80 100 120 Total Core Flow (f; of Roted)
Figure 6.1 l Reduced Flow MCPR LirWt for l Automatic Flow Control l (Base Case OLMCPR) _
l
)
EMF 96180 !
Revision 1 i
Page 614 4.0 . , , . . . . ,
I' c 3 GE9 OLMCPR a 1.61 [
3J c 3 GE10 OLMCPR = 1.56 3.6 - 1 ATRIUW-98 Offset -
OLMCPR = 1.51 ,
, 3,4
'3 32 :
g: so -
W 2., _
m -
} 2.6 y 2.4 -
u 32.2 e
CY. -
2.0 -
1.s -
1.6 -
1,4 . . . . . . . . . . . . . . . . . . . . . . . . . . . .
0 20 40 60 80 100 120 Total Core Flow (% of Roted)
EMF 9618C ,
Revision 1 Page 615 l
'1 i
j 2.6 . . . . . . . .
. i l
GE9 MCPR(f) Limit '
14 ~ GE10 WCPR(f) Limit ATRIUM-98 Offset 5 MCPR(f) Limit -
+ GE9 Results
= 23 -
x GE10 Resul's $
.E o ATRIUM-98 Offset -
Results e - 2.0 - -
a 1.s -
+
2 '
w . m
} 1.6 -
e
. m -
e -
% 1.4 -
4 a
~
8 .
12 - -
- e -
3,o . . . . . . , . . . . , , . . . , , .
O 20 40 60 80 100 120
, Total Core Flow (r. Roted)
Figure 6.3 l Reduced Flow MCPR Limit for l P.lanual Flow Control (SLMCPR = 1.10) l W
tI s
- EMF-96 180 Revision 1
. Page 71: :
7O EVALUATION OF EOD/EOOS CONDITIONS i
Reference 17 provides a discut,sion of operation with EOD/EOOS at Quad Cities and also .
provides generic (cycle independent) penalties for operation.in EOD/EOOS. The specific t EOC/EOOS conditions supported for Quad Cities are identified in Table 2.6. ;
4 1
The Reference 17 generic EOD/EOOS penaltiet are 0.04 for OLMCPR and 5 psi for ASME- ;
, overpressurizat;an. The 0.04 MCPR generic penalty is applicable with any combination of ,
3 .
FFTR, FHOOS, RVOOS, and all coastdown operating conditions. Other EOD/EOOS conditions 4
require no OLMCPR penalty. The EOD/EOOS pressure penalty is required for operation with
. coastdown conditions, while other EOD/EOOS conditions require no pressurization penalty.
I The generic EOD/EOOS penalties are reconfirmed with explicit analyses for Cycle 15 for the
.following reasons: (1) Reference 17 analyses consisted of an equilibrium core of ATRIUM 98 -
offset fuel while the OC2C15 core is approximately 1/3 ATRIUM 9B offset and 2/3 GE9/GE10 ~
and; (2) minor. plant parameter' changes were made and needed to be dispositioned (see
{
Section 3.0). The explicit analyses documented in the following sections confirm that the
. conclusions and EOD/EOOS penalties determined for the equilibrium ATRIUM 98 offset core of Reference 17 are applicable for the OC2C15 core.
7.1 Final Feedwater Temoerature Reduction Final feedwater temperature reduction (FFTR) at the end'of cycle can be used to extend full l- power operation of the cycle.' Analyses.were performed for a 100* F reduction in feedwater -
temperature. Results for FFTR operation are presented in Tables 7.1-7.3. The results confirra that the generic per$alties are bounding. The LRNB, TTNB, and MSIV closure events are non- ~4
. limiting because the reduced feedwater temperature causes a decrease in steam flow.
=
7.2 Coastdown
. , Coastdown operation occurs after EOFP where a gradual reduction in core power occurs as the' fuel depletes. Coastdown analyses assume an additional 1500 mwd /MTU full power
I EMF 96180 l Revislot.1 Page 7 2 exposure step after EOFP to provide for operation of 15% of rated power above the equilibrium xenon coastdown power level. It is the 1500 mwd /MTU exposure extension from EOFP that forces the need to establish the coastdown penalties. As explained in Reference 17, after EOFP + 1500 mwd /MTU the core power is conservatively assumed to decrease at a rate of 10% in rated power per 1000 mwd /MTU increase in exposure. Analyses at EOFP + 1500 mwd /MTU bound coastdown at higher exposures. Results for coastdown are presented in Tables 7.4-7.6. The results confirm that the generic penalties are bounding. In addition, coastdown operation for 100% power /108% flow at EOFP + 1500 mwd /MTU produced the largest lower plenum pressurization for ASME MSIV closure overpressurization analyses, As identified in Table 3.1 the limiting MSIV closure event was repeated with the SRV in service and the lowest opening pressure set point safety valve out of service (SVOOS). As seen in Table 7.6, SRVOOS bounds SVOOS.
