ML20247D299
| ML20247D299 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 03/14/1989 |
| From: | Cacciapouti R, Slifer B, Woehlke R YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20247D253 | List: |
| References | |
| YAEC-1683, NUDOCS 8903310033 | |
| Download: ML20247D299 (145) | |
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E b MICBURN-3/CABMO-3/TABIJCS 3/ SIMULATE-3 BENCHMARKING OF VERMONT YANEEE CYCLES 9 THROUGH IS March 1989 Principal Investigators: B. Y. Hubbard D. J. Morin J. Pappas R C. Potter 1 l Prepared By: MIN " R A. Wochlke, VY lead Engineer (Date) Reac^or Physics Group Approved By: /dd// J. Caccia 11. Manager '(Date) Reactor Physics Group Appmved By: M ' A B. C. Slifer, Directorh (Date) Nuclear Engineering Department Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 .
DISCLAIMER This document was prepared by Yankee Atomic Electric Company for its own use. 'Ihe use of information contained in this document by anyone other than Yr.nkee Atomic Electric Company is not authorized, and in regard to unauthorized use neither Yankee Atomic Electric Company or any of its ofDeers, directors, agents or employees assumes any obligation, responsibility or liability, or makes any warranty or representaiton, with respect to the contents of this document, or its accuracy or completeness.
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ABS'IRACT The MICBURN 3/CASMO-3/TABIES-3/ SIMULATE 3 code package is applied to thefive most recent cycles of Vermont Yankee, a General Electric BWR. The purpose of this benchmark is to demonstrate the suitability of these codes for plant support and licensing offuture Vermont Yankee cycles. Both cold critical startup and hot steady-statefull power operation are modeled. Both exhibit steady. consistent eigenvalues close to unity. Extensive 3 D compadsons are made between the model-calculated and plant-measured detector responses. The comparisons imply that the code package rproduces the power distributions of the plant with reasonable accuracy. The minor discrepancies that do exist are in a consenntive direction asfar as licensing applications of the codes are concemed. i
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TABLE OF CONTEN'IS Disclaimer .............................................11 Abstract ............................................. 111 List of Table s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v1 List of Figures ............................................vii Acknowledgements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix i 1.O INTRODUCTION AND
SUMMARY
.............................1 2.0 GENERAL CODE PACKAGE DESCRIF110N . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 MICBURN-3 ........................................2 2.2 CASM O-3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.3 TABLES- 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.4 SIMULATE-3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 VERMONT YANKEE OPERATING CHARACTERISTICS . . . . . . . . . . . . . . . . 7 3.1 Core Description .....................................7 3.2 Description of Cycles Modeled . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.3 Fuel Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.0 VERMONT YANKEE SPECIFIC MODEL , . . . . . . . . . . . . . . . . . . . . . . . . 16 4.1 MICBURN-3 Model Description . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.1.1 MICBURN-3 Sensitivity Studies . . . . . . . . . . . . . . . . . . . . . . . 17 4.2 CASMO-3 Model Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.2.1 CASMO-3 Depletion Cases . . . . . . . . . . . . . . . . . . . . . . . . . . 20 4.2.2 CASMO-3 Branch Cases . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.2.3 CASMO-3 Reflector Region Model . . . . . . . . . . . . . . . . . . . . . 21 4.2.4 CASMO-3 Sensitivity Studies . . . . . . . . . . . . . . . . . . . . . . . . 22
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4.3 TABLES-3 Construction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 4.4 SIMULATE-3 Model Description . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 4.4.1 SIMULATE-3 Hot Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 .. 4.4.2 SIMULATE-3 Cold Model . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.4.3 SIMULATE-3 Sensitivity Studies . . . . . . . . . . . . . . . . . . . . . . 27 5.0 VERMONT YANKEE PHYSICS MODEL RESULTS . . . . . . . . . . . . . . . . . . 41 k 5.1 Hot Model Eigenvalues . . . . . . . . . . ....................... 41 5.2 Cold Model Eigenvalues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 5.3 Hot Model Detector Comparisons . . . ....................... 43 5.3.1 Radial Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 , 5.3.2 Axial Average Comparison . . . . . . . . . . . . . . . . . . . . . . .... 45 5.4 Comparison of New Model to Current Model . . . . . . . . . . . . . . . . . . 46
6.0 CONCLUSION
S .......................................58
7.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 APPENDIX A Hot Depletion Statepoints . . . ....................... A-1 APPENDIX B Cold Critical Statepoints . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 APPENDIX C Channel Bowing at D-Lattice Plants . . . . . . . . . . . . . . . . . . . . C-1 APPENDIX D Hot Model-to-Plant Detector Comparisons . . . . . . . . . . . . . . . . D-1
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LIST OF TABLES Number Title e P3g,e, 3.1 Vennont Yankee Rated Operating Characteristics 10 3.2 Vermont Yankee General Core Description 11 3.3 Summary of Vennont Yankee Cycles Modeled 12 5.1 SIMULATE-3 Cycle Average Hot Eigenvalues 48 5.2 SIMULATE-3 Beginning of Cycle Hot Eigenvalues 49 5.3 SIMULATE-3 End of Full Power Life Hot Eigenvalues 49 5.4 VY Cold Critical Case Conditions and SIMULATE-3 Results 50 5.5 SIMUIATE-3 Nodal TIP Reading RMS Errors 51 5.6 VY Total TIP Uncertainties at Beginning of Cycle 51 5.7 Results of SIMULATE-3 Compared to SIMULA'IE 2 52 C.1 Channel Bowing vs. Burnup as a Function of Initial Bowing C-3 l l l
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LIST OF FIGURES Number Figure Egge 2.1 Process Stream of Current Physics Codes 5 2.2 Process Stream of Proposed New Physics Codes 6 3.1 Radial Map of Vermont Yankee Core 13 3.2 Axial Relationship of Fuel, Control Rods and Incore Detectors 14 3.3 Detail of Fuel, Control Rod and Incore Instrument Tube 15 4.1 Composition Regions Specified for Vermont Yankee MICBURN-3 Model 30 4.2 MICBURN-3 Sensitivity: Adding Water to the Buffer Region 31 4.3 MICBURN-3 Sensitivity: Adding Voids to the Moderator Region 32 4.4 CASMO-3 Representation of the Vermont Yankee Lattice 33 4.5 Cross Section View of the Vermont Yankee Control Rod Wing 34 4.6 CASMO-3 Representation of the Control Rod Wing 34 4.7 CASMO-3 Sensitivity: Change in K= Between 40 and 70 Groups 35 4.8 CASMO-3 Sensitivity: Change in K= Caused by 40 mil Channel Bow 36 4.9 SIMUIATE-3 Axial Representation of Vermont Yankee Fuel 37 4.10 Impact of Spacer Correction and Channel Bowing on SIMUIATE-3 Eigenvalue 38 4.11 Plant Average Axial TIP Trace vs. Plain SIMUIATE-3 Model 39 4.12 Plant Average Axial TIP Trace vs. Spacer Correction Model 40 5.1 SIMUIATE-3 Hot Eigenvalues for Cycle 9-13 vs. Core Exposure 53 5.2 SIMUIATE-3 Cold Eigenvalues for Cycles 9-13 vs. Core Temperature 54 5.3 SIMUIATE-3 Cold Eigenvalues for Cycles 9-13 vs. Core Exposure 55 5.4 SIMUIATE-3 Averaged TIP Integral Errors, S. D., and RMS Errors for 56 Cycles 9-13 5.5 SIMUIATE-3 Core Average Axial TIP Errors and S. D. for Cycles 9-13 57 A.1 Reload Decign of Cycle 9 A-2 A.2 Power History of Cycle 9 Showing TIP Statepoints A-3 A.3 Control Rod Inventory of Cycle 9 Showing TIP Statepoints A-4 A.4 Reactor Conditions for Cycle 9 Depletion Steps A-5 A.5 Reload Design of Cycle 10 A-7 A.6 Power History of Cycle 10 Showing TIP Statepoints A-8 l A.7 Control Rod Inventory of Cycle 10 Showing TIP Statepoints A-9 f I vit . l
LIST OF FIGURES Number Pfgure Pgge, A.8 Reactor Conditions for Cycle 10 Depletion Steps A-10 A.9 Reload Design of Cycle 11 A-12 A.10 Power History of Cycle 11 Showing MP Statepoints A-13 A.11 Control Rod Inventory of Cycle 11 Showing TIP Statepoints A-14 A.12 Reactor Conditions for Cycle 11 Depletion Steps A 15 A.13 Reload Design of Cycle 12 A-17 A14 Power History of Cycle 12 Showing TIP Statepoints A-18 A.15 Control Rod Inventory of Cycle 12 Showing TIP Statepoints A-19 l Reactor Conditions for Cycle 12 Depletion Steps A-20 A.16 A.17 Reload Design of Cycle 13 A-22 A.18 Power History of Cycle 13 Showing TIP Statepoints A-23 A.19 Control Rod Inventory of Cycle 13 Showing TIP Statepoints - A-24 A.20 Reactor Conditions for Cycle 13 Depletion Steps A-25 B.1 Cycle 9 Cold Critical Patterns B-2 B.2 Cycle 10 Cold Critical Patterns B-4 l B-6 B.3 Cycle 11 Cold Critical Patterns B.4 Cycle 12 Cold Critical Patterns B-7 B.5 Cycle 13 Cold Critical Patterns B-8 C.1 Axial Representation of Channel Bowing (Exaggerated) C-4 C2 Impact of Channel Bowing on Lattice (Exaggerated) C-5 D.1 Cycle 9 Averaged UP Integral Errors Standard Deviations, and RMS Errors D-2 D.2 Cycle 9 Core Average Axial TIP Comparison by TIP Set D-3 D.3 Cycle 10 Averaged TIP Integral Errors, Standard Deviations, and RMS Errors D 9 D.4 Cycle 10 Core Average Axial TIP Comprisons by TIP Set D 10 D.5 Cycle 11 Averaged TIP Integral Errors, Standard Deviations, and RMS Errors D-16 D.6 Cycle 11 Core Average Axial TIP Comparisons by EP Set D 17 D.7 Cycle 12 Averaged TIP Integral Errors, Standard Deviations, and RMS Errors D-22 D.8 Cycle 12 Core Average Axial TIP Comparisons by HP Set D-23 D.9 Cycle 13 Averaged TIP Integral Errors. Standard Deviations, and RMS Errors D 28 D.10 Cycle 13 Core Average Axial TIP Comparisons by MP Set D-29
