ML022560157
| ML022560157 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/30/2002 |
| From: | Framatome ANP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML022560157 (91) | |
Text
Attachment 3 Browns Ferry Nuclear Plant (BFN)
Transition To Framatome Advanced Nuclear Power (ANP)
Fuel Non-Proprietary Presentation
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Transition to Framatome ANP Fuel at Browns Ferry DRAFT F *Presentation to NRC 2*,Rockville, Maryland October 3, 2002 FA S~FRAMATOME ANP
1 to~
> Transition
> Transition Experience Approach
> Transition Cycle Analyses
> Safety Analysis
> Mixed Core Ant s
> Neutronics
> Thermal Hydraulics Methods Application for Power Uprate A
FRAMATOME ANP Agenda 9
FT
.......... 01
Transition pe e ce
.
F A
L UI I
I Transition Experience DRAFT A
FRAMATOME ANP
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> Framatome ANP (Richland) has performed new fuel vendor transitions for more than 60 reload cores
- 26 BWRs
- 34 PWRs
> Framatome AN isFr, Fnansition DRA~Fn~T cycles A
INFRAMATOME ANP Framatome ANP Transition Core ExperienceJ
r poll j
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Plant Reactor Type Year Fuel Type(s)
Big Rock Point GE 1971 1lxl1 Ka hi AEG 1972 6x6 Oyster Creek GE MkIl 1973 7x7, 8x8 Humboldt Bay GE 1973 6x6 KRB 7W 72 m
- 72 6x6 LaCrosse Als r#*k r1 75 10x10 Dresden-1DUEIX F, 1976 6x6 Oskarsham-1 ABB-Atom 1977 8x8 Lingen AEG 1977 6x6 Barsebeck-1 Asea-Atom 1979 8x8 Dresden-3 GE BWR 3 1982 8x8,9x9-2, ATRIUM-9 Chinshan-1 GEBWR4 1983 8x8 Dresden-2 GE BWR 3 1983 8x8, 9x9-2, ATRIUM-9 J
BWR Transition Core Experience A
FRAMATOME ANP
VON 2i, wj, jy Plant Reactor Type Year Fuel Type(s)
Grand Gulf GEBWR6 1983 8x8, 9x9-5 Kuosheng-1 GE BWR 6 1985 8x8, 9x9-2, ATRIUM-9, ATRIUM-10 Chinshan-2 GE BWR 4 1985 8x8 Susquehanna-1 GE BWR 4 1985 8x8, 9x9-2, ATRIUM-10 Kuosheng-2 GE BWR 6 1986 8x8, 9x9-2, ATRIUM-9 Susquehanna-2 GEl 16 8x8,9x9-2, ATRIUM-10 WNP-2 R-ATRIUM-9 Quad Cities-2 GEBWR3 1997 ATRIUM-9 Quad Cities-1 GE BWR 3 1998 ATRIUM-9 LaSalle-2 GEBWR5 1999 ATRIUM-9 LaSalle-1 GE BWR 5 1999 ATRIUM-9, ATRIUM-10 Grand Gulf GEBWR6 2001 ATRIUM-10 River Bend GEBWR6 2001 ATRIUM-10 A
FRAMATOME ANP J
BWR Transition Core Experience
BWR Transition Core Experience J
Recent BWR Experience
"> Quad Cities
"* ATRIUM TM-9B fuel coresident with GE9 and GEI0 fuel
"* Unit 2 startup in June 1997
"* Unit 1 startup in October 1998
"> LaSalle A''
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-
4
-
F m Unit 1 startu
> Grand Gulf 7
"* ATRIUMTm-10 fuel coresident with GEl11 fuel
"* Grand Gulf startup in May 2001 river Bend m ATRIUMm-10 fuel coresident with GEI11 fuel
- River Bend startup in October 2001 A
FRAMATOME ANP
- ATRIUMT_ -c
- Unit 2 startu I
F BWR Transitions Currently in Progress
> Columbia
"* ATRIUMTM-10 fuel coresident with SVEA-96 fuel
"* Columbia startup in May 2003
> Browns Ferry m ATRIUM TM-,
and GE13 fuel m Unit 3 startup in April 2004 A
FRAMATOME ANP BWR Transition Core Experience J
as to pp ac
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i Transition Approach DRAFT FRAMATOME ANP
Transition Core Approach
> Maintain current plant licensing basis to the extent possible
> Evaluate the introduction of ATRIUM TM -10 fuel per the requirements of 10 CFR 50.59
, Similar to approach used for any plant change y Similar to apl h
,f°rare design (except for scope)
> Identify plant sa a-nalyses potentially affected by a fuel or core design change
> Assess impact of fuel transition on plant safety analyses and repeat analyses if the conclusions of the analysis are potentially affected FRAMATOME ANP
Transition Core Approach if
> Technical specification changes generally limited to
"* References to NRC-approved methods used to determine thermal limits specified in the COLR
"* MCPR safety limit based on Framatome ANP methods 0
> COLR thermal based on analyses using NRC-approved
-'-4 of FRA-ANP fuel the transition core methods A
FRAMATOME ANP 9
Transition Core Approach Steps in Transition L- > Data collection
