ML20236T839
| ML20236T839 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/30/1998 |
| From: | Serell D, Will A SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17191A805 | List: |
| References | |
| EMF-96-141, EMF-96-141-R01, EMF-96-141-R1, NUDOCS 9807290063 | |
| Download: ML20236T839 (85) | |
Text
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Dresden Unit 3 Cycle 15 Reload Analysis Report I
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9807290063 900724 DR ADOCK 05000249 pDR-June 1998 Dresden Unit 3 Cycle 15 N -_-
ISSUED IN SPC-ND ON t im Siemens Power Corporation - Nuclear Division DoCUF NT STEM
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DATE: ()
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EMF-96-141 Revision 1 Issue Date:
Dresden Unit 3 Cycle 15 Reload Analysis Prepared by:
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A. W. Will, Engineer BWR Safety Analysis Nuclear Engineering
'and D. C. Serell, Engineer BWR Safety Analysis Nuclear Engineering June 1998 i.
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i Customer Disclaimer Important Notice Regarding Contents and Use of This Document Please Read CareMiy 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, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on its behalf:
a.
makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method 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 damages 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 rights 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 writing by Siemens Power Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document.
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EMF 96-141 Revision 1 Page ii Nature of Change
Title:
Dresden Unit 3 Cycle 15 Reload Analysis Superseded Issue:
O ltem Paae No Nature of Chance
GENERAL COMMENT
- The report is revised to include analysis results and licensing limits for new plant operating conditions. Specifically, the revised parameters are a steam flow of 9.90 Mlbm/hr, feedwater flow of 9.87 Mlbm/hr, control rod drive flow of 0.03 Mlbm/hr, and a feedwater temperature range of 340 F-360 F. Revised LOCA analysis results and extended 9x9-2 LHGR limits are included.
1 3
Added text describing extended LHGR limits for 9x9-2 fuel.
2 4
Changed SLMCPR Reference to 9.16.
3 5
Updated Thermal Power and Feedwater Flow Rate at SLMCPR. Added a footnote to the feedwater temperature in Section 3.3.1.
4 8
Added statements of stability analysis applicability with increased steam flow.
5 9
Replaced results in Section 5.1 with results from analyses performed with revised operating conditions. Revised footnote (a).
6 10 Updated ASME overpressurization results and added footnote (a).
7 11 Updated ACPR results and MCPR limits. Added text to footnote (a), and added footnote (c).
8 14 Extended steady-state LHGR limits for 9x9-2 fuel to 50.9 GWd/MTU and
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updated the reference to Reference 9.14.
9 16 Updated References 9.3,9.4, and 9.5 to current reports. Added References 9.14, 9.15, and 9.16.
10 19,20 Eliminated maximum F-eff value from Figures 3.2 and 3.3.
l 11 35 Updated Figure 7.1 consistent with item 8 above.
12 A-2 Added footnote to single-loop MCPR limit determination.
13 A-3 Added footnote to SLO MAPLHGR multiplier.
14 A-3 linciated Reference A.2 to latest revision. Added Reference A.3.
Changed items are further identified with ( l ) in left margins.
NOTE:
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EMF-96141 j
Revision 1
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Page iii Contents Section Pace
1.0 INTRODUCTION
............................................. 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS...
......................... 3 3.0 THERMAL-HYDRAULIC DESIGN ANALYSIS
.....................,,,.. 4 3.2 Hydraulic Characterization 4
l 3.2.1 Hydraulic Compatibility............................. 4 3.2.3 Fuel Centerline Temperature
......................... 4 i
3.2.5 B y p a s s Fl o w..................................... 4 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR) 4 3.3.1 Coolant Thermodynamic Condition..................... 5 3.3.2 Design Basis Radial Power Distribution 5
3.3.3 Design Basis Local Power Distribution.....
5 4.0 NUCLEAR DESIG N AN ALYSIS.................................... 6 4.1 Fuel Bundle Nuclear Design Analysis......................... 6 4.2 Core Nuclear Design Analysis.................,............ 7 4.2.1 Core Configuration
................................ 7 4.2.2 Core Reactivity Characteristics.
7 4.2.4 Core Hydrodynamic Stability 8
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5.0 ANTICIPATED OPERATIONAL OCCURRENCES
....................... 9 l
5.1 Analysis of Plant Transients at Rated Conditions 9
5.2-Analysis for Reduced Flow Operation 9
5.3 Analysis for Reduced Power Operation 9
5.4 ASME Overpressurization Analysis
.......................... 10 5.5 Control Rod Withdrawal Error.............................. 10 5.6 Fuel L o a din g Erro r...................................... 10 5.7 Determination of Thermal Margins........................... 11 t
6.0.
POSTUL ATED AC CIDENTS..................................... 12 l
6.1 Los s-of-Coola nt Accident................................. 12 t
6.1.1 Break Location Spectrum........................... 12 6.1.2 Break Size Spectrum.............................. 12 6.1.3 MAPLHGR Analyses 12 6.2 Control Rod Drop Accident................................ 12 l
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EMF 96141 t
r 1-Revision 1 l
Page iv Contents (Continued)
Section Pace 7.0 TECHNICAL SPECIFICATIONS..............
13 7.1 Limiting Safety System Settings......................
13 7.1.1 MCPR Fuel Cladding integrity Safety Limit............... 13 l
7.1.2 Steam Dome Pressure Safety Limit.................... 13 7.2 Limiting Conditions for Operation 13 7.2.1 Average Planar Linear Heat Generation Rate 13 7.2.2 Minimum Critical Power Ratio 14 7.2.3 Linear Heat Generation Rate......................... 14 8.0 METH O D O LO G Y R EFEREN C ES.................................. 15 9.0 ADDITIONAL REFERENCES.................
16 APPENDIX A Single-Loop Operation............................ A-1 I
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EMF-96-141 Revision 1 Page v I
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Tables Table Paae 4.1-Neutronic Design Values 21 I
l Figures Fiaure Paae 3.1 Design Basis Radial Power Distribution for SLMCPR Determination......... 18 3.2-Design Basis Local Power Distribution for SPC-ND ATRIUM-9B Fuel (SPCA9-3268-11GZ-80M) Uncontrolled at 15,000 mwd /MTU and 70% Void for SLMCPR Determination............................. 19 3.3 Design Basis Local Power Distribution for SPC-ND ATRIUM-9B Fuel
'(SPCA9-3398-6GZ-80M) Uncontrolled at 17,500 mwd /MTU and 70% Void f or SLMCPR Determination............................. 20 4.1 Dresden Unit 3 Cycle 15 SPCA9 343L-9G3.5 Enrichment Distribution 22 4.2 Dresden Unit 3 Cycle 15 SPCA9-344L-11G5.5 Enrichment Distribution 23 4.3 Dresden Unit 3 Cycle 15 SPCA9-378L-11G5.5 Enrichment Distribution 24 4.4 Dresden Unit 3 Cycle 15 SPCA9 378L-9G4.5 Enrichment Distribution 25 4.5 Dresden Unit 3 Cycle 15 SPCA9-362L-6G3.5 Enrichment Distribution 26 4.6 Dresden Unit 3 Cycle 15 SPCA9 362L-6G4.5 Enrichment Distribution
......27 4.7 Dresden Unit 3 Cycle 15 SPCA9 388L-6G5.5 Enrichment Distribution...... 28 4.8 Dresden Unit 3 Reload Batch DRC 8 Axial Fuel Assembly Design.......... 29 4.9 Dresden Unit 3 Reload Batch DRC-8A Axial Fuel Assembly Design......... 30
-4.10 Dresden Unit 3 Cycle 15 Reference Loading Map (Quarter Core S ym metric Loading ).......................................... 31 l
5.1 Starting Control Rod Pattern for Control Rod Withdrawal Analysis......... 32 5.2 Reduced Flow MCPR Limit for Manual Flow Control (SLMCPR = 1.08)
Applicable to ATRIUM-98 and 9x9-2 Fuel 33 5.3 Reduced Flow MCPR Limit for Automatic Flow Control Applicable to ATRIUM-9B and 9x9-2 Fuel.................................... 34 4
7.1 Protection Against Power Transient LHGR Limit for 9x9 2 Fuel........... 35 7.2 Protection Against Power Transient LHGR Limit for ATRIUM 9B Fuel....... 36 -
EMF-96-141 Revision 1 Page
- Dresden Unit 3 Cycle 15 Reload Analysis
1.0 INTRODUCTION
This report provides the results of the analysis performed by Siemens Power Corporation -
Nuclear Division (SPC-ND)in support of the Cycle 15 reload for Dresden Unit 3. This report is intended to be used in conjunction with the SPC-ND topical Report XN-NF-80-19(P)(A),
Volume 4, Revision 1, Application of the ENCMethodology to BWR Reloads, which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(P)(A), Volume 4, Revision 1. Methodology used in this report which supersedes XN NF-80-19(P)(A), Volume 4, Revision 1,is referenced in Section 8.0. The NRC Technical Limitations presented in the methodology documents, including the documents referenced in Section 8.0, have been satisfied by these analyses.
For Dresden Unit 3 Cycle 15, Commonwealth Edison Company (Comed) has responsibility for portions of the reload safety analysis. This document describes only the Cycle 15 analyses performed by SPC-ND: CemEd analyses are described elsewhere. This document alone does not necessarily identify the limiting events or the appropriate operating limits for Cycle 15.The limiting events and operating limits must be determined in conjunction with results from Comed analyses.
The Dresden Unit 3 Cycle 15 core consists of a total of 724 fuel assemblies, including 232 unitradiated DRC-8 and DRC-8A ATRIUM *-9B"' assemblies and 492 irradiated SPC-ND 9x9-2 and 9x9-2B assemblies. The reference core configuration is described in Section 4.2.
The design and safety analyses reported in this document were based on the design and operational assumptions in effect for 0 esden Unit 3 during the previous operating cycle.The ef fects of channel bow are explicitly accounted for in the safety limit analysis. Increased core i
ATRIUM is a trademark of Siemens.
EMF-96-141 Revision 1 Page 2 flow was not evaluated for Cycle 15. SPC ND has performed time step size sensitivity studies to assure that the numerics solution in the COTRANSA2 code converged.
