ML17285B056
| ML17285B056 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/31/1990 |
| From: | Krajicek J, Leonard S SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17285B053 | List: |
| References | |
| ANF-90-02, ANF-90-2, NUDOCS 9003080236 | |
| Download: ML17285B056 (56) | |
Text
ADVANCEDNUCLEARFUELS CORPORATlON ANF-90-02 Issue Date:
1/15/90 WNP-2 CYCLE 6 RELOAD ANALYSIS Prepared by J.
. Krajicek BWR S fety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services S.
L. Leonar BWR Neutronics Neutronics and Fuel Management Fuel Engineering and Technical Services January 1990 90030S0236 PDR ADOCK P
NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REQARDINQ CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development pro.
grams sponsored by Advanced Nuclear Fuels Corporation. It Is being submit.
ted by Advanced 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 Ad.
vanced Nuclear Fuels Corporation.fabricated reload fuel or other technical services provided by Advanced Nuclear Fuels Corporation for light water power reactors and It Is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, Informtttlon, and belief. The information con.
talned 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 Corporation In their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.
Advanced Nuclear Fuels Corporation's warranties and representations con cerning the subject matter of this document are those set forth in the agree.
ment between Advanced Nuclear Fuels Corporation and the customer to which this document 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 usefulness of the Information contained In this docu.
ment, or that the use of any Information, apparatus, method, or process disclosed In this document willnot 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.
ANF 3iiS.scca
<4
TABLE OF CONTENTS ANF-90-02 Page i Section
- 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 Design Basis Local Power Distribution
~Pa e
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 4.2.4 Core Hydrodynamic Stability I
5.0 ANTICIPATED OPERATIONAL OCCURRENCES
- 5. 1 Analysis Of Plant Transients At Increased 5.2 Analyses For Reduced Flow Operation 5.3 Analysis For Reduced Power Operation (SLO) 5.4 ASME Overpressurization Analysis 5.5 Control Rod Withdrawal Error 5.6 Loading Error for Reload Fuels 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 0
Core Flow
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Conditi ons 11 11 11 11 11 11 7.0 TECHNICAL SPECIFICATIONS
- 7. 1 Limiting Safety System Settings
- 7. 1. 1 MCPR Fuel Cladding Integrity Safety Limi
- 7. 1.2 Steam Dome Pressure Safety Limit 7.2 Limiting Conditions For Operation 7.2. 1 Average Planar Linear Heat Generation ANF 8x8 Fuel 7.2.2 Minimum Critical Pow'er Ratio
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Rate Limits For 12 12 12 12 12 12 12
ANF-90-0 Page 7.2.3 Surveillance Requirements 7.2.3. 1 Scram Insertion Time Surveillance 7.2.3.2 Stability Surveillance 7.2.3.3 Technical Specification LHBR Surveillance 9.0 ADDITIONAL REFERENCES 13 13 14 14 31
Qe ANF-90-02 Page iii LIST OF TABLES Table Pacae 4.1 NEUTRONIC DESIGN VALUES......................
15 LIST OF FIGURES
~Fi ere Pa(ac 3.1 3.2 3.3 3.4 3.5 3.7 3.8 4.1 4.2 5.1 5.2 5.3 7.1 A.1 A.2 A.3 RADIAL POWER HISTOGRAM FOR I/4 CORE SAFETY LIMIT MODEL WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-5 FUEL)
WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)
WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL)
WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)
WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 CENTRAL FUEL)
WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 PERIPHERAL FUEL)
WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL)
WNP-2 ANF-5 CYCLE 6 ENRICHED ZONE ENRICHMENT DISTRIBUTION WNP-2 CYCLE 6 REFERENCE LOADING PATTERN BY FUEL TYPE (ONE QUARTER OF SYMMETRICAL CORE LOADING)
WNP-2 CYCLE 6 CONTROL ROD WITHDRAWAL ANALYSIS INITIAL CONTROL ROD PATTERN REDUCED FLOW HCPR OPERATING LIMIT FOR NORMAL FEEDWATER TEMPERATURE REDUCED FLOW HCPR OPERATING LIMIT FOR FFTR OPERATION LINEAR HEAT GENERATION RATE (LHGR) LIHIT VERSUS AVERAGE PLANAR
- EXPOSURE, ANF 8X8 FUEL LHGR LIMIT FOR 9X9-IX FUEL LHGR LIMIT FOR 9X9-9X FUEL ANF 9X9-IX AND 9X9-9X MAPLHGR LIMITS.
17 18 19 20 21 22 23 24 25 26 27 28 29 30 A-5 A-6 A-7
ANF-90-02 Page iv WNP-2 CYCLE 6 RELOAD ANALYSIS REPORT The following Design Notebooks support the above titled document.