> Model construction and benchmark
> Core follow analyses
> Compatibility analyses
.I L*72
> Establish current licensing basis
> Disposition of events
> Plant transition
> LOCA analysis
> Core monitoring system parallel operation
[77#
> Criticality analyses
> Transition cycle design
> Initial cycle reload licensing analyses
> Licensing support
> Training plilili';FRAMATOME ANP
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> Obtain the plant data required to develop core neutronic, plant transient, and LOCA analysis models
> Framatome ANP can provide onsite representative to assist with data collection mentation and
> Utility provides plant operating
> Framatome ANP develops documentation necessary to describe source references for reactor data used in analysis models (design data package or calculation notebooks)
A FRAMATOME ANP Transition Core Approach Data Collection
177 40lo WON "POO Sources of Input Data
> NSSS reactor drawings
> Balance of plant drawings
> FSAR
> Techn
> NSSS FT
> Plant component technical manuals
> Coresident fuel design reports
> Parameter lists for previous safety analyses 9
Transition Core Approach Data Collection (Continued)
A FRAMATOME ANP
Transition Core Approach
-Thermal-Hydraulic Compatibility
> Thermal-hydraulic compatibility with reactor and coresident fuel is evaluated m Core pressure drop
- Core bypass flow Flow distrib "m Thermal mat7RAFT m Water channel flow
> Documented in the Fuel Design Report for the initial reload of a fuel design A
FRAMATOME ANP
"Transition Core Approach Establish Current Licensing Basis
> Licensing basis consists of all analyses performed to demonstrate that regulatory requirements are met
[
> Licensing basis is defined in documents, such as L FSAR Technical S
- fica s
m Core Operati g L
- Technical RV rntsandT "m Cycle Reload Licensing Reports Extended Operating Domain Reports (e.g., increased core flow operation)
, Equipment Out-of-Service Reports (e.g., feedwater heaters OOS)
Fol NOW -'
[ MAI.
- tools,
> The first transition cycle for ATRIUMTM-10 fuel at Browns Ferry will be Unit 3 Cycle 12 (BF3C12)
> An extended power uprate will also be implemented for BF3C12
> The licensi performed to si
- le all analyses
%r uprate A
FRAMATOME ANP Transition Core Approach Establish Current Licensing Basis (Continued) 9
Transition Core Approach Disposition of Events
> Framatome ANP reviews all event analyses identified in the current licensing basis
> Analyses performed to demonstrate compliance with fuel related design or licensing criteria are identified
> Fuel-related analyses are classified as Not impacte th ngn fFdesign Bounded by ece otltr event Potentially ilrL)'-.Ptaaeing Fr atome ANP methodology
> Rated and off-rated conditions considered
> Potentially limiting events are evaluated generically or reanalyzed each cycle using NRC-approved methodology
> Analyses consider operating conditions and plant parameters associated with the extended power uprate no=
FRAMATOME ANP
Transition Core Approach Plant Transition Safety Analysis
> Plant transient and accident analyses are performed prior to the initial transition cycle to support the introduction of Framatome ANP fuel m Equilibrium cycle design used in analyses m Potentially limiting events are analyzed m Analysis results may be used to disposition some events as non limiting and q
forVcl nalyses Expectednopmiratio LrI
'are determined for nominal opatici m Analyses performed for EOD and EOOS options
- Approach and basis for EOD/EOOS operating limits is established
> Summary of disposition of events is prepared m Identifies event analyses to be performed on a cycle specific basis A
FRAMATOME ANP
Transition Core Approach LOCA Analysis
> Break spectrum analyses are performed to determine the limiting break characteristics based on Framatome ANP methodology
- Break location 17
- Break configuration "Breaksize iRAFT
- ECCS failur
> MAPLHGR limit analysis m Performed for the limiting break characteristics determined from the break spectrum
