ML17139D101
| ML17139D101 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 08/12/1985 |
| From: | Campagnone M Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8508200451 | |
| Download: ML17139D101 (41) | |
Text
Docket Nos.:
5O>>387/388 AUG 1 2 1985 APPLICANT:
Susquehanna Steam Electric Station, Units 1 and 2 (SSES)
FACILITY:
Pennsylvania Power and Light Company (PP8L)
SUBJECT:
SUMMARY
OF MEETING ON PPSL'S INTERNAL DEVELOPMENT OF TRANSIENT ANALYSIS AND CORE PHYSICS CODES On August 8, 1985, representatives of PP8L met with the NRC staff in Bethesda, Maryland to present an overview of PP8L's internal capabilities of performing portions of future core reload analysis.
PP8L made a presentation to the staff which covered topics such as PP8L's model validation approaches, critical power ratio methodology, and the NRC submittal schedule.
Enclosure I contains a list of Attendees and Enclosure 2 is a copy of'he meeting handouts.
Enclosures:
As stated Mari-Josette Campagnone, Project Manager Licensing Branch No.
2 Division of Licensing D
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MEETING
SUMMARY
DISTRIBUTION
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Applicant 8 Service List
Docket Nos.:
50-387/388.
AUG 1 2 1985 APPLICANT:
Susquehanna Steam Electric Station, Units I and 2
(SSES)
FACILITY.:
Pennsylvania Power and Light Company (PP&L)
SUBJECT:
SUMMARY
OF MEETING ON PPSL'S INTERNAL DEVELOPMENT OF TRANSIENT ANALYSIS AND CORE PHYSICS CODES On August 8, 1985, representatives of PPSL met with the NRC staff in Bethesda, Maryland to present an overview of PP8L's internal capabilities of performing portions of future core reload analysis.
PPSL made a presentation to the staff which covered topics such as PPSL's model validation approaches, critical power ratio methodology, and the NRC submittal schedule.
Enclosure I contains a list of Attendees and Enclosure 2 is a copy of the meeting handouts.
Enclosures:
As stated Mari-Josette Campagnone, Project Manager Licensing Branch No.
2 Division of Licensing D
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none:pob WButler 8/)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 AUG 1 2 1985 Docket Nos.:
50-387/388 APPLICANT:
Susquehanna Steam Electric Station, Units I and 2
(SSES)
FACILITY:
Pennsylvania Power and Light Company (PP8L)
SUBJECT:
SUMMARY
OF MEETING ON PPSL'S INTERNAL DEVELOPMENT OF TRANSIENT ANALYSIS AND CORE PHYSICS CODES On August 8, 1985, representative's of PPSL met with the NRC staff in Bethesda, Maryland to present an overview of PP8L's internal capabilities of performing portions of future core reload analysis.
PPSL made a presentation to the staff which covered topics such as PP8L's model validation approaches, critical power ratio methodology, and the NRC submittal schedule.
Enclosure I contains a list of Attendees and Enclosure 2 is a copy of the meeting handouts.
Enclosures:
As stated Mari-Josette Campagnone, Project Manager Licensing Branch No.
2 Division of Licensing
Ilr. I!os.an
'I. Curtis Pennsylvania Vower a Light Company Susquehanna Stear., Electric Station Units 1
Im 2
cc:
Jay Silberc, Esq.
Shaw, Pittman, Potts, 5 Trowbriage 1800 M Street, N.
W.
llashington, D.C.
20036 Edward M. Nagel, Esq.
General Counsel and Secretary Pennsylvania Power S Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. William E. Barberich tlanager-Nuclear Licensing Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 I'r. R. Jacobs Resiaent inspector P.U.
Box 52 Shickshinny, Pennsylvania 18655 Mr. R. J.
Benich Services Project Ilanager General Electric Company 1000 First Avenue I'ing of Pressia, Pennsylvania 19406
~ Vir. Thomas I'i. Gerusky, Director bureau of Radiation Protection Resources Commonwealth of Pennsylvania P. 0.
BOx 20b3 Harrisburg, Pennsylvania 17120 Robert R. Alder, Esquire Office of Attorney General P.O.