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ACKNOWLEDGEMENTS The authors wish to acknowledge contributions by J. P. Gomld and N. Barbetta in the preparation of both text and figures for this manuscript. We also wish to thank D. M. VerPlanck and A. S. D!Giovine of Studsvik of America for their ideas and encouragement in this benchmarking effort. l l l . ix . f
1.0 INTRODUCTION
AND SUMM,__A,RY,,, _ This report benchmarks the MICBURN-3/CASMO-3/ TABLES-3/SIMUIATE-3 code package against several recent cycles of Vermont Yankee operatiry 2ata. The models are used to calculate eigenvalues and detector response data at nurn ous steady-state exposure points. These calculations are compared to plant measurements; i.e. criticality and Traversing Incore Probe (TIP) data. De consistency and accuracy of the comparisons validate these codes for reload design, steady state licensing and plant support applications. Section 2 provides a brief overview of the propose'd code package. With regard to overall function, MICBURN-3/CASMO-3/ TABLES-3/SIMUIATE-3 w!!! replace the current physics codes of MICBURN/CASMO-2/ TABLES-2/SIMUIATE-2. Differences between the two sets of codes are highlighted. Section 3 describes the Vermont Yankee cores and fuel types covered by the benchmark. Only the later cycles are modeled. The earlier cycles are no longer applicable to future plant operation, because they used low enrichment fuel and a less accurate form of incore instrumentation. Section 4 describes the construction of the Vermont Yankee model, included are the results of numerous sensitivity studies performed. Based on these results and our best engineering judgement, the model was kept as simple as possible while including all the Inajor effects. Thus, this benchmark is fully applicable to the intended use of these codes as
" production tools."
Section 5 provides the detailed results of the model-to-plant comparisons. Briefly, the hot eigenvalue for the five cycles is 0.9989 with a standard deviation of 0.0010. The cold l eigenvalue is 0.9968 with a standard deviation of 0.0017. We model produces nodal instrument readings which are compared to the plant for 223 sets of 'HPs. For these comparisons, the RMS differences in nodal TIP readings are 2.9%. In conclusion, the results of the MICBURN-3/CASMO-3/ TABLES.3/SIMUIATE-3 benchmark prove that this code package is superior to the current methods. It is suitable for all applications, including licensing, for which the current methods are approved. 1
2.0 GENERAL CODE PACKAGE DESCRF!10N ne proposed code package benchmarked in this report is intended to replace the current 5 physics codes of MICBURN'", CASMO-2 . TABLES 25, and SIMUIAE-25 (also mferred to as SIMUIAE-E, or SIMUIATE-YA). De lattice physics code CASMO 2 is virtually identical in I form, function and application to the original CASMO code described in the Reference 5 benchmark. The 3-D nodal code SIMUIAE-2 is virtually identical to the original SIMUIAE j described in the Reference 6 benchmark. TABLES-2 is not actually a physics code; it knks , CASMO-2 to SIMUIATE-2. All of the above codes have been approved by the NRC** for Yankee l Atomic Electric Company use in reload physics analysis of the Vermont Yankee Nuclear Power Station. This analysis includes core design, startup test predictions, generation of physics data for safety analysis as well as the direct safety analysis of certain quasi-static transients. Figure 2.1 shows the function of, and relationship between each of the curnmt physics codes. Notice that th? 3-D nodal code requires disc 6nt!nuity factom. In the past this was supplied by PDQ." Notice too, the gamma detector (11P) responses were supphed externally. These were generated with KENO.* In addition, user specilled " adjustment factom" such as albedos and thermal leakage correction factors were iterated upon to bring the 3-D model's power distribution into alignment with plant data. In contrast, the proposed MICBURN-3/CASMO-3/ TABLES-3/SIMUIATE-3 code package, shown in Figure 2.2, is greatly simplifled. The functions performed by the codes are virtually the same as their predecessors. However, discontinuity factors, and gamma '11P responses are now provided by CASMO-3. Also, the entire process of normalizing the 3D nodal code has been eliminated, nis is because of significant improvements made to several of the codes. Rese will be discussed briefly below. 2.1 MICBURN-3 De MICBURN-3 code is described in detail in References 10 and 11. Simnar to its predecessor, it calculates the burnup in a fuel pin containing an initially homogenous distribution of the bumable absorber, gadounium. The code supphes CASMO-3 with effective e.bsorption uoss sections for the gadohnium, homogenized over the fuel pin. We upgrade to MICBURN-3 is required because it us:s the same hbrary and group'- ctructure as CASMO-3. It also has maisy desirable featums its predecessor did not possess, 1 which allow it to model future, moc complicated fuel designs. 2.2 CASMO-2
. CASMO-3 (also referred .to as CASMO-3G) is described in detail in Reference 12. We q generic benchmark of the code is provided in Reference 11. Similar to its predecessor, !
CASMO-3 performs burnup calculations on an entire fuel assembly. The code handles -l l geometry consisting of cylindrical fuel rods of varying composition in a square pitch array. It j allows for fuel rods loaded with burnable poison, water gaps, water holes, boron steel curtains, ! channels and crucifonn control rods. Like its predecessor, CASMO 3 uses multi-group transport theory to calculate the 2-D space / energy distribution of flux within the bundle. It
. performs the depletion calculation and produces two-group cross sectiom, homogenized over the assembly, for input into SIMULATE-3.
However, CASMO-3 has several advantages over its predecessor. It can generate discontinuity factors for use in SIMUIATE-3. The production model uses 40 rather than 25 neutron energy groups. The nuclear library is based on the more modern ENDF/B-IV data - rather than ENDF/B III. It can calculate gamma detector responses. Finally, CASMO-3 can handle baffle / reflector regions, hafnium control blades, water cross and large water hole essembly designs. Therefore, this upgrade provides the abihty to model future, more compucated fuel and control rod designs. 2.3 IABLES-3 The TN3LES-3 code (also referred to as TABLES-3P) is described in detail in References IS and 14. It is a linking code between CASMO-3 and SIMULNIE-3. TABLES-3 processes CASMO 3's two-group cross sec : into two and three dimenatonal tables. SIMULATE-3 reads these tames and, accordirt to t] e local conditions, SIMULATE-3 can reconstruct the appropriate homogenized t a-group cross sections in each node. 1 3-
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{ Unlike its predecessor, TABLES-3 can handle discontinuity factors, gamma detector response data and other data SIMUIATE-3 requires. The input format of TABLES-3 is also easier to learn and use than that of its predecessor. 2.4 SIMUIATE-3
. i ne SIMUIAE-3 code (also called SIMUIAE-3P) is described in detail in References 14 l l
and 15. Reference 16 provides the generic benchmark. SIMUIA'IE-3 is a three dimensional i nodal analysis code. It models the steady state neutronics and thennal-hydraulic behavior of the core. It provides power, exposure, and void distributions, burnup, fission product and reactivity effects. However, beyond these functions, all shn11arity with its predecessor ends. D The predecessor. SIMUIA'IE-2, used the same nodalization as the familiar codes FIARE 'l and TRILUM*. The neutron balance equation was modified coarse mesh diffusion theory". It was a one-group method with a two-group approximation afforded by the use of thennal leskge correction factors. Calculations were only executed in the fueled region of the core; user adjustable albedos tenninated the neutron balance equation at the reflector interfaces. In contrast SIMUIATE-3 is a true two-group nodal code. This eliminates the need for thermal leakage correction factors. It uses the QPANDA" model which solves the three-dimensional, two-group neutron diffusion equation. The QPANDA methodology also assumes that the flux distribution is comprised of two pieces: global shapes (homogeneous emooth flux distribution) and local shapes (heterogeneous asserably flux distributions). This assumption allows assembly discontinuity factors (ADF's) to be edited from the same CASMO-3 calculations that produce two-group cross sections. When used in the QPANDA model, the ADFs alter the neutron currents between nodes, effectively eliminating spatial homogenization errors. The ADF concept is also applied in modeling homogenized baflie/ reflector nodes. %is eliminates the need for any user adjustments (albedos) in the reflector region, in summary, SIMUIATE-3 provides superior accuracy with fewer user-adjustable inputs than its predecessor. It also runs faster. r 4 t
b FIGURE 2.1 Process Stream of Current Physics Codes Code Name Function l Provide effective cross sections for MICBURN a pin as gadoiinium durns out i ! Y Provide assembly homogenized two-group
** ** **'" "* *2"*"
- d^t^ ^"d ' ^2 CASMO-2 peaking factors l l
Y Process cross sections, etc., into TABLES-2 functionaitzed tadies KENO Provide gamma detector responses User Supplied Adjustment pDQ Provide discontinuity factors Factors f JL Provide 3-D power, void, exposure and l SIMULATE-2 rasion product distribution for static reactor conditions j .,.