LSSE E-5003-593-1 E-5003-593-2 E-5003-593-3 E-5003-593-4 E-5003-595-1 E-5003-595-2 E-5003-595-3 E-5003-595-4 E-5003-596-5 E-5003-596-6 E-5003-595-7 E-5003-872-1 N&FH E-5003-N04-2 E-5003-N06-1 E-5003-N06-2 E-5003-N06-7 E-5003-N06-5 E-5003-N06-4 E-5003-N06-8 WNP-2 Cycle 6 XCOBRA Data DECKPL WNP-'2 Cycle 6 Heat Transfer Coefficients WNP-2 Cycle 6 XCOBRA-T Data DECKPL WNP-2 Cycle 6
COTRANSA Input Deck Preparation WNP-2 Cycle 6 104/106 LRNB RPT Operable NSS WNP-2 Cycle 6 104/106 LRNB RPT Inoperable NSS WNP-2 Cycle 6 104/106 LRNB RPT Operable TSSS WNP-2 Cycle 6 104/106 LRNB RPT Inoperable TSSS WNP-2 Cycle 6 47/106 FWCF RPT Operable NSS WNP-2 Cycle 6 47/106 FWCF RPT Inoperable NSS WNP-2 Cycle 6 ASME Overpressurization Transient WNP-2 Cycle 6
MCPR Safety Limit Analysis WNP-2 Cycle 6 Cycle Design
- Part II WNP-2 Cycle 6
COTRANSA Input WNP-2 Cycle 6 Fuel Mislocation Error Analysis WNP-2 Cycle 6 Control Rod Withdrawal Error WNP-2 Cycle 6 Control Rod Drop Analysis WNP-2 Cycle 6 Stability Analysis WNP-2 Cycle 6 Loss of Feedwater Heating
I I
Qe ANF-90-02 Page 1
- 1. 0 INTRODUCTION This report summarizes the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 6
reload for the Supply System Nuclear Project Number 2
(WNP-2).
The WNP-2 Cycle 6 core was designed and safety analyses were performed to support a reload batch size of 144 fresh assemblies and 56 assemblies reloaded from the spent fuel storage pool.
WNP-2 is scheduled to commence Cycle 6 operation in May 1990.
This report is intended to be used in conjunction with ANF topical report
~X---,414,R II I, "Appll I
V AXNCII Rdlpy 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 p
dl I
4 I
~XN-AP-
-I A,VI 4,R The WNP-2 Cycle 6 core will comprise a total of 764 fuel assemblies:
including 144 unirradiated assemblies of which 136 are ANF 8x8 and 8
assemblies are other vendors'ead test assemblies (LTAs);
2 ANF 9x9-IX irradiated lead fuel assemblies (LFAs);
2 ANF 9x9-9X irradiated LFAs; 560 irradiated ANF 8x8 assemblies; and 56 reloaded irradiated P8x8R assemblies fabricated by General Electric (GE).
The analysis reported in this document assumes the 144 unirradiated assemblies are ANF 8x8 fuel.
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 encompasses core flow up to 106% of the design basis value.
h
ANF-90-02 Page 2
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 6
of WNP-2 is bounded by the assumed power history in the fuel mechanical design analyses.
0
ANF-90-02 Page 3
- 3. 0 THERMAL"HYDRAULIC DESIGN ANALYSIS 3.1 Desi n Criteria
- 3. 1.3 Fuel Centerline Tem erature The fuel analyses using the LHGR curve of Figure 3.4 of Reference 9.8 show that the ANF 8x8 fuel centerline temperature will be below the melting point at 120% overpower.
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 draul ic Characteri zati on 3.2.3
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Calculated Bypass Flow Fraction (100% power/106% flow) 10.4%
3.3 MCPR Fuel Claddin Inte rit Safet Limit
'3.3. 1 Coolant Thermod namic Condition Core Power Core Inlet Enthalpy Steam Dome Pressure Feedwater Temperature 3900 MWt 525.6 Btu/ibm 1021 psia 414'F 3.3.2 Desi n Basis Radial Power Distribution See Figure 3. 1.
3.3.3 Desi n Basis Local Power Distribution See Figures 3.2, 3.3, 3.4, 3.5, 3.6, 3.7 and 3.8.
k 1
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4.0 NUCLEAR DESIGN ANALYSIS ANF-90-02 Page 4
- 4. 1 Fuel Bundle Nuclear Desi n Anal sis Assembly Average Enrichment Radial Enrichment Distribution Axial Enrichment Distribution 2.62 w/o U-235 Figure 4.1 Uniform 2.79 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 Note:
The reload includes 8 LTAs modeled as ANF 8x8 fuel.
4.2 Core Nuclear Desi n Anal sis 4.2.1 Core Confi uration
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Core Exposure at EOC5 (MWd/MTU)
Core Exposure at BOC6 (MWd/MTU)
Core Exposure at EOC6 (MWd/MTU)
Figure 4.2 18,400 12,800 18,700 4.2.2 Core Reactivit Characteristics BOC Cold k-eff, All Rods Out BOC Cold k-eff, Strongest Rod Out Reactivity Defect (R-Value),
% delta k/k Standby Liquid Control System (SBLC) 660 ppm Boron, Cold k-eff 1.1138 0.9868 0.0 0.9638
ANF Pag 4.2.4 Core H drod namic Stabilit
/Power %Flow State Points 65/45*
47/27.6**
42/23 8***
Deca Ratio COTRAN 0.46 0.86 0.83
- 45 percent flow - APRN Rod Block intercept point.
- Two pump minimum flow - 47 percent power.
- Natural circulation flow - APRN Rod Block intercept point.
ANF-90-02 Page 6
5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Transient Analysis Report Reference 9.3
- 5. 1 Anal sis Of Plant Transients At Increased Core Flow Conditions References 9.3 and 9,11 Limiting Transient(s):
Load Rejection Without Bypass (LRNB)
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 6 analyses results at increased core flow conditions are conservatively applicable to rated flow conditions.
Cycle 6 specific analyses of transient events were performed for two recirculation pump operation conditions, with the recirculation pump trip (RPT) in service and out of service, and for two scram conditions which are normal scram speed (NSS) and technical specification scram speed (TSSS).
Analyses were performed at end-of-cycle exposures.
Generic analyses were performed for FFTR to extend cycle operation (Reference
- 9. 11).
