ML16336A559

Donald C. Cook Nuclear Plant, Units 1 & 2, Revision 27 to Updated Final Safety Analysis Report, Chapter 3, Table 3.1-2

ML16336A559
Person / Time
Site:	Cook
Issue date:	10/24/2016
From:	Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16336A263	List: AEP-NRC-2016-42, Fire Protection Program Change Summary ML16336A225 ML16336A228 ML16336A238 ML16336A291 ML16336A292 ML16336A294 ML16336A295 ML16336A297 ML16336A298 ML16336A299 ML16336A300 ML16336A301 ML16336A302 ML16336A303 ML16336A304 ML16336A305 ML16336A306 ML16336A308 ML16336A309 ML16336A310 ML16336A311 ML16336A312 ML16336A313 ML16336A314 ML16336A315 ML16336A316 ML16336A317 ML16336A318 ML16336A319 ML16336A320 ML16336A322 ML16336A323 ML16336A324 ML16336A325 ML16336A327 ML16336A328 ML16336A329 ML16336A330 ML16336A332 ML16336A333 ML16336A334 ML16336A336 ML16336A337 ML16336A338 ML16336A339 ML16336A340 ML16336A341 ML16336A342 ML16336A345 ... further results
References
AEP-NRC-2016-42
Download: ML16336A559 (2)

Text

I I N N D D I I A A N N A A A A N N D D M M I I C C H H I I G G A A N N P P O O W W E E R R D D.. C C.. C C O O O O K K N N U U C C L L E E A A R R P P L L A A N N T T U U P P D D A A T T E E D D F F I I N N A A L L S S A A F F E E T T Y Y A A N N A A L L Y Y S S I I S S R R E E P P O O R R T TRevision:16.4Table:3.1-2 Page:1 of 2Unit 2ANALYTIC TECHNIQUES IN CORE DESIGNAnalysis TechniqueComputer Code Section ReferencedMechanical Design of Core InternalsLoads, Deflections, and Stress AnalysisStatic and Dynamic ModelingBlowdown code, FORCE, Finiteelement structural analysis code, andothers 14.3.3Fuel Rod DesignFuel Performance Characteristics (temperature, internalpressure, clad stress, etc.)Semi-empirical thermal model of fuelrod with consideration of fuel densitychanges, heat transfer, fission gasrelease, etc.Westinghouse fuel rod design model 3.2.1.3.1 3.3.3.1 3.4.2.2 3.4.3.4.2Nuclear Design1. Cross Sections and Group ConstantsMicroscopic dataMacroscopic constants for homogenizedcore regionsModified ENDF/B-V or ENDF/B-VI libraryPHOENIX-P 3.3.3.2 3.3.3.2Group constants for control rods withself-shieldingPHOENIX-P3.3.3.2Nuclear Design (Continued)2. X-Y Power Distributions, Fuel Depletion, CriticalBoron Concentrations, X-Y Xenon Distributions, 3D, 2-Group Nodal Expansion MethodANC 3.3.3.3 I I N N D D I I A A N N A A A A N N D D M M I I C C H H I I G G A A N N P P O O W W E E R R D D.. C C.. C C O O O O K K N N U U C C L L E E A A R R P P L L A A N N T T U U P P D D A A T T E E D D F F I I N N A A L L S S A A F F E E T T Y Y A A N N A A L L Y Y S S I I S S R R E E P P O O R R T TRevision:16.4Table:3.1-2 Page:2 of 2Unit 2ANALYTIC TECHNIQUES IN CORE DESIGNAnalysis TechniqueComputer Code Section ReferencedReactivity Coefficients3. Axial Power Distributions, Control Rod Worths,and Axial Xenon Distribution1-D, 2-Group Diffusion TheoryAPOLLO 3.3.3.34. Fuel Rod PowerIntegral Transport TheoryLASER 3.3.3.1Effective Resonance TemperatureMonte Carlo Weighting FunctionREPADThermal-Hydraulic Design1. Steady-StateSubchannel analysis of local fluidconditions in rod bundles, includinginertial and crossflow resistance terms,solution - progresses from core-wide tohot assembly to hot channelTHINC-IV 3.4.3.4.12. Transient Departure from Nucleate BoilingAnalysisSubchannel analysis of local fluidconditions in rod bundles duringtransients by including accumulationterms in conservation equations solutionprogresses from core-wide to hotassembly to hot channelTHINC-I (THINC-III) 3.4.3.4.1