FIGURE 2.2 Process Stream of Proposed New Physics Codes - Code Names Function i Provide effective cross sections for MICBURN-3 a pin as gadolinium burns out l l i Provide assembly homogenized two-group y cross sections, kinetics data and local CASMO-3 Peaking factors. Also, flux discontinuity data, reflector cross sections = 2d gamma detector responses. V Process cross sections, etc. into TABLES-3 functionalized tables 1P Provide 3-D power, void, exposure and [ SIMULATE-3 fl881on Product distributions for static reactor conditions. l 6- )
3.0 VERMONT YANKEE OPERATING CHARACTERISTICS Vermont Yankee (VY) is a General Electric BWR-3 which began cammerical operation in 1972. mg:nn:ng in Cycle 6 (1978), Vermont Yankee began a dunsition to axially zoned fuel ) assemblies with a longer active fuel length. Thus, starting with Cycle 6, all W cores have been zoned. Also, from Cycles 6 to 9. W had cores that possessed uneven active fuel at the top. Beginning with Cycle 9 (1981) W began a transition to longer cycles. This required larger .l reload batches, and later, higher enrichments and heavier burnable poison loadings. Wert ; have also been several changes in the thermo-mechanical design of the fuel rods in the past f ten years. All of these changes can be accommodated by the modeling capabilities of the l MICBURN-3/CASMO-3/ TABLES-3/SIMUIEIE-3 code package. 3.1 Core Description . Vennont Yankee is a D-lattice plant with a small diameter, high power density core. We rated operating characteristics are provided in Table 3.1. The core description is summarized in Table 3.2. A radial map of the W core is shown in Figure 3.1. All 368 channeled fuel essemblies are orificed to control flow. The shaded assemblies, shown in Figure 3.1, are tightly orificed to maintain a flow balance between the higher powered core interior and the lower powered edge. De intersections of the dashed lines show the centers of the control blades. Figure 3.1 also shows the locations of the various incore detectors. The benchmarking effort is primarily interested in the 20 instrumentation locations containing both the Traversing incore Probes frIPs) and the local power range monitors (LPRMs). De IERMs are fixed neutron detectors. At each of these 20 instrument locations t there are four LPRMs arranged at four axial levels as shown in Figure 3.2. We LPRMs are calibrated frequently; that is, normaltred to their adjacent TIPS. We three TIP machines (A, B, and C) are, in turn, normali=d to each other by use of a common core location shown in Figurt 3.1. Rus, the TIP readings are the single most important measure of the model's accuracy. All other measures of power distribution, including those at the plant, are either derived or inferred from the TIPS. l 7 1
Toward the end of Cycle 8 (1981), the TIPS were converted from neutron detectors to gamma sensing detectors. This significantly reduced a major component of measurement i uncertainty, namely 'TIP asymmetry." TIP asymmetry is caused by a disorientation of the TIP instrument tube within the so called " narrow-narrow corner" of the water gap between the fuel channels (see Figure 3.3). 'Ihe thermal neutron.'11Ps were very sensitive to this orientation, gamma-TIPS are not. In providing an axial shape, the gamma-TIP is less sensitive to uncertainties in the axial variation of voids up the channel. Finally, it is also less sensitive to the possible presence of voids in between the channels (bypass voiding). Neutron-TIPS detect primarily those neutrons emitted by the four nearest corner pins. The power in the assembly must be inferred from each corner pin by means of a multi-step process. In contrast, the gamma response function receives contributions from deep within each surrounding assembly. The power in adjacent assemblies is more directly inferred. In conclusion, comparisons between the model-calculated gamma-TIP readings and the plant-measured gamma-TIP readings provide the most direct and accurate test of the model's ability to reproduce the plant's power distribution. 3.2 Description of Cycles Modeled As discussed Vermont Yankee converted to gamma TIPS toward the end of Cycle 8. Thus, Cycle 9 was the first full cycle to use gamma-TIPS. It was also the first transition to longer cycles. 'Ihus, the benchmarking will examine Cycle 9 through the recently completed Cycle 13.
'Ihe previous cycles have little application to future plant operation and will not be presented here. A summary of the operation of Cycles 913 is provided in Table 3.3.
l More detail of each Cycle is provided in Appendix A. This includes the reload design of each of the cycles. As shown in Appendix A. Vermont Yankee employs a conventional (not Control Cell Core) loading scheme. This frequently results in very old assemblies alternating with fresh assemblies. The loading schemes are also low leakage; i.e., oldest fuel loaded on the core periphery. Both of these situations, in combination with VTs high power density and small core, result in steep, rapidly varying flux gradients. This can also be handled by the modehng capabilities of the code package. 1
1 3.3 Fuel Description I Table 3.3 lists the vendor designation of each fuel type modeled in the complete benchmarking effort. Detailed descriptions of the fuel are proprietary to the vendor and can i be found in Reference 21. However, as a general description of the fuel: 'Ihe fuel types modeled are all D lattice 8 x 8 arrays as shown in Figure 3.3. They include several different l enrichmmts, gadounium loadings and active fuel lengths. Some of the fuel is zoned . Udly, creating several lattices per fuel type. A " lattice" consists of any unique pin distribution of l enrichment or gadolinium in an axial slice of an assembly. Each lattice must be exphcity ; modeled. ! In addition to the complexity of the fuel designs, the vendor altered the thermo-mechanical properties of the fuel pins several times. For the fuel present in the core during this benchmark, the following is an approximate chronology: a) The vendor introduced hehum back filled pre-pressurized fuel (1979). b) Reduced pellet surface roughness and inside clad ! roughness (1983). c) Gradually increased pellet density (1983-1983). d) Increased pellet O.D. and decnased pellet / clad gap (1986). These changes were all directed toward reducing fuel j failures by mducing average fuel temperatures. The latter affects the overall Doppler defect on i a batch by batch basis. 'Ihis can also be handled by the model. i l r l l L
TABLE 3.1 - Vermont Yankee Rated Operating Characteristics q l
- Rated Power (MWth) ' 1593.0
. Average Power Density (kW/1) 49 (1) l Number of Assemblies 368 Equivalent Core Diameter (inches) 129.9
- Total Con Rated Flow (Mlb/hr) 48.0 i Core Bypass Flow (Mlb/hr) 5.2(2)~
Steam Flow Rate (Mlb/hr) 6.43 f Feedwater Flow Rate (Mlb/hr) 6.40 Feedwater Temperature (*F) - 372 ). Nominal Steam Dome Pmasure (psia) 1020 Nominal Reactor Average Pressure (psia) 1033
~
Core Inlet Enthalpy (Btu /lbm) 520 t Core Inlet Subcooling (Btu /lbm) 27 Core Exit Quality (% steam) 13.3 Notes: (1) Varies with active length of fuel. A full core of 150 inch fuel equals 48.92 kW/1. (2) Varles slightly from cycle to cycle. It depends on the number of assemblies with drilled lower Ue plates and the number of water rods per assembly. Most later cycles have about 5.2 Mlb/hr of bypass flow.