The loss of feedwater heating event was analyzed on a plant specific bounding value basis and the delta CPR results are bounding values for WNP-2.
ANF Pag Transient*
Haximum
% Power/
Haximum Haximum Pressure
% Flow Heat Flux %
Power
% ~si Delta CPR GE ANF Fuel Fuel
- LRNB, NSS RPT Inoperable LRNB, TSSS RPT Operable LRNB, TSSS RPT Inoperable FWCF7 NSS RPT Operable
- FWCF, NSS RPT Inoperable 104/106 104/106 104/106 122 122 132 553 483 592 47/106 163 47/106 56 217 104/106 117 397 1203 1212 1205 1189 1026 1023 0.24 0.25 0.28 0.30 0.27 0.30 0.31 0.34 0.20 0'3 0.26 0.28 LOFH N/A N/A N/A N/A 0.09 0.
5.2 Anal ses For Reduced Flow 0 eration Limiting Transient:
Recirculation Flow Increase References 9.3 and 9. 11 The 104%
power/106%
flow cases conservatively bound the reduced power 106% flow operation conditions.
References
- 9. 12, 9. 13, and 9.14 ANF has performed analyses for WNP-2 which demonstrate the safety of plant operation with a
single recirculation loop out of service for an extended period of time.
These analyses were performed for the most limiting transient
- events, the pump seizure accident and the loss-of-coolant-accident (LOCA) for the maximum extended power state during WNP-2 single-loop operation (SLO).
The transient analysis and pump seizure accident analysis are documented in Reference
- 9. 12, and the LOCA analysis is documented in
ANF-90-02 Page 8
Reference
- 9. 13.
The conclusions presented in these documents are applicable to future cycles with ANF fuel and have been reviewed by the U.
S.
Nuclear Regulatory Commission in Reference
- 9. 14.
The SLO limits from the USNRC review are summarized below.
SLO MCPR Operating Limit for ANF and GE fuel 1.35 Two-loop HAPLHGR limits which are shown in Section
- 6. 1.3 for ANF fuel apply during SLO.
For GE fuel the reduction of the MAPLHGR limit to a
value of 0.84 times the two recirculation loop operation HAPLHGR limit for SLO remains unchanged.
5.4 ASHE Over ressurization Anal sis Limiting Event Worst Single Failure Maximum Pressure Maximum Steam Dome Pressure References 9.3 and 9.11 HSIV Closure MSIV Position Scram Trip 1317 psig 1291 psig 5.5 Control Rod Withdrawal Error Initial Control Rod Pattern for CRWE Analysis Figure 5. 1 Rod Block Monitor Settin 106%+
107%
108%
Distance Withdrawn (ft) 5.5 6.5 8.0 Limiting Delta-CPR**
0.18 0.21 0.23 5.6 Loadin Error for Reload Fuels With Loadin Error Correctly Loaded Core Maximum LHGR, kW/ft Minimum HCPR 15.0 1.29 13.4 1.44
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- Rod Block Monitor Setting (RBH) of 106%.
- The limiting delta CPR for the control rod withdrawal error conservatively applies to all fuel in the core including the ANF Bx8 and GE Bx8.
The GE Bx8 fuel is in non-limiting core locations for Cycle 6.
ANF Pag 5.7 Determination Of Thermal Mar ins Summary of Thermal Margin Requirements All system transient results were analyzed at the more limiting increased flow conditions (106%)
rather than rated flow conditions (100%).
LRNB results for the more limiting power (design basis
'ondition -
104%) were used for this transient.
These calculated results are based on end of cycle conditions and increased core flow (106%).
Event Equipment 0 erational Status Delta CPR GE ANF Fuel Fuel Indicated MCPR Limit GE ANF Fuel Fuel Model LRNB LRNB RPT Operable, NSS RPT Inoperable, NSS
- 0. 24
- 0. 25 1.31
- 1. 31 COTRANSA/XCOBRA-T 0.28 0.30 1.36 1.36 LRNB RPT Operable, TSSS 0.27 0.30 1.36 1.36 LRNB FWCF RPT Inoperable, TSSS RPT Operable, NSS 0.31 0.34 1.40 1.40 0.20 0.21 1.27 1.27 FWCF RPT Inoperable, NSS 0.26 0.28 1.34 1.34 LOFH N/A 0.09 0.09 1.15 1.15 XTGBWR Notes:
1.
ANF 8x8 fuel MCPR limits bound GE 8x8 fuel for all conditions in the above table based on the calculations performed; thus it is appropriate to monitor both the ANF and GE fuels to the MCPR limit of the ANF fuel.
2.
For cycle extension with reduced feedwater temperature, add 0.02 to delta CPR/MCPR LRNB and FWCF transient results in the above table.
0
ANF-90-02 Page 10 MCPR Operating Limits At Rated Condition For Cycle Exposures Less Than 3750 MWd/MTU are based on the CRWE (100% To 106% Flow)
~Fuel T
RBM 1.24 1.24 HCPR Operating Limits At Rated Condition From 3750 HWd/MTU To EOC (100%
To 106% Flow) With Normal Feedwater Temperature
~Fuel T
e ANF GE MCPR Limit 1.31 1.31 MCPR Operating Limits At Rated Condition Beyond All Rods Out With Reduced Feedwater
'emperature (100% To 106%
Flow And Thermal Coastdown)
Point (EOCS)
~Fuel T
e ANF GE MCPR Limit 1.33 1.33 MCPR Limits at Off-Rated Conditions Figures 5.2 and S.3 Reduced Flow HCPR Limit References 9.3 and 9. 11
ANF-90-02 Page 11 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
- Two-Loop Operation and SLO)
References 9.5, 9.13 and 9.14 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
~ ~d 20,000 MAPLHGR
~kW ft 13.0 Peak Clad*
Tem erature
'F 1771 Peak Local*
MWR 0.51 6.2 Control Rod Dro Accident Dropped Control Rod Worth, mK Doppler Coefficient dk/kdT, 1/'F Effective Delayed Neutron Fraction Four-Bundle Local Peaking Factor Maximum Deposited Fuel Rod Enthalpy (cal/gm)
Reference 9.7 6.9
-10.0 x 10 6
0.0050 1.17 97
- For the ANF-4(6Gd2) fuel design.