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Analyses and limits presented in this report support operation with various equipment out of service (EOOS). The EOOS conditions addressed in this report include:
Feedwater heaters out of service Relief valve out of service Safety / relief valve safety function out of service (ASME events)
Up to 40% TIP channels (equivalent of up to 2 TIP machines) out of service Single-loop operation i
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- sur-EMF-96-141 Revision 1 Page 3
'2.0 FUEL MECHANICAL DESIGN ANALYSIS 1
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l Applicable SPC-ND Fuel Design Reports References 9.1,9.7,9.8,9.14 To assure that the power history for the fuel to be irradiated during Cycle 15 of Dresden Unit 3 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits have been specified. In addition, LHGR limits for Anticipated Operational i Occurrences have been specified in the references. The LHGR limits for 9x9 2 fuel were j extended from a planar exposure of 48.0 GWd/MTU (Reference 9.1) to 50.9 GWd/MTU l (Reference 9.14). Steady-state and transient LHGR limits are provided in Section 7.2.3 and t
l 'in Figures 7.1 and 7.2 for both ATRIUM-9B and 9x9 2 fuel. The bundle exposure limit of 40
. I GWd/MTU for the 9x9 2 fuel (Reference 9.1)is not changed.
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EMF-96-141 Revision ~1 Page 4 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 Hydraulic Characterization 3.2.1 Hydraulic Compatibility Component hydraulic resistances for the constitur it fuel types in the Dresden Unit 3 Cycle 15 core have been determined in single-phase flow tests of full scale assemblies. The hydraulic demand curves for SPC-ND 9x9-2 and ATRIUM 9B fuel in the Dresden Unit 3 core are provided in Reference 9.7 (Figure 5.2 in the reference).
3.2.3 Fuel Centerline Temperature 9x9 2 Reference 9.1, Figure 3.21 ATRIUM 9B Reference 9.7, Figure 3.2 3.2.5 Bvnass Flow l
J Calculated Bypass Flow Fraction at 10.9%
100% power /100% flow at EOC"'
l 3.3 MCPR Fuel Claddina Intearity Safety Lithit (SLMCPRJ 1
Safety Limit MCPR - 1.08S'"'
Reference 9.16 r
includes water rod / channel flow.
Analyses performed support a two-loop MCPR safety limit of 1.08 or greater. Operating S'
limits are based on the Technical Specification two-loop MCPR safety limit of 1.08.
includes the effects of channel bow, up to 40% of the TIP channels (equivalent of up.
to 2 TIP machines) out of service, a 2000 EFPH calibration interval, and up to 50% of the LPRMs out of ervice, i-
EMF 96-141 Revision 1 Page5 l
3.3.1 Coolant Thermodynamic Condition j'
Thermal Power (at SLMCPR) 4260 MWt l
Feedwater Flow Rate (at SLMCPR) 16.5 Mlb/hr l
Core Pressure 1030 psia l
Feedwater Temperature 340.1 F#
l 3.3.2 Desion Basis Radial Power Distribution i
Figure 3.1 shows the limiting radial power distribution used in the MCPR Fuel Cladding integrity Safety Limit analysis.
3.3.3 Desion Basis Local Power Distribution l
Figures 3.2 and 3.3 show the conservative local power distributions used in the MCPR Fuel Cladding Integrity Safety Limit analysis.
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l Disposition of increased feedwater temperature a
.:iated with increased steam flow l
.is presented in Reference 9.15.
EMF 96141 Revision 1 Page 6 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Desian Analysis Assembly Average Enrichment ATRIUM-9B (DRC 8) 3.26 wt%
i (DRC 8A) 3.39 wt%
Radial Enrichment Distribution SPCA9-343L-9G3.5 Figure 4.1 SPCA9 344L-11G5.5 Figure 4.2 SPC A9-378L-11 G5.5 Figure 4.3
)
SPCA9 378L-9G4.5 Figure 4.4 SPCA9 362L-6G3.5 Figure 4.5 SPCA9-362L 6G4.5 Figure 4.6 SPCA9-388L-6G5.5 Figure 4.7 Axial Enrichment Distribution Figures 4.8 and 4.9 Burnable Absorber Distribution Figures 4.8 and 4.9 j
Non-Fueled Rods Figures 4.1 -4.7 Neutronics Design Parameters Table 4.1 Maximum Lattice k,,*
l ATRIUM-9B References 9.9 and 9.10 i
The ATRIUM-9B is bounded by the referenced analysis.
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EMF 96141 Revision 1 Page 7 4.2 Core Nuclear Desian Analysis 4.2.1 Core Configuration Figure 4.10 Core Exposure at EOC14, mwd /MTU 26,292 (nominal value)
Core Exposure at BOC15, mwd /MTU 14,527 (from nominal EOC14)
Core Exposure at EOC15, mwd /MTU 26,386 (licensing basis)
NOTE:
Analyses in this report are applicable to a core exposure of 26,386 mwd /MTU. Generic coastdown analyses (References 9.6 and 9.11) are applicable for Cycle 15 provided full power capability is lost prior to reaching a core exposure of 26,386 mwd /MTU.
< Cycle 15 short window exposure to be furnished by Comed. >
4.2.2 : Core Reactivity Characteristics
< This data is to be furnished by Comed. >
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EMF 96141 Revision 1 Page 8 4.2.4 Core Hydrodynamic Stability The results of the evaluation of decay ratio for several points along the current exclusion region boundary of the power / flow operation map are shown below. These results show that the maximum STAIF decay ratio throughout the cycle occurs at the intercept of the APRM rod block line and the natural circulation flow line. This analysis was performed using the design basis step-through control rod pattern projection, hence,it explicitly models the effects of Cycle 15 exposure. The calculated decay ratios are for demonstrating and tracking relative I stability from cycle to cycle. These results are applicable for increased feedwater temperature
! conditions associated with a 1 % increase in rated steam flow. The resulting lower core inlet l subcooling will have an insignificant effect on stability.
Decav Ratio (ADR'*)
% Power / % Flow State Points Global Reaional _
1.
58 / 33""
0.97 0.87 2.
67 / 41k' O.84 0.77 l
3.
73 /45"'
O.76 (.10) 0.72
(.04) 4.
63 /45'"
0.52 0.48 5,
34 / 28"'
O.41 (.11) 0.35
(.00) 6.
39 /38'8' O.23 0.21 For reactor operation under ccMitions of coastdown, feedwater heaters out of service, and single-loop, it is possible that higher decay ratios could be achieved than are shown for normal operation. Operation under these conditions will be acceptable in Cycle 15 as long as
' operating procedures and precautions defined by the NRC (Reference 9.12) and BWROG (Reference 9.13) for interim Corrective Actions are followed.
l, DRey,3-DRevu values are in parentheses.
APRM rod block line - naturr,? <irculation flow.
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APRM rod block line - two-pump minimum flow.
APRM rod block line - 45% flow.
100% rod line - 45% flow.
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70% rod line - natural circulation flow.
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'8' 70% rod line - two pump minimum flow.
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EMF-96-141 Revision 1 Page 9 L
5.0 ANTICIPATED OPERATIONAL OCCURRENCES l
Applicable Generic Transient Analysis Report Reference 9.2 l
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5.1 Analysis of Plant Transients at Rated Conditions References 9.3,9.6 and 9.11 Limiting Transients: Load Rejection Nr Bypass (LRNB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LFWH) j Maximum Peak Maximum Power Flow Heat Flux Neutron Pressure Event (W
R
(%)
Flux (%)
fosia) 6CPR*
Model l
LRN B*'
100 100 128 670 1298 0.35/0.36 COTRANSA2 i
FW C F*'
-100 100 133 528 1166 0.37/0.38 COTRANSA2 LFWH
'd8 id) 88) id) id:
'd) 885 l
l 5.2 Analysis for Reduced Flow Operation Reference 9.3 Limiting Transient:
Recirculation Flow' Increase Transient (RFIT)
(Pump Run-Up Event) 5.3 Analysis for Reduced Power Ooeration Reference 9.3 l
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Limiting Transient:
Feedwater Controller Failure (FWCF)
~ ACPR results for second-cycle 9x9-2/first-cycle ATRIUM 98 fuel types with revised l
operating conditions.
Based on Technical Specification limiting scram performance parameters.
- )
Feedwater heaters out of service (100'F reduction in feedwater temperature).
This data to be furnished by Comed.
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l EMF-96-141 Revision 1 Page 10 l 5.4 ASME Overoressurization Analysis,*'
Limiting Event MSIV Closure Worst Single Failure Valve Position Scram l
Maximum Pressure (Lower Plenum) 1339 psig l
Maximum Steam Dome Pressure
1313 psig 5.5 Control Rod Withdrawal Error Starting Control Pattern for Analysis Figure 5.1 i
< This data is to be furnished by Comed. >
J 5.6 Fuel Loadina Error
< This data is to be furnished by Comed. >
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Overpressurization results are for operation with revised operating conditions.
S' Coastdown operation requires a generic pressure penalty of 5.0 psid.
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Max mum steam dome pressure corresponds to the analysis that resulted in the l
maximum vessel pressure.
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EMF-96-141 Revision 1 Page 11 5.7 Determination of Thermal Marains Summary of Thermal Margin Requirements Power Flow Event 4%)
f%)
A C P R'*'
MCPR Limit
l LRNB 100 100 0.35 / 0.36 1.43 / 1.44'*'
l FWCPc' 100 100 0.37 / 0.38 1.45 / 1.46
l CRWE l
l MCPR Operating Limit at Rated Conditions
Prior to End of Licensing Basis Exposure (Section 4.2.1) 1.46 l
During Coastdown
- 1. 5 0"'
MCPR Operating Limits at Off-Rated Conditions'*'
Reduced Flow MCPR Limits:
Manual Flow Control Figure 5.2 Automatic Flow Control Figure 5.3 l
Values for second cycle 9x9-2/first-cycle ATRIUM-98 fuel types with revised operating l
conditions.
Based on plant Technical Specification two-loop MCPR safety limit of 1.08 and Technical Specification limiting scram performance parameters.
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Feedwater heaters out of service (100 F reduction in feedwater temperature),
j This data is to be furnished by Comed.
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The presented limits are applicable to both ATRIUM-9B and 9x9-2 fuelin the Cycle 15 l
core. These limits may need to be increased if Comed CRWE analysis results are more l
- limiting, j
Generic MCPR penalty of 0.04 is added to the MCPR operating limit to support I
coastdown operation beginning at EOFP (References 9.6 and 9.11).This penalty is not
(
necessary if the station elects to monitor to the core thermal power limit in Figure 2.1 l
l in Reference 9.11. If the 0.04 adt er is applied, the core thermal power limit provided l
in Figure 2.2 in Reference 9.11 must be maintained.
