ML17279A168
| ML17279A168 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/31/1987 |
| From: | Busselman G, Krajicek J, Ward G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17279A161 | List: |
| References | |
| XN-NF-87-25, NUDOCS 8704030092 | |
| Download: ML17279A168 (35) | |
Text
8704030092 S70327 PDR ADOCK 05000397
- . P P R ADNWCEDNUCL.EAR FUELS CORPORATION XN-NF-87-25
, Issue Date: 3/2//87 WNP-2 CYCLE 3 RELOAD ANALYSIS Prepared By: 4R Z3'VS7 E. ajicek, Sr. Engineer Date BWR Sa ety Anlaysis Concur: 8/2V a G. N. War , Manager Date Reload Licensing Approve:
G. J. Busselman, Manager Date Fuel Design Approve s/Pif-
. J. Fegerico, Man ger i/ 4( /3 Date g
V Neutronics and Fue Management Approve: P(~~< ~/g,g/t';
E. W iamson, Manager Date Licensing and Safety Engineering Approve:
G. L. Ritter, Manager Date Fuel Engineering and Technical Services Concur:
N. Morgan, Manager Date Customer Services Engineering llh AIIAFFILIATEOF KRAFTWERK VHIOII Qw Kwv
NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Advanced Nuclear Fuels Corporation. It is being submitted by Ad.
vanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Advanced Nuclear Fuels Corporation-fabricated reload fuel or other technical services provided by Ad-vanced Nuclear Fuels Corporation for light water power reactors and it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, informa-tion, and belief. The information contained herein may be used by the U.S.
Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Cor-poration in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.
Advanced Nuclear Fuels Corporation's warranties and representations concern.
ing the subject matter of this document are those set forth in the agreement bet-ween Advanced Nuclear Fuels Corporation and the customer to which this docu-ment is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:
A. Makes any warranty, or representation, express or im-plied, with respect to the accuracy, completeness. or use-fulness of the information contained in this document, or that the use of any information, apparatus, method, or pro-cess 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, ap-paratus, method. or process disclosed in this document.
XN NF-F00.766 (1/87)
ZN-NF-87-25 TABLE OF CONTENTS Section ~Pa e
1.0 INTRODUCTION
2.0 FUEL MECHANICAL DESIGN ANALYSIS
- 3. 0 THERMAL HYDRAULIC DESIGN ANALYSIS ..
3.1 Design Criteria.
3.1.3 Fuel Centerline Temperature.
3.2 Hydraulic Characterization.
3.2.5 Bypass Flow 3.3 MCPR Fuel Cladding Integrity Safety Limit 3.3.1 Coolant Thermodynamic Condition.
3.3.2 Design Basis Radial Power Distribution. 3 3.3.3 Design Basis Local Power Distribution..
4.0 NUCLEAR DESIGN ANALYSIS..
4.1 Fuel Bundle Nuclear Design Analysis.
4.2 Core Nuclear Design Analysis.
4.2.1 Core Configuration..
4.2.2 Core Reactivity Characteristics......
5.0 ANTICIPATED OPERATIONAL OCCURRENCES 5.1 Analysis Of Plant Transients At Increased Core Flow Conditions.. '.6 5.2 Analyses For Reduced Flow Operation.