Box 2357 Harrisburg, Pennsyl vani a 17120 Mr. William tlatson Allegheny Elec. Coorperative, Inc.
212 Locust Street P. 0.
Box 1266 Harrisburg, Pennsylvania 17108-1266 Ilr. Anthony J.
Pi etrofitta, General Manager Power Production Engineering and Construcxiun Atlantic Electric 1199 Black Horse Pike Pleasantville, NJ 08232 Regional Administrator, Region i U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406
Enclosure 1
Attendees Name H. Campagnone W. Hodges J.
G. Refling C.
R.
Lehmann S. B. Sun Richard Lobel Walter L. Brooks John H. Emmett Jerome Stefanko Jack Fiskew Andrew Dipzel Daniel Fieno John t1. Kulick Tony Roscioli Kenneth Knoll R. Sgarro Or anization NRC NRC PP&L PP&L NRC NRC NRC PP&L PP&L NAI PP&L NRC PP&L PP&L PP&L PP&L
En c.)OSur< 4 MEETING OBJEGTIVE To present PPKL's in-house reload analysis methods, our model validation approaches, our Critical Power Ratio Methodology, and the NRC submittal schedule.
AGENDA I
I.
INTRODUCTION (J. S. Stefanko) o Organization Structure o
Ultimate Objective/Overall Plan Schedule o
Corporate Committment II. CORE PHYSICS A. Experience B. NFKSE Procedural Controls C. Methods (J. M. Kulick)
(K. C. Knoll) o Code Sequence, Preliminary Results o
Safety Limit MCPR D. Topical Report o Benchmarking Activities o Applications of Licensing Basis Methods E. Generation of Transient Analysis Data
AGENDA CONT'D III. TRANSIENT ANALYSIS A. Experience B. Methods o System Model o Hot Bundle Model o Gap Conductance o Transient 6 CPR C. Topical Report
. o Benchmarking o RETRAN Code Uncertainty o Sample Reload Analysis IV. SCHEDULE (J. G. Refling}
(A. J. Roscioli}
(C. R. Lehmann)
(A. J. Roscioli)
.(J. S. Stefanko)
E H
V.P. En ineering 6 Construction-Nuclear N. W. Curtis Mana er-Nuclear Licensing W. E. Barberich Mana er-Nuclear Plannin 6 Controls Manager-Nuclear Plant Engineerin PROJECT MANAGEMENT R. A. Mazzini A. M. Male Assist. Pro'ect Director-'SES R. H. Featenby Mana er-Nuclear Desi n Mana er-Nuclear Fuels 6
S stems Engineerin Mana er-Engineering 0 erat A. W. Metzger J.
S. Stefanko I. E. Keister
CA's 482, 483, 484, 485 Manager-Nuclear Fuels 6
S stems Engineering Jerome S. Stefanko Consulting Engineer Paul R. Hill CA-482 CA-483 CA-484 CA-485 Su vervisor-Core M 't John M. Kulick Su exvisor-En ineerin Anal.
Jack G. Refling Su ervisor-Econ.
6 Contracts Robert J.
McKeon Su ervisor-S stems Engineering Michael B. Detamore
PP&L CORE MANAGEMENTGROUP o
General Description of Susquehanna SES Plant o
Core Management Group Organization o
Experience o
NFKSE Procedural Controls o Analytical Methods o
Benchmarking Activities/Topical Report o
Generation of Transient Analysis Data
GEN RAL DESCRIPTI N OF SUSQUEHANNA SES PLANT O erator:
Pennsylvania Power 5 Light Co.