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EMF-96-141 Revision 1 Page 12 6.0 POSTULATED ACCIDENTS j
6.1 Loss-of Coolant Accident l
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1 6.1.1 Break Location Soectrum Reference 9.4 i
6.1.2 Break Size Soectrum i
Reference 9.4 i.
i 6.1.3 MAPLHGR Analyses Reference 9.5 The MAPLHGR limits of Reference 9.5 are valid for the Dresden Unit 3 9x9 2 (ANF 5, ANF 6, I
DRC-7) and ATRIUM-9B (DRC-8 and DRC-8A) fuels for Cycle 15 operation. MAPLHGR limits
- are presented in Section 7.2.1.
Limiting Break:
Double-Ended Guillotine Pipe Break Recirculation Pump Suction Line t
1.0 Discharge Coefficient LPCI Valve Failure - DBA Single Failure L
Peak clad temperature, peak metal water reaction (MWR), and total core MWR are 1920*F, l
1
< 1.09% locally, and < 0.12% core wide, respectively for 9x9 2 fuel with flow measurement uncertainties. For ATRIUM-98 fuel, PCT, peak MWR, and total core MWR are.1.838*F, I
e
. < 0.80% locally, and < 0.12% core wide, respect.ively with flow measurement uncertainties.
The 9x9 2 fuelis the limiting fuel type for Cycle 15.
I g 6.2 Control Rod Drop Accident 4
f.
< This data is to be furnished by Comed. >
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EMF 96-141 Revision 1
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Page 13 1
7.0 TECHNICAL SPECIFICATIONS p
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7.1 Limitino Safety System Settinas 7.1.1 MCPR Fuel Claddina intearity Safety Limit i
MCPR Safety Limit (all fuel) 1.0 8"' *'
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'7.1.2 Steam Dome Pressure Safetv Limit 1
Pressure Safety Lirnit 1345 psig 7.2 Limitina Conditions for Operation 7.2.1 Averaae Planar Linear Heat Generation Rate i
Planar Average 9x9 2 ATRIUM 9B Exposure MAPLHGR MAPLHGR (GWd/MTU)
(kW/ft)
(kW/ft) 0 12.5 13.5 3
15 12.5 13.5 20 11.'9 13.5 55 7.7 9.3 8.7 60 As long as bundle exposures are within the range of exposures considered in the LOCA analyses, the specified MAPLHGR limits remain valid during coastdown operation, 1
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3 Analyses perf ormed support a two-loop MCPR safetylimit of 1.08 or greater. Operating limits are based on the Technical Specification two-loop MCPR safety limit of 1.08.
I includes the effects of channel bow, up to 40% of the TIP channels (equivalent of up to 2 Tt machines) out of service, a 2000 EFPH calibration interval, and up to 50% of the LPRMs out of service.
EMF-96-141 Revision 1 Page 14 7.2.2 Minimum Critical Power Ratio i
{
Rated Conditions MCPR Limit Based on Technical Specification Scram Times f
1 Off Rated Conditions MCPR Limits:
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Manual Flow Control Figure 5.2 Automatic Flow Control Figure 5.3 Figure 1 of Reference 9.14 l 7.2.3 Linear Heat Generation Rate and Figure 2.1 of Reference 9.7 r
Steady-State LHGR Limits ATRIUM 98 Fuel'*'
9x9-2 Fuel Planar Average Planar Average Exposure LHGR Exposure LHGR (GWd/MTU)
(kW/ft)
(GWd/MTU)
IkW/ft) 0.0 14.5 0.0 14.4 5.0 14.5 15.0 14.4 25.2 10.8 55.0 9.1 48.0 7.2 i
50.9 6.7 The transient linear heat generation rate curve is Figure 2 of Reference 9.14 for 9x9 2 an l
Figure 2.2 of Reference 9.7 for ATRIUM-9B. These figures are presented in the report Figures 7.1 and 7.2 f or convenience.
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Based on limiting results.from Section 5.7 or analyses within Comed's scope of f.
responsibility. The MCPR operating limit is based on a Technical Specificatio I
MCPR safety limit of 1.08 and the limiting ACPR for Cycle 15.
J ATRIUM-9B planar exposure is limited to 55 GWd/MTU based on a peak pelle exposure of 60 GWd/MTU.
EMF 96-141 Revision 1 Page 15 8.0 METHODOLOGY REFERENCES See XN-NF 8019(P)(A), Volume 4, Revision 1 for a complete bibliography, i
i COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis, f
8.1 ANF-913(P)(A), Volume 1, Revision 1, Suppleinents 1,2,3 and 4, Advanced Nuclear Fuels Corporation, August 1990.
8.2 Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors, XN NF-524(P)(A), Revision 2,and Supplements, Advanced Nuclear Fuels Corporation, j
November 1990.
B.3 -
ANFB Critical Power Correlation, ANF-1125(P)(A), Supplements 1 and 2, Advanced Nuclear Fuels Corporation, April 1990.
Advanced Nuclear Fuels Methodology for Boiling WaterReactors: Benchmark Results 8.4 for the CASMO-3G/MICROBURN-B Calculation Methodology, XN-NF-80-19(P)(A),
Volume 1 and Supplement 3, Appendix F and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.
STAIF: A Computer Program for BWR Stability Analysis in the Frequency Domain, EMF-8.5 CC-074(P)(A), Volume 1 and Volume 2, Siemens Power Corporation Nuclear Division, July 1994.
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EMF-96-141, l
i Revision 1 l
Page 16
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9.0 ADDITIONAL REFERENCES L
i 9.1 Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, l
XN-NF-85-67(P)(A), Revision 1, Exxon Nuclear Company, Inc., September 1986.
I 9.2 Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF 71(P), Revision 2, including Supplements 1 through 3(P)(A), Exxon Nuclear Company, Inc., March 1986.
l 9.3 Dresden Unit 3 Cycle 15 Plant Transient Analysis Withincreased Steam Flow, EMF j l
047, Siemens Power Corporation - Nuclear Division, June 1998.
1 l 9.4 LOCA Break Spectrum Analysis forDresden Units 2 and 3, EMF-97-025(P), Revision 1, l
Siemens Power Corporation Nuclear Division, May 1997.
l 9.5 Dresden LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM *-9B and 9x9 2 Fuel, l
EMF-98-007(P), Siemens Power Corporation - Nuclear Division, January 1998, 9.6 Dresden Units 2 and 3 Generic Coastdown Analysis, EMF-92149(P), Siemens Power Corporation - Nuclear Division, October 1992.
9.7 Fuel Design Report for Dresden 3, Cycle 15 A TRIUM*-9B Fuel Assemblies, EMF 040(P), Revision 1, Siemens Power Corporation - Nuclear Division, August 1996.
l 9.8 Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X BWR Reload fuel, ANF-89-014(P)(A) Revision 1 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1991.
9.9 Criticality Safety Analysis for Dresden Units 2 and 3 Spent Fuel Storage Pool, EMF-94 098(P), Revision 1, Siemens Power Corporation - Nuclear Division, January 1996.
9.10 Criticality Safety Analysis for A TRIUM"-9B FuelDresden and Quad Cities New Fuel Storage Vaults, EMF-96-148(P), Revision 1, Siemens Power Corporation - Nuclear Division, September 1996.
l 9.11 Dresden Units 2 and 3 Generic Coastdown Analysis for A TRIUM*-98, EMF 92-149(P),
Supplement 1, Revision 1, Siemens Power Corporation - Nuclear Division, September 1996.
9.12 "Long-Term Solutions and Upgrade of interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors," NRC Generic Letter 94-02, U.S.
Nuclear Regulatory Commission, July 11,1994.
9.13 "BWR Owners' Group Guidelines for Stability Interim Corrective Action," BWR Owners' Group Letter BWROG-94078, June 6,1994.
l 9.14 Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Extension of LHGR Curve for Higher l
Planar Exposure for 9x9 2 Fuel at Dresden," JHR:97:200, May 20,1997.
l.
1 EMF-96-141 i
Revision 1 f
Page 17 9.0 ADDITIONAL REFERENCES (Continued) j j 9.15 Letter, J. H. Riddle (SPC) to R. J. Chin (Comed), " Clarification of Design Operating l
Conditions in Transient, LOCA and Safety Limit Analyses," JHR:97:088, March 3,1997.
l 9.16 Letter, D. E. Garber (SPC) to R. J. Chin (Comed), "Dresden Unit 3 Cycle 15 SLMCPR l
l Results with Updated ATRIUM *-98 Additive Constants and Various Additive Constant l
l Uncertainty Approaches," DEG:98:116, April 6,1998.
l l
l t
i l
l 1
l
EMF 96-141 Revision 1 Page 18 100 50 80 70 me mgw CD -
- o. 50 t.
o
.o e
E 3
z 3
'20
~
'10 i
p 0
.0
.2
.4
.6 2
1.0 1.2 1.4 1.6 Radicl-Power Peaking Figure 3.1 f
Design Basis Radial Power Distribution for SLMCPR Determination
]
o l
.i i.
