ML20065K696
| ML20065K696 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 10/31/1990 |
| From: | Charles Brown SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML14184A759 | List: |
| References | |
| ANF-89-164(NP), NUDOCS 9012040097 | |
| Download: ML20065K696 (86) | |
Text
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ADVANCED NUCLEAR FUELS CORPORATION uw.
MECHANICAL LICENSING REPORT FOR H.B. ROBINSON HIGH THERMAL PERFORMANCE FUEL ASSEMBLIES OCTOBER 1990 i
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ADVANCED NUCLEAR FUELS CORPORATION ANF 89 164(NP)
Revision 0 Issue Date: 10/26/90 MECHANICAL llCENSING REPORT FOR H.B. ROBINSON 111GH THERMAL PERFORMANCE FUEL ASSEMBLIES Compiled by:
J hhqf C. A. Brown, Manager PWR Design Fuel Design October 1990 icp
CUSTOWER OtSCLAIMER tesPORTANT NOTICE REGAMOING CohTENTS AND Utt CP THIS DOCunstNT M.IAss atAD CAREPULLY Advanced Nucteer Puees Corporacon's warrerines and representatens con.
coming me suotect mener of me occument are those set fortn in tne Agreement between Advanced Nucteer Fuees Corporaten and me Catomer cursuant to weien me document e issued. Ms.v;p, exceot as otnervnse excressay pro.
need in sucn Agreement. neitner Acvenced Nuclear Fuese Corporation nor any person acong on Rs bened mance any werranty or reorosentaten, encrossed or imodied, witn roepect to tne accuracy, completeness, or usefuiness of tne infor.
menon contemec in the docurnent, of that the use of any mtorrnation. accarstus.
metnod or process ciacesed m mis cocument edi not mfnnge onvateiy ownec ngryte or annumes arty 16anditme with respect to me use of any information. ao-paratus, metnod or procese ciacioeod in mis occument.
The insormenon contemed hereen a for me sees use of Catomer-in omor to awed imperment of ngnia of Advanced Nucteer Puese Corporaten m patente or irtwencone wnicn may os inciuced in the informanon contamec e tnis occument, me recoorst. Dy its acceptance of this cocument, agrees not to puoleen or make pulpic use (m the retent use of tne term) of sucn mtormacon until so aumorued m wnnng Oy Advanced Nucteer Fueis CorDoration or untd after sin (6) montne todoweg termmanon or expiration of tne aforesaic Agreement anc any extensen thereof, uniese otnerwise expreesty proviceo en me Agreement. No ngnts or ticonses e of to any ostents are amoiied Dy me fumisning of inis cocu.
ment.
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- ANF-89-164(NP)
Revision-0:
Page.i 5-
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- f.
-TABLE OF CONTENTS SECTION-
?.ME -
l.0" INTRODUCTION
........=..-.................-..
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-2.0^
SUMMARY
2 3.0' DESIGNLDESCRIPTION
-4 3 li Fuel. Assembly Description............ -.... >......
4-3s2: Fuel lRodDescription=....._....e.... -
5-
- 4.0- FUEL-ASSEMBLY ENVIRONMENT...-..-......'.....,.....
17
. 4.1' Reactor Coridi tions _ for Design.. _...... -..
-4.2: Steady State Power Historles.
....... -17L 4
- 4;31, Power Ramps.-....__
18
'4244 0uty Cycles...
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18-
'5'O :0ESIGN EVALUATION.
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(5;1L: Fuel System 0amagel Evaluation ' :. -.. --:...
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J5.1.2-l Steady-State-Strain..--_.._._.. c.
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LS.I.3.. Cyclic: Fatigue 28 8
5.1.'4= Fretting Wear...._...._,...-,............,
30 5.1.5$ExternalCorrosion-...............-
.34 '
5.1.6-Rod: Bowing
..-.36 l5.1.7: Axial' Growth _..................-...........
36' 1
ANF-89-164(NP)
Revision 0 Page ii TABLE OF CONTENTS (CONTINUED)
SECTION EaQ1 5.1.8 Fuel Rod Gas Pressure.......=................
37 5.1.9 Assembly Liftoff 37 5.2 Fuel Rod Failure Evaluation.........,
38 5.2.1 Internal Hydriding 38 5.2.2 Creep Collapse 38 5.2.3 Overheating of Cladding......................
39 5.2.4 Overheating of Fuel. Pell ets....................
39 5.2.5 Transient Stress and Strain....................
39 5.2.6 Cladding Rupture 40 5.2.7 Fuel Rod and Assembly Damage from External Forces.........
41 6.0 TESTING, INSPECTION, AND SURVEILLANCE PLANS.............
74 6.1 Testing and Inspection of New Fuel 74 6.2 Post-Irradiation Surveillance..............
74
7.0 REFERENCES
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ANF-89-164(NP)
Revision 0 u-Page lii 2
LIST OF TABLES -
laELE Eani
- 3.2 FUEL R00 CHARACTERISTICS 7
14.1 DUTY CYCLES.....
1.
20 15.1'
SUMMARY
OF' LIMITING-STEADY-STATE _ CLADDING STRESSES 47 5.2 ' MAXIMUM RAMP: STRESS AND STRAIN.
SUMMARY
- 5.3 - CALCULATED. VALUES, FOR STEADY-STATE STRAIN,
.CORROSIONi-AND HYDROGEN ABSORPTION.....
49 5.4 SPACER CHARACTERISTICS AT 0PERATING TEMPERATURE OF 650'F..,..:. ; 50 I UI0t TUBE COMBINED SEISMIC LOCA' STRESSES...............