7.3 Combined Final Feedwater Temocrature Reduction /Coastdown Results for combined FFTR/Coastdown are presented in Tables 7.7-7.9.The results confirm that the generic penalties are bounding.
7.4 Feedwater Heater (s) Out'of Service The feedwater heater out of service (FHOOS) scenario assumes a 100*F reduction in the feedwater temperature. Operation with FHOOS is similar to operation with FFTR except that the reduction in feedwater temperature can occur at any time in during the cycle. Results for FHOOS are presented in Tables 7.10-7.12. The results confirm that the generic penalties are bounding. The LRNB, TTNo, and MSIV closure events are non-limiting because the reduced feedwater temperature causes a decrease in steam flow.
7.5 Combined Feedwater Heaters Out of Service /Coastdown Results for combined FHOOS/coastdown are presented in Tables 7.13-7.14. The results
~
confirm that the generic penalties are bounding.
4 EPAF 98180 Revision 1 Page 7 3 Table 7.1 Quad Cities Unit 2 Cycle 15 Final Feedwater Temperature Reduction MCPR Results and Comparison to Base Case (TSSS Insertion Times)
Peak Peak Maximum Neutron Heat Vesset/
Flux Flux Dome Pressure Change in Transent Power / Flow (% tated) f% rateds (neen) (ACPR)*' ACPR From Base Case *'
LRNB 100 /108 524 129 1267/1234 0.40 / 0.36 / 0.33 0.05 / 0.04 / 0.01 LRNB 100/100 504 129 1267/1237 0.39 / 0.34 / 0.31 -0.05 /-0.04 /-0.01 LRNB 100 / 87 451 128 1268 /1241 0.32 / 0.28 / 0.26 -0.05 / 0.04 / 0.02 TTNB 100/108 513 129 1264 /12.31 0.40 /0.36 / 0.32 0.05 / 0.04 / 0.02 TTNB 100/100 496 128 1264 /1233 0.38 / 0.34 / O.30 0.05 /-0.04 / 0.02 TTNB 100 / 87 445 127 1265 /1239 0.32 /0.28 / 0.25 0.05 /-0.04 / 0.02 FWCF 100/108 506 139 1156/1122 0.47 /0.42 / 0.39 0.00 / O.00 /0.02 FWCF 100/100 488 139 1154 /1123 0.45/0.41 / 0.38 0.01 / 0.00 / 0.03 FWCF 100 / 87 433 137 1150 /1123 0.39 /0.36 / 0.33 0.00 / O.01 /0.03 Lower plenum S'
Third cycle GE9/second cycle GE10/first cycle ATRIUM 9B fuel
EMF 96180 Revision 1 Page 7 4 Table 7.2 Quad Cities Unit 2 Cycle 15 Final Feedwater Temperature Reduction MCPR Results and Comparison to Base Case (NSS Insertion Times)
Peak Peak Mannnum Neutron Heat Vessell Flux Flux Dome Pressure Change in Treasent Power / Flow (% rated) (% rated) lose) ( ACPR)*' ACPR From Base Case *8 LRNB 100 /108 518 129 1232 /1199 0.40 /0.36 / O.33 0.05 /-0.04 /-0.01 LRNB 100/100 499 128 1232 /1202 0.38 /0.34 /0.30 -0.05 /-0.04 / 0.02 LRNB 100 /87 444 127 1232 /1206 0.31 /0.28 /0.25 0.05/-0.04/ 0.02 TTN8 100 /108 507 128 1229 /1196 0.40 /0.36 / 0.32 -0.05 / 0.04 / 0.02 TTNB 100/100 488 128 1229 /1199 0.38 /0.34 / 0.30 -0.05 / 0.04 /-0.01 TTNB 100/87 434 126 1230 /1203 0.31 /0.27 /0.25 -0.05 /-0.05/-0.01 FWCF 100/108 500 139 1136/1102 0.46 /O.42 / 0.39 -0.01 / 0.00 /0.02 FWCF 100/100 481 138 1133 /1102 0.45/0.40 /0.37 0.00 /0.00 /0.03 FWCF 100 / 87 423 136 1129/1102 0.38 /0.35 / 0.33 0.00 / 0.01 / 0.04 Lower plenum Third-cycle GE9/second cycle GE10/first cycle ATRIUM-9B fuel