ANF-90-02 Page 12
- 7. 0 TECHNICAL SPECIFICATIONS
- 7. 1 Limitin Safet S stem Settin s
- 7. 1. 1 MCPR Fuel Claddin Inte rit Safet Limit HCPR Safety Limit 1.06
- 7. 1.2 Steam Dome Pressure Safet Limit.
Pressure Safety Limit 1346 psig 7.2 Limitin Conditions For 0 eration 7.2. 1 Avera e Planar Linear Heat Generation Rate Limits For ANF 8x8 Fuel Bundle Average Exposure MWd MTU 0
5,000 10,000 15,000 20,000 25,000 30,000 35,000 MAPLHGR
~kW ft 13.0 13.0 13.0 13.0 13.0 11.3 9.4 7.9 For single-loop operation these limits also apply to ANF Fuel when using a
SLO HCPR operating limit of at least 1.35.
7.2.2 Minimum Critical Power Ratio Rated Condition HCPR Operating Limit Up To 3750 HWd/HTU Exposure (100% To 106% Flow)
~Fuel T
e ANF GE Limit 106%
RBM 1.24 1.24
ANF Page Rated Conditions MCPR Operating Limits From 3750 MWd/MTU To EOC (100% To 106% Flow)
~Fuel T
e
'NF GE Limit 1.31 1.31 Thermal Coastdown and FFTR Rated Condition MCPR Operating Limit Beyond All Rods Out Point With Reduced Feedwater Temperature (100% to 106% Flow)
~Fue1 T
e ANF GE Limit 1.33 1.33 Reduced Flow MCPR Limit (all cycle exposures)
Figures 5.2 and 5.3 Single-Loop Operation (SLO)
MCPR Limit (all cycle exposures),
~Fue1 T
e Limit ANF 1.35 GE 1.35 7.2.3 Surveillance Re uirements 7.2.3. 1 Scram Insertion Time Surveillance The ANF reload safety analyses were labeled NSS (Normal Scram Speed) 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 (see Section 5.7).
ANF-90-02 Page 14 Position Inserted From Full Withdrawn Average Rod Time In Seconds As Defined In Footnote*
Notch 45 Notch 39 Notch 25 Notch 5
0.404 0.660 1.504 2.624 7.2.3. 2 Stabi1 it Survei ll ance Core hydrodynamic stability analyses support stability regions specified in the Technical Specifications Amendment No.
71.
This Amendment defines regions on the power flow map which preclude operation as well as regions where operation is allowed with surveillance provided by the enhanced stability monitoring system, A'NNA'.
The results of these analyses support operation below a
line defined by the following power/flow points:
42/ Power/23.8/
- Flow, 47/
Power/27.6/
- Flow, and 65/
Power/45/
Flow (see Section 4.2.4).
The stability line defined by these points bounds the regions in the Amended Technical Specifications.
'.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.
- Slowest measured average control rod insertion time to specified notches for each group of four control rods arranged in a 2x2 array.
ANF Page TABLE 4. 1 NEUTRONIC DESIGN VALUES Fuel Pellet Fuel Material Density, g/cc
% of T.D.
- Diameter, inch Enriched Fuel Natural Fuel U02 Sintered Pellets 10.36 94.5 0.4055 0.4045 Fuel Rod Fuel
- Length, inch Cladding Material Clad, I.D., inch Clad, O.D., inch 150 Zircaloy-2 0.414 0.484
~FI 1 11 Number of Fuel Rods Number of Inert Water Rods Fuel Rod Enrichments Fuel Rod Pitch, inch Fuel Assembly Loading, kgU 62 Figure 4.1 0.641 176.0
TABLE 4. 1 NEUTRONIC DESIGN VALUES (Continued)
ANF-90-02 Page 16 Core Data Number of Fuel Assemblies Rated Thermal
- Power, HW Rated Core Flow, Hlbm/hr Core Inlet Subcooling, Btu/ibm Reactor Pressure, psia Channel Thickness, inch Fuel Assembly Pitch, inch Water Gap Thickness (symmetric),
inch 764 3323 108.5 19.0 1008.