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TABLE 3.2 Vermont Yankee Geneml Core Ce criction
< 7.gg) -
Number of Assemblies 368 .; Cold Assembly Pitch (inches) 6.0 144-150 I Active Fuel Height (inches) 8x8. ' Fuel Rod Array. Lattice Type D
' Fuel Pellet Matedal Sintered UO. f Fuel Clad Material Zr _,
Channel material . Zr-4 ' ') Channel Thickness (mils)~ 80 , Spacer Material . Zr-4 and Inconel Number of Spacers 7 Movable Control Rods Number of Rods 89 Shape - Cruciform Cold Control Rod Pitch (inches) 12.0 Control Material Height (inches) 143 Control Material Compacted B.C in S.S. tubes and sheath Incore Instrumentation Source Range Monitors . 4 Intermediate Range Monitors 6 Power Range Detector Locations 20 LPRMs (4 per location) 80 TIP Machines 3 P ., . . E TABLE 3.3 j Stimmary of Vermont Yankee Cycles Modeled Cycle 9 Cycle 10 Cycle 11 Cycle 12 Cycle 13 l r
.l Operation Dates: .: BOC 12/1/81 6/17/83 8/6/84 6/30/86 10/2/87 ]
EOC 3/5/83 6/15/84 9/21/85 8/7/87 2/11/89 l 4 As-loaded Core Weight (Short Tons): Initial H.M. 74.15- 74.13 74.25 74.44 '74.82 L ) : Core Average Burnups (Mwd /St): 9,192 10.463 10,418 9,820- 8,613 [ BOC EOFPL 16,595 17.185 16,733 16,358 16,830 EOCL .18,137 17,806 18,283 17,949 18,307 Thermal Capacity Factor While Operating: [ CF (%) 90.9 93.6 89.2 94.3 91.4 i Number and Type of Fuel Assemblies leaded: Fresh 120 108 104 120 136
'lype P8DPB289 P8DPB289 P8DPB289 P8DPB289 BP8DRB299 1 Cycle 80 120 108 104 120 Type P8DPB289 P8DPB289 P8DPB289 P8DPB289 P8DPB289 2 Cycle 96 80 120 108 104 Type P8DPB289 P8DPB289 P8DPB289 P8DPB289 P8DPB289 3 Cycle 60 + 12 60. 36 36 8 Type 6.7 8289 P8DPB289 P8DPB289 P8DPB289 P8DPB289
+ 8D274H L
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{ > 4.0 VERMONT YANKEE SPECIFIC MODEL The benchmarking of Vennont Yankee was performed with the intention of producing a production tool for reload analysis. Therefort, the input to each of the codes in the package was kept as simple as possible. However, to avoid the trap of oversimplifying, numerous l sensitivity studies were performed at each step in the modeling process. Only those detads which caused a substantial improvement in the final model were retained. 4.1 MIgURN-3 Model Description ? ! The MICBURN-3 code microscopically depletes the gadolinium in the gad pin. For the fuel types modeled in this benchmark, MICBURN-3 cases were run for each gad pin with a ) l different w/o gadohnium. For the BP8DRB299 fuel type, the same w/o gadolinium existed in ! two separate Um enrichment pins. Therefore, two separate cases were run. This proved to be l unnecessary (see Section 4.1.1, MICBURN-3 Sensitivity Studies). I Figure 4.1 shows the geometry of the Vermont Yankee MICBURN-3 model. In this figure, four separate composition regions can be seen. These regions are defined for the cold (20*C) lattice. MICBURN-3 thermally expands them and modifles the material densities automatically. The fuel region requires the pellet radius (R,), the pellet density, the w/o Um and the w/o t gadolinium of the gad pin. The VY model uses the default geometric specificatiori for the fuel region: the fuel region has been divided into 10 " macro-regions" for the transport calculation, and 20 " micro-regions" for the burnup calculation. The average fuel temperature specified for the fuel region was generated with the in-house fuel behavior code, FROSSTEY"* ** **'. t_ The cladding region requires the outer radius of the cladding, (Rc). It is defined as a single macro-region. s MICBURN-3 breaks the moderator region into two macro-regions, although the composition is homogeneous. The moderator is specified as saturated water at T.c = 560'K. The default ~ value of zero voids in the moderator is used. This irgion represents the water immediately
surrounding the gad rod. The moderator radius, (I(), creates a circle whose area is equal to the pin pitch squared. Spacer material is ignoitd in the moderator region and also in the
' buffer region.
l The buffer region is also split into two macro regions of homogeneous material. The buffer
- represents everything else surrounding the gad pin and its immediate moderator, ne buffer
. region is created by homogenizing the remaining fuel pins, cladding and water of the fuel 1 gasembly into one region. For the W model, the average uranium enrichment of the assembly is assumed for the remaining fuel pins. The buffer region is assumed to be free of volds and the effects of other gad pins. W uses the suggested buffer radius (14), of 5 cm.
> The prima 2y output of MICBURN-3 consists of effective gadolinium microscopic absorption cross sections for the burnable absorber (gad pin) as a function of the absorber number density, N . The latter is a pseudo-number density which takes into account the transition chain for the various gadolinium isotopes. The output of the production model is in 40 energy groups, consistent with CASMO-3. i 4.1.1 MICBURF-3 Sensitivity Studies ) A number of sensitivity studies were performed with MICBURN-3 to determine what , significantly altered the output. The output studied was the effective Gd microscopic absorption i cross section as a function of Nm. ne first study examined the effect of increasing the enrichment of Um in the gadolinium pin by 1.0 w/o. nis had negligible impact on the burnable absorber cross sections. A second study examined the effect of increasing the enrichment of Um in the buffer region by 1.0 w/o. This also had negligible impact.
Conclusions:
MICBURN 3 cases do not need to be run for different enrichment pins having the same gadolinium content. Alst , using the bundle average enrichment in the buffer region is a valid input assumption. 1 - The only variable which seems to have the slightest impact on MICBURN 3 (other than w/o gadohnium) is the water density. The water density directly affects the snectrum, which in 3 tum, affects the relative rates at which the various Gd isotopes burn out. Figure 4.2 shows f the impact of replacing half the fuel in the buffer region with water rods. This softens the spectrum, which preferentially burns out more Gd, during the early stages of depletion. The l
remaining- Gd isotopes have smaller thermal absorption cross-sections. Werefore, the microscopic absorption cross section versus Nm is reduced. The fast cross section is not visibly affected. Hardening the spectrum has just the opposite effect. Figure 4.3 shows that running MICBURN-3 at 40% volds causes a slight change in thermal microscopic cross section in the ] cpposite direction. However, when carried through to the CASMO-3 level, running i MICBURN-3 at 40% voids has negligible impact on the results. Therefore, for the production model,we use the recommended default value of zero volds for all MICBURN-3 cases. 4.2 CASMO-3 Model Description { The CASMO-3 code (also called CASMO-3G) is used to calculate the burnup and 2-D flux and energy distributions within each of the lattices required by the benchmark. CASMO 3 can use either a 70 or 40 neutron energy group library. The Vermont Yankee production model uses 40. This is only for calculations at the pin cell level. For the macro group calculation (see Reference 12), the VY model uses the default collapse to 23 energy groups. For the 2-D l calculation, it uses the default collapse to 7 energy groups. The VY lattice geometry is shown in Figure 4.4. The lattice is synunetrical along the diagonal so only half the bundle is modeled. The bundle is surrounded by a channel of 80 mils.. Water outside the channel is specified separately from water inside the channel. De water inside the channel may contain volds, depending on the case. It definitely contains spacers. The amount of spacer material for the assembly is specified as if it is homogenized over all the water inside the channel. Figure 4.4 shows the mesh spacing of the unit cells for the calculation of the 2-D {. l distribution of flux and local power within the lattice. %e mesh spacing is the pin pitch, which is specified cold (20*C). CASMO-3 increases the pitch for hot problems by using the , thermal expansion coefficient of the spacers (Zr-4). Likewise, all other dimensions are input for the cold lattice. CASMO-3 uses the Tnm, or To with internal material expansion coefficients to expand the dimensions and adjust the material densities. f
The unit cells require seveml arrays to specify them.- One such array distinguishes between fuel rods and water rods. These are treated separately at the pin cell level in CASMO-3. For the fuel rods, the radil of the fuel region, gap region and cladding region are i specified. CASMO-3 automatically surrounds the fuel rod with an appropriate moderator, or moderator and buffer region, as required. Water rods require the dimensions of the water inside and the cladding. For W. the water inside the water rods is specified as being free of j volds. Unlike its predecessor, CASMO-3 allows several different size water rods to be specified. f l The fuel rods require more than the dimensions. For each fuel rod type, the U. I enrichment and, if appmpriate, w/o gadolinium is specified. Although CASMO-3 contains an internal Gd library, separate MICBURN-3 cases (see Section 4.1) were run to provide the cpecific gadolinium cross sections for each lattice / gad pin type. The pellet stack density of each fuel rod type is required. This varies with the w/o gadolinium, which creates a less dense pellet matrix. For the W model, the gad pellet density is calculated using a vendor correlation. The final pellet densities for all fuel rods are checked by running a zero depletion CASMO-3 case. This provides a weight for the lattice which is checked against the average bundle weight. The pellet densities are adjusted, as required. ) One step up from the pin cell calculation is the' macro-group calculation performed on fuel rods along the edge of the assembly. This calculation improves the homogenized cross sections for these rods and provides a neutron source for calculating the channel, gap, and control rod cross sections. Thus, CASMO-3 requires the dimensions of the channel, the narrow ) and wide water gaps. Only half of each water gap is specified; running through the middle of each water gap is another minor symmetry line. Together with the diagonal, these form the boundaries of the infinite lattice problem. When a rodded case is run, half of a control blade is included in the wide water gap as shown in Figure 4.4. The control blade is modeled as two regions. 'Ihe first region is the central steel shaft; the second region is the active absorber region. The B.C powder in the control blade is contained within thin tubes of stainless steel bound by a steel sheath, as shown in Figure 4.5. The blade is modeled in CASMO-3 as follows: The f B.C powder is tnated like a strong absorber " pin" placed in the blade matrix according to the l
I tube pitch. This blade matrix is taken to be a smeared mixture of stainless steel and water as shown in Figure 4.6. In the VY model, the absorbing region is considered to be the area starting with the first B 4C pin an'd going to the last B C pin. The thickness of the tubing is . omitted from the first and last B.C pin. The remaining tube thickness on one end is accounted for as part of the central steel shaft. The little bit of tubing and steel sheath at the very tip of the blade wing is ignored. l 4.2.1 CASMO-3 Deoletion Cases Each lattice in the benchmark required four separate depletion cases. Three were uncontrolled: one each, at 0%,40%, and 70% voids. He fourth case was a control rod history depletion. It was run with the control rod inserted at 0% volds. The void fractions of 0%,40%, and 70% volds roughly represent conditions at the bottom, middle and top of the core 7 respectively. The control rod history depletion was run at zero volds because the BWR control mds are bottom entry and spend most of their lives in a low void environment. 6 All depletion cases were run at hot reactor average conditions. The moderator temperature was'specified at T e = 560"K. 21s provides the appropriate water density for the water and . vapor (volds) mixture specified in the channel and also for the solid" water specified in the water rods and bypass region (gaps). The fuel temperatures for CASMO-3 were calculated for o typical fuel rod at reactor average power u.ing FROSSTEY. A set of identical depletion cases were performed for each lattice modeled. ) The depletion cases use the CASMO-3 default timestep scheme. That is, the depletion
- time steps are close together (.5 GWd/Mtu) until the g'adolinium has burned out. Afterwards, the steps are gradually increased (up to 5 GWd/Mtu). The output edited from each of these depletion steps includes: two-group cross sections and infinite multiplication constant, assembly power distribution, conversion ratios, kinetics data (delayed neutron fractions, etc.),
gamma detector response data, xenon and samarium yields, and assembly discontinuity factors.