Location of Plant:
Near Berwick, Pa.
o Two BWR-4 Reactors (60 KW/L)
Unit 1: Initial Criticality - 9/10/82 Commercial - 6/8/83 EOC-1
- 2/9/86
. BOC-2
- 6/8/86 Unit 2: Initial Criticality - 5/8/84 Commercial - 2/12/86 o
Reactor Description o 3293 MWth (1050 MWe) o 251lnch Vessel o 2 Loop Jet Pump Recirculation o 764 Fuel Assemblies (8X8) o Active Fuel Height = 150 inches
CORE MANAGEMENTGROUP Supervisor - Core Management John M. Kulick RELOAD APPLICATIONS NUCLEAR FUEL MANAGEMENT PRO ECT ENGINEERS'ohn H. Emrnett Robert M. Rose OPERATIONS SUPPORT
'UCLEAR FUEL MANAGEMENT PROJECT ENGINEERS John P. Spadaro
+
William J. Weadon Eric R. Jebsen METHODS DEVELOPMEN l NUCLEAR FUEL MANAGEMENT PROJE T ENGINEERS Kenneth C. Knoll ANALYST-ENGINEERING SUPPORT Denise S. Showalter
" GROUP LEADER
PP&L GORE MANAGEMENTGROUP EXPERIENCE o
Plant Operations Support o
Core Follow o
Target Control Rod Patterns o
Estimated Criticals o
Pre-maneuver Predictions o
Fuel/Core Design o Vendor Verifications, Alternative Designs o
Core Monitoring Systems o Off-Line PCS Code o
Implementation of POWERPLEX CMS o
Other Applications
Core Average Tip Comparison at 5.918 GWD/MT 180 160 140 120 x h X X
~
~
CL 100 80 0
60 X
0 X X X
X 40 20 Legend
~
NODE-B+
X TIP. Measurement 0
I I
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 2223 24 Core Axial Node'
Form 208.1 4
Bob Boesch
, Susq.
SES RE: TARGET CONTROL ROD PATTERN F'pr Full Power A2 Susquehanna SES Unit ~, Cycle 61 59 57 55 53 51 47 45 43 41 39 37 35 33'1 29 MCPR Control Rod Sequence:
A1 A2 B1 B2.
0 02 04 06 08 10 12 14 16 18 20 22 24 26 28 30 32 42 42 Reactor Operating Conditions:
7.+%
Core Avg. Exp. =
7.464'WD/ST Core Pwr.
I evel = 8988 MW 99.9',
Total Core Flow'=
95 6 Mlb/hr Reactor Pressure =
991.2 psia Ak+
Inlet Subcooling =
24-'71 BTU/Ibm Corresponding Equilibrium Xenon
- assumed, unless specifically noted here:
'42 36 14 36 MFLCPR LOC MLHGR MFLPD LOC MAPLHGR MAPRAT LOC 1.542 I4 1 ~ 546 1'59
~ 824
~ 821
~ 815 13-46 45-48 13-48 10'7 10'5
~ 811
~ 802 11.02I
~
823 13-48-6 47-46-6
/~48 9.77
~ 806 13-48-6
.804
~47-46-'otes/Comments:
- 1) Due to modeling biases and differing operatlhg conditions, some adjustment to the target pattern may be required Prepared by:
J.
P.
sgadaro
'8/10/84
NF&SE PROCEDURES Nuclear De artment Instructions:
o NDI-QA-7.2.2, "Design Activities Related to Fuel/Core Performance" o
NDI-QA-7.2.1, "In-Core Fuel Management In Support of SSES Operations" En ineerin Procedures:
o DC 120.0, "Computer Code Documentation 5 Control" o
DC 220.0, "Documentation L Verification of Calculations"
CORE PHYSICS METHODS CODE SEQUENCE MICBURN CPM2 NORGE-B2 COP HIN TO POWERPLEX CMS SIMULATE-E (FIBWR)
PDQ o CORE DEPLETION e POWER DISTRIBUTIONS o ASSEMBLY FLOWS o CORE K-eff o CPR'(XN-3) o KW/FT TIPPLOT SAS STATISTICAL ANALYSIS RESULTS
CORE AVERAGE TIP COMPARISON AT 5.918 GWD/MT 180 160 140 120 V)
I 100 LIJ O
80 M
CL I
60 xx
~ g x x x
40 20 Legend
~
SIMULATE X
TIP Measurement 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21222324 Core Axial Node
BENCHMARKINGACTIVITIES o
TIP, Cold Critical, Pressure Drop Comparisons o
S1/C1 o
S1/C2 o
S2/C1 o
S1/C3 (for available data) h o K-eff vs Exposure o