i l-
l EMF-96141 Revision 1 Page 19 Contr01 Rod Corner o
n t
0.997 0.995 1.036 1.005 1.087 1.073 1.078 1.005 1.063
)
r 0
I 0.995 0.974 1.026 1.000 0.878 1.068 1.057 0.866 1.051 1
R 1.036 1.026 1.001 1.031 1.041 1.102 1.023 1.004 1.082 i
C o
1.005 1.000 1 031 0.869 0.973 1.033 r
Internal 1.087 0.878 1.041 Water 1.075 0.764 1.012 Channel 1.073 1.068 1.102 1.028 0.937 0.998 l
1.078 1.057 1.023 0.869 1.075 1.028 0.761 0.932 1.007 1.005 0.866 1.004 0.973 0.764 0.937 0.932 0.759 1.046 1.063 1.051 1.082 1.033 1.012 0.998 1.007 1.046 1.022 Maximum Local Power:
1.102 Figure 3.2 Design Basis Local Power Distribution for SPC-ND ATRIUM 9B Fuel (SPCAS 326B-11GZ-80M)
Uncontrolled at 15,000 mwd /MTU and 70% Void for SLMCPR Determination I
I l
l EMF 96-141 Revision 1 Page 20 Contr01 Rod Corner o
l n
t 1.005 0.983 1.053 1.028 1.086 1.075 1.081 1.026 1.044 r
o l
0.983 0.990 1.029 0.916 0.986 1.064 1.057 0.887 1.056 l
R i-1.053 1.029 1.008 1.040 1.044 1.000 1.025 1.005 1.083 l
C f
o 1.028 0.916 1.040 0.890 0.969 1.034 r
n Internal f
1.086 0.986 1.044 Water 1.070 0.955 1.010 1
Channel 1.075 1.064 1.000 '
1.030 0.773 1.008 l
1.081 1.057 1.025 0.890 1.070 1.030 0.803 0.936 1.016 1
1.026 0.887 1.005 0.969 0.955 0.773 0.936 0.803 0.956 1.044 1.056 1.083 1.034 1.010 1.008 1.016 0.956 0.961 i
)
Maximum Local Power:
1.086 l
l Figure 3.3 Design Basis Local Power Distribution for I
SPC-ND ATRIUM 9B Fuel (SPCA9 339B-6GZ-80M)
Uncontrolled at 17,500 mwd /MTU and 70% Void for SLMCPR Determination i
a _ ______ ___ __ _
. EMF 96141 Revision 1 Page 21 Table 4.1 Neutronic Design Values I
Number of Fuel Assemblies 724 Rated Thermal Power, MWt 2527 Rated Core Flow, Mlbm/hr 98.0 Core inlet Subcooling, Btu /lbm 22.7
Moderator Temperature, *F 546 Channel Thickness, inch 0.080 Channel internal Face to-Face Dimension, inch 5.278 Fuel Assembly Pitch, inch 6.0 Wide Water Gap Thickness, inch 0.750 Narrow Water Gap Thickness, inch 0.374 Control Rod Data
Absorber Material B,C Total Blade Span, inch 9.810 Total Blade Support Span, inch 1.580 Blade Thickness, inch 0.312 i
Absorber Rods Per Blade 84 i
Absorber Rod OD, inch 0.188 Absorber Rod ID, inch 0.138 l
Absorber Density, % of theoretical 70 Based on actual operating experience.
i The control rod data represents original equipment control blades at Dresden which were modeled in the licensing analyses. Dresden FSAR Sec: ion 4.6.2.1 indicates that reactivity characteristics of replacem nt control blades closely match originct equipment blades.
EMF-96141 Revision 1 Page 22 Contr01 Rod Corner o
n 1
2 3
3 4
4 4
3 3
t 1.800 2.200 2.700 2.700 3.300 3.300 3.300 2.700 2.700 l
o 1
2 5
4 4
5 6
6 5
4 l
2.200 2.700 3.300 3.300 2.700 4.060 4.060 2.700 3.300 R
0 3
4 4
4 4
6 6
6 6
d 2.700 3.300 3.300 3.300 3.300 4.060 4.060 4.060 4.060 3
4 4
a 3
6 6
2.700 3.300 3.300 2.700 4.060 4.060 r
n Internal e
4 5
4 6
5 6
r 3.300 2.700 3.300 Water 4.060 2.700 4.060 Channel 4
6 6
6 6
6 3.300 4.060 4.060 4.060 4.060 4.060 4
6 6
3 6
6 5
6 6
3.300 4.060 4.060 2.700 4.060 4.060 2.700 4.060 4.060 3
5 6
6 5
6 6
5 6
2.700 2.700 4.060 4.060 2.700 4.060 4.060 2.700 4.060 3
4 6
6 6
6 6
6 4
2.700 3.300 4.060 4.060 4.060 4.0.60 4.060 4.060 3.300 1
Rods ( 1) 1.800 wt% U-235 2
Rods ( 2) 2.200 wt% U-235 3
Rods (10) 2.700 wt% U-235 4
Rods (18) 3.300 wt% U-235 5
Rods ( 9) 2.700 wt% U-235 + 3.50 wt% Gd 9 23 6
Rods (32) 4.060 wt% U-235 l
Figure 4.1 Dresden Unit 3 Cycle 15 SPCA9 343L 9G3.5 Enrichment Distribution w____________________________.
EMF-96-141 Revision 1 Page 23 Contr01 Rod Corner i
o 1
2 3
3 4
4 4
3 3
1.800 2.200 2.700 2.700 3.300 3.300 3.300 2.700 2.700 r
o I
2 5
4 4
5 6
6 5
4 2.200 2.700 3.300 3.300 2.700 4.060 4.060 2.700 3.300 R
o d
3 4
4 4
4 6
6 6
6 2.700 3.300 3.300 3.300 3.300 4.060 4.060 4.060 4.060 C
o r
3 4
4 5
6 6
n 2.700 3.300 3.300 2.700 4.060 4.060 e
Internal 4
5 4
6 5
6 Water 3.300 2.700
'3.300 4.060 2.700 4.060 l
Channel 4
6 5
6 6
6 3.300 4.060 4.060 4.060 4.060 4.060 4
6 6
5 6
6 5
6 6
3.300 4.060 4.060 2.700 4.060 4.060 2.700 4.060 4.060 3
5 6
6 5
6 6
5 6
2.700 2.700 4.060 4.060 2.700 4.060 4.060 2.700 4.060 3
4 6
6 6
6 6
6 4
2.700 3.300 4.060 4.060 4.060 4.060 4.060 4.060 3.300 1
Rods ( 1) 1.800 wt% U-235 2
Rods ( 2) 2.200 wt% U-235 3
Rods ( 8) 2.700 wt% U-235 4
Rods (18) 3.300 wt% U-235 2.700 wt% U-235 + 5.50 wt% Gd 0 E
Rods (11) 23 4.060 wt% U-235 6
Rods (32) l Figure 4.2 Dresden Unit 3 Cycle 15 SPCA9 344L-11G5.5 Enrichment Distribution i
f l
EMF-96141 Revision 1 Page 24 Contr01 Rod Corner o
1 2
3 3
4 4
4 3
5 1.800 2.500 3.100 3.100 3.300 3.300 3.300 3.100 2.700 l
r l
0 I
2 6
-7 7
6 8
8 6
7 i
R 2.500 3.100 3.860 3.860 3.100 4.390 4.390 3.100 3.860 l
o d
3 7
7 7
7 8
8 8
8 3.100 3.860 3.860 3.860 3.860 4.390 4.390 4.390 4.390 l
o l
r 3
7 7
6 8
8 n
3.100 3.860 3.860 3.100 4.390 4.390 e
Internal 4
6 7
3.300 3.100 3.860 W ter 8
6 8
4.390 3.100 4.390 l
Channel 4
8 8
8 8
8 3.300 4.390 4.390 4.390 4.390 4.390 4
8 8
6 8
8 6
8 8-l l
3.300 4.390 4.390 3.100 4.390 4.390 3.100 4,390 4.390 t
3 6
8 8
6 8
8 6
8 3.100 3.100 4.390 4.390 3.100 4.390 4.390.
3.100 4.390 l
5 7
8 8
8 8
8 8
7 2.700 3.860 4.390 4.390 4.390 4.390 4.390 4.390 3.860 l
1 Rods ( 1) 1.800 wt% U-235 2
Rods'( 2) 2.500 wt% U-235 3
Rods ( 6) 3.100 wt% U-235 4
Rods ( 6) 3.300 wt% U-235 5
Rods ( 2) 2.700 wt% U-235 6
Rods (11) 3.100 wt% U-235 + 5.50 wt% Gd 0 23 7
Rods (12) 3.860 wt% U-235 8
Rods (32) 4.390 wt% U-235 l.