EMF-96180 Revision 1 Page 7 5 Table 7,3 Quad Cities Unit 2 Cycle 16 Final Feedwater Temperature Reduction ASME Overpressurization Analysis Results and Comparison to Base Case (TSSS Insertion Times)
Peak Peak Meximum Chan0e in VesseP'/
Neutron Heat VesseF'/ Dome Pressure From Flux Flux Dome Piessure Bose Case I
Trarwent PowerNiow (% rated) f% rated) (nain) (nell MSIV 100/108 271 128 1299 / 1266 24 / 24 MSIV 100/100 271 126 1299 /1269 23 / 23 MSIV 100 /87 2.9 124 1299 /1273 22 / 21 s
Lower plenum
EMF 96180
- Revision 1 Page 7 6 Table 7.4 Quad Cities Unit 2 Cycle 15 -
~
Coastdown Operation MCPR Results and Comparison to Base Case (TSSS Insertion Times)
Peak Peak Maximum Neutron Heat Vessel *'/
Flux Nu Dome Pressure Change in im Power /Aow (% rated) (% rated),, tonen) (ACPR)*' ACPR From Base Case *'
LRNB 100 /108 643 136 1309 /1275 0.49 /0.43 / 0.37 0.04 /0.03 / 0.03 LRNB 100 /100 601 135 1308 /1277 0.46 / 0.41 / 0.35 0.02 /0.03 /0.03 LRNB 100 / 87 496 132 1306 /1279 0.37 / 0.33 / 0.30 0.00 / 0.01 / 0.02 TTNB 100 /108 635 136 1307 /1273 0.48 /0.43 / 0.37 0.03 /0.03/0.03 TTNB 100 /100 595 135 1306 /1275 0.46 / 0.40 / 0.35 0.03 /0.02 /0.03 TTNB 100 /87 493 132 1305 /1277 0.37 /0.33 / 0.30 0.00 / O.01 / 0.03 FWCF 100 /108 618 140 1205 /1171 0.49 / 0.44 / 0.39 0.02 /0.02 /0.02 FWCF 100 /100 571 139 1202 /1171 0.47 / 0.42 / 0.37 0.01 / 0.01 / 0.02 FWCF 100 /87 457 134 1198 /1170 0.37 /0.34 / O.30 -0.02 / 0.01 / 0.00
'*' Lower plenum
. Third cycle GE9/second cycle GE10/first cycle ATRIUM-98 fuel
EMF-96180 Revision 1 Page 7 7 Table 7.5 Quad Cities Unit 2 Cycle 15 Coastdown Operation MCPR Results and Comparison to Base Case (NSS Insertion Times)
Peak Peak Maximum Neutron Heat Vessel"'l Flux Flum Dome Pressure Change in Transgat. hwerSlow (% rated) (% rated) tosin) (A CPR)*' ACPR From sese Case *'
LRNB 100 /108 633 135 1278 /1244 0.48 /0.43/ 0.37 0.03 /0.03 / 0.03 LRNB 100 /100 589 134 1278 /1247 0.45 / 0.40 / 0.35 0.02 /0.02 / 0.03 LRNB 100 / 87 477 131 1276 /1250 0.36 / 0.32 / 0.27 0.00 / 0.00 / 0.00 TTNB 100 /108 623 135 1277 /1243 0.48 /0.43 /0.37 0.03 / 0,03 / 0.03 TTNB 100 /100 590 134, 1276 /1245 0.45 /0.40 / 0.34 0.02 / 0.02 / O.03 TTNB 100 / 87 472 131 1275 /1248 0.36 / 0.32 / 0.27 0.00 / 0.00 / 0.01 FWCF 100 /108 606 139 1178 /1143 0.49 / 0.44 / 0.39 0.02 / O.02 / O.02 FWCF 100 /100 557 138 1175 /1143 0.46 / 0.41 / 0.37 0.01 / 0.01 / 0.03 FWCF 100 / 87 437 133 1169 /1142 0.36 /0.33/0.29 0.02 / 0.01 / 0.00
. Lower plenum Third cycle GE9/second cycle.GE10/first cycle ATRIUM 98 fuel
a x - - - -
i . .