5.4 ASME Overpressurization Analysis.
5.5 Control Rod Withdrawal Error Fuel Loading Error 5.7 Determination Of Thermal Margins.
6.0 POSTULATED ACCIDENTS............
6.1 Loss-Of-Coolant Accident 6.1.1 Break Location Spectrum.
6.1.2 Break Size Spectrum 6.1.3 MAPLHGR Analyses 6.2 Control Rod Drop Accident
ii- XN-NF-87-25 TABLE OF CONTENTS (Continued)
Section ~Pa e 7.0 TECHNICAL SPECIFICATIONS 12 7.1 Limiting Safety System Settings 12 7.1.1 MCPR Fuel Cladding Integrity Safety Limit 12 7.1.2 Steam Dome Pressure Safety Limit 12 7.2 Limiting Conditions For Operation 12 7.2.1 Average Planar Linear Heat Generation Rate Limits For ANF 8x8 Fuel 12 7.2.2 Minimum I
Critical Power Ratio. 12
, 7.2.3 Surveillance Requirements. 13 7.2.3.1 Scram Insertion Time Surveillance 13 7.2.3.2 Stability Surveillance..... 14 7.2.3.3 Technical Specification LHGR Surveillance.. 14 9.0 ADDITIONAL REFERENCES..... 24 APPENDIX A. A-1
XN-NF-87-25 LIST OF TABLES Table ~Pa e 4.1 Neutronic Design Values. 15 LIST OF FIGURES
~Fi uue ~Pa e 3.1 Radial Power Histogram For 1/4 Core Safety Limit Model....... 17 3.2 WNP-2 Cycle 3 Safety Limit Local Peaking Factors (ANF Fuel).. 18 4.1 WNP-2 Cycle 3 Enriched Zone Enrichment Distribution. 19 4.2 WNP-2 Cycle 3 Reference Loading Pattern By Fuel Type (One Quarter Of Symmetrical Core Loading) 20 5.1 WNP-2 Cycle 3 Control Rod Wi.thdrawal Analysis Initial Control Rod Pattern. 21 5.2 Reduced Flow MCPR Operating Limit 22 Linear Heat Generation Rate (LHGR) Limit.Versus Average Planar Exposure, ANF 8x8 Fuel 23
XN-NF-87-25
1.0 INTRODUCTION
This report summarizes the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 3 reload for the Supply System Nuclear Project Number 2 (WNP-2). WNP-2 is scheduled to commence Cycle 3 operation in June 1987. This report is intended to be used Volume 4, Rev. 1, "Application of the ENC Methodology 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 single loop operation.
The WNP-2 Cycle 3 core will comprise a total of 764 fuel assemblies, including 148 ANF 8x8 unirradiated assemblies, 128 once irradiated ANF 8x8 assemblies, and 488 twice irradiated P8x8R assemblies fabricated by General Electric (GE).
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 WNP-2 during the previous operating cycle which encompass core flow up to 106% of the design basis value.
XN-NF-87-25 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: Reference 9.8 The expected power history for the fuel to be irradiated during Cycle 3 of VNP-2 is bounded by the assumed power history in the fuel mechanical design analyses.
XN-NF-87-25 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.1 Desi n Criteria 3.1.3 Fuel Centerline Tem erature
. The LHGR curve in Figure 3.4 of Reference 9.8 shows that the ANF 8x8 fuel centerline temperature is protected for 120% over power. The LHGR curve in Reference 9.8 is greater than 120% above the LHGR limit curve in Reference 9.1. Therefore, fuel centerline melt is protected for all ANF 8x8 exposures within the bounds of the referenced LHGR curves.
'3.2 H draulic Characterization Calculated Bypass Flow Fraction 11.6%
MCPR Fuel Claddin Inte rit Safet Limit 3.3.1 Coolant Thermod amic Condition Core Power 3844 HWt Core Inlet Enthalpy 526.4 BTU/ibm Steam Dome Pressure 1030 psia Feedwater Temperature 420'F 3.3.2 Desi n Basis Radial Power Distribution See Figure 3.1
XN-NF-87-25 3.3.3 Desi n Basis Local Power Distribuito See Figure 3.2
XN-NF-87-25 4.0 NUCLEAR DESIGN ANALYSIS 4.] Fuel Bundle Nuclear Desi n Anal sis Assembly Average Enrichment 2.72 w/o U-235 Radial Enrichment Distribution Figure 4. 1 Axial Enrichment Distribution Uniform 2.89 w/o U-235 with 6-inch top and bottom natural uranium blankets Burnable Poisons Figure 4.1 Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 4.2 Core Nuclear Desi n Anal sis 4.2.1 Core Confi uration Figure 4.2 Core Exposure at EOC2 (MWD/MTM) 12,153 Core Exposure at BOC2 (MWD/MTM) 9,639 Core Exposure at EOFP3 (MWD/MTM) 15,103 4.2.2 Core Reactivit Characteristics BOC Cold K-effective, All Rods Out 1.1257 BOC Cold K-effective, Strongest Rod Out 0.9882 Reactivity Defect/R-Value, 0 delta k/k 0.0 Standby Liquid Control System (SBLC) 0.9722 Reactivity, 660 PPM Boron, K-effective
I XN-NF-87-25 5.0 NTICIPATED OPERATIONAL OCCURRENCES Applicable Transient Analysis Report Reference 9.3 5.1 Anal sis Of Plant Transi'ents At Increased Core Flow Conditions Reference 9.3 Limiting Transient(s): Load Rejection Without Bypass (LRWB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LOFH)
Transient analyses for WNP-2 Cycle 2 anticipated operational events showed that delta CPR values at design basis conditions are bounded by delta CPR values at design basis power (104%) and increased core flow conditions (106%). Thus Cycle 3 analyses results at increased core flow conditions are conservatively applicable to rated flow conditions.