0.100 6.00 0.522 Control Rod Data Absorber Haterial Total Blade Span, inch Total Blade Support
- Span, inch Blade Thickness, inch Blade Face-To-Face Internal Dimension, inch Absorber Rods Per Blade Absorber Rod Outside Diameter, inch Absorber Rod Inside Diameter, inch Absorber Density,
% of Theoretical B4C 9.75 1.58 0.260 0.200 76 0.188 0.138 70
90 80 70 60 C) 50 C3 40 30 20 10 0
0.2 0.4 0.6 0.8 1
1.2 RRDIRL PDN PERKING FIGURE 3. 1 RADIAL POWER HISTOGRAM OR I/4 CORE SAFETY LIHIT HODEL
Qa ANF-90-02 Page 18
~
- 0.936 :.0.977
- 1.023
- 1.015
- 1.011
- 1.041
- 1.076
- 1.052
~
- 0.977
- 1.011
- 0.907
- 1.042
- 1.035
- 0.932
- 0.962
- 1.075
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- 1.023
- 0.907
- 1.017
- 0.988
- 0.974
- 0.996
- 0.931
- 1.040
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- 1.042
- 0,988
- 0.000
- 0.850
- 0.972
- 1,033
- 1.009
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- 1.011
- 1.035
- 0.974
- 0.850
- 0.000
- 0.985
- 1.038
- 1.011
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- 1.041
- 0.932
- 0.996
- 0.972
- 0.985
- 1.012
- 0.901
- 1.043
~
1.076
- 0.962
- 0.931
- 1.033
- 1.038
- 0.901
- 0.976
- 1.078
- 1.052
- 1.075
- 1.040
- 1.009
- 1.011
- 1.043
- 1.078
- 1.054 FIGURE 3.2 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-5 FUEL)
ANF Page
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- 0,928
- 0.958
- 1.002
- 1.004
- 1.002
- 1.019
- 1.032
- 1.000
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- 0.958
- 0.998
- 0.938
- 1.047
- 1.042
- 0.952
- 0.974
- 1.032
~
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- 1.002
- 0.938
- 1.032
- 1.015
- 1.002
- 1.015
- 0.952
- 1.019
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1.004
- 1.047
- 1.015
- 0.000
- 0.906
- 1.002
- 1.041
- 1.002 1.002
- 1.042
- 1.002
- 0.906
- 0.000
- 1.014
- 1.046
- 1.004
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- 1.019
- 0.952
- 1.015
- 1.002
- 1.014
- 1.030
- 0.937
- 1.022
~
1.032
- 0.974
- 0.952
- 1.041
- 1.046
- 0.937
- 0.979
- 1.034 1.000
- 1.032
- 1.019
- 1.002
- 1.004
- 1.022
- 1.034
- 1.002 FIGURE 3.3 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF-4 FUEL)
ANF-90-02 Page 20
- 0.958
- 0.964
- 1.000
- 1.026
- 1.025
- 1.000
- 1.064
- 0.958
- 0.964
- 0.981
- 1.049
- 0.922
- 1.031
- 1.047
- 1.019
- 0.964 1.000
- 1.049
- 1.015
- 1.004
- 0.996
- 1.010
- 0.937
- 1.000 1.026
- 0.922
- 1.004
- 0.000
- 0.937
- 0.995
- 1.031 1.026 1.025
- 1.031
- 0.996
- 0.937
- 0.000
- 1.001
- 0.972
- 1.027
- 1.000
- 1.047
- 1.010
- 0.995
- 1.001
- 1.014
- 1.051
- 1.042
- 0.964
- 1.019
- 0.937
- 1.031
- 0.972
- 1.051
- 0.974
- 1.029
- 0.958
- 0.964
- 1.000
- 1.026
- 1.027
- 1.042
- 1.029 1.004 FIGURE 3.4 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-3 FUEL)
ANF Page
~
- 0.967
- 0.969
- 0.997
- 1.019
- 1.019
- 0.996
- 0.968
- 0.966
~4
~
- 0.969
- 0.981
- 1.044
- 0.932
- 1.030
- 1.042
- 1.013
- 0.968
~
- 0.997
- 1.044
- 1.017
- 1.008
- 1.001
- 1.012
- 0.944
- 0.997
~
- 1.019
- 0.932
- 1.008
- 0.000
- 0.947
- 1.000
- 1.030
- 1.019
~
~
1.019
- 1.030
- 1.001
- 0.947
- 0.000
- 1.006
- 0.976
- 1.020
~
~
- 0.996
- 1.042
- 1.012
- 1.000
- 1.006
- 1.017
- 1.047
- 1.032
~
- 0.968
- 1.013
- 0.944
- 1.030
- 0.976
- 1.047
- 0.975
- 1.020
- 0.966
- 0.968
- 0.997
- 1.019
- 1.020
- 1.032
- 1.020
- 1.003 FIGURE 3.5 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-2 FUEL)
ANF-90-02 Page 22
~
- 0.984
- 0.976
- 0.995
- 1.012
- 1.012
- 0.995
- 0.976
- 0.984
~
- 0.976
- 0.982
- 1.034
- 0.940
- 1.024
- 1.033
- 1.007
- 0.976
~
~
- 0.995
- 1.034
- 1.015
- 1.008
- 1.002
- 1.011
- 0.950
- 0.995
~
~
- 1.012
- 0.940 1
F 008