EMF 96180 Revision 1 Page_7 8 Table 7.6 Quad Cities Unit 2 Cycle 15 Coastdown Operation ASME Overpressurization Analysis Results and Comparison to Base Case (TSSS Insertior. Times)
Peak Peak Maximum Change in Vessel *'/
Neutron Heat Vessel"'/ Dome Pressure From Flux Flux Dome Pressure seee Case Trenesent Power / Flow f % rated) f% rated) (nein) ,
fosi)
MSlV 100 /108 318 132 1326 /1293 3/3 MSIV 100 /108 318 132 1323 /1290 NA MSIV 100 /100 315 130 1325 /1295 3/3
'*: Lower plenum -
'*' SVOOS. Results demonstrate that SRVOOS is more limiting than SVOOS.
EMF 96180 Revision 1 Page 7 9 Table 7.7 Quad Cities Unit 2 Cycle 15 Combined FFTR/Coastdown MCPR Results and Comparison to Base Case (TSSS Insertion Times)
Peak Peak Maximum Neutron Heat Vester*'t Flum Flux Dome Pressure Change in Transsnt PowerMiow (% rated) 1% rated) (Daio) (A CPR)*' ACPR From Base Case *'
LRNB 100 /108 571 131 1276 /1243 0.42 / 0.38 / 0.36 -0.03 /-0.02 / 0.02 LRNB 100/100 538 131 1277 /1246 0.39 /0.36 / 0.33 0.05 /-0.02 / 0.01 TTNB 100/108 559 131 1273 /1240 0.42 / 0.38 / 0.36 0.03 / 0.02 / 0.02 TTNB 100/100 529 130 1273 /1243 0.39 /0.36 / 0.33 0.04 / 0.02 / 0.01 FWCF 100/108 538 140 1161 /1127 0.47 /0.43 /O.41 0.00 / 0.01 / 0,04 FWCF 100/100 502 139 1158 /1127 0.45 / 0.41 / 0.39 0.01 / 0.00 / 0.04 Lower plenum Third cycle GE9/second cycle GE10/first cycle ATRIUM 9B fuel
EMF.96180 Revision 1 Page 710 Table 7.8 Quad Cities Unit 2 Cycle 15 Combined FFTR/Coastdown MCPR Results and Comparison to Base Case (NSS Insertion Times)
Peek Peak Maximum Neutron Heat Vesser/
Flux Flum Dome Pressure Change in Transent _ Power / Flow _ (% rated) (% toted) _ tosiot _
(ACPRi*' ACPR From Base Case *'
LRNB 100/108 563 131 1239/1206 0.42 / 0.38 /0.36 0.03/-0.02 /0.02 LRNB 100/100 530 130 1239 /1208 0.39 /0.35 /O.33 0.04 / 0.03 / 0.01 TTNB 100/108 550 130 1236 /1204 0.41 / 0.38 /0.35 0.04 / 0.02 /0.01 YrNB 100/100 517 129 1236 /1206 0.38 /0.35 / 0.33 0.05/ 0.03/0.02 FWCF 100/108 528 140 1139 /1105 0.46/O.43 /0.41 -0.01 / 0.01 / 0.04 FWCF 100/100 491 138 1136 /1105 0.44 / 0.41 / 0.38 -0.01 / 0.01 / 0.04
( '*' Lower plenum -
- Third cycle GE9/second cycle GE10/first cycle ATRIUM 9B fuel l
l
_ _ ~- _ _ _ _ _ _ _ _ __ _ _ . __
EMF 96180 Revision 1 Page 711 Table 7.9 Quad Cities Unit 2 Cycle 15 Combined FFTR/Coastdown ASME Overpressurization Analysis Hesults and Comparison to Base Case (TSSS Insertion Times) .
Peak Peak Maximum Change in Vessel */
Neutron Heat Vessel */ Dome Pressure From Flux Flux Dome Pressurs Base Case Transent Powe'/ Flow (% rated) (% rated) foam) (Dei)
MSIV 100 /108 287 128 1305 /1272 18 / 18 MSIV 100 /100 286 120 1304 /1274 18/ 18 l
l l
I h
- Lower plenum
EMF 96180.