- Demonstrate that all 10 CFR 50.46 criteria are satisfied for MAPLHGR limit versus exposure
FRAMATOME ANP
> Technical specification updates
> FSAR updates
> Fuel data for EPGs DRAFT Transition Core Approach Licensing Support A
FRAMATOME ANP I
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VIA
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"> POWERPLEX CMSS
"> Use of Framatome ANP design and safety methodologies
"* Formal classroom training on codes/methods
"* Work experience training at Framatome ANP (utility engineer at Richland to° Dprg FA MA O E Slid iiFRAMATOME ANP Transition Core Approach Traininq
st y e ye Q.; 't V'
I m Transition Cycle Analyses DRAFT A
FRAMATOME ANP
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> Results from the disposition of events define the safety analyses required for the transition cycle to address the in fuel and core design change
> A preliminary disposition of events has been performed for the pre-power uprate Browns Ferry licensingasis
> The final d to support the
- s analyses performed A
SIFRAMATOME ANP J
Transition Cycle Analyses
WIN DRAFT A
FRAMATOME ANP I-Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate)
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A FRAMATOME ANP Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate)
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A FRAMATOME ANP Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate)
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A FRAMATOME ANP Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate) I i
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A FRAMATOME AMP Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate)
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FRAMATOME ANP Transition Disposition of i
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A FRAMATOME ANP I-Transition Cycle Analyses Disposition of Events (preliminary pre-power-uprate)
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FRAMATOME ANP I-Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate) i
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FRAMATOME ANP I-Transition Cycle Analyses Disposition of Events (preliminary pre-power uprate)
Transition Cycle Analyses Disposition Summary Analyses for Initial Reload
> Mechanical design
> Nuclear design SStability m Cold shutdown margin
> Thermal hydrai Hydraulic co t
"- MCPR safety limit m MCPRf (slow flow excursion)
> ASME over-pressurization d
> ATWS m Over-pressurization n Standby liquid control system A
FRAMATOME ANP
K 4I Mi, 1t
- 1 Analyses for Initial Reload (continued)
> Criticality analyses
"* New fuel storage
"* Spent fuel storage
> Abnormal operational transients
"* Load rejectioRA
- Turbine trip by
"* Loss of feedwater heating
"* Inadvertent ECCS pump startup
"* Control rod withdrawal error
"* Fuel assembly mislocation
"* Fuel assembly misorientation
"* Startup of idle recirculation loop
"* Feedwater controller failure FT A
FRAMATOME ANP Transition Cycle Analyses Disposition Summary.
J
- a - - - -
-.. 010
r ATKI OVA-Analyses for Initial Reload (continued)
> Design basis accidents
"* Control rod drop accident
"* Loss of coolant accident 0
> Emergency op(
m Fuel dependent input parameters
> Post-fire safe shutdown A
FRAMATOME ANP Transition Cycle Analyses Disposition Summary T
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Lil Sach I
oi1, 1 MO Safety Analysis Methodology DRAFT AM FRAMATOME ANP
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> Introduction of Framatome ANP fuel requires confirmation that fuel-related design and licensing criteria continue to be satisfied
> Potentially limiting events are analyzed using Framatome ANP methodo0l
> No new metho to be required for the transitiot Cycle 12 Browns Ferry Unit 3 A
SEFRAMATOME ANP Safety Analysis Methodology J
- KOM,
> Major computer codes m XCOBRA (XN-NF-79-59(P)(A))
- Predicts the steady-state performance of BWR cores at various operating conditions and power distributions