51
- 5.5 G
5.6L FUEL =RODiCOMBINED' SEISMIC-LOCA STRESSES...-.........,..
52 1
.1
'i
ANF-89-164(NP)
Revision 0 Page iv LIST OF FIGURES FIGURE EASI 3.1 FUEL BUNDLE ASSEMBLY 306719.....................
8 3.2 CAGE ASSEMBLY 306718 10 3.3 SPACER ASSEMBLY HTP 306652 13 3.4 INTERMEDIATE FLOW MIXER 306656 14 3.5 U02 FUEL ROD ASSEMBLY 304545 15 UO -Gd 023 FUEL R0D ASSEMBLY 305634 16 3.6 2
4.1 U02 DESIGN HISTORIES: R0D POWER,.........,,.......
21 UO 0ESIGN HISTORIES: PEAK PELLET..................
21 4.2 2
4.3 NAF DESIGN HISTORIES: R00 POWER 22 4.4 NAF DESIGN HISTORIES: PEAK PELLET,,................
22 5.1 UO2 ROD STRAIN ANALYSIS: CLAD CREEPDOWN...............
53 5.2 NAF R00 STRAIN ANALYSIS: CLAD CREEPDOWN~...............
53 5.3 FATIGUE DESIGN CURVE FOR IRRA.DIATED ZIRCALOY 54 5,4 PWR SPACER R00 WITHDRAWAL FORCE...................
55 5.5 PWR TOP SPACER ROD WITHDRAWAL FORCE.................
56 5.6 HTP DEFLECTION ANALYSIS - UO2 - NOMINAL.........
57 5.7 HTP DEFLECTION ANALYSIS - UO2 - DESIGN CASE..........
58 5.8 HTP OEFLECTION ANALYSIS - NAF NOMINAL............
59 5.9 HTP DEFLECTION ANALYSIS - NAF - DESIGN CASE...
60
B ANF-89-164(NP)
Revision 0 Page v LIST OF FIGURES (CONTINUED)
FIGURE pag {
5.10 ANF FRETTING WEAR TEST DATA 61 5.11 UO2 ROD STRAIN ANALYSIS: HAXIMUM CORROSION.............
62 5.12 NAF R0D STRAIN ANALYSIS: HAXIMUM CORROSION.............
62 5.13 PWR R00 GROWTH...........................
E3 5.14 PWR ASSEMBLY GROWTH 64 5.15 U02 GAS PRESSURE HISTORIES:.00 POWER 65 5.16 002 GAS PRESSURE HISTORIES: PEAK PELLET 65 5.17 UO2 GAS PRESSURE ANALYSIS: FISSION GAS RELEASE,..........
66
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5.18 U02 GAS PRESSURE ANALYSIS: GAS PRESSURE 66 5.19 UO2 R00 GAS PRESSURE ANALYSIS: FUEL-TEMPERATURE 67 5.20 NAF GAS PRESSURE-HISTORIES: R00 POWER 68 5.21 NAF GAS. PRESSURE HISTORIES: PEAK PELLET 68 5.:22 NAF GAS PRESSURE ANALYSIS: FISSION GAS RELEASE,
69 5.23 NAF GAS PRESSURE ANALYSIS: GAS PRESSURE 69 5.24 NAF GAS PRESSURE ANALYSIS: FUEL TEMPERATURE 70 5.25 PRESSURE LOSS COEFFICIENTS FOR SMV AND HTP FUEL ASSEMBLIES.,....
71 5.26 SPACER DYNAMIC STRENGTH 72 5.27 SPACER-GUIDE TUBE CONNECTION AX1AL LOAD DEFLECTION TEST....
73
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ANF-89-164(NP)
.evision 0 Page 1 MECHANICAL LICENSING REPORT FOR H.B. ROBINSON HIGH THERMAL PERFORMANCE FUEL ASSEMBLIES
1.0 INTRODUCTION
The purpose of this, report is to provide the design description and evaluation of the Advanced Nuclear Fuels Corporation (ANF) 15x15 high thermal performance fuel assemblies for the H,B Robinson reactor starting with Reload ANF-11.
The report describes the revised fuel assembly, which utilizes high thermal performance spacers (HTPs), intermediate flow mixers (IFMs) and a small hole debris resistant lower tie plate.
The U02 and U0 -Gd 023 fuel rods 2
are unchanged from previous reloads.
The results of the mechanical analyses are provided to show compliance with the requirements of Section 4.2 of the Standard Review Plan (NUREG 800).
The fuel was evaluated for a peak assembly burnup value of 52.5 GWd/MTV and for an increased F limit of 1.79 and a 4H Fg limit of 2.55.
The evaluation of the rods and fuel assembly was performed in accordance with the USNRC approved methodologyl supplemet.'.ad with calculations in accordance with referenced documents ANF-88 133(P) and ANF 89-060(P)2,3, 1
ANF-89-164(NP)
Revision 0 Page 2 2.0 SUMARY The mechanical analysis shows the fuel assembly will meet the design criteria for the expected operating conditions and postulated accidents to the design assembly burnup of 52.5 GWd/MTV.
The key results of the fuel system damage evaluation are:
The maximum steady-state cladding and assembly component stresses are within the ASME Boiler and Pressure Vessel Code limits.
The maximum steady-state cladding strain is well below the 1%
design limit.
The cladding and assembly component fatigue usage factors are below the design limit.
Fretting wear of the spacers and fuel rods is precluded.
Corrosion of the fuel rod and the fuel assembly structural components is below the design limit.
Fuel rod bowing will be limited so that it has no impact on thermal margins.
Axial growth of the fuel rods and fuel assemtily is accommodated within the design clearances.