Revision 1 Page 712 Table 7.10 Quad Cities Unit 2 Cycle 15 Feedwater Hooter Out of Service MCPR Results Comparison to Base Case (TSSS insertion Times)
Peak Peak Maximum N uiron Hoet vesseral Flux Flou Dome Pressure Change in iranseent Fower/ Flow f% rated) (% rated) (oeial (ACM)** _ACM From Base Case *'
FWCF 100 /108 491 138 1155 /1121 0.46 /0.42 / 0.37 -0.01 / 0.00 / 0.00 FWCF 100/100 478 138 1153 /1122 0.44 / 0.40 / 0.35 0.02 / 0.01 / 0.00 FWCF 100 / 87 432 136 1149 /1122 0.39 / 0.36 / 0.32 0.00 / 0.01 / 0.02 r
4
'*' Lower plenum 4
- Third-cycle GE9/second cycle GE10/first cycle ATRIUM 9B fuel 1
EMF 96180 Revision i Page 713 Table 7.11 Quad Cities Unit 2 Cycle 15 Feedwater Heater Out of Service MCPR Results and Comparison to Base Case (NSS Insertion Times)
Peak Peak Menimum Neutron Heat Vesser/
F6ux Fluu Dome Pressure Change in Transent Power / Flow f% rn (% rated) (osm) (A CPR)*' ACPR From Base Case *'
FWCF 100 /108 486 138 1135 /1101 0.45 / 0.41 / 0.37 0.02 /-0.01 / 0.00 FWCF 100 /100 472 137 1132 /1101 0.44 /0.40 /0,35 -0.01 / 0.00 / 0.01 FWCF 100 /87 423 136 1128 /1101 0.38 / 0.35 / 0.31 0.00 / 0.01 / 0.02
Lower plenum Third-cycle GE9/second cycle GE10/first cycle ATRIUM 98 fuel
c EMF 96 ' 30 l Revision 1 Page 714 Table 7.12 Quad Cities Unit 2 Cycle 15 Feedwater Heater Out of Service ASME Overpressurization Analysis Results and Comparison to Base Case (TSSS Insertion Times)
Peak Peek Menimum Change in Vessef*'l Neutron Heat Vessel */ Dome Pressure Fet,m Fluz Flux Dome Pressure Base Case Transent Power / Flow (% rated) (% rated) (osio) fosi) ,
MSIV 100 /108 266 127 1298 /1265 25 / 25 MSIV 100 /100 267 126 1298 /1268 24 / 24 MSIV 100 / 87 266 123 1298 /1272 23 / 22 e
- Lower plenum i
. . . - . . .. . - - - . ~. . .- .. .
EMF 96180 Revision 1 Page 715 Table 7.13 Quad Cities Unit 2 Cycle 15 i Combined FHOOS/Coastdown MCPR Resutts and Comparison to Base Case (TSSS Insertion Times)
Peak Peak Maximum Neutron Heat Vessel *'/
Fium Flux Dome Proceurs Change in Trenoient Powef/ Flow (% toted) (% toted) (Dein) (ACPR)*' 6CPR From Base Case *'
FWCF 100/108 525 140 1159/1125 0.48 /0.43 / 0.39 0.01 / 0.01 /0.02 FWCF 100/100 499 139 1156/1125 0.45 /0.42 / 0.38 0.01 / 0.01 /0.03 FWCF 100 / 87 430 137 1152/1125 0.39 /0.36 / 0.33 0.00 / 0.01 /0.03
Lower plenum
'*' Third cycle GE9/second cycle GE10/first cycle ATRIUM 98 fuel W-
EMF 96180 Revision 1 Page 716 Table 7.14 Quad Cities Unit 2 Cycle 15 Combined FHOOS/Coastdown MCPR Results and Comparison to Base Case (NSS insertion Times)
Peak Peak Maximum Neutron Heat Vesset"'/
Fium Flum Dome Pressure Change in Transent PowerSlow (% rated) (% rated) (p6.at ($CPH)*' 6CPR From Base Ceae*'
FWCF 100/108 518 140 1138 /1104 0.47 / 0.43 /0.39 0.00 / 0.01 /0.02 l FWCF 100/100 490 139 1135 /1104 0.45 /0.41 / 0.38 0.00 / 0.01 / 0.04 FWCF 100 / 87 418 136 1130 /1103 0.38 / 0.35 / O.32 0.00 / 0.01 / 0.03 l
l l
l
\
'S L Lower plenum
'S Third cycle GE9/second cycle GE10/first cycle ATRIUM 9B fuel
EMF 96180 Revision 1 Page 81
8.0 REFERENCES
1.- Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Quad Cities Unit 2 Cycle 15 Calculation Plan Final Version,' JHR:96:187, May 17,1996.