- SAFLIM2 (ANF-524(P)(A) Revision 1 and Supplements 1 and 2) 9.Used tod ter n abM safetylimit for a core design
> Other methodolDRAF T
"* SPCB CPR correlation (EMF-2209(P)(A) Revision 1)
"* Application of CPR correlation to coresident fuel (EMF-2245(P)(A))
A FRAMATOME ANP J
Thermal-Hydraulic Analysis Methodology
> Major computer codes m CASMO-4/MICROBURN-B2 (EMF-2158(P)(A))
- Assess impact on thermal limits during localized or quasi-steady-state events, such as control rod withdrawal error and fuel loading error
- COTRAN and 2) 9 Determi
- STAIF (EN
- Calculat
'ther method m Generic los Revision 1)
(XN-NF-80-19(P)(A) Volume 1 and Supplements 1 in c pRA*
- ding ontrol rod drop accident 1F-CC-074(P)(A) Volume 4) te core and channel decay ratio (frequency domain) ology ss of feedwater heating analysis (ANF-1358(P)(A)
)
A FRAMATOME ANP J
>0 Neutronic Analysis Methodology
N ow "A
l t ip A 1 six 9
> Major computer codes m RODEX2-2A (XN-NF-82-06(P)(A) Revision 1 and Supplements 2,4,5) 9 Calculate gap conductance for core and hot channel
- XCOBRA ()
o Determin(
m COTRAN, (N-NF-79-59(P)(A))
l and Suplplements 2, 3, 4)
- Calculate reactor system and core response during transient events
- XCOBRA-T (XN-NF-84-105(P)(A) Volume 1 and Supplements 1, 2)
- Calculate ACPR for each hot channel.in the core A
FRAMATOME ANP Transient Analysis Methodology
LOCA Analysis Methodology
> Major computer codes RODEX2-2A (XN-NF-82-06(P)(A) Revision I and Supplements 2,4,5) j
. Determine initial fuel characteristics U RELAX (EM)
- "e Determin o
0s s du g the blowdown phase e Determine hot channel response during the blowdown phase
. Determine reactor system response during the refill and reflood phases (time of core reflood)
HUXY (XN-CC-33(P)(A) Revision 1)
SCalculate PCT and metal-water reaction rate at limiting axial plane A
FRAMATOME ANP
1-F, ' -i
4 J
Summary
"> The transition process is well understood by Framatome ANP
"> The transition approach described is applicable with either original or uprated power (i.e., is independent of core power ievei)
> Framatome AN t
,r Au4"methods, and technology to support a smooth transition to ATRIUM TM-10 fuel at Browns Ferry A
FRAMATOME ANP Transition Core Approach
- - 1 '- ° Mixed Core Analyses DRAFT A
FRAMATOME ANP
VISA
- OVA,
- -kilo.
ANA" 9
> Analyses are performed to confirm that all design and licensing criteria are satisfied
> Analyses explicitly include each fuel type in the core m Performed using generically approved methodology m Cycle-spi m Input
> Thermal limits are established for each fuel design
> During operation, each fuel type is monitored against applicable limits A
FRAMATOME ANP Mixed Core Analyses Licensing Approach
Mixed Core Analyses Analysis Considerations
> Mechanical Analyses
> Neutronic Analyses
> Thermal-Hydraulic Analyses
> CPR Correlation
> Transient Analyses
>*Stability Analyses
> LOCA Analyses
> Core Monitoring FRAMATOME ANP
1Ln RICO
> Fuel mechanical limits for both FRA-ANP fuel and coresident fuel are verified for cores designed by FRA-ANP m Mechanical limits (steady-state and transient) for coresident fuel will be supplied to FRA-ANP by TVA
- FRA-ANPpI steady-stat I during AQOs it core designs meet exceeded
- FRA-ANP has experience checking transient LHGR, MOP, and TOP limits for GE fuel in previous transitions A
FRAMATOME ANP Mixed Core Analyses Mechanical Analyses J
L AN
[,7 lo
> Thermal-hydraulic analyses explicitly address mixed cores N Hydraulic characteristics are determined for each fuel type in the core
- Hydraulic characteristics are based on flow test measurements for both FRA-ANP fuel and coresident fuel
"* Measure nt e or all fu types with the FRA-ANP flow loop to eli na ri il testini
"* The tests result in consistent assembly and component pressure drop information for each fuel type in the core
"* Flow tests for GE14 fuel completed in February 2002
- XCOBRA computer code is used in thermal-hydraulic analyses