Peach Bottom Comparisons
.o PDQ vs CPA/I Comparisons
TOPICAL REPORT CONTENT o
Introduction o
Description of PPKL Analytical Methods o
Methods Validation 5, Uncertainty Determination o Methods Applications (sample analyses) o Appendix: Statistical Analysis Methods
APPLICATIONS OF CORE PHYSICS LICENSING BASIS METHODS o
Plant Operations Support o
Reload Safety Evaluations o
Generation of Transient Analysis Neutronic Data Q
Analysis of RWE, LOFWH, Fuel Loading Error o.Support of. Other Licensing Actions
GENERATION OF TRANSIENT ANALYSIS DATA o 'Kinetics Parameters
{P,V) o 2-Group Macroscopic Cross Sections o
Core Power Distributions o
Local Peaking Factors o
Control Rod Worths o
Core Bypass Flow Fractions
ENQINEERINQ ANALYSIS ORQANI2ATION REACTOR SYSTEMS TRANSIENT ANALYSIS J. G. Refling Supervisor-Engineering Analysis POWER & AUXILIARYSYSTEMS TRANSIENT ANALYSIS FUEL METHODS DEV &
PERFORMANCE ANALYSIS 4
A. J. Roscioli" S. A. Somma" C. R. Lehmann "
F. E. Grim M. A. Chaiko W. A. OeLorme L. M. Olson A. L Reid D. A. Matchick S. Seyedhosseini T. S. Yih Co-Op Co-Op
" GROUP LEADER
PP&L ENGINEERING ANALYSIS GROUP EXPERIENCE o
Plant Startup Test Support o
Unit 'I Pre-Test Predictions o
Unit 1 Post-Test Analysis o
Unit 2 Pre-Test Predictions o
Engineering Test Support o
FSAR Transient Analysis Comparison o
Plant Operations Support o
T-10 (Loss Of Single Transformer} Event o
Moisture Separator Analysis o
Consulting o
Licensing Support o
Feedwater Line Break Analysis o
Emergency Operating Procedure Review o
Consulting
L
~
a CL 20 X Plant RETRAN 12 aD
~
~.9 0
lL 6
CO I
X Plant RETRAN XX XX
'R 0
1000 lL E
CO 5
500 0
0 Time (s) 0 0
Time (s)
C a
a 8
0 IaR 50 40 30 20 10 0
X Plant RETRAN 125 E0 100 75 e
50 25 8
Ia 0
1100 1050
'R 1000 I
0 950 N
E 900 XXX)<
X Plant RETRAN E
O E00 E
-10 0
20 Time (s) 850 0
20 Time (s) lL 3
C0 a
8 X Plant RETRAN 8000 'R IL 4000 a 8
C0
~l 2000 I CC 50 X
m u-30 Cl 3
20 C
.9 t) 10 X Plant RETRAN ca 0
4000 u.
Cla 3
C0 2000 9 8
K 0
0 20 Time (s) 0 0
Time (s)
TRANSIENT ANALYSIS CODE STRUCTURE SIMULATE 3-D SIMTRAN 1-D CROSS SECTIONS FUEL & REACTOR SYSTEM DATA RETRAN (SYSTEM MODEL)
UPPER/LOWER PLENUM T-H CONDITIONS NORMALIZEDPOWER RETRAN (HOT BUNDLE)
NODAL FLOWS & HEAT FLUXES SYSTEM PRESSURE CPR CODE TRANSIENT b, CPR
RETRAN SYSTEM MODEL FEATURES o
Nodalization o
Trip Logic o
Controllers o
1-D Kinetics for Pressurization Transients o
Point Kinetics for Non-Pressurization Transients o
Bubble Rise Model for Separators o
Non-Equilibrium Volume in Upper Downcomer o
Direct Heating of Bypass
110 II0 121 QII0 8
QI>>
'IOo y
)
hieThl cToh AC I IICIC
~TCAA ATAAh evrtLT ~V Q>>I
~00 m0>>00>>
Q rlc OTHO tItse OEio IIIOOLC ONACOtCII JCT tired h
Ii0 ISO Ooo te IIVI AC-I2 rC \\0 IO AC-S e
e Ac-e
'o 0
Ac-T ss~
iC 2 2
AC I OUTI Ie2
'8 ISI JET tIAit O22I tieir 01 SCHAAOC tltII40
%0 Oa STO T
Q~
Q~
Q~
VILLE N.K 10 Io lel ISO ISI 211 211 IRDIcIJma pQp ae anna rlraa LIXED h ACCIACULATIatat AAO eoCTION tltIAA LAOt ~
SSES RETRRN MODEL
RETRAN HOT BUNDLE MODEL o
Bypass Channel 5 Bundle Channel o
Leakage Paths o
Finger Springs o
LTP Holes o
Direct Moderator Heating o
Benchmarked vs Exxon/GE Data 5 Calc's 4!