- 2. COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis, ANF 913(P)(A), Volume 1, Revision 1, Supplements 2, 3, and 4, Advanced Nuclear Fuels _ Corporation, August 1990,
- 3. Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors, ANF 524(P)(A), Revision 2/ Supplement 1, Revision 2/ Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.
- 4. ANFB Critic 4/ Power Correlation, ANF 1125(P)(A), Supplements 1 and 2, Advanced Nuclear Fuels Corporation, April 1990.
- 5. Advanced Nuclear Fuels Methodology for Bolling Water Reactors: Benchmark Results for CASMO 3G/ MICROBURN-B Calculation Methodology, XN NF 8019(P)(A),
Volume 1, Supplement 3, Appendix F and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.
- 6. Exxon Nuclear Methodology for Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary Description, XN NF 80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Company, Inc., January 1987.
- 7. Quad Cities Unit 2 Cycle 15 Principal Transient Analysis Parameters, EMF 96-015, Revision 1, Siemens Power Corporation - Nuclear Division, August 1996.
- 8. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN NF 79 71(P), Revision 2, including Supplements 1 through 3(P)(A), Exxon Nuclear Company, Inc., March 1986.
- 9. XCOBRA T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN NF 84105(P)(A), Volume 1 and Supplements 1 and 2, Exxon Nuclear Company, Inc., February 1987,
- 10. RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model, XN-NF 81-58(P)(A), Revision 2, and Supplements 1 and 2, Exxon Nuclear Company, Inc., March 1984.
- 11. Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Disposition of the Feedwater Control System Change in Quad Cities Unit 2 Final Letter Report," SPCCWE:96:040/
JHR:96:048, February 14,1996.
- 12. Principal Reload Fuel Design Parameters Quad Cities 2, QCB-1, Cycle 15 (A TRIUM'-
SBliEMF-96120, Revision 0, Siemens Power Corporation Nuclear Division, June 1996.
EMF.96180 Revision 1
- Page 8 2
8.0 REFERENCES
_ (Continued)
- 13. Impact of Failed / Bypassed LPRMs and TIPS and Extended LPRM Calibration Intervalon Radial Bundle Power Uncertainty, EMF 1903(P), Revision 2, Siemens Power ,
Corporation Nuclear Division, October 1996.
l 14. . Quad Cities Station 2 Technical Specifications, as amended.
l 15. ANFB Critical Power Correlation Uncertainty for Limited-Data Sets, EMF-1125 l Supplement 1 Appendix D, Siemens Power Corporation Nuclear Division, April 1997,
- 16. Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Responses to NRC Request for Additional Information for Application of the ANFB Correlation to Co Resident GE Fuel at LaSalle and Quad Cities," JHR:96:235, June 28,1996.
- 17. Quad Cities Extended Operating Domain (E00) and Equipment Out of Service (E00S)
Safety Analysi:: for ATRIUM
l 18. Quad Cities Unit 2 Cycle 15 Reload Analysis, EMF 96-177, Revision 2, Siemens Power j Corporation Nuclear Division, April 1997,
- 19. Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Draft Principal Transient Analysis Parameters for Quad Cities Unit 2 Cycle 15," JHR:96:052, February 16,1996.
- 20. Application of ANFB CriticalPower Correlation to Coresident GEFuelat the Quad Cities and LaSalle Nuclear Power Stations, EMF 95 049(P), Siemens Power Corporation -
Nuclear Division, October 1995,
- 21. Application of ANFB CriticalPower CriticalPower Correlation to Coresident GEFuelfor Quad Cities Unit 2 Cycle 15, EMF 96-051(P), Siemens Power Corporation - Nuclear Division, May 1996.
- 22. Letter, R. J. Chin (Comed) to J. H. Riddle (SPC), "GE10 Channel Bow Data
' Assumptions for Quad Cities MCPR Safety Limit Calculations," NFS:BND:96105, September 19,1996.
- t 23. Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Quad Citier. Unit 2 Cycle 15 MCPR l Safety Limits for ATRIUM
- 98 Fuel increased Additive Constant Uncertainty,"
l JHR:97:155, April 18,1997.