Figure 4.3 J
Dresden Unit 3 Cycle 15 SPCA9-378L-11G5.5 Enrichment Distribution l
EMF-96-141 Revision 1 Page 25 Contro1 Rod Corner o
l n
1 2
3 3
4 4
4 3
5 t
1.800 2.500 3.100 3.100 3.300 3.300 3.300 3.100 2.700 r
2 6
7 7
6 8
8 6
7 2.500 3.100 3.860 3.860 3.100 4.390 4.390 3.100 3.860 R
0 3
7 7
7 7
8 8
8 8
d 3.100 3.860 3.860 3.860 3.860 4.390 4.390 4.390 4.390 3
7 7
3 8
8 g
3.100 3.860 3.860 3.100 4.390 4.390 p
Internal n
e 4
6 7
8 6
8 i
Water r
3.300 3.100
~3.860 4.390 3.100 4.390 Channel 4
8 8
8 8
8 3.300 4.390 4.390 4.390 4.390 4.390 4
8 8
3 8
8 6
8 8
3.300 4.390 4.390 3.100 4.390 4.390 3.100 4.390 4.390 3
6 8
8 6
8 8
6 8
3.100 3.100 4.390 4.390 3.100 4.390 4.390 3.100 4.390 5
7 8
8 8
8 8
8 7
2.700 3.860 4.390 4.390 4.390 4.390 4.390 4.390 3.860 1
Rods ( 1) 1.800 wt% U-235 2.500 wt% U-235 2
Rods ( 2) 3 Rods ( 8) 3.100 wt% U-235 4
Rods ( 6) 3.300 wt% U-235 5
Rods ( 2) 2.700 wt% U-235 6
Rods ( 9) --
3.100 wt% U-235 + 4.50 wt% Gd 0 22 3.860 wt% U-235 7
Rods (12) l 8
Rods (32) 4.390 wt% U-235 Figure 4.4 Dresden Unit 3 Cycle 15 g
SPCA9-378L-9G4.5 Enrichment Distribution I
EMF 96-141 Revision 1 Page 26 Control Rod Corner o
n 1
2 3
3 4
4 4
3 5
t 2.000 2.300 3.000 3.000 3.530 3.530 3.530 3.000 2.750 o
1 2
3 4
6 4
7 7
6 4
2.300 3.000 3.530 3.000 3.530 4.300 4.300 3.000 3.530 R
0 3
4 4
4 4
4 7
7 7
d 3.000 3.530 3.530 3.530 3.530 3.530 4.300 4.300 4.300 3
6 4
3 7
7 3.000 3.000 3.530 3.000 4.300 4.300 r
n Internal e
4 4
4 7
7 7
r 3.530 3.530 3.530 W ter 4.300 4.300 4.a00 Channel 4
7 4
7 6
7 3.530 4.300 3.530 4.300 3.000 4.300 4
7 7
3 7
7 3
7 7
3.530 4.300 4.300 3.000 4.300 4.300 3.000 4.300 4.300 3
6 7
7 7
6 7
3 4
3.000 3.000 4.300 4.300 4.300 3.000 4.300 3.000 3.530 5
4 7
7 7
7 7
4 3
2.750 3.530 4.300 4.300 4.300 4.300 4.300 3.530 3.000 1
Rods ( 1) 2.000 wt% U-235 2
Rods ( 2) 2.300 wt% U-235 1
3 Rods (12) 3.000 wt% U-235 4
Rods (21) 3.530 wt% U-235 5
Rods ( 2) 2.750 wt% U-235 6
Rods ( 6) 3.000 wt% U-235 + 3.50 wt% Gd 0 l
23 7
Rods (28) 4.300 wt% U-235 Figure 4.5 Dresden Unit 3 Cycle 15 I
SPCA9-362L-6G3.5 Enrichment Distribution
I l
EMF-96-141 Revision 1 Page 27 Contro1 Rod Corner o
l n
1 2
3 3
4 4
4 3
5 t
2.000 2.300 3.000 3.000 3.530 3.530 3.530 3.000 2.750 r
i o
1 2
3 4
6 4
7 7
6 4
2.300 3.000 3.530 3.000 3.530 4.300 4.300 3.000 3.530 i
R 0
3 4
4 4
4 4
7 7
7 d
3.000 3.530 3.530 3.530 3.530 3.530 4.300 4.300 4.300 l
3 6
4 3
7 7
g 3.000 3.000 3.530 3.000 4.300 4.300 r
n Internal e
4 4
4 7
7 7
r 3.530 3.530 3.530 Water 4.300 4.300 4.300 Channel 4
7 4
7 6
7 3.530 4.300 3.530 4.300 3.000 4.300 4
7 7
3 7
7 3
7 7
3.530 4.300 4.300 3.000 4.300 4.300 3.000 4.300 4.300 l
3 6
7 7
7 6
7 3
4 3.000 3.000 4.300 4.300 4.300 3.000 4.300 3.000 3.530 5
4 7
7 7
7 7
4 3
2.750 3.530 4.300 4.300 4.300 4.300 4.300 3.530 3.000 1
Rods ( 1) 2.000 wt% U-235 2
Rods ( 2) 2.300 wt% U-235 3
Rods (12) 3.000 wt% U-235 4
Rods (21) 3.530 wt% U-235 5
Rods ( 2) 2.750 wt% U-235 6
Rods ( 6) 3.000 wt% U-235 + 4.50 wt% Gd 0 23 7
Rods (28) 4.300 wt% U-235 l,
i Figure 4.6 I
i Dresden Unit 3 Cycle 15 SPCA9 362L-6G4.5 Enrichment Distribution
EMF-96-141-Revision 1 Page 28 Control Rod Corner o
n 1
2 3
4 5
5 5
4 3
t 2.000 2.300 3.000 3.300 3.780 3.780 3.780 3.300 3.000 o
1 2
4 5
6 5
7 7
6 5
2.300 3.300 3.780 3.530 3.780 4.550 4.550 3.530 3.780 R
o 3
5 5
5 5
5 7
7 7
d 3.000 3.780 3.780 3.780 3.780 3.780 4.550 4.550 4.550 C
4 6
5 4
7 7
o 3.300 3.530 3.780 3.300 4.550 4.550 r
Internal n
e 5
5 5
8 7
7 Water r
3.780 3.780 3.780 4.300 4.550 4.550 Channel 5
7 5
7 6
7 3.780 4.550 3.780 4.550 3.530 4.550 5
7 7
4 8
7 9
7 7
3.780 4.550 4.550 3.300 4.300 4.550 3.530 4.550 4.550 4
6 7
7 7
6 7
9 5
3.300 3.530 4.550 4.550 4.550 3.530 4.550 3.530 3.780 3
5 7
7 7
7 7
5 4
3.000 3.780 4.550 4.550 4.550 4.550 4.550 3.780 3.300 1
Rods ( 1) 2.000 wt% U-235 2
Rods ( 2) 2.300 wt% U-235 3
Rods ( 4) 3.000 wt% U-235 4
Rods ( 8) 3.300 wt% U-235 5
Rods (21) 3.780 wt% U-235 6
Rods ( 6) 3.530 wt% U-235 + 5.50 wt% Gd 0 I
23 7
Rods (26) 4.550 wt% U-235 8
Rods ( 2) 4.300 wt% U-235 l
9 Rods ( 2) 3.530 wt% U-235 Figure 4.7 Dresden Unit 3 Cycle 15 SPCA.9-388L-6G5.5 Enrichment Distribution L_______
EBAF-96-141 Revision 1 Page 29 SPC A9-3268-11 GZL-80M SPCA9-3268-11GZH-80M l
(DRC-8 Type L)
(DRC-8 Type H)
\\
l Natural Uranium Blanket Natural Uranium Blanket SPCA9 343L-9G3.5 SPCA9-343L-9G3.5 s
SPC A9-344L-11 G 5.5 SPCA9-344L-11 G 5.5 SPCA9-376L-9G4.5 SPCA9-378L-11G5.5 Natural Uranium Blanket Natural Uranium Blanket Figure 4.8 Dresden Unit 3 Reload Batch DRC-8 Axial Fuel Assembly Design
2 GMF-96-141 Revision 1 I
Page 30 SPCA9-339B 6GZ-80M (DRC 8A)
Natural Uranium Blanket SPCA9-362L-6G3.5 SPCA9-362L-6G4.5 SPCA9 388L-6G5.5 Natural Uranium Blanket F
Figure 4.9 Dresden Unit 3 Reload Batch DRC-8A A-3=' Fuel Assembly Design
__._.___A,__-
j EMF 96-141 Revision 1 Page 31 E2 H0 F1 D2 D2 HO G1 D2 D2 H0 F1 E2 C2 D2 A3 I
HO G1 10 F1 HO F1 10 G1 JO G1 HO JO 10 D2 83 F1 10 D2 JO G1 10 F1 10 F1 HO G1 10 D2 D2 A3 D2 F1 JO D2 E2 G1 JO D2 D2 D2 JO F1 10 D2 A3 D2 H0 G1 E2 D2 H0 G1 C2 E2 H0 C2 H0 C2 E2 A3 H0 F1 10 G1 H0 E2 HO G1 JO G1 HO 10 C2 A3 G1 10 F1 JO G1 HO E2 H0 G1 10 F1 G1 A3 D2 G1 10 D2 C2 G1 H0 G1 E2 JO G1 F1 B3 D2 JO F1 D2 E2 JO G1 E2 G1 HO D2 C2 A3 HO G1 HO D2 H0 G1 10 JO HO A3 A3 A3 F1 H0 G1 JO C2 H0 F1 G1 D2 A3 A3 E2 JO 10 F1 H0 10 G1 F1 C2 A3 C2 10 D2 10 C2 C2 A3 A3 A3 XY D2 D2 D2 D2 E2 A3 X = Fuel Type A3 B3)A3 A3 A3 Y = Cycles Irradiated A
84 SPC 9x9-2 3.13 wt% U-235 ANF-5L B
1.2 SPC 9x9-2 3.13 wt% U-235 ANF-5H C
48 SPC 9x9-2 2.95 wt% U-235 ANF-6L D
116 SPC 9x9 2 2.95 wt% U-235 ANF-6H E
52 SPC 9x9 2 2.95 wt% U 235 ANF-6A F
72 SPC 9x9 2B 3.13 wt% U-235 DRC-7L G
108 SPC 9x9-2B 3.13 wt% U-235 DRC 7H H
104 SPC ATRIUM-9B 3.26 wt% U-235 DRC-8L I
/2 SPC ATRIUM-98 3.26 wt% U-235 DRC-8H J
56 SPC ATRIUM-9B 3.39 wt% U-235 DRC-8A Figure 4.10 j
Dresden Unit 3 Cycle 15 Reference Loading Man (Quarter-Core Symmetric Loading)
EMF-96-151 Revision 1 Page 32 l
< This data is to be furnished by Comed. >
l Figure 5.1 Starting Control Rod Pattern for Control Rod Withdrawal Analysis l
l L_____.______.____---_--_-----
EMF 96-141 Revision 1 Page 33 l
14 Reduced Flow MCPR Limit l
0 ATRIUM-98 Results 22 o
9x9-2 Results
.t E 2.0
- 3 1
~ o 1.s 2
3 2
g k 1.6 8
D e
t 9
oa c 1.4 9
c:
Q 1.2 -
D 1.0 O
20 40 60 80 100 120 Total Core Flow (% Rated) j i
i i
Figure 5.2 Reduced Flow MCPR Limit for i
Manual Flow Control (SLMCPR = 1.08)
Applicable to ATRIUM.9B and 9x9-2 Fuel 1
c EMF-96-141 Revision 1 Page 34 i
l l
l t -
l -
3.0 l
C 3 MCPR = 1.46 e
o MCPR = 1.50 ;
2.8
^
^ MCPR = 1.55
,e 2.6 E
~
3
% 14 l
CL O
2 3: 2.2
_o l
u.
A 2.0 o
a t
o
% 1.8 l
1.6 l
6 1.4 t
t i
t 0
20 40 60 80 100 120 l
Totcl Core Flow (% Rated) l l
l l
I l
Figure 5.3 i
I Reduced Flow MCPR Limit for Automatic Flow Control Applicable to ATRIUM-98 and 9x9 2 Fuel
l
-7 EMF-96141 Revision 1 Page 35 s
22 j
'20
- (0,0,19.2)
~
L_
l 18 (25.4,16.9) l 16 l
i 14 i
l i
\\
g 12 (43.2.10.8)
(48.o,1o.o)
- 0: 10-
.O I
(50.9,9.5)
.J B
l 6
4 l-2 O
t i
1 f
f f
t t
f I
t 0
5 10 15 20 25 30 35 40 45 50 55 60 l
Planor Exposure, _GWd/MTU I'
(-
l~
l l
l-Figure 7.1 l
Protection Against Power Transient LHGR Limit
--)
for 9x9-2 Fuel I
EMF-96 141 I
Revision 1 Page 36 1
22
- (0 U 19 4)
(150194) 20 i
18
.)
i 16 14 N
3ll: 12 (55.0.12.3)
Ct' 10 or
.t B
6 4
1 2
0 i
i t
O 5
10
-15 20 25 30 35 40 45 50 55 60 Pianor Exposure, GWd/MTU Figure 7.2-L l
Protection Against Power Transisr IC Limit for ATRIUM 98 Fuei i
i
EMF-96-141 Revision 1 Page A 1 i
APPENDIX A j
Single-Loop Operation l
l l
l A.1 ANTICIPATED OPERATIONAL OCCURRENCES l
l
- Analyses have been provided which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended period of time (Reference A.1). These
(
analyses confirm that during single-loop operation, the plant cannot reach the normal bundle l
power levels and nodal power levels that are possible when both recirculation systems are in i
operation. The physical interdependence between core power and recirculation flow rate l
inherently limits the core to less than rated power. Because the SPC-ND fuel was designed to be compatible with the coresident fuelin thermal-hydraulic, nuclear, and mechanical design l
performance, and because the SPC ND methodology has given results which are consistent with those of the previous analyses for normal two-loop operation, the analyses performed by the NSSS supplier for single-loop operation are also applicable to single loop operation with fuel and anr' 7es provided by SPC-ND.