The fpel rod internal pressure remains belou the ' reactor system pressure throughout life.
Fuel assembly liftoff will not oc.eur during normal operation.
The fuel rod failure evaluation shows that the fuel rod design will operate without failure during normal operation and anticipated transients, and the fuel performance is properly accounted for in the accident analyses.
ANF-89-164(NP)
Revision 0 Page 3 The key results of the fuel rod failure evaluation are:
Internal hydriding is prevented by controlling hydrogen in the manufagture of the fuel.
Evaluation of the fuel rod behavior shows that cladding creep collapse will not occur.
Adequate cooling exists to prevent overheating of the cladding.
Fuel melting will not occur during normal operation and anticipated operational occurrences (A00s).
The transient circumferential strain is within the If, design limit.
The impact of cladding rupture is incorporated in the 1.0CA analysis.
Fuel coolable geometry and the capability for control rod insertion is shown to be maintained during seismic and LOCA events.
e ANF-89-164(NP)
Revision 0 Page 4 3.0 DESIGN DESCRIPTION The H.B. Robinson 15x15 fuel assemblies each contain 204 fuel rods,1 instrument tube, and 20 guide tubes.
The fuel assembly differs from previous reloads 5 in that six of the seven standard bi-metallic spacers are replaced 4
by all Zircaloy high thermal performance (HTP) spacers and 3 intermediate flow mixers (IFMs) are added.
The.. movable stainless steel upper tie plate with Inconel 718 leaf springs is unchanged from earlier reloads while the lower tie plate is replaced with a small hole debris resistant design.
3.1 Fuel Assembly Descriotion The high thermal performance (HTP) fuel assembly for H.B.
Robinson incorporates the same fuel rod design as provided in previous reloads.
The fuel assembly cage is revised, whereby, at six of the seven spacer locations, the bi-metallic grid spacer is replaced by an all Zircaloy HTP spacer.
At three mid span locations, in limiting DNB regions toward the top of the fuel assembly, intermediate flow mixers are added.
The bi metallic spacer at the bottom of the assembly is modified so that the fuel rod backup dimples cover the full spacer periphery, for increased resistance to handling damage.
The fuel assembly outline is shown in Figure 3.1.
The cage canstruction is shown in Figure 3.2.
The HTP spacer is shown in Figure 3.3, the IFM in Figure 3.4.
There are several design differences that make the HTP and IFM grids an improvement over the ANF bi-metallic design.
The change from an Inconel spring to a Zircaloy spring reduces the neutron capture cross-section of the spacers.
The internal strips of both HTP and IFM grids contain slanted channels that create a swirling flow pattern in the coolant.
This flow pattern increases coolant mixing and thus improves the heat transfer from the fuel rods. Also, the HTP spacer is structurally stronger than the bi-metallic spacer.
The stainless steel lower tie plate was revised by changing from large to small flow holes.
This significantly increases the debris resistance of I
l
ANF-89-164(NP)
Revision 0 Page 5 the fuel assembly by limiting the size and quantity of potentially damaging foreign objects which can pass through the lower tie plate and lodge in the fuel rods.
The upper tie plate, the guide tube latching system, and the holddown spring are unchanged.
To accommodate differential guide tube to fuel rod growth to higher burnup, more space was provided between the upper and lower tie plates.
This was achieved by removing 0.2 inch from the legs of the lower tie plate and adding 0.2 inch to the guide tube length.
Table 3.1 summarizes the characteristics of the fuel assembly.
3.2 Fuel Rod Descriotion The fuel rod is unchanged from earlier H.B. Robinson reloads.
The fuel rod cladding is cold worked and stress relieved Zircaloy 4 tubing.
Each fuel rod contains a center column of either enriched V02 or enriched UO Gd 023 fuel 3
pellets with an upper and lower column of natural UO2 pellets that act as axial blankets.
The pellets are cold pressed and sintered to 9e of the theoretical solid density.
The upper plenum region contains an inconel X-750 compression spring.
The upper and lower ends of "m rods are seal welded with Zircaloy-4 plug-type end caps.
The fuel rn, are pressurized with helium gas.
Figure 3.5 is a drawing of the U02 fuel rod assembly.
Figure 3.6 is a drawing of the V0 -Gd 023 fuel rod assembly.
The U0 -Gd 023 fuel rod assembly 2
2 differs from the U02 rod in that the axial blankets account for 12 rather than 6 inches-at the stack ends.
Table 3.2 summarizes the characteristics of the fuel rods and pellets.
ANF-89-164(NP)
Revision 0 Page 6 TABLE 3.1 FUEL ASSEMBLY CHARACTERISTICS Characteristic Array 15x15 Number of Fuel Rods 204 Number of Guide Tubes 20 Instrument Tube 1
Fuel Rod Pitch (in) 0.563 Rod to Rod Spacing (in) 0.139 Rod to Guide Tube Spacing (in) 0.079 Number of Bi Metallic Spacers 1
High Thermal Performance (HTP) Spacers 6
Intermediate Flow Mixers (IFMs) 3 Guide Tube OD (in) 0.544 Guide Tube 10 (.in) 0.511 Dashpot 10 (in) 0.455 Spacer Nominal. Envelope (in)
Lower Tie Plate Envelope (in)
.8.424 Upper Tie Plate Envelope (in) 8.406 Overall Length less Spring _(in) 159.710 Overall length (in) 161.300
ANf-89-164(NP)
Revision 0 Page 7 TABLE 3.2 FUEL ROD CHARACTERISTICS Claddina yAlqg 0.0. (in) 424 1.D. (in)
.3640 Length (in) 151.64 Fuel Column Pellet 0.0 (in)
.3565 Percent.of Theoretical Density (%)
94.0 Enriched Length (in)
UO2 Rods 132.0 UO -Gd 023 Rods 120.0 2
Axial Blankets Length, Each End (in)
UO2 Rods 6.0 UO -Gd 023 Rods 12,0 2
Total Active length (in) 144.0 Rad Assembiv Total Fuel Mass (kg) 002 Rods 2.403 V02 w/4% Gad 2.375 UO2 w/6% Gad 2.361 Fill Gas Pressure (psia He) 305
- 4
ANF-89-164(NP)
Revision 0 Page 8 Pages 8 - 16 have been deleted.