- 24. Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Updated Transient Power History Data for Confirming Mechanical Limits for GE Fuel for Quad Cities Unit 2 Cycle 15,"
JHR:96:389, October 4,1996.
w -
l F.MF-96 180 Revision 1 Page A i Appendix A Margin to Unpiped Safety Valves
=
EMF 90180 Revision 1 Page A 1 ,
Appendix A Margin to Unpiped Safety Valves SPC performed analyses for Quad Cities Unit 2 Cycle 15 to determine the margin between peak steam line pressure and the lowest set point of the unpiped safety valves. Quad Cities procedures require a 60 psi margin for the main steamisolation valve closure - unpiped safety valve margin (MSIVC USM) analysis. The load rejection no bypass - unpiped safety valve margin (LRNB USM) analysis was also performed. Based on analyses performed to support reload licensing of Cycle 15, the limiting initial conditions for steam line pressurization occur '
at 100% core power and 87% core flow (100%P/8': %F). The lowest nominal set point for a Quad Cities unpiped safety valve is 1254.7 psia.
Because the unpiped safety valve margin analyses are not licensing analyses (margin requirements are specified in Quad Cities operating procedures), some of the conservatism normally assumed in COTRANSA2 analyses is relaxed. The MSIVC USM analysis with direct scram results in a fairly mild reactor pressurization. The relief valves have sufficient capacity to depressurize the reactor once the valves actuate. The MSIVC USM analysis with direct scram was performed with Technical Specification scram speed and scram delay. The same relief valve set points, stroke times and delays used in the thermal margin licensing analyses were used for the MSIVC USM analyses. Analyses were performed with the SRV not credited.
' Analyses were performed for 100%P/87%F at EOFP and EOFP+ 1500 mwd /MTU to cover coastdown operation (Reference A.1). For the MSIVC-USM transient, the calculated peak steam line pressure is 1134.7 psia. This results in a calculated margin of 120.0 psi to the lowest unpiped safety valve set point as shown in Table A.1. The required 60 psi margin is met.
For the LRNB USM analysis, nominal relief valve set points, best-estimate relief valve opening times, and best estimate scram speed are used. All relief valves are assumed to be operable.
~
A best estimate relief valve opening delay time of 1.25 seconds and a stroke time of 0.20 second were used in the analyses baser en values from Reference A.2. Analyses are
EMF 96180 Revision 1 Page A 2 performed with and without the relief function of the SRV Scram insertion is based on plant- .
specific data (NSS insertion times) provided by Comed in Reference A.2.
Refer to Table A.1 for the results of the LRNB USM analyses. Analyses were performed for state points 100%P/108%F,100%P/100%F, and 100%P/87%F for exposures of EOFP and EOFP+ 1600 mwd /MTU to cover coastdown operation.
Quad Cities analyses indicate that a 1% decrease in rated core power increases pressure . ,
margin approximately 4 psi (Reference A.3).
6 6
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e EMF 9618C Revision 1 Page A 3
. Table A.1 Margin to Opening Unpiped Safety Valve Results Maximum SRV Pressure Margin Transient Exposure Power / Flow (psia) (psi)
LRNB USM EOFP 100 /108 1236.5 18.2 LRNB USM EOFP+ 1500 mwd /MTU 100 /108 1242.6 12.1 LRNB USM SRVOOS EOFP 100 /108 1245.2 9.5 LRNB USM SRVOOS EOFP+ 1500 mwd /MTU 100 /108 1251.4 3.3 LRNB USM EOFP 100 /100 1238.3 16.4 LRNB USM EOFP+ 1500 mwd /MTU 100 /100 1243.7 11.0 LRNB USM SRVOOS EOFP 100 /100 1247.0 7.7 LRNB USM SRVOOS EOFP+ 1500 mwd /MTU 100 /100 1252,5 2.2 LRNB USM EOFP 100 /87 1239.8 14.9 LRNB USM EOFP + 1500 mwd /MTU 100 /87 1244.3 10.4 LRNB USM SRVOOS EOFP 100/87 1248.7 6.0 LRNB USM SRVOOS EOFP+1500 mwd /MTU 100 /87 1253.2 1.5 MSIVC USM. EOFP 100 /87 1134.7_ 120.0 MSIVC-USM EOFP+ 1500 mwd /MTU 100/87 1134.7 120.0
"- EMF 96-180 Revision 1 Page A 4 A.1 REFERENCES
^ A.1 Quad Cities Extended Operating Domain (EOD) and Equipment Out of Service (EOOS)
Safety Analysis /or ATR/UM* 9B Fuel, EMF 96 037(P), Revision 1, Elemens Power Corporation Nuclear Division, September 1996.