A.2 IV.JrR SAFETY LIMIT It is conservative to use the reduced flow two-loop operating MCPR limit or full fl,ow MCPR operating limit plus 0.01 (whichever is greatest) for single-loop operations. These limits conservatively bound all transients from SLO conditions. The reduced flow MCPR limit is to 1
protect against boiling transition during flow excursions to maximum two-pump flow; excursions to such 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.
i L
l.
t
EMF 96141
)
Revision 1 Page A-2 t
A.3 REDUCED FLOW OPERATION i
l l
i It is conservative to use the reduced flow two-loop operating MCPR limit or full flow MCPR operating limit plus 0.01 (whichever is greatest) for single-loop operations. This method is applied for operation up to end of full power and for coastdown. These limits conservatively bound all transients from SLO conditions. The reduced flow MCPR limit is to protect against boiling transition during flow excursions to maximum two-pump flow; excursions to such high flows are not possible during single-loop one-pump operation. Thus, conservatively maintaining I
this two-loop limit assures that there is even more thermal margin under single-loop conditions l
than under two-loop full power / full flow conditions.
A.4 SINGLE LOOP PUMP SElZURE ACCIDENT l
I l
Pump seizure is a postulated accident where the operating recirculation pump suddenly stops rotating. This causes a rapid decrease in core flow, a decrease in the rate at which heat can l
be transferred from the fuel rods, and a decrease in the critical power ratio. COTRANSA2.
l XCOBRA, and XCOBRA-T are used to calculate the MCPR for SPC-ND fuel during a pump seizure from single-loop operation.
l l
COTRANSA2 was used to simulate system response to a pump seizure in single-loop j
operation. The core was assumed to be operating at the MCPR, limit at 58% core flow. The operating recirculation pump speed was reduced to zero causing a sudden decrease in active jet pump drive flow. During the event the inactive jet pump diffuser flow went from negative flow to positive flow.
Thermal-hydraulic analysis using SPC-ND safety limit methodology has shown that less than 1
10% of the rods in the core would experience boiling transition during the event. The Dresden FSAR indicates that a LOCA with a failure of 45% of the fuel rods in the core results in an off-site dose of less than 10% of 10CFR100 limits. Therefore, the two-loop manual flow control l
MCPR, limit below 58% flow provides the required protection such that any postulated fuel failures would not result in exceeding a small fraction of the 10CFR100 requirements.
g j '*'
This conservative determination of MCPR limits remains applicable for the revised j
cperating conditions analyzed in Reference A.3.
I
EMF-96-141 Revision 1 Page A-3 A.5 MAPLHGR LIMITS SPC-ND performed LOCA analyses from single-loop conditions and determined an appropriate SLO MAPLHGR multiplier of 0.90 for 9x9-2 and ATRIUM-98 fuel. The ECCS analysis results are presented in Reference A.2. All calculations were performed with the NRC-approved EXEM/BWR ECCS Evaluation Model according to Appendix K of 10CFR50.
i I
l
(
A.6 REFERENCES l
t
'A.1 Dresden Unit 3 Cycle,'1 Plant Transient Analysis, ANF-87-096, Advanced Nuclear Fuels Corporation, September 1987.
i I A.2 Dresden LOCA-ECCS Analysis MAPLHGR Limits for A TRIUM"-98 and 9x9-2 Fuel-l Single-Loop Analysis, EMF-98-OO7(P) Supplement 1, Siemens Power Corporation -
l Nuclear Division, January 1998.
l A.3 Dresden Unit 3 Cycle 15 Plant Transient Analysis With increased Steam Flo w, EMF l 047, Siemens Power Corporation - Nuclear Division, June 1998.
l i;.
l-
)
i j
The SLO MAPLHGR multiplier of 0.90 remains applicable for the revised operating j
conuniuns analyzed in Reference A.3.
EMF 96141 l-Revision 1 l
4 i
i Dresden Unit 3 Cycle 15 Reload Analysis 1
i Distribution D. E. Garber, 38 (15) i Notification List (E-Mail Notification).
D. J. Braun O. C. Brown D. G. Carr
(.
R. J. DeMartino -
l M. E. Garrett J. L. Maryott R.R.Schnepp l
D. C. Serell A. W. Will l^
J. A. White i
t l
I
1 l
t Dresden Unit 3 Cycle 15 Excerpts from Plant Transient Analysis Report i
June 1998 Dresden Unit 3 Cycle 15 c_-____--_
EMF 97-047 Page 1-3 j
i
'l 12 0 l
1 l
100/100 g
j 100 1
APRM Rod Block Line j
/
t
/
e. 80 o
-c:
/
$ 60 c.
100". Rod Line u'e a
o j
40 Minimum Pump Speed Natural t-o u
Circulation 20 b.
I l'
0 0
10 20 30 40 50 60 70 80 90 100
[
Core Flow, Percent of Rated j.
l l
i i
Figure 1.1 I
Dresden Unit 3 i
Operating Power / Flow Map l
l
EMF 97-047 Page 2 3 I'
i l
l-Table 2.1 Dresden Unit 3 Cycle 15 k
ACPRs at Rated Power With increased Steam Flow i
ACPR'*
.9x9-2 ATRIUM-98 Load Rejection No Bypass
100% Power /100% Flow 0.35 0.36 100% Power / 87% Flow 0.33 0.36 Feedwater Flow Controller Failure
l 100% Power /100% Flow 0.37 0.37 100% Power / 87% Flow 0.34 0.37 100% Power /100% Flow - FHOOS" 0.37 0.38 100% Power / 87% Flow - FHOOS'"
0.35 0.38 Loss of Feedwater Heating l
l
' ACPRs presented are for second-cycle 9x9-2 fuel and first-cycle ATRIUM-98 fuel.
ACPR based on Technical Specification scram performance.
FHOOS -feedwater heaters out of service (100'F reduction in feedwatertemperature).
Analysis of the LFWH is the responsibility of Comed for Dresden Unit 3 Cycle 15.
a-_____
EMF 97-047 Page 2-4 i
Table 2.2 Dresden Unit 3 Cycle 15 Thermal Margin Summary With increased Steam Flow i
MCPR Operatina Limit
Transient OLMCPR for 9x9-2/ ATRIUM-98 Up to EOFP Coastdown Feedwater Controller Failure 1.46 1.50*
)
(100%P /100%F - FHOOS)
(100%P / 87%F - FHOOS) i Maximum Pressurization (psig)
Transient Steam Dome Lower Plenum Steam Lines
_ MSIV Closure Without 1318
1344'"
1318
(100%P /100%F) l-l h
i Based on a plant technical specification two-loop MCPR safety limit of 1.08 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 Table 2.1.
Generic MCPR penalty of 0.04 is added to the MCPR operating limit to support coastdown operation beginning at EOFP (References 17 and 18). This penalty is not necessary if the station elects to monitor to the core thermal power limit in Figure 2.1 in Reference 17. If the 0.04 adder is applied, the core thermal power limit provided in Figure 2.2 in Reference 17 must be maintained.
Generic pressure penalty of 5.0 psid was added to the results from the limiting end of L
full power case to support coastdown operation (Reference 17).
r i
EMF 97-047 Page 2-5 l
Table 2.3 Dre:, den Unit 3 Cycle 15
Results of Plant Transient Analysis With increased Steam Flow Maximum Maximum Maximum Core Average Vessel *'/
Neutron Flux Heat Flux Dome Pressure Event
(% of Rated)
-(% of Rated)
(osia)
Load Rejection 670 128 1298 /1271 No Bypass (100%P /100%F)
Load Rejection 570 127 1299 /1277 No Bypass
-(100%P / 87%F)
Feedwater Flow 528 133 1166/1137 Controller Failure FHOOS (100%P /100%F)
Feedwater Flow 478 132 1163 l'1138 Controller Failure FHOOS (100%P / 87%F)
Feedwater Flow 620 132 1204 /1173 Controller Failure (100%P /100%F)
Feedwater Flow 530 130 1200/1174 Controller Failure (100%P / 87%F)
MSlV Closure 326 128 1339 /1313 ASME Analysis (100%P /100%F)
MSIV Closure 320 125 1339 /1316 l
ASME Analysis (100%P / 87%F)
Bounding state points.
L Lower plenum pressure.
.t EMF 97 047 Page 3 7 Table 3.1 Dresden Unit 3 Design Reactor and Plant Conditions Reactor Thermal Power 2527 MWt Total Core Flow 98.0 Mibm/hr Core Active Flow 87.3 Mlbm/hr Core Bypass Flow
- 10.7 Mlbm/hr Core inlet Enthalpy 523.0 Btullbm Vessel Pressures Steam Dome 1020 psia Core Exit (upper plenum).
1030 psia E
Lower Plenum 1053 psia l
. Turbine Pressure 964 psia Feedwater/ Steam Flow *'
9.9 Mlbm/hr 4
Feedwater Enthalpy 321.1 Stu/lbm Recirculating Pump Flow (per pump) 17.8 Mlbm/hr t.
i i-l:
l Includes water rod / channel. flow.
Feedwate, flow is set equal to steam flow because control rod drive flow is not modeled separately in the analysis.