. These are accessed by TABLES-3 for processing into SIMULATE-3 input.
20 I
4.2.2 CASMO-3 Branch Cases i I Since the reactor is dynamic, it is not sufficient to run only the depletion cases. These
. 1 provide appropriate results only if the void fraction, or control rod position remains flxed. The depletion cases provide an average spectral history (so called " void history" or " control history")
of a node. However, at any instant in time, the local conditions change. It is germane to the theory of the model, from this point onward, that these " instantaneous effects" are separable. That is, they can be individually calculated and added together later to recreate the proper cross sections, etc., for each individual reactor node. The instantaneous effects are generated by means of branch cases run from various restart exposure points in the void or control history depletion cases. Essentially, for each void history depletion, branch cases are run to each of the other void levels not covered by the depletion. A lower fuel temperature branch case is run. A rodded branch case is run. From the control rod history depletion, an unrodded branch case is run. All of these changes are assumed to be instantaneous, except that CASMO-3 calculates equilibrium xenon. / In order to model the cold lattice, two uncontrolled and two controlled branch cases were run from each of the 0%, 40% and 70% void history depletions. The moderator for the cold cases is set to zero volds. The moderator and fuel temperatures are set equal to each other. The controlled and uncontrolled cases are run at each of two cold branch temperatures. Control history is ignored in the cold model because the flux, during cold conditions, is peaked ) at the top of the core. There is almost no control history at the top of the core. 4.2.3 CASMO-3 Reflector Region Model In order for CASMO-3 to calculate the aflector region cross sections, it needs a source of ) neutrons. The majority lattice of the 8DPB289 assembly was chosen as the neutron soume. The reflector was specified in tenns of its thickness and its homogen! zed material properties. l Data was calculated for the top, bottom, and radial reflectors in the com. 'Ihe cmss-sections at the top of the core were calculated for a 15 cm. section of plenum at three
different volds (0%, 40%, and 70%). He bottom reflector was specified as two adjoining thicknesses. Together, these thicknesses encompassed: lower end plugs, lower tie plate, lower - channel and the control blade handle. The coolant in the bottom reflector was specified at 0% volds. De radial reflector was also specified as water at 0% voids. No other material was included in the radial reflector because VY has an oversized core shroud. Therefore, them ) is no significant structure near the periphery of the core. - In a nimiler manner, reflector cross sections were generated for the cold model at Tuoo = 293*K. 1 4.2.4 CASMO-3 Sensitivity Studies A few sensitivity studies were performed with the CASMO-3 model. F1mt, the 70 group energy library was tested. (This required a 70 group MICBURN-3 to be run). He two-group K-infinities from the 40 and 70 group cases were virtually the same. The differences between the two are plotted in Figure 4.7. As shown, the differences start at 0.5 milli-K and drop off
-rapidly. This minor difference is not worth the cost of running in 70 groups.
s-Another sensitivity study addressed the impact of channel bowing on the t-ttice physics. A discussion of channel bowing in BWRs is presented in Appendix C. Briety. the channel L bowing, in the interior portion of a D-lattice plant, shifts the lattice toward the narrow-narrow corner, affecting the reactivity. For this study, a modest deflection of 40 mils (half the channel thickness) was assumed. This was modeled by simply adding 1 mm to the wide gap and subtracting 1 mm from the narrow gap. The standard depletion cases were re-run for one of the middle lattice sections. ) The results were quite remarkable as shown in Figure 4.8. On the average, the deflection resulted in a 5 milli K decrease in nactivity for the lattice. Figure 4.8 shows the loss in reactivity due to bowing at each of the void levels. As might be expected, the high void region suffers a greater loss. This is because the bypass water makes a greater contribution to the overall moderation when the lattice is voided. These results were dramatic enough to warrant re-calculation of all the VY lattices, f including branch cases. The zones at each end of the fuel assemblies were not re-calculated, tince the channel is fixed in the core near the top and bottom. Wenfore, bowing will cause l little, if any, deflection of these lattices. The " bowed lattice cross sections were then checked cut in a SIMULATE-3 sensitivity study (Section 4.4.3). l 4.3 TABLES-3 Construction TABLES-3 (also called TABLES-3P) reads user specified CASMO 3 punch files and creates tables of parameterized input for SIMULATE-3. Separate TABLES-3 cases are run for each lattice type used in the benchmark. Because the W model is segregated into a hot model and a cold model, separate sets of TABLES-3 cases are run for each condition. However, the cutput from both is loaded into a single library that SIMULATE-3 can access. The hot model cnd cold model are distinguishable on the basis of ' major" and " minor" flags in the library. The tables stored in the library include two-group cross sections, detector response functions, kinetics data, assembly discontinuity factors, etc. To simplify the remaining discussion, only cross-section data will be referred to. As discussed in Reference 13. TABLES-3 constructs its cross sections according to the general formula: E (A, B, C ... Z) = Em (Ao, Bo, Co ... ) )
+ A E (A, B, C) + AE (B, C, D) ... + AE (X, Y, Z)
The user specifies the source of the base cross sections (usually one of the history depletions) and the source of the partial cross sections. Each of the partial cross sections can only have up to three variables. All other variables are fixed. The partial cross sections are derived by comparing various CASMO-3 branch cases to their history depletions, or comparing them to each other. For the W hot model, the variables "A ---> Z" which are used to construct the cross f sections, E(A, B, C .... Z), are exposure, void history, instantaneous voids, fuel temperature, I ' i instantaneous control rod, and control rod history. The base case selected for the hot model l base cross section set is the unrodded 0% void history depletion. ( For the W cold model, the variables used to construct the cross sections are exposure, ! void history, moderator temperature, and instantaneous control rod. The base case for the cold model cross section set is the unrodded, cold branch case from the 70% void history depletion. l
This case was chosen because most of the flux, during a cold critical, is in the top 1/4 of the f core. The void history in the top of the core is closest to 70% volds: therefore, less interpolation will be required in the area of importance. After TABLES-3 constructed its hbrary of cross sectional tables, the latter were tested out f I using the " AUDIT' function of SIMUIA'IE-3. The AUDIT function allows the user to exercise the cross sectional variables for cach lattice at the specific CASMO-3 depletion or branch case conditions. The results were compared against the CASMO-3 two-group K-infinities and M'. Overall, the agreement was excellent; this proved that the tables were constructed properly. I I 4.4 SIMUIATE-3 Model Description t l l SIMULATE-3 (also called SIMUIATE-3P) is a sophisticated nodal code with relatively simple input. The simplicity of the input stems from the fact that it allows for few user adjustments when setting up a model. Unlike its predecessor, 'SIMUIATE-2, there are no adjustable neutronics parameters in SIMUIATE-3. In essence, the user apecifies the actual reactor geometry, fuel assembly characteristics, and operational data when developing the rnodel. Nearly all variations in model design are introduced by means of the cross sections. Some minor leeway exists in the specification of the thermal-hydraulics parameters. The SIMULATE-3 model of VY is developed in two parts. These are the hot and the cold model. They differ in the temperature range at which the cross sections were generated, and in the thermal-hydraulics of the core. In both cases, the geometry and fuel specifications of ) the reactor are identical. The geometry of the core begins with the nodalization: The core is divided into 27 axial nodes per assembly and one radial node per assembly. The nodes are 6" on a side. A single node of reflector material is added around the sides of the core by specifying a pseudo-assembly consisting of nothing but reflector material. The SIMULATE-3 system allows for the possibility of modeling an uneven height active core by means of a flexible fuel assembly definition. Figure 4.9 gives two examples: The 150" fuel is defined as having a bottom reflector node, a bottom fuel zone node, 23 nodes of middle fuel zone and a top fuel zone node followed by a single top reflector node. The 144" fuel, on the other hand, has a bottom reflector node. )
l 24' nodes of fuel zone and two nodes of top reflector. By means of these zones (including
.. reflector zones) the appropriate cross sections can be associated with each part of the core.