- 0.000
- 0.956
- 1.002
- 1.025
- 1.014
~
- 1.012
- 1.024
- 1.002
- 0.956
- 0.000
- 1.007
- 0.981
- 1.015
~
~
- 0.995
- 1.033
- 1.011
- 1.002
- 1.007
- 1.016
- 1.039
- 1.024
~
0
- 0.976
- 1.007
- 0.950
- 1.025
- 0.981
- 1.039
- 0.976
- 1.016
- 0.984
- 0.976
- 0.995
- 1.014
- 1.015
- 1.024
- 1.016
- 1.009 FIGURE 3.6 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 CENTRAL FUEL)
ANF Page
~
- 0.984
- 0.976
- 0.995
- 1.012
- 1.012
- 0.995
- 0.976
- 0.984
~
- 0.976
- 0.982
- 1.034
- 0.940
- 1.024
- 1.033
- 1.007
- 0.976
~
~
4
- 0.995
- 1.034
- 1.015
- 1.008
- 1.002
- 1.011
~
- 0.950
- 0.995
- 1.012
- 0.940
- 1.008
- 0.000
- 0.956
- 1.002
- 1.025
- 1.014
~
1.012
- 1.024
- 1.002
- 0.956
- 0.000
- 1.007
- 0.981
- 1.015
~
- 0.995
- 1.033
- 1.011
- 1.002
- 1.007
- 1.016
- 1.039
- 1.024
~
- 0.976
- 1.007
- 0.950
- 1.025
- 0.981
- 1.039
- 0.976
- 1.016
- 0.984
- 0.976
- 0.995
- 1.014
- 1.015
- 1.024
- 1.016
- 1.009 FIGURE 3.7 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (ANF XN-1 PERIPHERAL FUEL)
ANF-90-02 Page 24 1.03 1.00
.99
.99
.99
.99 1.00 1.03 1.00
.97
.99 1.02 1.03
. 1.03
.99 1.00
.99
.99 1.02 1.01 1.02
.91 1.03
.99
.99 1.02 1.01
.91
.00 1.02 1.02
.99
.99 1.03 1.02
.00 1.02 1.01
.99 99
.99 1.03
.91 1.02 1.01
.98
.99
.99 1.00
.99 1.03 1.02
.99
.99
.97 1.00 1.03 1.00
.99
.99
.99
,99 1.00 1.03 FIGURE 3.8 WNP-2 CYCLE 6 SAFETY LIMIT LOCAL PEAKING FACTORS (GE FUEL)
ANF Page
~
LL L
ML M
M M
~
o
~
M: ML*: H: H: M:
ML*
~
ML.
~
M(*
~ H: H:
. H: H; M
~
~
M
~
H H:
W H:
H H
H M
W H:
H M
M H
H H
H:
ML*
ML: ML*:
M
~
H H
ML*
M:
M M:
RODS ML RODS M
RODS H
RODS ML* RODS W
RODS
( 1)
( 5)
( 9) 21)
(20)
( 6)
( 2) 1 ~ 50 W/0 U235 2.00 W/0 U235 2.50 W/0 U235-2.64 W/0 U235 3.43 W/0 U235 2.50 W/0 U235 + 2.00 W/0 GD203 INERT WATER ROD.
FIGURE 4.1 WNP-2 ANF-5 CYCLE 6 ENRICHED ZONE ENRICHMENT DISTRIBUTION
10 12 13 15 E
C C
J H
F C
J H
D C
J H
D D
J H
0 D
J F
F D
J C
J A
A H
C H
F J
D H
0 E
D H
0 E
D J
D J
0 H
C H
C J
D H
C H
C J
0 H
F H
F J
F H
0 F
F J
F A
A A
H 0
D J
H C
C J
H D
D J
J C
C 0
H 0
D J
H F
C J
F E
F C
H D
0 J.
H C
C J
H 0
F J
H D
D J
J G
I C
0 F
H C
A C
ANF-90-02 Page 26 Fuel Number of
~T e
Assemblies Descri tion Bundle Enrichments A
C D
E F
G H
I J
56 128 148 24 128 4
128 144*
GE 8X8 TYPE II 1.76 w/o U-235 (Reinserted From Cycles 1
8 2)
ANF SX8 2.72 w/o u-235 (Cycle 2)
ANF SX8 2.72 w/o U-235 (Cycle 3)
ANF SXS 2.72 w/o U-235 (Cycle 4)
ANF SXS 2.64 w/o U-235 (Cycle 4)
ANF SXS 2.64 w/o U-235 (Cycle 5)
ANF SXS 2.62 w/o U-235 (Cycle 5)
ANF 9X9 Lead 2.53/2.59 w/o U-235 (Cycle 5)
ANF 8X8 2.62 w/o U-235 (Cycle 6)
This includes 8 Test Assemblies from other vendors.
FIGURE 4.2 WNP-2 CYCLE 6 REFERENCE LOADING PATTERN BY FUEL TYPE (ONE QUARTER OF SYMMETRICAL CORE LOADING)
59 55 51 43 2
6 10 14 18 22 26 30 34 38 42 46 50 54 58 00 40 00 36 16- --
00 16 36 ANF Pag 59 47 43 39 00 35 31 40 27 16 00 00 36 36 00 16 00* --
36 00 00 39 35 40 31 27 23 00 16 00 36 00 16 00 23 19 36 16 00 16 36 00 40 00 19 is 11 2
6 10 14 18 22 26 30 34 38 42 46 50 54 58 Control Rod Being Withdrawn Rod Position in Notches Withdrawn Full In 00 Full Out =--
FIGURE 5.1 WNP-2 CYCLE 6 CONTROL ROD WITHDRAWAL ANALYSIS INITIAL CONTROL ROD PATTERN
1.6 1.5 NOTE: The 3ECPR operating limitshall be the greater of the rated condition MCPR operating limitor the value for reduced flow from this curve 1.4 1.3 1.2 20 30 40 50 60 VO 80 90 100 11 0 TOTAL CORE RECIRCULATING FLOW (%RATED)
FIGURE 5.2 REDUCED FLOW HCPR OPERATING LIMIT FOR NORMAL FEEDMATER TEMPERATURE U ll I
cQ LD lD O I
NO Co M
- 1. 6 NOTE:
The HCPR operating limit shall be the maximum of this cnrve or the rated condition HCPR operating limit.