A.2 Quad Cities Unit 2 Cycle 15 Principal Transient Analysis Parameters, EMF 96-015, Revision 1, Siemens Power Corporation - Nuclear Division, August 1996.
- A.3 Letter, M. L. Hymas (SPC) to R. J. Chin (Comed), "Dresden and Quad Cities Unpiped Cafety Valve Margin Analyses," MLH:96:025, May 23,1996.
h-a k
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EMr.90180 Revision 1 issue Date: 4/18/97 Quad Cities Unit 2 Cycle 15 Mont Transient Analysis Distribution D. J. Braun, 34
- 0. C. Brown, 34 D. G. Carr, 34 M. E. Garrett, 34 J. G. Ingham, 36 D. M. Knee, 34
<- J. L. Maryott, 36 J. H. Riddle, 38 (8 customer copies)
R. R. Schnepp, 34 4
P. E. Smith, 34 J. A. White, 31
{ A. W. Will, 34 Document Control (2)
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ATTACHMENT 2 (Affidavit Pursuant to 10CFR2.790)
- i. *
- A F FID AVIT STATE OF WASHINGTON ) )
) sp. '
COUNTY OF BENTON ) ;
I, J. S. Holm being duly sworn, hereby say and depose:
- 1. I am Manager, Product Licensing, for Siemens Power Corporation ("SPC"), and as such I am authorized to exocute this Affidavit.
- 2. I am f amiliar with SPC's detailed document control system and policies which guvern the protection and control of inforrnation.
- 3. I am f amiliar with the report EMF 96180(P) Revision 1, " Quad Cities Unit 2 Cycle 15 Plant Transient Analysis," April 1997, referred to as " Document" forwarded to the U.S. Nuclear Regulatory Commission by letter L. W. Pearce (Comed) to the U.S. Nuclear
{lcaplatorv Commission, "Q2C15 Core Operatino Lirnits Reoort Revision 1 " November 4.199_7.
Some information contained in this Document has been classified by SPC as proprietary in accordance with the control system and policies established by SPC for the control and protection of information.
- 4. The Document contains information of a proprietary and confidential nature and is of thu type customarily held in confidence by SPC and not made available to the public.
Based on my experience, I am aware that other companies regard information of the kind contained in the Document as proprietary and confidential.
- , * .- l l
i D. The Document has been made available to the U.S. Nuclear Regulatoiy 3 L !
Commission in confidence, with the request that the information contained in the Document will l i
not be disclosed or divulged. - i l !
- 6. The Document contains information which is vital to a competitive advantage ;
i
' i
- 7. The information contained in the Document is considered to be proprietary by i
SPC because it reveals certain distinguishing aspects of SPC licer. sing methodology which secure competitive advantage to SPC for fuel design optimization and marketability, and
. includes information utilized by SPC in its business which affords SPC an opportunity to obtain a .
competitive advantage over its competitors who do not or may not know or use the information contained in the Document.
- 8. The disclosure of the proprietary information contained in the Document to a competitor would permit the competitor to reduce its expenditure of money and manpower and r
to improve its competitive position by giving it valuable insights into SPC licensing methodology and would result in substantial harm to the competitive positicn of SPC.
- 9. The Document contains proprietary information which is held in confidence by SPC and is not available in public sources.
- 10. In accordance with SPC's policies governing the protection and control of information, proprietary information contained in the Document has bcen made available, on a limited basis, to others outside SPC only as required and under suitable agreement providing for nondisclostre and limited use of the information,
- 11. SPC policy requires that proprietary information be kept in a secured file or area and distributed on a needi to know basis.
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- 12. Information in this Document provides insight into SPC licensing methodology developed by SPC. SPC has invested significant resources in developing the methodology as well as the strategy for this application. Assuming a competitor had available the same background data and incentives as SPC, the competitor might, at a minimum, develop the information for the same expenditure of manpower and money as SPC.
THAT the statements made hereinabove are, to the best of my knowledge,
.information, and belief, truthful and complete.
FURTHER AFFIANT SAYETH NOT. '
/'
SUBSCRIBED before me this M day of b(J*k .1997. CD M
- 4. $
+0 TARP
- A %9thO h h er,M A q 1(Y\. $ 0p' o
- wa Sue M. Galplo NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 02/27/00