\\
i l
EMF-57-047 Page 3-8 i
Table 3.2 i
Dresden Unit 3
. Significant Parameter Values Used in Analysis l
High Neutron Flux Trip 3032.4 MWt Control Rod Insertion Time 3.5 sec / 90% inserted
i Time to Deenergize Pilot Scram 200 msec Solenoid Valves i
Time to Sense' Fast Turbine Control 80 msec
- Valve Closure i
Time From High Neutron Flux Trip 290 msec"'
To Control Rod Motion Turbine Stop Valve Stroh Time 100 msec Turbine Stop Valve Positi T.w 90% open Turbine Control Valve Stroke Time 150 msec (total)
Core Average Fuel / Cladding Gap
- 789.2 Btu /hr-ft: *F Conductance (cycle-specific value) l 1
1 f.
includes a 0.2-second time delay to deenergize scram pilot valve solenoids.
Includes a 50-msec delay for RPS logic transfer and a 30-msec delay until signal is received by RPS logic.
includes a 90 msec delay for signal to reach solenoid valves and a 200' msec delay for pilot scram solenoid valves to deenergize.
~ Calculated by SPC for the Cycle 15 core using RODEX2.
u_-___--____
l EMF-97 047 Page 3-9 Table 3.2 Dresden Unit 3
('
Significant Parameter Values Used in Analysis i
(Continued) l Safety / Relief _ Valve Performance Settings
l Target Rock Safety / Relief Valve (1 valve) l Capacity Per Valve (relief) 155.0 lbm/sec at 1120 psig*'
Capacity Per Valve (safety) 159.5 lbm/sec at 1112.4 psig'"
l Power Relief Valves Capacity (4 valves)'*
Capacity Per Valve 155.0 lbm/sec at 1120 psig Safety Valves Capacity (8 valves)
Capacity Per.Vaive 171.8 lbm/sec at 1277.2 psig Target Rock Valve Delay / Stroke 967/200 msec'"
i Power Relief Valves Delay / Stroke 967/200 msec MSIV Stroke Time '
3.0 sec MSIV Position Trip Set Point 90% open Condenser Bypass Valve Performance l.
Total Capacity 1085 lbm/sec i
Delay to Opening (from demand) 150 msec i
l Opening Time (entire bank, maximum demand) 1.0 see Fraction of Energy Generated in Fuel 0.965*'
I 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) 3311 lbm/sec Recirculating Pump Trip Set Point 1250 psig i
Steam Dome Pressure Valve set points are given in Reference 16.
The relief valve mode of the Target Rock SRV is conservatively modeled with Dresser J
RV flow capacity and set points.
For ASME overpressurization event, Target Rock SRV safety function is not credited.
One relief valve at the lowest set point is not credited.
Reference 12.
L-
EMF 97-047 Page 3-10 Table 3.3 Control Characteristics
Sensor Time Constants Pressure 100 rnsec 1
Steam Flow /Feedwater Flow 250 msec Level 1 sec i
i l
Feedwater Control Mode Single Element
Feedwater 100% Mismatch Water Level Error 45 in L.
Pressure Regulator Settings l
Lead 1.0 sec Lag 6.0 sec
(
Gain 3.33%/psid Bypass Flow Signal Bias 3.0 %
Combined Steam Flow limiter 1.05 Sening Turbine Maximum Steam Flow 2816.67 lbm/sec Recirculation Flow Control Mode Manual l
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 the -control system in the analysis model results in a more realistic calculated plant response. The representative parameters used in the analysis may not be bounding, but their effects on pressure and thermal margins are insignificant.
Dresden Unit 3 plans to have a modification in the f eedwater control system. Dresden
'icensmy analyses are insensitive' to the feedwater control system algorithms or settings. Single-element mode provides slightly more conservative results compared to manual or three element control mode for all events (Reference 1 U.
m b
1 EMF 97-047 Page 311 Table 3.4 Dresden Unit 3 Cycle 15 Comparison of LRNB and TTNB Results With increased Steam Flow
'i Maximum Maximum'*
Maximum Core Average Vessel Neutron Flux Heat Flux Pressure State Point
(% of Rated)
-(% of Rated)
(psig)
ACPR*'
100% Power /100% Flow LRNB 669.7 128.4 1297.9 0.35 / 0.36 TTNB-664.8 128.4 1298.3 0.35 / 0.35 l
r i
4 i
Lower plenum pressure.
Values for second cycle 9x9-2/first-cycle ATRIUM-9B fuel. LRNB ACPRs are 0.0002
. higher than corresponding TTf;B ACPRs.The difference in the ATRIUM 98 results is due to rounding.
EMF 97-047 Page 31,2 Table 3.5 Turbine Bypass Valve Degradation Study ACPR Results With increased Steam Flow Bypass Valve Second-Cycle First-Cycle Delay Time
9x9-2 ATRIUM 98 msec A( ACPR)
A(ACPR)
50 0.000 0.000 100 0.010 0.010 150 0.017 0.019 250 0.026 0.029 350 0.031 0.033 450 0.035 0.037 550 0.039 0.041 700 0.042 0.044 No Bypass 0.042 0.044 I
l l
Delay is relative to time of TSV full closure (TSV closure takes 105 msec).
Relative to ACPR for case with 50-msec delay (FWCF st 100% power /87% flow with FHOOS in Table 2.1).
1 i
L__.____._____._.________________
(-
EMF 97-047 Page 313 1
Table 3.6 input for MCPR Safety Limit Analysis Fuel-Related Uncertainties
l Statistical Parameter Source Document Treatment ANFB Correlation *'
References 4,15, 20, 21, and 25 Convoluted Radial Peaking Factor
References 13 and 21 Convoluted Local Peaking Factor Reference 5 Convoluted Assembly Flow Rate Reference 14 Convoluted Channel Bow Local Reference 3 Convoluted Peaking Factor Plant Measurement Uncertainties Uncertainty Statistical Parameter Units Value'd' Percent Treatment Feedwater Flow Rate Mlbm/hr'
- 16. 5 "'
2.306 Convoluted Feedwater Temperature F
340.1'"
2.36 Convoluted Core Pressure psia 1030 1.42 Convoluted Total Core Flow Mlbm/hr 98.0 2.50 Convoluted Core Power MWt 4260"'
Allowed to vary with heat balance Fuel related uncertainties are proprietary and can be found in the indicated references.
Additive constant uncertainty values are used.
Radial peaking factor uncertainty includes allowances for up to 40% (equivalent of 2 TIP machines) of the TIP machines out of service (with POWERPLEXS-Il CMSS SUBTIP methodology), LPRM recalibration interval up to 2000 EFPH, and LPRM failures up to 50% with POWERPLEXS-ll CMSS bypass methodology on or off.
'd '
Values are from analysis to support a two-loop MCPR safety limit of 1.08 provided in Reference 25. The Cycle 15 SLMCPR is 1.08 from Technical Specifications.
Feedwater flow rate and core power were increased above design values to attain l-desired core MCPR for safety limit evaluation, consistent with Reference 3 roetovoology.
Disposition of increased feedwater temperature associated with increased steam flow is presented in Reference 26.
r-_________--__.
I I
EMF 97-047 Page 3-14 10004 Core Power
_ _ _H_e_at_ _Fl_ux_ _ _ _ _ _
Core Flow
~~
St_eam Flo_w Feedwater Flow 0
oooo-
=
5x
-o m.
Cv 2
c.
2oco -
-w_______
__________m
)
f
([ \\ [(,M [N/
I
- ~~
s o.
s l
-N lo 4.o So 4
to 2.o Time, seconds t
l l:
l-Figure 3.1 Load Rejection No Bypass at 100/1CO -
Key Parameters l.
Lu ____-- -__ _ _ _-_ --_
l l
EMF 97-047 Page 3-16 son l
son
-G i
Ca L.
I 3 =co -
0 3:
's a
e 1
10D -
D D
1'D
$0
$0 O
gp Time, seconds Figure 3.2 Load Rejection No Bypass at 100/100 -
Vessel Water Level (Referenced to instrument Zero) l l
m__________
EMF 97-047 Page 3-16 1350.0 1300.0 -
O
'iBc.
6 8 1250D-E E
.g.
12004 -
E D'
so d 11504 -
"3 E.
11004 -
i 10504
's s'n b
3B 44 Sb l'
Time, seconds i
l' Figure 3.3 l-Load Rejection No Bypass at 100/100 -
H Vessel Pressure Response
i EMF 97 047 i
Page 317-l
)
l l
l SooD -
RV.
1 Volve
_ _ _R_V_. _ _2_. V_o_lv_es_ _.
]
SRV,1 Vdve q
1 SV, - 2 V_olves i
4oco.
l
,SV,,2 Volves _ _ _
I y
f
- < soao-E.
I a
i I
t' i'
2 i
L 1
8 :oon-l 3
1 i
ie i/
I!
il if l'
too.o -
Il c
i I
.o.
.o 14 2.0 10 44 54
. Time, seconds i
F.:9ure 3.4 1
Load Rejection No Bypass at 100/100 -
Safety / Relief Valve Flows I:.
I
(.
L l
u_____._______..
____m
EMF.97 047 Page 3 :w.o CORE POWER
___H_ EAT _ FLUX _____.
~ ~ ~ ~ ~
STEAM FLOW
.F_EED.W A TE. IR F.L_OW~~
~
,,,o.
t
'y O
\\
e
-/.
o s
g
'o f
~~~_.....
h)r s,
'*3.... _
,aan g
-l i
u t
I r
n.
s
\\
\\
f I
60.0 -
') 4^ ~
\\
l l
I
)
o
.0 1bD 2bb 3bD ~
eb.O DOo Time, secones 900.0 CORE POWER
___H_ EAT _F.L.UX _____.
p
. ST_EAM FLOW ecco.
-FEEDWATER FLOW _._
m 400D -
o 2
U t
- mo.
Q
~
?...... C ' %.. % -
~~~'==---n___-----__________.
q f1 g
f \\/,( %
\\
b=
t
/
l l
l 1
-:54 I
44 4 45.0 44.0 47.0 44LC 4G4 Time, seconds Figure 3.5 Feedwater Controller Failure at 100/100 FHOOS -
Key Parameters
r EMF-97 047 Page 3-19 1
80.0 i
o I
eon -
.E i
s
- 2 J
w
= 4o.0 -
Ow
-e 20 4 -
1
.0 4
1d.0 Od.0 3d.D 4d.O SO4 Time, seconds 70.0 65.0 -
W m
c~-
J w= ss.O-Oa
.> som-4s.0 -
l son 4do 4N.0 48.0 40.0 44 4 4$m Time, seconds l
Figure 3.6 l
Feedwater Controller Failure at 100/100 FHOOS.