3O 4O SO 6O 7O BO SO 1OO TQ TRL CQRE BEC I RCULR T f NG FLQH
(%
FIGURE 5.3 REOUCEO FLOW MCPR 0 TING LIMIT FOR FFTR OPERATION
~
~
~
~
~
~
~
~
~
~ ~
~
~
~
0
~
~
~
~
~
~
. PERHISSIBLE REGION OF
~
OPERATION
~
~
10000
>0000 30000 40000 Average Planar Exposure {MNID/MT)
~
= ~
~
~
60000
~+R 0
lb.62 610 16.62 2,680 lg.lO 6230 1471 7.940 1$.19 l0,470 14.13 13,220 14.06 l6.990 14.06 18,780 14.00 21,690 13.93 24,420 13.93 27,280 13.08 30,160 12.24 33,060 11.40 3b,960 10.47 38.900 9.bb 41.830 8.66 44,760 7.77 FIGURE 7.1 LINEAR HEAT GENERATION RATE (LHGR) LINIT VERSUS AVERAGE PLANAR EXPOSURE, ANF SXS FUEL U rl PP I
tD LD 8 O I
lA O ON
I il I
ANF-90-02 Page 31 9.0 9.1 9.2 ADDITIONAL REFERENCES S.
F.
- Gaines, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Rl dF I,"~X---,EI I,EN I
Cp
- Richland, WA 99352, January 1982.
R.
H.
- Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling Ilk R
I,"X~N-Ny- ->>,R 11 2,E N
I 4
p y,
- Richland, WA 99352, November 1981.
9.3 J.
E.
- Krajicek, "WNP-2 Cycle 6 Plant Transient Analysis,"
ANF-90-01, Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, January 1990.
E.R d.
E.
X
'I k,
"LOCA B k Ep f
BIIR 5," X~1, E
Nuclear
- Company, 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 H.
H.
- Smith, "Generic Hechanical Design for Exxon Nuclear Jet Pump BWR Rl d
F I,"
~XB-Ny- ->>,
R 11 I, Eppl I I, 1
N Company, Inc., Richland, WA 99352, March 1985.
~
~
9.7 "Exxon Nuclear Methodology for Boiling Water Reactors-Neutronics Methods f
ld dA I i,'"~,
I d ppl, E
Nuclear Company, Inc., Richland, WA 99352, May 1980.
9.8 "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"
~XN-NF-E-FA, R
11 I,
E N
I 4
p y,
I
., Rill d,
NA
- 99352, September 1986.
9.9 9.10 9.11 "Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods f
P ld A
ly I,"
~XN-NF-
-I A, 21 I,
Eppl I
d 2,
Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.
J.
B.
- Edgar, Letter to
- WPPSS, Supplemental Licensing Analysis
- Results, ENWP-86-0067, Exxon Nuclear
- Company, Inc.,
- Richland, WA
- 99352, April 15, 1986.
J.
E. Krajicek, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction,"
XN-NF-87-92 and XN-NF-87-92, Supplement 1,
Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, June 1987 and Hay 1988.
- 9. 12 J.
E.
- Krajicek, "WNP-2 Single Loop Operation Analysis,"
ANF-87-119, Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, September 1987.
- 9. 13 J.
E. Krajicek and T. Tahvili, "WNP-2 LOCA Analysis For Single Loop Operation,"
ANF-87-118, Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, September 1987.
I ANF Page
- 9. 14 Letter, R.. B.
- Samworth, USNRC, to G.
C.
- Sorensen, WPPSS,
Subject:
Issuance Of Amendment No.
62 To Facility Operating License No.
NPF-21-WPPSS Nuclear Project 2
August 5, 1988.
ANF-90-02 Page A-l APPENDIX A 9X9-IX AND 9X9-9X LEAD FUEL ASSEHBLIES (LFAs)
A.I INTRODUCTION Evaluations were performed for two ANF 9x9-IX and two ANF 9x9-9X Lead Fuel Assemblies (LFAs) in the WNP-2 Cycle 5 core (Reference A. 1)* consistent with ANF methodology
("Exxon Nuclear Hethodology for Boiling Water Reactors,"
XN-NF-80-19). This evaluation established a licensing basis for these LFAs in the Cycle 5 core:
The discussion which follows summarizes the evaluation for the Cycle 6
core and updates the information provided for Cycle 5
where required to confirm that it remains applicable to WNP-2 Cycle 6.
The four ANF 9x9 LFAs in the Cycle 6 core will have negligible effects upon core wide transient performance, because they represent such a
small
~
~
~
~
fraction of the core and are not in high power locations.
Cycle specific analyses for these LFAs have confirmed that the Cycle 6
8x8 fuel HCPR operating limits remain applicable to the LFAs.
Fuel specific LHGR and HAPLHGR limits that were developed in Reference A. 1* for these LFAs are presented in this appendix.
A. 2 NUCLEAR DESIGN The average enrichment and enrichment distribution for the 9x9-IX and 9x9-9X fuel assemblies have been selected to match, as closely as possible, the neutronic performance of the four 8x8 XN-3 2.64 w/o U-235 reload assemblies included in the Cycle 5 reload.
Thus for Cycle 6,
the LFAs are loaded in locations which are eighth core symmetric to the four 8x8 XN-3 assemblies loaded in Cycle 5.
A comparison of the neutronic parameters for the ANF 9x9 LFAs to the parameters for the ANF 8x8 XN-3 assemblies is given in Appendix A of Reference A.l*.
- A.1 J.