Analyses of transient events were performed with the recirculation pump (RPT) in service and out of service, with normal scram speed (NSS), technical specification scram speed (TSSS), and at exposures of end-of-cycle and at end-of-cycle -2000 MWD/HTU (4150 lSD/HTU) as shown in following table.
The loss of feedwater heating event was analyzed on a generic basis and the delta CPR results are bounding values.
XN-NF-87-25 Delta CPR
% Power/ Maximum Maximum Maximum GE ANF Transient* Flow Heat Flux Power Pressure Fuel Fuel LRWB, NSS 104/106 115% 295% 1165 psig 0.25 0.23 RPT Operable LRWB, NSS 104/106 121% 390% 1175 psig 0.31 0.28 RPT Inoperable LRWB, TSSS RPT Operable 104/106 '21% 370% 1170 psig 0.33 0.29 LRWB, TSSS 104/106 127% 440% 1183 psig 0.37 0.33 RPT Inoperable LRWB, TSSS 104/106 112% 304% 1167 psig O.ll 0.12 RPT Inoperable end-of-cycle minus 2000 MWD/MTU FWCF, NSS 47/106 54% 156% 1015 psig 0.26 0.24 RPT, Operable FWCF, NSS 47/106 57% 205% 1020 psig 0.31 0.29 RPT Inoperable FWCF, TSSS 47/106 55% 172% 1020 psig 0.30 0.27 RPT Operable LOFH N/A N/A N/A N/A 0.09 0.09 5.2 Ana ses For Reduced Flow 0 eratio Reference 9.3 Limiting Transient: Recirculation Flow Increase ASME Over ressurization Anal sis Reference 9.3 Limiting Event MSIV Closure Worst Single Failure MSIV Position Scram Trip
XN-NF-87-25 Maximum Pressure 1313 psig Maximum Steam Dome Pressure 1285 psig 5.5 Control Rod Withdrawal Error Initial Control Rod Pattern for CRWE Analysis Figure 5.1 Rod Block ANF GE Monitor Settin Distance Withdrawn Delta-CPR Delta-CPR (ft) 106%* 4.5 0. 20 0.23 107% 4.5 0.20 0.23 108% 5.0 0.22 0.25 5.6 Fuel Loadin Error Delta CPR 0.13 5.7 Determination Of Thermal Mar ins Summary of Thermal Margin Requirements All system transient results at the more limiting increased flow conditions (106%). LRWB results for the more limiting power (design basis condition - 104%) for this transient.
FWCF results for the more limiting power (minimum allowable - 47%)
condition for this transient.
- Rod Block Monitor Setting (RBM) of 106% for Cycle 3.