"* Each fuel assembly type is individually represented in the model
"* Assembly flow is calculated for each fuel type in core A
FRAMATOME ANP Mixed Core Analyses Thermal Hydraulic Analyses F
F1-,. 7 tie, too
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> Critical power is calculated for each assembly using local fluid conditions
"* The FRA-ANP CPR correlation is applied to both the FRA-ANP and coresident fuel
"* The NRC-approved process for developing additive constants (correlation described in 9 OperatingMN"diti~ns eor coresident fuel is ind expected operation
"* TVA provides calculated critical power performance data for GE coresident fuel
"* The calculated critical power data and the corresponding operating conditions are used to determine additive constants and uncertainty A
FRAMATOME ANP Mixed Core Analyses CPR Calculation J
Mixed Core Analyses MCPR Safety Limit
> MCPR safety limit analysis specifically addresses mixed core configuration
- FRA-ANP CPR correlation used with fuel type specific additive constants n Core power is increased until the limiting assembly is at the MCPR safety limit
- Monte Carlo ac*
diterent parameters important for
~ca Rpo fcaEl 5
thnumber of rods in boiling transition for each assembly in the core SFRA-ANP correlation uncertainties for each fuel type are used
- The safety limit is calculated/verified for each cycle design Calculations are performed throughout the cycle to ensure that the safety limit is bounding for the entire cycle A
FRAMATOME ANP
Mixed Core Analyses Transient Analysis
> Transient response is determined for each fuel type m COTRANSA2 calculations determine the upper and lower core plenum boundary conditions
- Plenum boundary conditions are applied to core in XCOBRA-T calculation e The c*The core i maaeflow paths
- Each fuel is c preAntel a hot
-sembly with appropriate geometry and thermal-hydraulic characteristics V7
- Transient local fluid conditions are calculated for each fuel ty pe for use in CPR correlation m Transient CPR is calculated for each fuel type FRAMATOME ANP
Mixed Core Analyses LOCA Analysis
[,K
> LOCA limits (MAPLHGR) are established for each fuel type w LOCA/ECCS analyses are performed to determine limits for t
FRA-ANP fuel 9 System response e Hot channel response
- Coresident f Ictn System r cted by change in the fiel design Pc 9 Core volumes and stored energy represent only a small fraction of the
.,total system values and are similar with either fuel type 9 Hydraulic differences between fuel types are small relative to the pressure drops that exist during a LOCA
- To support operation or plant changes that impact LOCA analyses, FRA-ANP can perform LOCA analyses for coresident fuel A
FRAMATOME ANP
r I L I
> Cycle design and licensing evaluations explicitly consider the mixed core configurations
> Limits are established for each assembly design type
> Operation withi cycle Mixed Core Analyses Summary A
FRAMATOME ANP
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"> BWR core transition experience
"> Transition steps/mixed core analysis
"> Neutronic methods
"> Applicability of neutronics methods to power uprates DRAFT A
FRAMATOME ANP Presentation to Cover 9
-
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L
F.
[4 F.
7 Reactc Reactor Type Chinshan 1 Chinshan 2 Kuosheng 1 Kuosheng 2 River Bend Grand Gulf 1 Susquehanna 1 Susquehanna 2 LaSalle 1 LaSalle 2 Columbia GS Browns Ferry 2 Browns Ferry 3
- BWR/4 BWR/4 BWR/6 BWR/6 BWR/6 BWR/6 BWR/4 BWR/4 BWR/5 BWR/5 BWR/5 BWR/4 BWR/4 Power, MWt or
(% Rated Increase)
Neutronics
[C]
1775 (0.0)
Full
[C]
1775 (0.0)
Full
[C]
2894 (0.0)
Full
[C]
2894 (0.0)
Full OTmN A Ftl
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.34 4 *(4. 1 Pf r 'Fal
[C]
3441' (4.5)
Partial
[C]
3489* (5.0)
Partial
[C]
3489* (5.0)
Full
[C]
3486*(4.9)
Full
[D]
3952* (20.0)
Full
[D]
3952* (20.0)
Full Safety Full Full Full Full Prtial Partial Full Full Full Full Full Lates Fue
- Latest power uprates (BF3 uprate in 2004, first reload FRA-ANP fuel).