E. Krajicek, "WNP-2 Cycle 5 Reload Analysis," ANF-89-02, Revision I,
~
~
~
~
~
Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, Harch 1989.
ANF Page The LFAs were included in the core-wide stability analysis reported in Section 4.2.4.
Local instability tests were performed on 9x9 leqds in a
BWR-3, ORNL/TH-9054; no detectable difference was noted in stability performance relative to the co-resident 8x8 fuel.
The control rod withdrawal error analysis was performed for the Cycle 6
core containing the 9x9 LFAs.
The delta CPR and the ANF HCPR operating limit values for cycle exposures up to 3750 HWD/HTU, reported in Sections 5.5 and 5.7, respectively, conservatively apply to all fuel in the core including the 9x9 LFAs.
A.3 ANTICIPATED OPERATIONAL OCCURRENCES The analyses of the WNP-2 Cycle 6 limiting pressurization transients have explicitly included ANF 8x8, ANF 9x9 LFAs, and GE P8x8R fuels.
The calculated delta CPR values for the ANF 9x9 LFAs for Cycle 6 follow.
The delta-CPRs calculated for the 9x9 LFAs are slightly higher than those calculated for ANF Bx8 fuel due primarily to the shorter fuel time constant of the small diameter fuel rods in the LFAs.
Transient*
Equipment 0 erational Status
/ Power/
/ Flow ANF 9x9 LFAs Delta CPR 9X IX Fuel Fuel LRNB LRNB LRNB LRNB FWCF FWCF RPT Operable, NSS RPT Inoperable, NSS RPT Operable, TSSS RPT Inoperable, TSSS RPT Operable, NSS RPT Inoperable, NSS 104/106 104/106 0.31 0.31 0.36 0.36 104/106 47/106 47/10 0.41 0.42 0.27, 0.27 0.33 0.35 104/106 0.37 0.38
ANF-90-02 Page A-3 The XN-3 correlation is being used to monitor the 9x9 LFAs.
This correlation has been evaluated to show sufficient conservatism in application to the LFAs such that the increased transient delta
- CPRs, relative to the
- 8x8s, are more than offset.
The Cycle 6 Safety Limit Analysis considered the LFAs such that the HCPR safety limit of 1.06 is also applicable to the LFAs.
Therefore, the ANF 9x9 LFAs can continue to be monitored to the ANF Bx8 fuel limits which are summarized below.
Event LRNB LRNB LRNB LRNB Equipment 0 erational Status RPT Operable, NSS RPT Inoperable, NSS RPT Operable, TSSS RPT Inoperable, TSSS Recommended HCPR Limits for 9x9 LFAs 1.31 1.36 1.36 1.40 A.4 POSTULATED ACCIDENTS Since heatup is primarily a
planar and not an axial phenomena, the appropriate bundle power limit that is derived from a
LOCA analysis is the peak bundle planar power.
The ANF 9x9 LFAs have better cooling during LOCA conditions relative to an ANF 8x8 fuel assembly due to the lower stored energy in the fuel
- rods, a greater surface area provided by the larger number of, fuel rods, and more inert surface from the central water channel.
- Thus, a
LOCA analysis for the ANF 9x9 LFAs would yield lower Peak Cladding Temperatures (PCTs) and metal-water reactions than an ANF Bx8 assembly at the same bundle peak planar power.
The HAPLHGR limits for the ANF 9x9 LFAs restrict the peak bundle planar power to that analyzed for the ANF Bx8 fuel and assure that the USNRC criteria are met for the ANF 9x9 LFAs in Cycle 6.
A. 5 TECHNICAL SPECIFICATIONS All operational limits used for ANF 8x8 fuel are applicable to the ANF 9x9 LFAs except for fuel type specific MAPLHGR limits and the 9x9-IX and 9x9-9X LHGR limits.
The LHGR limits for. the 9x9-IX and 9x9-9X LFAs are shown 4
~y r
ANF Page in Figures A. 1 and A.2 respectively, and the HAPLHGR limits for the LFAs are shown in Figure A.3.
The numerical values of Figure A.3 are 0.861 (62/72) times the HAPLHGR values of Section 7.2. 1.
The LFA single-loop operation (SLO) limits are bounded by the two-loop operation limits.
Al QO CX:
CD Ol CQ 70, 5.5 0
io 20 30 40 PLANAR EXPOSURE, GMD/MTU 50 60 URll
'I Q5 LD 70 CJl M FIGURE A.l LHGR LINIT FOR 9X9-IX FUEL
13.1Q OJ l
QO tX CD Ql CO 15.5, 13.1 70, 6.1 0'0 50 20 30 40 PLANAR EXPO GND/MTV FIGURE A. 2 LHGR LI IT FOR 9X9-9X FUEL 60 70
5 0
10 15 20 25 30 ASSEH8LY AVERAGE BURNUP, GMD/NTH FIGURE A.3 ANF 9X9-IX AND 9X9-9X HAPLHGR LINITS
~ g 1
fl r
t II I
H
ANF-90-02 Issue Date: 1/15/90 WNP-2 CYCLE 6 RELOAD ANALYSIS Distribution:
D. J.
Braun 0.
C.
Brown R.
E. Collingham R. A. Copeland W. S. Dunnivant L. J.
Federico J.
G.
Ingham S.
E. Jensen F.
B. Skogen/R.
B. Stout J.
E. Krajicek S.
L. Leonard J.
L. Naryott L. A. Nielsen A. Reparaz R.
S.
Reynolds L. G. Riniker G. L. Ritter H.
E. Williamson Y. U. Fresk/WPPSS (51)
Document Control (5)
I I