The cross-sections for the SIMULATE-3 model are loaded from the hbrary created by j
. TABLES-3 using a library specification card. This library specification associates a cross section set with each zone of fuel (and reflector) specified in the fuel definition cards.
In addition to the nodalization of the model and the specification of the fuel types, several other parameters are required. For VY, the following were specifled: serial number and location of each fuel type; fuel batch assignments (for fuel management editing); locations and hydraulic specification of core orificing; and, the location of control rods and in-core l Instrumentation. The control rods were specified as being 144" long even though the control material only l reaches to 143" (see Figure 3.2). This approximation introduces very little error because there are at least two inches of steel above the B C region of the control rod which provide significant epithermal absorption. The in-core instrumentation ('I1Ps) cover only 144" of the active fuel length of the core (see Figure 3.2). Thus, the top node of the active fuel region is not instrumented. The SIMULATE-3 code expects all the fueled region of the core to be instrumented. The standard Vermont Yankee TIP string reading has 24 values. SIMULATE-3 would " stretch" these 24 values over the 25 axial nodes if we did not supply a dummy 25th value at the end of each data string. This dummy value is ignored when calculating the statistics of the plant-to-model TIP comparisons in Section 5.4. However, SIMULA'IE-3 continues to calculate a pseudo 'I1P reading for the 25th node. The latter is visible in some of the plots. If the user specifies the location of the spacers in the core, SIMULATE-3 will use the
" spacer correction" model. When the spacer correction model is turned on, SIMULNIE-3 l edjusts the thermal-hydraulic calculation down stream of each spacer to account for voiding l caused by the pressure drop in those areas. The SIMUIATE-3 model was tested without, and
- _ with, spacer correction.
L 4.4.1 SIMULATE-3 Hot Model { The hot model performs thermal-hydraulics calculations. SIMUIATE 3's internal heat balance is input by means of interpolation tables of various reactor water flows, temperatures, pressure drops, etc. The heat balance for the W model was largely derived from fits to actual plant measurements. Some of the values, such as carry-under fraction, were inferred. The fraction of bypass flow and several other pressure drop parameters were calculated by pm . sei, ms Based on this heat balance, the SIMUIATE-3 model calculates inlet subcooling. As an alternative, the subcooling calculated by the plant process computer can be input directly. In this case, the hot model requires a pressure correction to convert plant subcooling (calculated at steam dome pressure) to SIMUIATE-3 subcooling (calculated at reactor average pressure). SIMUIATE-3 was tested both ways. For the calculation of volds up the channel, the Vermont Yankee model uses the standard EPRI-Void correlation". For fuel temperatures, the W model uses fuel temperatures generated with FROSSTEY. These can be input to SIMUIATE as a function of exposure on the assembly. However, to keep the model simple, the fuel temperature in the W model is kept fixed with expor2re, it only varies with fuel type and relative nodal power. The test of the W hot model consists of a TIP to TIP depletion of the five cycles of the benchmark. On the order of 50 TIP sets were used for each cycle. They are all gamma TIP measurements and nearly all are at equilibrium xenon. Appendix A contains the details of each depletion TIP step as well as descriptions for each core. The input to SIMUIATE-3 for each depletion step includes: core average exposure, power, flow rate, inlet subcooling, steam dome pressure, and control rod positions. SIMUIA'IE-3 depletes to where a TIP set was taken and calculates a corresponding detector response. This response can then be compared to the plant measured data. Appendix D presents the comparisons of the model to plant TIP measurements. Section 5.3 summarizes the hot model eigenvalue behavior.
.m.
4.4.2 SIMUIATE-3 Cold Model The cold model uses the thermal-hydraulics of the hot model. However, it is not sensitive to the heat balance, except for the following Some flow and pressure must be specified. This is required, because the SIMUIATE-3 model insists on a tiny amount of core thermal power to provide a neutron source. For a given core moderator temperature, voiding will be calculated to occur unless some flow and pressure (greater than the saturation pressure) is specified. The test of the VY cold model consists of statepoint benchmarks to plant measured cold cdticals. Each cold benchmark is done as a branch SIMUIATE-3 case following the hot model depletion to the appropriate exposure point in the cycle. A fission product depletion step always precedes the cold case, to account for the shutdown period before the critical was measured. This is necessary to get the right xenon and samarium concentrations. Each cold benchmark case is modeled at the plant measured conditions of: core average exposure, reactor average temperature (fuel and moderator are at the same temperature), and control rod positions. Reactor pressure is usually atmospheric; however, some of the criticals required a higher saturation pressure because they took place above 212*F. The reactor period, recorded at the time of the critical, is converted to worth and used to correct the model's eigenvalue. A total of 24 cold criticals occurred at VY during the five cycles modeled in the benchmark. Every one of them was modeled. Appendix B gives the details of each cold critical. The model's accuracy in the cold benchmark is determined by a steady consistent eigenvalue among all the cold entical cases. Section 5.2 of this report presents the SIMUIATE-3 model results for all the cold benchmarks. f SIMUIATE-3 Sensitivity Studies 4.4.3 l l A number of sensitivity studies were performed with SIMU1 ATE-3. The first two merely confirmed the accuracy of our me.ans of constructing the model. The last three illustrate what the model is sensitive to. This helps identify those areas v.hich require caution when analyzing
)
I
future fuel designs or changes to those codes (FROSSTEY and FIBWR) which supply auxilhary input for the SIMUIATE-3 model. SIMUIATE-3 will either accept fuel bundle weights, or it will calculate the same based on a pass-through of the CASMO-3 lattice weights. As discussed (see Section 4.2) the CASMO-3 ttack densities were explicitly adjusted to give weights very close to the long term average weights of the various fuel types involved in the benchmark. Yet, batches of the same fuel type, loaded at different times in the benchmark, varied from 182.4 KgHM in early cycles,up , to 184.3 KgHM in later cycles. This was caused by the fuel vendor gradually increasing the pellet density (see Section 3.3). As a test, the actual fuel batch weights were installed in , SIMUIATE-3. This did not improve the results of the model relative to the pass through of f CASMO-3 lattice weights. Therefore, the production model will continue to use the generalized , CASMO-3 weights. < l l Another sensitivity study tested whether using the internal heat balance in SIMUIATE-3 altered the benchmark results relative to direct input of the plant subcooling. Again, the 1 results were virtually indistinguishable. Therefore, the plant subcooling will be used when it is available: the heat balance will be used when it is not. One input parameter that does affect the SIMUIATE-3 results significantly is the fuel temperature. Based on the user's definition of average fuel temperature in SIMUIATE-3, the latter will interpolate (or extrapolate) between the CASMO-3 fuel temperature results to get less (or more) Doppler feedback. FROSSTEY is used to calculate individual fuel temperatures for each fuel type. This was input to SIMUIATE-3 as a value which varies only with power. To test the importance of the fuel temperatures, the VY model was exercised by simply increasing the values of the average fuel temperature assigned to each fuel type by 80*C. The Doppler f:edback of the model behaved predictably and produced a 2.3 milli-K reduction in core K effective.