l
ANF-89-164(NP)
Revision 0 Page 17 4.0 FUEL ASSDBLY ENVIRONMENT This section summarizes the reactor conditions used in the evaluation of the fuel rods and fuel assembly including the steady-state, transient, and cyclic conditions.
4.1 Reactor Conditions for Desian Core thermal power 2300 MWth System pressure 2250.0 psia Number of assemblies 157 Nominal measured core mass flow rate 109 x E6 lb/hr Maximum core flow for holddown analysis 113 x E6 lb/hr Flow fraction for heat transfer
.955 Core inlet temperature 546.3*F Core outlet temperature 603.0*F Maximum overpower 187.
Fraction of heat from fuel reds
.974 Core average LHGR 5.98 kW/ft Maximum allowable pellet factor, Fg 2.55 Maximum allowable rod factor, F I'79 AH Peak assembly burnup 52.5 GWd/MTV 4.2 Steady-State Power Histories Pcwer histories were created consistent with the methodology of Reference 1.
Four power histories were used in the mechanical design evaluation of the U02 fuel rods.
These histories represent the projected high power rods in each of the three cycles, and the high power rod in the peak exposure fuei assembly.
Three power histories were provided for use in the analyses of the fuel rods containing gadolinia.
- Again, these histories represent the projected gadolinia bearing rods with the high power in each of the three cycles.
Each history follows a rod throughout its three-cycle residence in the reactor.
The histories are consistent with operation at limits of FAH "
1.79 ' and Fg - 2.55 and a peak assembly exposure of 52.5 GWd/MTV.
Each of these design histories has been plotted and the following list specifies the figure where it can be found and the index used as its designator.
ANF-89-164(NP)
Revision 0 Page 18 Rod History Deseriotion f,11gr.g J,nitz High First Cycle U02 4.1, 4.2 1
High Second Cycle, V02 4 ), 4.2 2
High Assembly Exposure, V02 4.2, 4.2 3
High Third Cycle UO2 4.1, 4.2 4
-High First Cycle, NAF*
4.3, 4.4 1
High Second Cycle, NAF 4.3, 4.4 2
High Third Cycle, NAF 4.3, 4.4 3
These histories are analyzed with the RODEX2 code for the design. criteria pertinent to steady-state operation, and are used to establish the initial conditions for power ramps.
4.3 Power Ramos In order to - determine the stress and strain of the cladding during
- transients, ramps to maximum -LHGR conditions are added to the steady state power histories at various times during. the irradiation.
4.4 Duty Cycles The fatigue usage f9ttors are determined for the design cycling conditions using the stresses determined in the ramping evaluation.
Neutron Absorber Fuel (U0 -Gd 0 )
2 23 l
ANF-89-164(NP)
Revision 0 Page 19
ANF-89-164(NP)
Revision 0 Page 20 TABLE 4.1 DUTY CYCLES 1.
Current and Anticioated Practice a.
Weekly valve operating test 100% to 70%
hold @ 70% for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 70% to 100% 9 PREMACX rate b.
Twice/ month Steam Generator leakage test 100% to hot standby hold 9 hot standby for one day hot standby to 30%
hold 0 30% for 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 30% to-100% 0 PREMACX rate c.
Once/6 months-Steam Generator inspection 100% to O power cold for one week O power to 30%
hold 0 30% for 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 30% to 100% 0 PREMACX RATE 2.-
Load Follow iga % to 50% to 100% 0 PREMACX rate - 2/ day 3.
Arbitrary Cycles a.
10 scrams / year b.
Step load decrease of 95% - 2/ year c.
Step load increase from 0 power to 30% - 2/ year d.
Step load increase of 20% power - 1/ week e.
30% to 100% @ PREMACX rate - 2/ year
ANF 89 164(NP)
Revision 0 Papt.11
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ANF 89 164(NP)
Revision 0 Page 22 I
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ANF-89-164(Np)
Revision 0 Page 23 5.0 DESIGN EVALUATION Fuel rod and assembly design analyses and tests were performed in accordance with the requirements of Chapter 4.2 of the Standard Review Plan (SRP) and the ANF design bases and methodology presented in References 1, 2 &
3.
5.1 Fuel System Damaae Evaluation The following paragraphs discuss the ability of the high thermal performance fuel assemblies to meet the fuel system damage criteria described in Section 3.1 of SRP 4.2.
The criteria apply to normal operation and anticipated transients.
5.1.1 Steady State Stress a)
Fuel Rod The cladding steady state stress analysis was performed considering primary and secondary membrane and bending stresses due to hydrostatic
- pressure, flow induced vibration,
- ovality, spacer contact.
pellet clad interaction (PCI),
thermal and mechanical
- bow, and thermal gradients.
Stresses were calculated for various combinations of the following conditions and locations:
BOL (beginning of life) and EOL (end of life)
Cold and hot conditions At mid span and at spacer locations At both the inner and outer surfaces of the cladding The applicable stresses in each orthogonal direction were combined to l
calculate the maximum stress intensities which were then compared to the riesign criteria 6 The results of the analysis indicate that all stress values -
are within acceptable design limits for all combinations of conditions and locations. A summary of the stress intensity results is shown in Table 5.1.