XN-NF-87-25 Delta CPR MCPR Limit Equipment GE ANF GE ANF Event 0 erational Status Fuel ~Fue ~Fue Fuel Model LRWB RPT Operable, NSS 0.25 0.23 1.31 1.29 COTRANSA/XCOBRA-T LRWB RPT Inoperable, 0.31 0.28 1.37 1.34 NSS LRWB RPT Operable, TSSS 0.33 0.29 1.39 1.35 LRWB RPT Inoperable, 0.37 0.33 1.43 1.39 TSSS LRWB RPT Inoperable, 0.11 0.12 1.17 1.18 TSSS, EOC -2000 MWD/MTU FWCF RPT Operable, NSS 0.26 0.24 1.32 1.30 FWCF RPT Inoperable, 0.31 0.29 1.37 1.35 NSS FWCF RPT Operable, TSSS 0.30 0.27 1.36 1.34 LOFH N/A 0.09 0.09 1.15 1;15 XTGBWR MCPR Operating Limits At Rated Condition For Cycle Exposures Less Than EOC -2000 MWD/MTU (100 To 106% Flow)
~Fuel T e MCPR Limit 107 RBS ANF 1.26 GE 1.29 MCPR Operating Limits At Rated Condition From EOC -2000 MWD/MTU To EOC (100 To 106$ Flow)
~Fuel T
10 XN-NF-87-25 MCPR Limits at Off-Rated Conditions Figure 5.2 Reduced Flow MCPR Limit Reference 9.3
XN-NF-87-25 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident 6.1.1 Break Location S ectrum Reference 9.4 6.1.2 Break Size S ectrum Reference 9.4 6.1.3 MAPLHGR Anal ses (ANF Fuel) Reference 9.5 Limiting Break: Split Break in the Recirculation Suction Piping With an Area Equal to Sixty Percent of the Double-Ended Cross-Sectional Pipe Area Bundle Average Exposure MAPLHGR Peak Clad Peak Local MWD MTM ~kw ft Tem erature 'F NWR 0 13.0 1765 0.49 5,000 13.0 1766 0.48 10,000 13.0 1765 0.47 15,000 13.0 1772 0.47 20,000 13.0 1788 0.54 25,000 11.3 1699 0.34 30,000 9.4 1521 0.17 35,000 7.9 1397 0.10 6.2 Control Rod Dro Accident Reference 9.7 Dropped Control Rod Worth, -mK 10.5 Doppler Coefficient dK/KdT, 1/'F -9.5 x 10 6 Effective Delayed Neutron Fraction 0.0050 Four-Bundle Local Peaking Factor 1.23 Maximum Deposited Fuel Rod Enthalpy (cal/gm) 170.
12 ZN-NF-87-25 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitin Safet S stem Settin s 7.1.1 MCPR Fuel Claddin Inte rit Safet Limit MCPR Safety Limit 1.06 7.1.2 Steam Dome Pressure Sa et Limit Pressure Safety Limit 1346 psig 7.2 Lim tin Cond tions For 0 eration 7.2.1 Avera e Planar Linear Heat Generation Rate Limits For ANF 8x8 Fuel Bundle Average Exposure MAPLHGR MWD MTU ~Ku Fu 0 13.0 5,000 13.0 10,000 13.0 15,000 13.0 20,000 13.0
'25,000 11.3 30,000 9.4 35,000 7.9 7.2.2 Minimum Critical Power Ratio Rated Condition MCPR Operating Limit Up To EOC -2000 HWD/MTU Exposure (100 To 106% Flow)
~Fuel T e Limit 107 RBS ANF 1.26 GE 1.29
13 XN-NF-87-25 Rated Conditions MCPR Operating Limits From EOC -2000 MWD/MTU To EOC (1008 To 106% Flow)
~Fuel T e Limit ANF 1. 30 GE 1.32 Reduced Flow MCPR Limit (all cycle exposures) Figure 5.2 7.2.3 Surveillance Re uirements 7.2.3.1 Scram Insertion Time Surveillance The ANF reload safety analyses were performed using the control rod insertion times shown below which are based on plant data. In the event that plant surveillance shows these scram insertion times may be exceeded, the plant thermal margin limits are to default to the values which correspond to the technical specification (TSSS) control rod scram times.
Position Inserted From Average Rod Time In Seconds Full Vithdrawn As Defined In Footnote*
Notch 45 0.404 Notch 39 0.660 Notch 25 1.504 Notch 5 2.624 The limiting transient using technical specification control rod scram times is the generator load rejection without bypass. The respective MCPR values for ANF and GE fuel during Cycle 3 are 1.35 and 1.39 using the technical specification control rod speeds with the recirculation pump trip operable.