Conclusion:
It is very important to be aware of any changes to the fuel behavior modeling code FROSSTEY. Likewise, any future fuel vendor changes to the fuel thermo-mechanical design which may affect average fuel temperature need to be evaluated. Even subtle changes, such as producing a smoother pellet surface can significantly affect the gap conductance and thereby change the average fuel temperature. The last two sensitivity studies were a test of the spacer correction model in SIMUIATE-3 and a test of the channel bowing cmss sections developed at the CASMO-3 level (see Section 4.2.4). Wese both had significant impact on the eigenvalue behavior of the model. Figure 4.10 gives an example of how the eigenvalue of Cycle 10 behaved as the two changes were made in Cuccession. Each one tended to bring the average eigenvalue closer to-1.0. 'Ihe channel bowing cross sections tended to flatten the "eigenvalue drift" the model exhibits as the cycle pmgresses. j Figures 4.11 and 4.12 illustrate the impact of the spacer correction model on the axial < l average of all the TIP readings. Notice in Figure 4.11 the obvious location of spacers in the l plant data. The plain, uncorrected, EPRI-void thermal hydraulic model, however, is smoothly j l varying. Notice in Figure 4,12 that the spacer correction model does, to some degree, exhibit I the characteristic dips in the
- IIP readings, especially at the top of the core.
Conclusions:
The SIMULATE-3 model with spacer correction and bowed channel cross sections is the preferred method of modeling Vermont Yankee. In the next section the results of this final model will be reported. 2e . iis, i,a- , ,, , - - -
FIGURE 4.1 ) i i Composition Regions Spectfled for Vermont Yankee MICBURN-3 Model C R,
%3N )
\
S e Notes Macro-Regions Micro-Regions l l R, = Fuel Pellet Composition 10 20 ,- Re = Fuel Cladding Composition - 1 N/A R. = Pin Moderator Composition 2 N/A R. = Buffer Region Composition 2 N/A 30 -
FIGURE 4.2 MICBURN.3 SensitMty: Adding Water to the Buffer Region
\
i 11 . l i
- , .,- l
. \
BASE CASE ~~.. '~.. THERMAL
'.r............4........-***...........-
EXTRA WATER RODS
, id] IN BUFFER REGION P .
O d O cc U U id - E : o
-o O
o - 5 5 - cn FAST
$f.
4 DIRECTION OF DEPLETION Td , , v . , , 30.0 35.0 40 0 00 50 *00 15 0 20.0 25.0
*1CT' BA NUMBER DENSITY
FIGURE 4.3 i 4 i MICBURN 3 SensitMty: Adding Voids to the Moderator Regfen J 1 i i 1 1d .
.... - THERMAL
~
' 40% VOIDS **...
.... -.... h 1d-7 BASE CASE O . ,
P .
~
x 0 O lI: E : 8
.m o -
e . U 5 -
< FAST m
16
,2 DIRECTION OF DEPLETION 1d 00 50 10.0 15 0 20 0 25 0 30.0 35.0 wo BA NUMBER DENSTY *16' i
L i
FIGURE 4.4
. CASMO-3 Representation of the Verrnont Yankee lattice I
i l control blade (if inserted) wide water gap outside of channel . channel wall water gap along inner channel wall N fuel pin cell (based on pin pitch) x
\
N N N~ (outside channel) narrow water g Bap
\
[ r 33 )
FIGURE 4.5 Cross Section View of the Vermont Yankee Control Rod Wing , n. 7, Stainless Steel
/ cnt ral Stru::ere s/ '
N stuee. Shee: steu st u e e.. steu Clad N -/-- --- ----
\ / .
300GO '/////M//1 W////X FIGURE 4.6 CASMO-3 Representation of the Control Rod Wing steel and veter absorber (Eg 9 V V/ N
/.
)l.*.
5.!(l*.
. g b ,
steel control region : "I"E enuafon reg tip (ignored) 34
jf i I-FIGURE 4.7 f-CASMO-3 SensitMtv: Change in K= Between 40 and 70 Grouns I. to 1 o f T h-O b O v l 0. D O 0.0 tr C o D x W C O b i b $ 10 0 $ EXPOSURE (GWD/T) I
FIGURE 4.8 CASMO-3 Sensitivity: Change in K= Caused by 40 mil Channel Bow 1
-I O 01 l
o 07. VOID HISTORY x 407. VOID HlSTORY q I C1 707. VOID HISTORY i o l I l 1 5 3 3 . O Z p 1 o
@ 0 00
> 6 8
-D x
N . s # : i
-0 01 25 30 35 40 0 B 10 lb 20 EXPOSURE (GWD/T) l l
........._m.. _ _ _ - _ - - - - - - - - - - _ _ _ . _ _ _ _
FIGURE 4.9 l S!MUIATE-3 Axial Representation of Vermont Yankee Fuel 150' FUEL ' 144' FUEL I TR e TR - T P REiFLECTOR ' ZONE' ; TR BR - BOTTOM REFLECTOR 'ZOME' Z1 - BOTTON FUEL ZOME 15Er FUEL ; Z2 - NDOLE FE ZONE, N FE l Z3 Z3 - TOP FUEL ZONE, ISr FLEEL Z4 - ACTIVE FUEL.~ 144' FUEL t ns !v "J v h 4 OJ OJ n, 22 Z4 21 1 \ . - - BR BR [
)
FIGURE 4.10 i Imnact of Spacer Correction and Channel Bowing on SIMUIATE-3 Eigenvalue q l i 1010 . SIMULATE-3 Cycle 10 g Plain Model 1.008 a y D Drif t = 0.0030 AK D Db D D tb $
- 1006 g 8'E d D0 DDh D 0 U oo o%Pm 1004- Drift = 0.0033a K a er Correction o
U g o o 3 l y y t002-o oo m v gug v
>>o m o -
. . A 1000 AA " Ah"AA Drift = 0.00244 K AAA with A
A AA A A 4 A3 bpacer correction
^A ^ * ^ .1 1 and 0998 ^- "" Bowed Cross Sections -
^^^A 41" 0.996 19.0 20.0 90 10.0 11.0 12.0 13.0 14.0 15.0 16.0 17.0 18.0 CORE AVERAGE EXPOSURE (GWD/ST) 1
FIGURE 4.11 E Plant Average Axial TIP Trace vs. Plain SIMUIATE-3 Model 1 l l 1 25 - 24 - N0 23 - 22 - 21 - S ' 20 - 19 - i
- l 18 -
17 - 3* 16 - 15 - KEY S w 14 -
- O 13 -
O l i Z 12 - S Spacer Location *
- Plant TIP Readings
- g11- 3 q10-
- 9-
- 8-
- 7- S 6-
- 5-
- 4- b 3-2-
1-O.dO '
' bd5 ' ' ' b.50 ' bj5' ' ' [0b ' ' ' bb ' ' ' l.5b ' ' ' I.75 ' ' ' 2.00 NORMALIZED TIP TRACE VALUE l
l l
FIGURE 4.12 ' Plant Average Arial TTP Trace vs. SIMUIATE-3 Scacer Correction Model I N ; i 4 - , N ' S EO
'o ~
S l N
$- - +
S +
@ _ +
e-W + O$ Z KEY 3 + J
$N
- S Spacer Location s
("O' + Plant TIP Readings S v-
+
CO - + S + (D - +
+
4 - + S \ h N - I I i l l l l I l l 1 1 I l 1 I I I o 1 0.0 0.5 1.0 1.5 2.0 NORMALIZED TIP TRACE VALUE l
L I 5.0 VERMOVT YANKEE PHYSICS MODEL RESULTS > The final evaluation of the entire code package is based upon the performance of the SIMUIA'IE-3 code. 'Ihe final Vermont Yankee model contains cross sections which reflect the shift due to channel bowing. The thermal hydraulics model uses the EPRI-void model, corrected for local spacer effects. SIMUIATE-3 is benchmarked against hot and cold critical Ctatepoints from Cycles 9-13. The model is judged on two main figures of mertt: l o Eigenvalues - must be reasonably close to 1.0. They must be steady, with small standard l deviations. Any trends exhibited must be gradually varying. j o TIP Traces - generated by the model must compare closely to the plant TIP traces, both I radially and axially. The errors must be small and steady. Any trend exhibited in the I errom must be gradually varyag. ; A model that meets both these figures of merit is suitable for making predictions. These can be short range, as in plant support; or, long range, as in licensing. 5.1 Hot Model Eigenvalues i. The benchmark went back to the end of Cycle 8 to pick up a set of exposure, void, and
. control history arrays. The latter were developed during the benchmark of the old CASMO/SIMUIATE model presented in Reference 6. As such, the EOC8 history arrays loaded into SIMUIATE-3 were slightly inconsistent with the new CASMO-3 cross sections. The differences, however, proved not to be that great. The old cross section history effects disappeared rapidly during the depletion of Cycle 9. Therefore, the beginning of Cycle 9 results are included in the total model results. With over 200 statepoints modeled, the impact of the BOC9 results on the statistics is negligible.