8
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t ANF-89-164(NP)
Revision 0 Page 24 a
b)
Guide Tubes The ' control rod guide tubes tie the assembly structure together_ and provide channels for the ' insertion of the control rods.
The bottom section of the guide tube consists of a reduced diameter to produce a dashpot action when-the control rods approach the end of their travel during a reactor trip.
The
' guide tube stresses due to differences in frictional loads between HTP and bi-metallic spacers _and the fuel rods which result from differential thermal s
expansion- ~ between' the fuel rods and the guide tubes are examined.
The stresses due'to-control rod insertion are-unchanged-from the reference analysis 7.-
i Axial Stresses:
L As the power-level.of the reactor is increased, differential thermal expansion between the Zircaloy guide tubes and the hotter Zircaloy clad fuel tends to put.the guide tubes in tension.
Af terL a period at power, vibration loads tend to reduce or eliminate differential thermal expansion loads.
With a reduction _in power, di_fferences in temperature between-the guide _-tubes and fuel ro'ds decrease and cause compression loading-on the guide tubes.
' The magnitude of_ guide tube loading depends on the differential expansion-
.i between_ the fuel rods.and guide tubes -and upon whether or not the fuel-rods
, slip throughtthe spacers.
If the force developed by the differential thermal' upansion= times the guide tube stiffness exceeds' the friction force. -slippage occurs first at the outer spacers with the friction force'being exerted on the inner spans.
The spacer friction forces are v.:umulated toward the center of the assembly -until the accumulated. force - equals the expansion-force.
The weight of the assembly and-the holddown force are also considered.in the calculations.
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ANF-89-164(NP)
Revision 0 Page 25 t
Buckling:
As the power level in the reactor i s reduced, the difference in temperature between the guide tubes and fuel rods 1ecreases causing
. compress ve s resses resulting - in a buckling type load in the guide tube i
t spans.
The differential thermal expansion was conservatively considered as an external load and-not limited by differential strain.
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ANF-89-164(NP)
Revision 0 Page 26 l
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Soacer The spacer grid attachments must withstand-the frictional forces '.due to:
the differential _ thermal expansion between the fuel rods and the guide tubes.
The grids are attached to the guide tubes and instrument tubes by resistance spot welds.
The' HTP spacer represents the limiting case because-its. friction forces are greater than those in the IFM, and because the HTP spacers are-
ANF-89-164(NP)
Revision 0 Page 27 attached to each guide tube in four places, whereas. the bi-metallic spacer is attached in eight places.
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ANF-89-164(NP)
Revision 0 Page 28 d)
Tie Plate Streng_th Structural tests have been made on both the upper and lower tie plates for the H.B. Robinson assemblies.
i 5.1.2 Steady-state Strain The cladding steady-state strain was evaluated with the approved RODEX2 8
code.
The code considers the thermal hydraulic environment at the cladding surfaces; the coolant and rod pressures; and the thermal, mechanical, and compositional state of the fuel and cladding.
The calculations are performed on a
time incremental basis with conditions updated at each calculated increment so that the power history and path dependent processes can be modeled.
The axial dependence of the power and burnup distributions are handled by dividing the fuel rod into a numter of axial and radial regions.
Power distributions can be changed at any desired time, and the coolant and cladding temperatures are readjusted in all the regions.
All the performance models, e.g., the deformations of the fuel and cladding and gas release, are calculated at successive times during each period of assumed constant power generation.
... ~....... - - -. - -.
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ANF-89-164(NP)
Revision 0 Page P.9 j
4 i
i 5.1.3 Cyclic Fatiaue a)
Fuel Rod The stress results from the ramping analysis (Section 5.2.5) were used to evaluate the cladding fatigue damage through EOL due to the cyclic power variations defined in f able 4.1.
For each of these repetitive operations, high and low reactor power levels were identified.
Fatigue damage was L
evaluated for-the specified number of repetitive operations as the reactor power _ is cycled back and forth between these two levels.
l L
.The transient stress results were evaluated to determine the ' fatigue-usage for each cycle-based on the O'Donnel-and Langer9 design curve shown in Figure' 5.3.
These results were accumulated to determine _ the total fatigue usage factor.which is tabulated in Table 5.2.
- b)
Guide Tube The-variation in stresses due to differential expansion at beginning of life 1(BOL) are-used in determining the fatigue. usage throughout,the life-of the' guide tubes.-
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Revision 0 Page 30 c)
Soacer The grid attachment welds must withstand cyclic stresses.
The operating duty cycles defined in Table 4.1 were used to estimate the number of cycles.
that wo_uld be applied to'the joint.
5.1.4-Frettina Wear I
a).
Bi-Metallic Soacer Each grid spacer is Ldesigned to support the fuel rod. without fretting wear'throughout the design life of the fuel.
For the'bi metallic ' spacer, this.
is assured - by~ ~ maintaining. a e positive spring force greater than the flow induced lateral vibration-forces.
. Spring Relaxation:,
During the irradiation of a fuel assembly there is a. gradual reduction of--
the holding force' exerted by. the Inconel spacer spring in the bi metallic spacer.
- The force reduction is due-- to irradiation inducidTspacer spring relaxation and to creepdown of _ the fuel; rod cladding. ~ The rate of change'is
.more rapid' early in burnup ?with' a gradualLdecrease in the rate of change throughout irradiation.