Vessel Water Level
- -,....-n.-.
1 EMF-97-047 Page 3-20 t 200.0
.9g t iso.o -
e
),
C et
/
1 h t ioo.o -
.E I
i3 z
N. soson -
ioooo A
1bb 2bD Jb.o ebb sod Time, seconds
]
12co.o i
{
l
.9g i sso.o -
a b
n.
11och -
~E b.3 i
z, a
C iosoo-l 5000D 44 464 edo 4Y.D 454 44.0 1
Time, seconds Figure 3.7 Feedwater Controller Failure at 100/100 FHOOS -
Vessel Pressure Response L_________-____________
f EMF 97-047 Page 3-21 Soon RV, 1 Volve
_ _ _R,V _ 2_V_qves,,,
3
,,,,,,S RV, 1_ _V_ol_v e SV, 4oo.o -
- 2 Volves SV.
2 Volves O
R soon -
E&
I e-oC
}
% 200 4 -
l
~6 t oom -
l
- O a
s
.o to.o 204
.no 404 504 Time, seconds 5004 i
RV, 1 Volve R,V,3,, 2_y,olves,,,
SRV, 1 Valve 4com -
- 2 Voives SV,
,SV, 2 Volves O
i
< sooo -
l E
l G
i I
% 2004 -
3>
{
soon -
A 44.0 47.0 edLa 4GD um 43.0 Time, seconds Figure 3.8 Feedwater Controller Failure at 100/100 FHOOS -
Safety / Relief Valve Flows
EMF-97-047 Page 3 22.
100 90 80 70 -
c1o m
j 60 U3 o
50 u
v
.o 4o E
p Z
30 20 10 0
.0
.2
.4
.6
.8 1.0 1.2 1.4 1.6 Radici Power Peaking l
l l
Figure 3.9 Design Basis Radial Power Distribution for SLMCPR Determination i
I L._______
l EMF-97-047 Page 3-23 l
l Cont o1 R00 Coe ne-1 O
n t
0.99-0.995 1.036 1.005
^ 087 1.073 1.078 1.005 1.063 r
0 1
0.995 0.974 1.026 1.000 0.878 1.068 1.057 0.866 1.051 R
o d
1.036 1.026 1.001 1.031 1.041 1.102 1.023 1.004 1.082 C
0 1.005 1.000 1.031 0.869 0.973 1.033 r
Internal 1.087 0.878 1.041 Water 1.075 0.764 1.012 r
Channel 1.073 1.068 1.102 1.028 0.937 0.998 1.073 1.057 1.023 0.869 1.075 1.028 0.761 0.932 1.007 1.005 0.866 1.004 0.973 0.764 0.937 0.932 0.759 1.046 1.063 1.051 1.082 1.033 1.012 0.998 1.007 1.046 1.022 1
Maximum Local Power:
1.102 l
1 i
l Figure 3.10 Design Basis Local Power Distribution for SPC-ND ATRIUM-9B Fuel (SPCA9-326811GZ-80M)
Uncontrolled at 15,000 mwd /MTU and 70% Void for SLMCPR Determination I
l L_________.___..__..____
J
EMF 97-047 Page 3-24 Contro1 Rod Corner
.o n
1.005 0.983 1.053 1.028 1.086 1.075 1.081 1.026 1.044 r
o
'l 0.983 0.990 1.029 0.916 0.986 1.064 1.057 0.887 1.056 I
R-1;053 1.029 1,008 1.040 1.044 1.000 1.025 1.005 1.083 l
C o
1.028 0.916 1.040 0.890 0.969 1.034 r
l Internal n
1.086 0.986 1.044 Water
.1.070 0.955 1.010 r
Channel 1.075 1.064 1.000 1.030 0.773 1.008 1.081 1.057 1.025 0.890 1.070 1.030 0.803 0.936 1.016 1.026 0.887 l'005 0.969 0.955 0.773 0.936 0.803 0.956 I
1.04d 1.056 1.083 1.034 1.010 1.008 1.016 0.956 0.961 Maximum Local Power:
1.086 1
i i
Figure 3.11 Design Basis Local Power Distribution for i
i SPC-ND ATRIUM 9B Fuel (SPCA9 339B-6GZ-80M)
Uncontrolled at 17.500 mwd /MTU and 70% Void for SLMCPR Determination I
t I
I
LMF 97-047 Page 4-3 i
Table 4.1 Dresden Unit 3 Cycle 15 Results Summary of ASME Overpressurization Analyses With increased Steam Flow Maximum Pressurization (psig)
J 1
Transient Steam Dome Lower Plenum MSIV Closure (100%P /100%F) 13 T2.9 1338.7 (100%P / 87%F) 1315.6 1338.7 TSV Closure (100%P / 87%F) 1315.9 1338.5 i
TCV Closure (100%P / 87%F) 1315.9 1338.5 l
1 I
L l
i EMF-97-047 Page 4 4 soar Core Power
_ _H_e_at_ _Fl_ux _ _ _ _
Core Flow Steam Flow Feedwate Flow sono-I a
(T.
o 200.o -
1E 8
t c.
goo.o
%'N s
%=.%
.o..
'}G\\f;&\\/\\,W
\\
-=
2.0 AD 6.0 34 10.0
.0 -
. ime, seconds T
l l
Figure 4.1 MStV Closure at 100/100 -
Key Parameters l
}.
,(
s_
EMF 97-047 Page 4' 5 l
i I
14004 L
i O.
1 tsoc.o-c i
),
I di-
- h. 120co-E ba 0
l J
l 2
m-E 1100.0-l i'
1' s
d a
J 6
4 2,0 44 64 80 104 Time, seconds N
l 1
l t'
Figure 4.2 I
' j MSIV Closure at 100/100 Vessel Pressure Response O_____________________
- m_.
j EMF 97 047 Page 4 6 1
l.
I 1
l 40.0 l
I-1 l-y
.G.
l A
$a u
i l
2 mn-
-o it e
I..
E 10D -
l D
.0 2.0 44 SD 54 10.0 l
Time, seconds l'
i Figure 4.3 l-MSlV Closure at 100/100 -
l Vessel Water Level (Re'ferenced to Instrument Zerol tl' l
L'
.~
l.
t I-I.
f-EMF 97-047 l
l Page 4-7 i
I i
2 Vo
- - - - - - - - -.l.v e s _.
SV.
2 Volves SV 2 Vojves 2
.SV.
2 Valves o
8 s soco.
/
, ~ ~,
E
/
p 1-i E
-O l
'*..,.'~.%
l--
~[
I l I
y y 2aan.
a l l*;
m>
.I I l II I I i
,I
,N 1004 -
l'jj iI f,
i i
l t
i A
4 s
4 6
27 -
Time, seconds Figure 4.4 MSIV Closure at 100/100 Safety Valve Flow ' Rates l
(SRV Assumed inoperable)
.s 6
1 4
'l
~
EMF-97-047 Page 5-4 I
)
i l
l Table 5.1 Automatic Flow Control Excursion Path
(
4 Recirculating Flow Power
(% of Rated)
(% of Rated) 100 100 90 92 80 84 i
70 76 60 68 50 60 40 52 30 44 i
l l
i l
1 I
EMF 97-047 Page 5 5 l
1 i
Table 5.2 Reduced Flow MCPR Limits for.
Automatic Flow Control (ATRIUM 9B and 9x9-2 Fuel) l
)
-i Recirculation.
MCPR, Limit MCPR, Limit MCPR, Limit j
Flow for for for
(% of Rated)
OLMCPR = 1.46 OLMCPR = 1.50 OLMCPR = 1.55 100 1 46 1.50 1.55 30 2.29.
2.35 2.45 j
0 2.79 2.87 2.97 j
f i
I l
4 I
so W MCPR = I.46
.j i
28 WCPR e 1.30 8
l
_ uc,R - i.33 7.
3 x
g2 N
b' s 22 E
2 24 l
'N 5u i.
i 1
0 20 40 60 to SCO t:0 I0tes Core Flow (*; Rotec)
NOTE: Larger view of the above graph is found on page 5-10 (Figure 5.1).
i 4'
EMF-97-047 Page 5~-6 l
t Table 5.3 Manual Flow Control Excursion Path i
i I
l 1
Recirculating Flow Power j
(% of Rated)
(% of Rated) i 110 120 l
100 111 I
i 90 102 i
80 93 70 83 60 74 50 65 40 56 30 47 l
~
EMF 97-047 l
Page 5-7
- Table 5.4 Reduced Flow MCPR Limits for Manual Flow Control (ATRIUM-9B and 9x9 2 Fuel)
Recirculation Flow MCPR
(% of Rated)
Limit 100 1.18 30 1.86 0
2.35 14 a
e i
i a
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NOTE: Larger view of the.above graph is found on page 511 (Figure 5.2).
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EMF-97 047
.s Page 5-8 i
F 3.0 c
c MCPR = 1.46 2.8 0 MCPR = 1.50 ;
4 a MCPR = 1.55
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Figure 5.1 Reduced Flow MCPR Limit for Automatic Flovi Control t
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I EMF-97-047 Page 5-9 14 Reduced Flow MCPR Limit a
ATRIUM-98 Results 2.5 -
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O 20 40 60 80 Total Cdre Flow (7. Rated)
Figure 5.2 1
Reduced Flow MCPR Limit for l-Manual Flow Control (SLMCPR = 1.08) r (ATRIUM 9B and 9x9 2 Fuell
EMF 97 047 Page A-4' i
l Table A.1 SLO Reactor and Plant Conditions Reactor Thermal Power (81.3%)
2054.5 MWt Total Recirculation Flow (58%)
56.84 Mlbm/hr Core Bypass Flow 6.07 Mlbm/hr
l Core inlet Enthalpy 507.8 Btu /lbm Vessel Pressures r
Steam Dome 993.8 psia Core Exit 999.0 psia i
Lower Plenum i
1012.6 psia Turbine Pressure 958.1 psia Feedwater/ Steam Flow 7.80 Mlbm/hr t
Feedwater Enthalpy 294.2 Btu /lbm l
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EMF.97 047 Page A-5 1
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EMF 97-047 Page A 7 l
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