A FRAMATOME ANP FRA-ANP BWR Fuel Management and Licensing Workscopes (August '02)
Cycle
- Length, Months 18 18 Latest Fuel Type ATRI UMTM-10 ATRIUMTM-10 ATRI UMTM-1 0 ATRI UMTM-10 ATRIUMTM-10 ATRIUMTM-10 ATRIUMTM-10 ATRI UMTM-10 ATRIUMTM-1 0 ATRI UMTM-10 ATRI UMTM-1 0 ATRIUMTM-1 0 ATRIUMTM-10 i
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18 18 18 18 24 24 24 24 24 24 24 i
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Transition Core Approach Steps in Transition
> Data collection (neutronic data-fuel, core, TIP, startup, operating)
,-J-*
> Model setup - CASMO-4/MICROBURN-B2 (mixed core effects accounted for by explicit modeling of all fuel in the core)
L
> Benchmark previous 4-5 cycles (core follow)
> Define core hot, t
ff
> Establish curreLas
> Disposition of events
> Provide neutronic input to transient, LOCA models
> Prepare core monitoring system input for Cycle N-1 parallel operation IBI A
FRAMATOME ANP
NS,
> Perform criticality safety (new and spent fuel) and fuel handling accident analyses
> Develop neutronic fuel design for Cycle N
> Perform Cycle N neutronic safety reload licensing analyses
> Provide startup)
> Licensing supp4 fCycle N A
FRAMATOME ANP Transition Core Approach Steps in Transition (Continued) i
FRA-ANP Neutronics Safety Analysis
> Approved methods generally defined in XN-NF-80-19(P)(A) or subsequent submitted and approved Topical Reports z-, 3LA
> Browns Ferry FSAR Analyses
- Cold shutdown margin (Section 3.6)
, Reactor core stability_(Sections 3.6 Standby boro l4APcT m Loss of feedwater heating (Section 14.5.3.1) i S,-*
Control rod withdrawal error (Sections 14.5.4.1)
Fs U Fuel assembly mislocation error (Section 3.6.4.2.3) 25
- Fuel assembly misorientation error (Section 3.6.4.2.3)
- Control rod drop accident (Section 14.6.2)
FRAMATOME ANP
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- city, fly".
EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validatbn of CASMO-4/MICROBURN-B2, October 1999
> "The CASMO-4 MICROBURN-B2 code system is approved as a replacement used in NRC-a sLiURN-B code system sing methodology and in FRA-ANP BWR core monitoring applications. Such replacements shall be evaluated to ensure that each affected methodology continues to comply with its SER restrictions and/or conditions."
A FRAMATOME ANP CASMO-4/MICROBURN-B2 Neutronic Code System l Topical and SER
Neutron Flux Solution
-> Full two group neutron diffusion equation Advanced Nodal Expansion Method (ANEM) m Accurate flux distributions for highly heterogeneous cores L
(mixture of various mechanical designs, heavy gadolinia loading, long fuel cycle)
Robust convergence for rated conditions, off-rated Conditions, or cold critical c m Fast solutionLce net currents implemented in a multi-level iteraion method)
E Needed for pin power reconstruction
-> Green's function nodal method (semi-analytic nodal method)
Microscopic Cross Section/Depletion Method K,2 > Core history effects Void history (moderator density history) m Fuel temperature history m Controlled depletion history SSpectral histF rF
>*Tracking the effs o these his ory parameters on macroscopic cross sections is a difficult task m The solution is the microscopic cross section/depletion method A
FRAMATOME ANP
Microscopic Cross Section/Depletion All i006 BATT
> Key nuclide depletion chains
"* U-235 -> U-236 -> Np-237 --> Pu-238 -> U-238 -> Np-239 -
Pu-239 -> Pu-240 -- Pu-241 -- Pu-242
"* Pu-241 -> Am-241 -- Pu-242 -- Cm-242 m Am-241 m Cm-242 + RAFT m Gd-154 -> Gd-155 --> Gd-156 -+ Gd-157 -> Gd-158 Gd-eff -> (n,y)
- 1-135--> Xe-135
- Pm-149 -> Sm-149 or A
FRAMATOME ANP Method (Continued)
U
Pin Power Reconstruction (PPR)
> Extensive qualification efforts m Large database of benchmarking of cycles in US, Europe and Taiwan FRA-ANP, GNF and West. (ABB) fuel in database m Quad Cities gamma scan data m Gamma sca ed9orF9X9 and ATRIUM Tm-10 fuel from Gu dr n i h
er ensity BWR in So.
Germany)
IN Colorset comparisons with CASMO-3, CASMO-4, MCNP
"> Currently, it is the only pin power reconstruction (PPR) model simulator in the industry benchmarked against recent, modern fuel gamma scan data for both U02 and MOX fuel bundles
_______A FRAMATOME ANP
'4
I 17 9
> Multiple control blade type modeling in core m Up to 5 different control blade types for different control blad e locations in the core can be handled m Can be an important feature in handling new control blades replacing ornD.