The: clad creepdown eventually reachesf a-minimum-L
. diameter whens firm pellet clad interaction 'is established.
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e
ANF-89-164(NP)
Revision 0 i
Page 31 1
1 Flow Vibration Force:
The spring force, F,,,
required to counteract the maximum lateral acceleration forces due to flow induced vibrations to the extent of preventing lift-off of the rod from the dimples 4
dspan calculated from Paidoussis's equation 10 El-Bending stiffness of clad
=
L Span length
=
9
ANF-89-164(NP)
Revision 0 Page 32 End of Life Bi-Metallic Spacer Spring Force:
i At the end-of-life, the spacer spring must overcome the force due to j
flow induced ' vibration.
The fuel rod cladding will have relaxed during irradiation so there are no forces due to thermal or mechanical bow.
- b)
HTP and IFM Soacers The prevention of. fretting corrosion in the HTP and IFM spacers is 3
demonstrated in accordance with the method!, of reference ANF 89 060 by a combination of analysis and fretting tests.
The design analysis determines n
the projected maximum end of life gap considering spring relaxation, clad creepdown, minimum fuel rod diameter, and minimitm initial spring deflection at the beginning of-life.
Flow' test data are used to confirm that fretting corrosion will not occur for the largest possible projected gaps.
r
pisi e
i e
ANF-89-164(NP)
Revision 0 Page 33 Interference Analysis:
The spring interference (or gap) throughout life is determined with an incremental calculation Flow Test Results:
Flow tests of up to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> have been used to determine the fretting characteristics associated with the bi metallic design for various combinations of spring / clad interferences (or gaps).
i ANF-89-164(NP)
Revision 0 j
Page 34 l
l i
i i
5.1.5-External' Corrosion
. a)
Fuel Rod The waterside corrosion and hydrogen pickup in the cladding were evaluated with RODEX2 for the seven power histories used in the steady state strain analysis.
-s y
v ww-r-.m
.-m.
.+5
-c,-
y
ANF-89-164(NP)
Revision 0 Page 35 b) fjtt) Assembly Cace The spacer and guide tube components have been analyzed for oxidation and 2
hydrogen pickup using the MATPRO corrosion prediction model in R00EX,
0 9
._._.___m____.____
ANF-89-164(NP)
Revision 0 Page 36 5.1.6 Rod Bowina The impact of rod bowing on the high thermal performance fuel assembly MDNBR values was evaluated 10 using the ANF Rod Bowing Methodologyll.
5.1.7 Axial Grpy,1h a)
Fuel Rods Axial extension of Zircaloy fuel rods in a reactor is primarily due to irradiation induced growth.
The assembly is designed to assure that an axial clearance will exist between the i'Je1 rods and the upper and lower tie plates at the end of life.
9 ANF-89-164(NP)
Revision 0 Page 37 b)
Fuel Assembly The minimum clearance between the tie plates and the core plates at beginning of life is 0.723 inches.
Fuel assembly growth data are based on ANF measured data shown in Figure 5.14.
5.1.8 Fuel Rod Gas Pressure Calculation of the gas pressure within a fuel rod is performed with the R00fX2 code.
Power histories for the gas pressure analysis were developed from each of the corresponding power histories shown in Section 4.0 in I
accordance with the USNRC approved methodology.
i 5
ANF-89-164(NP)
Revision 0 Page 38 5.1.9 Assembly Liftoff Since the holddown spring characteristics and the overall fuel bundle length for the high thermal performance design is the same as the bi-metallic fuel design, and since the HTP assembly pressure drop is slightly below that of the bi-metallic design, the existing holddown system for the 15x15 bi-metallic assembly design will provide ample holddown for the high thermal performance fuel.
5.2 Fuel Rod Failure Evaluation The following paragraphs discuss the ability of the high thermal fuel performance assemblies to operate without failure during normal operation and anticipated transients, and the accounting for fuel red failures in accident analyses.
5.2.1 Jatfrnal Hvdridina Hydriding as a cladding failure mechanism is precluded by controlling the level of moisture and other hydrogenous impurities in the rod during fabrication.
This is done by careful cleaning and drying of the clad, and by careful moisture control of the pellets during fuel fabrication.
l ANF-89-164(NP) l Revision 0 Page 39 5.2.2 Creeo collaose Creep collapse calculations were performed with the RODEX2 and COLAPXI2 codes.
4 collapse will not occur.
5.2.3 Overheatina of Claddina As stated in SRP tection 4.2, adequate cooling is assumed to exist when the thermal margin criterion departure from nucleate boiling ratio (DNBR) is satisfied.
The method employed to meet the DNBR design limit has been reviewed by the NRC as part of a thermal hydraulic codes and' methods.
The Cycle 14 reload analysis report gives the details of this analysis and results.
5.2.4 Overheatino of Fuel Pelle u The prevention of fuel failure from overheating of the fuel pellets is accomplished by ensuring that the peak linear heat generation (LHGR) during normal operation and anticipated operational occurrences (A00s) does not result in calculated fuel centerline melting.
This analysis is performed as part of the plant transient analysis (Chapter 15 analyses).
1
4 4
ANF-89-164(NP)
Revision 0 Page 40 5.2.5 Transient Stress and Strain a)
Normal Ooeration in order to determine the stresses and strains of the cladding during transients, ramps to maximum LHGR conditions were applied to the steady state histories eriodically throughout the irradiation.
The ramp LHGR levels were determined by assuming that a pellet from a rod at the F limit can be ramped to Fg, where F3g is the maximum allowable rod AH peaking factor in the core (1.79) and Fg is the maximum allowable pellet peaking factor in the core (2.55) l l-I r
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.v.,
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ANF-89-164(NP)
Revision 0 Page 41 5.2.6 Claddina Ruoture ANF's cladding rupture model for LOCA analysis is described in a generic report (XN NF 82 07) entitled, " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model."