- Slowest measured average control rod insertion time to specified notches for each group of four control rods arranged in a 2x2 array.
14 XN-NF-87-25 7.2.3.2 Stabilit Surveillance Acceptable surveillance procedures for potentially unstable operation shall be instituted in the portion of the operating power-flow map bounded by the 80%
flow control line and 45% of rated recirculation flow.
7.2.3.3 Technical S ecification LHGR Surveillance The Technical Specification linear heat generation rate (LHGR) limit versus average planar exposure for ANF 8x8 reload fuel is shown in Figure 7.1. This figure was developed from information contained in Reference 9.1, and the region of permissible operation is shown.
15 XN-NF-87-25 TABLE 4.1 NEUTRONIC DESIGN VALUES Fuel Pellet Fuel Material U02 Sintered Pellets Density, g/cc 10.36 0 of T.D. 94.5 Diameter Enriched Fuel 0.4055 Natural Fuel 0.4045 Fuel Rod Fuel Length, inches 150 Cladding Material Zircaloy-2 Clad, I.D., inches 0.414 Clad, O.D., inches 0.484 Fuel Assembl Number of Fuel Rods 62 Number of Inert Water Rods Fuel Rod Enrichments Figure 4.1 Fuel Rod Pitch, inches 0.641 Fuel. Assembly Loading, KgU 176.0'
XN-NF-87-25 TABLE 4.1 NEUTRONIC DESIGN VALUES (Continued)
Core Data Number of Fuel Assemblies 764 Rated Thermal Power, MW 3323 Rated Core Flow, Mlbm/hr 108.5 Core Inlet Subcooling, BTU/ibm 19.0 Reactor Pressure, psia 1008.0 Channel Thickness, inch 0.100 Fuel Assembly Pitch, inch 6.00 Water Gap Thickness (symmetric), inch 0.522 Control Rod Data Absorber Material B4C Total Blade Span, inch 9.75
.Total Blade Support Span, inch 1.58 Blade Thickness, inch 0.260 Blade Face-To-Face Internal Dimension, inch 0.200 Absorber Rods Per Blade 76 Absorber Rod Outside Diameter, inch 0.188 Absorber Rod Inside Diameter, inch 0.138 Absorber Density, % of Theoretical 70.0
WNP-2 CYCLE 3 DESIGN BRS IS RRDIRL POWER 12 10 M
lad Cl 8
CQ C3 cr 6 43 0.2 0.0 0.6 0.8 1 1.2 1.6 BUNDLE PONER I.FlCTOR Figiire 3.1 Radial Power llistogram For 1/4 Core Safety I.imit Model
18 XN-NF-87-25 LL L ML M M ML L LI 0.91 0.95 1.01 1.05 'f.05 1.01. 0.95 0.91 L ML H ML H L 0.96 0.9? 1.07 0.89 1.D7 0.95 ML H H H H ML 1.0'f 'f.02 ML'.9'f 1.0?