~
o Driven By NSSS Model o
Core Power o
Upper 5 Lower Plenum Pressures o Output to MCPR Code o
Nodal Flow o
Nodal Heat Flux
RETRAN HOT BUNDLE MODEL UPPER PLENUM CHIMNEY (UTP Grid) 12.5'CTIVE FUEL REGION VOLUMES UPPER BYPASS 13.694'104' (LTP Grid)
(Finger Springs)
MIDDLE BYPASS
.687'5 I
(CSP)
LOWER BYPASS 367)
(ORIFICE)
LOWER PLENUM
CPR METHODS o XN-3 o
Benchmarked vs Exxon Method
Statistical Analysis (Limiting Pressurization Events) o
Response
Surface Methods o
Convolute Errors o
Fitting Error o
Code Uncertainty o
Scram Time o
Scram Delay o
Initial Steam Flow o
Select 95/95 d CPR o
Deterministic Analysis (Non-Limiting Events}
GAP CONDUCTANCE METHODS o
ESCORE o
Hot Bundle o
Maximize Hgap / Maximum Cycle Value o Axial Average o
Core Average o Typical Rod Power History o
Minimize'Hgap o 3-D SIMULATE - BU 5 Power Distribution o
Nodal Average: Function of BU
4 0
~
4 t
I BENCHMARKING o
Susquehanna Startup Tests and Plant Data o
Pressure Regulator Setpoint Changes o
Feedwater System Level Setpoint Changes o
Loss of Feedwater Heater o Two Recirculation Pump Trip o
One Recirculation M/G Set Trip o
Pressure Regulator Failure" (CLOSED) at Natural Circulation o
End-of-Cycle Generator Load Rejection" o
Peach Bottom Turbine Trip Tests 0
TT1%
o TT2".
o TT3" 1-D Kinetics Benchmarks
'ETRAN GODE UNGERTAINTY o
Point Kinetics o
Bounding Input Values for Safety Analyses o
Conservative Void, Doppler. and Scram Reactivity Multipliers o
1-D Kinetics o Uncertainty Based on Benchmark Calculations o 6 CPR Calculated Based on Plant Transient Data o 6 CPR Calculated Using Analysis Methodology o
5 Benchmark Cases Used to Calculate 6 CPR/ICPR Uncertainty o Assuming a Normal Distribution, the 95~/o Confidence Level is Estimated Using a Chi-Squared Distribution [ ie, o = S x F (n,n) ]
o RETRAN Code Uncertainty is Input to Monte Carlo h, CPR
SAMPLE RELOAD ANALYSIS o
Susquehanna 1 Cycle 2 Input Parameters o
Transients Presented in Topical o
Generator Load Rejection W/0 Bypass o
Feedwater Controller Failure o
MSIV Closure with High Flux Scram (Overpressure Analysis) o Sensitivity Studies o 6 CPR Calculation
OVERALL SCHEDULE o
Submit Topical Reports - Third Quarter 1986 o
Core Physics o
Transient Analysis Includes - Model Description
- Results of Benchmarking
- Code Uncertainty &.
Statistical Analysis
- Sample Reload Analysis
- Fuel Performance o
Request NRC Review Complete By June 1987 o
To Support FIRST PPKL Submittal
- January 1988
0-II I
R 4
N