This report adopted the NUREG 0630 data base and modeling.
The NRC has reviewed this report and found it acceptable.
The impact of rupture using this accepted model is implicit in the LOCA results reported in the reload analysis report for Cycle 14.
5.2.7 Fuel Rod and Assembly Damaae from External Forces The fuel assembly cage must assure that fuel assembly coolable geometry, and the ability to insert control rods, is maintained under anticipated accident conditions.
Axial and lateral loading conditions due to seismic and LOCA events have previously been analyzed and reported in the XN NF 76-47 (P)(A)15 for the 15x15 fuel assembly with bi metallic spac'ers.
Due to the incorporation of different fuel assembly components, namely HTP and IFM spacers and the debris resistant lower tie
- plate, the seismic LOCA evaluation reported in the reference has been readdressed.
The differences between the reference and the revised fuel assembly design have been analyzed to show that the characteristics of the new fuel design are equivalent to or improved over the reference design.
9
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ANF-89-164(NP)
!=
Revision 0.
Page 42 i
ti 1
)
a)
Core Evaluation 1
)
4 e
.. i L
b)
HTP Scacar
~The strength characteristics of the HTP spacer are compared with the data for the bi metallic spacer and the simulated Westinghouse spacer-considered in l
the' reference. analysis.in Table ' 5.4.
The comparative HTP and bi metallic j
spacer ' load-deflection characteristics are shown in Figure' 5.26.
r
- The grid spacer strength' was evaluated in accordance with 'SRP 4.2 l Appendix A guidelines whereby, the maximum' allowable crushing load is the'95?._
n
. confidence lower limit on. the meaniof1the crush test measurements.
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........, _ - -. ~ - -. - _. _ _ _,.... -... - -
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h ANF-89-164(NP)
Revision 0 Page 43 f
i I
c) Jiuide Tubes lateral Bending Stresses:
t I
n_
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s
...,,v,--.,
,,...,,,,.,,,.,,n,,
,,.. -.,,,,, -,, -,, -.,.,,,e,,
,,n,,..-
.g
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ANF-89-164(NP)
Revision 0 i
Page 44 1
l s
A-Combined Stresses:
t E
h
+mm now
,-rp.2
-m e
m
+e me-
- ee MI.ee m.
l ANF-89-164(NP)
Revision 0 Page 45 Buckling:
d)
Lower Tie Plate The purpose of the lower tie plate is to axially support the fuel assembly during normal and accident conditions, e)
Fuel Rod Stresses The use of HTP spacer cages may have a slight effect on the resultant rod stresses due to increased beginning of life fuel rod to spacer friction and a potentially larger effect due to the increased strength of the HTP grid spacer. To determine the effect of these factors on the fuel rod seismic LOCA stressest each stress component was individually examined.
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1 ANF-89-164(NP)
Revision 0 Page 46 S
1 p M-1
.-.-. - - - - -. -.-.... -.....~. -
~. -
P
~
ANF-89-164(NP)'
Revision 0 Page 47 t
i t
P t
TABLE 5.1 SUMMlRY OF LIM! TIN 0 STEADY-STATE CLADDING STRESSES
{
Primary Primary l
Membrat'.
Net Primary and Secondarv Location ligg itgig igtij d
M R
L(11),
R i
r 5
4
)
.. i i
1 I
Eniti:
MS = Criteria /Stresszin. tensity <
1 1,
Primary: Membrane Stress.
Criteriai= lower valueLof: 2/3;YJor 1/3 U-Y =. yield: strength'(ksi)
U = ultimate strength.(ksi).
\\.
Net Primary Stress.= Primary Membrane + Primary Bending
- 1...
- Criteria = lower value of Y or 1/2 U Primary t.Second'ary Stresses Criteria
-lower value of 2 Y or U
~
4 g-d y
9 -m-gm>
- e v s ra-me
-s--
v:
= = =e et et**-tm--w e sm e
---wi*mc***v=4
>e-=--%e-
=----+=dee-+
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=*=-e
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- v
ANF-89-164(NP)
Revision 0 Page 48 TABLE 5.2 MAXIMUM RAMP STRESS AND STRAIN
SUMMARY
Maximum Maximum Total Hoop Uniform Stress Strain Fatigue
.No. of Power History (ksil
(%)
Usaae Cycles BOC ramp with a pellet chip
O q-d ANF-89-164(NP)
Revision 0 Page 49 3
TABLE 5.3 CALCULATED VALUES FOR STEADY STATE STRAIN, CORROSION, AND HYDROGEN ABSORPTION Circumferential Peak Local Peak Hydrogen Power Outward Creep 0xide Absorption Historv S.ttfin (% 0. Rad d
_(mic rons)
(com) 4 9
9
l ANF-89-164(NP)
Revision 0 Page 50 TABLE 5.4 SPACER CHARACTERISTICS AT OPERATING TEMPERATURE OF 650'F.
Dynamic Oynamic Stiffness (1b/in)
Strenath (1bs) t
)
1 ANF-89-164(NP)
Revision 0 Page 51 TABLE 5.5 GUIDE TUBE COMBINED SEISMIC LOCA STRESSES Stress at lower end of Lo'ad Condition
($$
e Stress criteria.from ASME Boiler & Pressure Vessel Code,Section III' Division I, Appendix F, 1980.
ANF-89-164(NP)
Revision 0 Page 52 TABLE 5,6 FUEL R00 COMBINED SEISMIC LOCA STRESSES
-Load Condition S
Strgs si) 3 hoop Capability of HTP spacer.