'f.01 0.99 1.01 1.01' H M '
ML'.89 H H 1.05 1.0'f' 0.00 0.91 0.99 1.04 :f.D5 M H H M W H M M 1.06 1.04 0.99 0.91 0.00 1.00 086 1.04 ML H H H H H H M 1.01 1.07 1.01 0.99 'f.00 f.01 1.07 1.07 L M ML H M H ML ML 0.95 1.03 081 1.04 0.95 1.07 0.97 1.05 LL L ML M M M ML L 0.91 0.95 1.01 1.05 1.04 1.07 1.06 1.01 Figure 3.2 WNP-2 Cycle 3 Safety Limit Local Peaking Factors (ANF Fuel)
19 XN-NF-87-25 ML ML LL ML ML0'c H H H H H ML>'<:, ML M H H H: H I"I: M H: H H H H H M ML':: ML LL. L . ML: M M ML LL RODS ( 3) 1.50 W/0 U235 L RODS,( 7) 2.00 W/0 U235
'11 RODS ( 9) 2.57 W/0 U235
".I RODS '(16) 2.94 W/0 U235 H RODS (22) 3.54 W/0 U235 ML>'ODS ( 5) 2.57 W/0 U235 + 2.00 W/0 GD203 W RODS'( 2) INERT WATER ROD Figure 4.1 WNP-2 Cycle 3 Enriched Zone Enrichment Distribution
20 XN-NF-87-25 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 I I 1 8 8 8 8 0 I 8 8/ 0/ BJ 8) 8 D I C I I I I 2 8 D 8 8 8 I 0 8 8/ 8/ 0/ 8/ 0( D I 8 I I I 3 8 8 C 8 C 8 ) C 8 / C 8 i 8 I I 4 8 8 8 D 8 D I 8 8) 8/ 0 I D I A l
5 D i 8 C 8 C 8 i C 8 i C 8' I
6 8 D 8 8 8 I 0 I 8 8 I D I 8 0 I 0 i A 7 8 8 C 8 C 8 C 8 A 8 8 8 8 0 i 8 8 8 D I 8 D i 8 D I I I 9 0 8 C 8 C 8 C 8 / C 8) Dl G I I I 10 8 0/ 8 8 8 D I 8 0 i 8 0) Bi 8 I
11 8 8 C 8 C 8 C Bt 0/
12 8 D 8 D Bl 0 I 8 0 I C
.I 13 Di 8 C 8 C 8 C 8 ) A 14 CD,'8 0) 8 0 I B 15 A 8 8 A 8) A A h'6 Fuel Number 432 of GE GE Bx8 Type 8x8 Type II 1.76 v/o U-235 (Cycle 1)
III 2.19 v/o U-235 (Cycle 1) 128 hNF 8x8 2.72 v/o U-235 (Cycle 2) 148 hNF 8x8 2.72 v/o U-235 (Cycle 3)
Figure 4.2 WNP-2 Cycle 3 Reference Loading Pattern By Fuel Type (One Quarter Of Symmetrical Core Loading)
21 XN-NF-87-25 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59 00 -- 36 -- 00 55 51 51 47 20 -- 14 -- 00 -- 14 -- 20 47 43 39 -'- 00 00 -- 20 00 -- 14 00 -- 39 35 35 31 -- 36 00 "'- 20 -- 12 20 -- 00 36 -- 31 27
- 27 23 -- 00 00 -- 20 00~ -- 14 00 -- 23 1.9 19
'15 14 -- 00 14 "-- 20 15 00 -- 36 -- 00 3
2 6 !.0 14 1.8 22 26 30 34 38 42 46 50 54 58
"'ontrol Rod Being Withdrawn Rod Position in Notches Withdrawn Full in = 00 Full out =--
Figure 5.1 WNP-2 Cycle 3 Control Rod Withdrawal Analysis Initial Control Rod Pattern
- 1. 6 NOTE: The flCPR operating limit shall be the maximum of this curve or the rated condition HCPR operating limit.
30 40 50 60 70 80 90 100 110 TGTBL CQRE. RECIRCULATING f LGN (% BRTEO)
I Figure 5.2 Reduced Flow MCPR Operating Limit I 00 V
I Vl
18 ~LgiR 0 15.82 610 16.62 14 2.680 15.10 6,230 14.7l
~ ~ ~ ~ ~ ~ ~ ~ ~ ~
7,940 1$ .19 10,470 14.13 12-13,220 14.06 h
16,990 14.06 18.780 14.00 10-I',
21,590 13.93 PERMISSIBLE 24,420 13.93 REGION OF 27.280 13h08 8- ". OP ERAT ION h
30,150 12.24 33,060 11.40 36,960 10.47 38.900 9.66 10000 20000 30000 40000 60000 41,830 8.66 44,760 7.77 Average Planar Exposure (MWD/MTj Figure 7.1 Linear Heat Generation Rate (LHGR) Limit I Versus Average Planar Exposure hrf ANF 8x8 Fuel 00 I
I
'Ln
24 XN-NF-87-25 9.0 ADDITIONAL REFERENCES 9.1 S. F. Gaines, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR RLAA.
Richland, WA 99352 (January 1982).
9.2 R. H. Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling Richland, WA 99352 (November 1981).
9.3 J. E. Krajicek, "WNP-2 Cycle 3 Plant Transient Analysis," XN-NF-87-24, Revision 1, Exxon Nuclear Company, Inc., Richland, WA 99352 (March 1987).