FT TYR7AFT A
FRAMATOME ANP Control Blade Modeling
-7 r Tan It1ý ý Oio Reactor Fuel Loaded Year Comments
- 1. Gundremmingen-B LTAs (8) 1992 Burnup to 71 GWd/MT
- 2. Gundremmingen-B Reloads 1995 Gamma scan data obtained in 1998
- 3. Olkiluoto-1,2 LTAs (8 each) 1996
- 4. BBK-1,OSK-3, FOR-1 LTAs (4 each) 1996
- 5. Philippsburg-1 elo s
- 6. Olkiluoto-1 IcAPT966 LOCAL cold criticals at
- 7. Isar-1 Re o L
1997B
- 8. Forsmark-1 Reloads 1999
- 9. Forsmark-3 Reloads 1999
- 10. Ringhals-1 Reloads 1999 ATRIUM TM-1O in Europe A
FRAMATOME ANP
- I-F
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DRAFT I
A FRAMATOME ANP BWR Cold Critical Predictions
Ir
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9 MICROBURN-B2 Core Follow DRAFT I
A FRAMATOME ANP
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J DRAFT I
A FRAMATOME ANP i-MICROBURN-B2 Cold Critical Predictions
r I
1' DRAFT j
A FRAMATOME ANP I-BWR/6 Cycle 12 Core-ave. TIPs (811 MWd/iMTU)
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40h cr) 0 6-I Za
ar ITW I-BWRI6 Cycle 12 Core-ave. TIP (7,913 MWd/MTU)
DRAFT I
A FRAMATOME ANP i
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DRAFT I
A FRAMATOME ANP BWRI6 Cycle 12 Core-ave. TIP (11,758 MWd/MTU)
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- Power, Ave. Bundle Calculated with Reactor MWt (%
- Power, CA4/MB2, Fuel/Cycle Reactor Size, #FA Uprated)
MWt/FA
- Cycles Licensig**
Uprate Comments KKI-1 (Isar) 592 2704 (5 0) 46 8
X Uprate planned KKP-1 (Philippsburg) 592 2575 (0.0) 44 13 (x)
KKK (Kruemmel) 840 3690 (0.0) 44 17 (x)
OL-1 (TVO) 500 2500 (15 7) 50 10 X
For 3 cycles operation Forsmark-3 700 3300 (9.3) 4 7 4
X Other European 444-1892-2928
<4.4 21 X
676 1_1 Gundremmingen-B,C 784 4100 (6 8) 52 23 (x)
Lic Applied for Cofrentes 624 10 5
Operation start in 2002 Leibstadtl 648 3515 (IF 49X For 1 cycle operation River Bend 624 3085J 6 45x Grand Gulf 1 800 (17rMC)ix LaSalle 1.2 764 3489* (5 0) 4 6 6
In progress Columbia GS 764 3486* (4.9) 4 6 8
In progress Browns Ferry 2/3 764 3952* (20 0) 5 2 In progress Latest power uprates. (BF3 uprates in2004)
- (x)=currently fuel licensing only (Europe)
A FRAMATOME ANP CASMO-4/MICROBURN-B2 Experience 9
Neutronic Codes Applicability to Power Uprate
> FRA-ANP neutronic methodology applicable to handle uprated BWR cores I-:,
- Technically rigorous treatment of phenomena Very well benchmarked (>>100 cycles, gamma scan data for 5*x:*s ATRI UTM -10)
- Experienane powers higher than Browns
> Increased steam flow from power uprate comes from increased power in normally lower power assemblies in the core running at higher power levels. High powered assemblies in uprated cores will be subject to the same LHGR, MAPLHGR, MCPR, and cold shutdown margin limits and restrictions as high powered assemblies in non-uprated cores A
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> Fuel reload licensing analysis are performed to assure that all fuel design safety limits are satisfied for the current operating cycle
> BWR operation is typically limited by the transient MCPR performance of
> Framatome AN correlation limits so the fundamental range of assembly conditions must remain within the same parameter space under uprate conditions A
FRAMATOME ANP tically impose the Power Uprate Considerations
Power Uprate Impact on Core Design I 5*
> To meet critical power limits, the increase in core thermal power L. *is achieved by "flattening" the radial power distribution and raising the average assembly power
- Due to small increases in the Safety Limit MCPR and transient ACPR the pe ass ly pweipated to be lower for the power u ate s*
g0 ntat thesame steady-state margin to op g9
> This trade-off between core design flexibility and operating power level assures that the local assembly conditions remain within the operating experience of Framatome ANP methodology FRAMATOME ANP
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I FRAMATOME ANP Assembly Comparisons at Limiting Step-through Conditions I --
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> Since the bundle operating conditions are within the envelope of hydraulic testing and operating experience, the following statements can be made:
m The hydraulic models used to compute the core flow distribution and local void content remain valid the nodal reactivity and
- The neutronic characteristics computed by the steady-state simulator and used in safety analysis remain valid A
FRAMATOME ANP Core Design Implications on Methodology 0
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> The trade-off between radial peaking and core power does have several minor impacts that need to be considered Increased core power increases the power generated in the bypass