Radial stress is assumed 0 psi for conservative stress _ intensity calculation.
1 J
ANF-89-164(Np)
Revision 0 page 53 Pages 53 :73 have been deleted, t
+
.o.
ANF-39-164(NP)
Revision 0 Page 74
.i 6.0 TESTING, INSPECTION, AND SURVEILLANCE PLANS 6.1' Testina and'Insoection of New Fut],
As described in'.SRP Section 4.2, testing and inspection plans - for new fuel should include verification of significant fuel design parameters. ANF's high thermal performance fuel design is tested and inspected during manufacture to comply with the engineering and. quality control requirements of ANF's reload parts list, drawing and specifications.
)
The-requirements of ANF's quality control program are provided in ANF 1, which addresses design,' process, quality, procurement and document.contrci applicable to design and manufacture - of. fuel system component parts, fuel pellets,. rods and. assemblies.
Fuel inspections. vary for the different component parts' and may include dimensions,. visual appearance, audits-of test I
reports, material certification. and - non-destructive examinations.
Pellet
. inspection 'is ~ performed for dimensional characteristics such as diameter, densit'y, length,'and squareness of ends.
Fuel rods. ' absorber. rods, upper and' lower - plates, _ 'and. spacer grid inspections ' consist of'-non destructive examination L techniques. such as leak -testing, weld inspection, ~ and dimensional l measurements.
Process; control. procedures are described in detail. 'In addition, for any.
tests. and ' inspections - performedi by other vendors, ANF reviews-the. quality control. procedures andLinspection plans to ensure-that they are equivalent to those' described. in ANF-1c and (are1 performed properly.
The requirements-are changed from those for - the existingiH.B.L Robinson fuel to tne 'high thcrmal performance fuel only-in the. dimensional and fabrication details -and inspection plans of the revised and new-HTP and IFM components.
- 6'2 Post frradiation Surveillance R-
+-
w
- er-
ANF-89-164(NP)
Revision 0 Page 75 assemblies which are currently in their first cycle will be examined after irradiation t) verify that the fuel assemblies are performing as anticipated.
1 l
ANF-89-164(NP)
Revision 0 Page 76. 0 REFERENCES.
" Qualification of-Exxon Nuclear Fuel for Extended Burnup (PWR)", XN NF-
- l..
82-06(P)(A),.Rev. 1 & Supplements 2, 4, 5, Exxon Nuclear Company, October 1986.;
2.
ANF_88-133(P), " Qualification of Advanced Nuclear Fuel's PWR Design'.
_ Methodology, for Rod Burnups of 62 GWd/MTV", Advanced Nuclear Fuels Corporation, September 1988.
_ 3.
~ ANF 89 060(P), ~" Generic Mechanical Design Report High Thormal Performance Spacer and Intermediate Flow Miur," Advanced Nuclear Fuels Corporation, April 1989.
4.
'XN 75-39, ' Generic Fuel Design-for 15x15 Reload Assemblies for~
Westinghouse P1 ants", Exxon Nuclear Company, September-1975.
- 5.
XN NF_-87-17(P), " Mechanical Design Report for -H.B. - Robinson - Extended D
Burnup Assemblies", Exxon Nuclear Company, March 1987.
6.-
.-ASME Boiler and Pressure Vessel Codes,Section III, 1968 Edition",-
American Society of Mechanical Engineers, 345_ East 47th Street, New-York, NY, 10017.
7?
~XN-75-25, "H.B.- Robinson Fuel Design Report Volume 1 Mechanical and-Thermal Hydraulic'Oesign for Cycle: 4", June 1975.
8.
XN NF-81-58(P), Rev.
2, Suppl ements.1-3, "RODEX2 - Fuel-Rod-Thermal Advanced Nuclear-Fuel s -'
w.
Mechanical-Response Evaluation' Model",
' ~
Corporation, July 1988.
9.
' W. J.. O'Donnel and 8.. F. Langer, " Fatigue Design Bases y for Zircaloy -
Components"', Nuclear Science and-Engineering, Volume 20, January 1964 L
- 10. AHF 89-165(P), " Thermal Hydraulic -. Compatibility ' Analysis -of -ANF High
~
L Thermal Performance Fuel for H.B. Robinson Unit -2",10ctober-1989. -
ll. L XN 75-32(A),. Supplements 1, 2, - 3. and 4,
" Computational Procedures-for.
Evaluating Fuel Rod Bowing", October 1983'.
- 12. 'JN-72-23,
" Cladding - Collapse Calculational Procedure", Jersey Nuclear
~ Company, November 1972..
- 13. XN NF-1068(P),
" Precondition and Maneuvering Criteria -for Advanced Nuclear Fuels' PWR Fuel (PREMACX-PWR)",. Advanced Nuclear F'uel s l
Corporation, June 1987.
I l
s m
-_--_7_____
o 4
ANF-89-164(NP)
Revision 0 Page 77 t
- 14. XN-NF-573, "RAMPEX: Pellet-Clad Interaction Evaluation Code for Power Ramps", Exxon Nuclear Company, April 1982.
15.
XN-76-47(P)(A), " Combined Seismic LOCA Mechanical Evaluation for Exxon Nuclear 15x15 Reload Fuel", January 1982.
I 5
ANF-89-164(NP)
Revision 0 Issue Date: 10/26/90 MECHANICAL LICENSING REPORT FOR H,B. ROBINSON HIGH THERMAL PERFORMANCE FVEL ASSEMBLIES Distribution HG Shaw/HB Robinson (1/10) 4