Nuclear Company, Ines Inc., Richland, WA 99352 (December 1985).
9.5 D. J. Braun, "WNP-2 LOCA-ECCS Analysis, MAPLHGR Results," XN-NF-85-139, Exxon Nuclear Company, Inc., Richland, WA 99352 (December 1984).
9.6 M. H. Smith, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Company, , Richland, WA 99352 (March 1985).
9.7 "Exxon Nuclear Methodology for 'Boiling Water Reactors-Neutronics Metehods Nuclear Company, Inc., Richland, WA 99352 (May 1980).
9.8 "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
99352 (September 1986).
9.9 R. G. Grummer, "A Generic Analysis of the Loss of Feedwater Heating Company, Inc., Richland, WA 99352 (February 1986).
9.10 J. B. Edgar, Letter to WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.
A-1 XN-NF-87-25 APPENDIX A SINGLE LOOP OPERATION (SLO)
The NSSS supplier, General Electric (GE), has provided analyses which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended period of time. These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power 'evels than are allowed when both recirculation systems are in operation. The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power. Because the ANF fuel was designed to be compatible with the coresident fuel in thermal hydraulic, nuclear, and mechanical design performance, and because the ANF methodology has given results which are consistent with those of the previous analyses for normal two-loop operation, the analyses performed by GE for single loop operation are also applicable to single loop operation with fuel and analyses provided by ANF.
With a single recirculation loop in operation, the GE analyses supported continued operation with an increase of 0.01 in the MCPR safety limit.
Because of the similarity between 'the ANF and GE fuel types making up the core, and because of the similarity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value 'is also appropriate for operation with ANF fuel and analyses. For Cycle 3 operation with both recirculation loops in operation, the MCPR safety limit is 1.06, which is the same value as was used for the previous cycle. For Cycle 3 operation with a single recirculation loop in service, the MCPR safety limit's 1.07, which is also the same value as was used for the previous cycle.
A-2 XN-NF-87-25 The consequences of core-vide transients at the reduced power and flow conditions necessitated by single loop operation are bounded by the consequences of these events at rated conditions. The additional conservatism imposed by the reduced flow MCPR operating limits specified in the main body of this report assures that the MCPR safety limit will not be violated during anticipated operational occurrences with a single recirculation loop in service. No modification to the delta-CPR defining the rated conditions MCPR operating limit is required, and the reduced flow MCPR limit curve remains conservatively applicable during single loop operation. Because the reduced flow MCPR limit curves are based on equipment performance which physically cannot happen during single loop operation, the added conservatism present in the curves compensates for the penalties associated with increased uncertainties in the MCPR safety limit and control rod drive performance. The reduced flow MCPR limit curves are applicable without modification during single loop operation.
To support operation of WNP-2 with a core composed of GE P8xSR and ANF SxS fuel with a single recirculation pump operating, ANF recommends the conservative use of GE fuel MAPLHGR limits for the similar GE P8x8R fuel design with a multiplier of 0.84 applied for single loop operation. The basis for this recommendation is as follows:
The phenomena which require the reduction in MAPLHGR limits are a result of operation of the WNP-2 system with a singl'e active recirculation loop, and are equally applicable to both GE and ANF fuel designs; and The analytical methods used by GE have yielded conservative MAPLHGR limits relative to the MAPLHGR limits obtained using the approved ANF analytical methods.
Therefore, applying the more conservative GE MAPLHGR limits to ANF fuel provides a limit which assures conformance with the criteria of 10 CFR 50.46.
XN-NF-87-25 Issue Date: 3/27/Q7 WNP-2 CYCLE 3 RELOAD ANALYSIS Distribution:
R. E. Collingham L. J. Federico J. G. Ingham S. E. Jensen T. H. Keheley J. E. Kraj icek T. L. Krysinski J. L. Maryott J. N. Morgan A. Reparaz G. L. Ritter B. T. Stiles G. N. Ward H. E. Williamson J. B. Edgar/WPPSS (50)
Document Control (5)
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