"* "Flatter" power distribution will impact the Safety Limit CPR
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"* Increased power will increase the core pressure drop and impact the steady-state and transient jet pump performance
"* Increased power results in increased steam flow rate, which impacts the steam-line dynamics
"* Increased power results in higher decay heat for LOCA analysis t
A FRAMATOME ANP Impact of Core Design
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> The steady-state simulator directly computes the core bypass conditions m Bypass flow is modeled explicitly in the parallel channel thermal hydraulic model for the core
> Design impact: The lower tie plate flow hole size will be verified to-preclude increased bypass voiding rNo change to the bypass model will be needed for power uprate 0 Bypass enerl function of in i6 determined as a and control fraction 9
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FRAMATOME ANP Power Uprate Impact on the Core Bypass m
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> The SAFLIM2 methodology computes the number of rods in boiling transition based on the licensing power and flow distributions
> The methodology uses the "flattest" radial distribution from the licensing basis s(
> The impact computed
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r-will be directly gn impact: A slight increase in the Safety Limit MCPR is cted No change to the safety limit model will be needed for power uprate 9
A FRAMATOME ANP Power Uprate Impact on Safety Limit CPR J
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> The STAIF code is used to compute the stability characteristics of the core
"* STAIF does not rely on a correlation between channel and global decay ratios to protect the regional mode
"* STAIF computes the regional mode directly using the actual state-point eir&%al paltiolw, kow1act will be directly computed
> STAIF has been brenchmarked against full assembly tests, as well as global and regional reactor data. The impact of the "flatter" core design on stability limits will be directly computed based on the projected operating conditions rNo change to the stability model will be needed for power uprateJ A
FRAMATOME ANP Power Uprate Impact on Stability
-Power Uprate Impact on Jet Pump SCalculations
> The Framatome ANP jet pump model is based on a mechanistic 5 control volume model using 2-dimensional equations of motion
> The model was verified for all six possible flow modes
- 1/6th scale si tedat INEL
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- tTAAjet ump
- Single phase flow test on the full scale Dresden 2 jet pumps
"> "Comparisons with tests resulted in very good agreement in pump pressure drop predictions," XN-NF-80-19 Volume 2 SER The jet pump model will correctly address the impact of power uprate FRAMATOME ANP
PoWer Uprate Impact on Steam-line Dynamics
> The COTRANSA2 transient simulator uses a nodal model utilizing a one-dimensional three-equation gas dynamics formulation for the steam-line and is not restricted to a particular steam flow rate
> The bypass flow capacity is effectively-reduced in terms of percent of stea Bypass capat intolutefw rate in COTRANSA2, so no change is required
> Plant specific modifications to the safety/relief and turbine control valve characteristics are input directly to the model The steam-line model remains valid for uprate conditionsJ A
FRAMATOME ANP
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~Power Uprate Impact on LOCA
> Increase in decay heat due to power uprate is well defined and does not require code modification
> Large-break LOCA results are determined primr-ily-lrom-the hot bundle initial power TBLOCA results are
- Since this is expected to
- No significant difference in the accident progression or range of conditions
> Small-break LOCA may be influenced by the higher decay heat rLOCA models remain valid for power uprate )
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"> Maintaining margin to fuel design safety limits imposes restrictions on the range of operating conditions an assembly may experience during transients
"> Increasing the core thermal power is accommodated by radial power flattening so that individual assembly conditions deviate only slightly fro f
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tir xp rience The -FramatomD FP iice inghthod directly assess the identified impacts of power uprate on operating limits without modification A
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