Figure 5.1 shows the eigenvalues for the five cycles of the benchmark. The data is consistent and tightly packed. There is a slight upward " drift" in the eigenvalues that starts in the middle of each cycle. However, this drift meets the criterion that any trends be slowly f varying. It will not affect our ability to make critical predictions with the model. L
Table 5.1 shows the average hot eigenvalue for each cycle with the standard deviation. l The average hot eigenvalue for all cycles is 0.9989 .0010. This meets the figure of merit of being reasonably close to unity. The overall standard deviation is small and consistent within the individual cycles. The standard deviations, for each cycle, primarily reflect the little amount of drift in the eigenvalue over each cycle. Within a given cycle, the eigenvalue is very well behaved. It varies gradually with exposure as the previous example in Figure 4.10 illustrates. Therefore, this model is suitable for hot critical predictions durag plant operation. Table 5.2 shows the hot startup eigenvalue at the beginning of each cycle. This was derived by averaging the first five equihbrium statepoints of each cycle. The average of the BOC hot eigenvalues is 0.9985, for all the cycles, with a standard deviation of 0.0005. This cmall standard deviation indicates good consistency among the cycles. In licensing applications, this provides a high level of confidence in the calculation of BOC hot excess reactivity. Thus, predictions of startup control rod patterns will be quite accurate. Table 5.3 shows the hot eigenvalue at EOFPL in each cycle. This was the eigenvalue of the depletion statepoint nearest to the start of coastdown. As with the BOC eigenvalues, the ) EOFPL values are consistent among all the cycles. The average of all the EOFPL hot eigenvalues is 0.9998 with a standard deviation of 0.0006. Consistency in these eigenvalues is important for predicting the EOFPL exposure when doing a licensing analysis, because the results of certain licensing transients, especially the pressurization transients, are sensitive to the EOFPL calculated. 5.2 Cold Model Eigenvalues As stated in Section 5.1, the beginning of Cycle 9 was initiated using old cross section histories. The impact this had on the hot model results was negligible because of the large number of hot statepoints. However, because of VTs exceptional operation, there are very few cold statepoints to benchmark against. The BOC9 statepoints represent a significant portion of available cold criticals. Therefore, the three cold criticals near BOC9 are given half the weighting of the remaining cold critical comparisons in deriving the cold model eigenvalue.
)
I The results of the cold critical benchmark are presented in Table 5.4. The actual control ! rod patterns for the criucals are given in Appendix B. The weighted cold critical eigenvalue for the SIMUIA'IE-3 model is .9968 with a standard deviation of .0017. To be sure that there is no bias in the cold model, the eigenvalues were grouped by A sequence and B sequence. The results of the A sequence cases are .9968 g .0015. The results of the B sequence cases are .9971 .0018. Statistically, there is no meaningful difference between sequences. The cold eigenvalue is also plotted against core temperature and core average exposure in Figures 5.2 and 5.3. There are no obvious trends in the cold model, with regard to either of these two variables. 5.3 Hot Model Detector Comparisons As described in Section 3.1, all plant " measurements" of 5 ter distribution ultimately depend on the TIP traces. SIMUIATE-3 produces a set of gamma based detector responses rt specific instrument locations. These can be compared with the plant-measured gamma-TIP traces, at the same locations. 1 The simplest comparison is a one-for-one comparison, or nodal comparison. Each of the ) 20 plant TIP traces has 24 readings, spaced evenly over the 144" of the TIP tube. This works out to one reading for every six inch node. SIMUIATE-3, however, has 25 six inch nodes. It produces, and expects to receive for plant data, 25 TIP readings per trace. To prevent inconsistency, the plant data is supplied to SIMUIATE-3 with a dummy 25th value at the top of each trace. The dummy value is zero. Therefore, at the top of each trace, the code always calculates 100% error. In the following statistical comparisons of model-to-plant, the 25th node " readings" are ignored. With the 25th node values removed, both the plant and SIMUIATE-3 TIP traces were normalized. That is, all 20 x 24 nodal values were summed and divided by 480 to obtain one global normalization factor for the plant a id one for SIMUIATE-3. This was divided into each nodal instrument reading. This normahzation is necessary because the plant uses arbitrary ) units for its TIP readings which diffcr from SIMULATE-3's units by a factor of about 100.
}
Following the normalization, comparisons were made between the SIMUIA'1E-3 TIP j l readings and the plant 'I1P readings at each of the 480 nodal (24 x 20) locations. The 480 comparisons provide an RMS error for each "TIP set"; 1.e., each exposure statepoint where TIPS were taken. De nodal RMS errors were statistically summed for each cycle and for all five cycles. The results are given in Table 5.5. For the five cycles the total RMS nodal error is - l 2.94%. ! Dis RMS error is bainning to approach the accuracy Itmitations of the '11Ps themselves. A measure of the instrumentation accuracy is the " Total TIP Uncertainty" provided in the VY Startup Test Reports"""* **'. Total TIP uncertainty is simply half of the RMS differences _ ; ebserved between mirror symmetric TIPS when the plant starts up in a symmetric rod pattern. Table 5.6 provides the plant total TIP uncertainty measured at BOC for the cycles included in the benchmark. Statistically combining the results for all cycles produces a five cycle TIP uncertainty of 1.74%. The model's 2.94% RMS error is very close to the intrinsic average instrumentation uncertainty of 1.74%. On the surface, this agreement appears to be sufficient. However, the nodal RMS erTors do not provide enough information tojudge the accuracy of the model. There i may still be some residual problems, either radially or axially, which get lost in the nodal RMS ' ctatisucs, because the vast majority of comparison points are good. Tojudge the model further, the radial and axial shapes of the instrument readings were individually examined for each TIP set. 5.3.1 Radial Comparisons Integrated TIP readings were created at each of the 20 instrument locations by adding up the 24 normalized nodal readings for each individual string. The resulting integral reading l fcr the given string : almost proportional to the relative reactor power in the four adjacent assemblies. By comparing the SIMUIATE-3 integrated '11Ps to the plant integrated TIPS, a map may be constructed showing the radial error. The integral ' LIP errors at each TIP location for each TIP set have been averaged and mapped for each cycle. These radial maps can be found in Appendix D. De errors at each location can also be averaged for all five cycles. The resulting map is shown in Figure 5.4 he 44 )
average radial errors are very close to zero. Therefore, on the average, the SIMUIA'IE-3 model is reproducing the plant radial power distribution quite well. However, the errors at any given location will make random step changes in going from cycle to cycle. 'Ihis may be seen in the specific cycle maps found in Appendix D. Oncc the step change is established, at BOC, the variation at any given T1P location is small and slowly varying with exposure. In theory, this step change at BOC is caused by. random shifts in the orientation of the
'ITP string and random variations in adjacent channel bowing created during fuel shuffling.
This changes the TIP asymmetry from cycle to cycle. Thus, the TIP integral errors shift at each location following refueling. Over the long haul (say, these five cycles), the impact of random
' IIP asymmetry should cancel. This is shown by the small average ermrs of Figure 5.4. Most radial locations in Figure 5.4 show an average error less than 1%. 'The worst location (near the periphery) is less than 2.2%.
As far as future predictions are concerned, any model of a future cycle may exhibit the magnitude of errors shown in the radial maps of Appendix D. At any given location, this could run 4-5%. Most of this error is due to TIP asymmetry. I 5.3.2 Axial Average Comparisons In a manner similar to the creation of integrated TIP readings, the TIP readings may be integrated for axial planes of the core. The resulting reading is almost proportional to the relative power in the given plane of the core. The SIMUIATE-3 core average (p'anar) TIP readings may then be compared to the plant core average (planar) ' IIP readings. Appendix D provides axial average comparisons for all TIP sets in the benchmark. The axial average comparisons, given in Appendix D, show ffood overall agreement between the model and the plant. However, there is a trend in the axial shapes exhibited by each cycle modeled. The SIMUIA'IE-3 model tends to underpredict the power in the bottom of the core, especially toward end of full power hfe. Figure 5.5 summa'izes the results of the axial average TIP trace comparisons made over } ell five cycles. Figure ' . 5 plots the average axial error in the planar TIP readings versus axial node. The i standarr deviations are also plotted. The bmadening of the standard deviations
. a near the bottom of the model is caused by the aforementioned tendency of the model to underpredict the bottom. The direction of the arrow indicates the general direction of the trend from beginning of cycle to end of full power life. (Note: the relatively large errors in the very bottom node should not cause any alarm. The power in the bottom of the core varies between 30% and 40% of reactor average. Therefore, it is never significant from a margin-to-limits ctandpoint.)
The trend shown in Figure 5.5 was also exhibited by the previous SIMULATE-2 model. It does not affect the licensing applications of the code. The limiting licensing transients near EOFPL are the pressurization transients. These transients are most sensitive to the scram reactivity function. The model's tendency to lose power in the bottom towards EOFPL will simply weaken the calculated (bottom entry) scram reactivity. The net result is that the pressurization transients will be calculated to be more severe than reality; more conservative 3 licensing limits will be generated. 5.4 Comparison of New Model to Current Model One of the means of judging the new code package is to compare it to the current code ) package of: MICBURN/ CASMO-2 / TABLES-2 / SIMULATE-2. The latter code package is also judged by the performance of the final code in the process stream; namely, SIMULNIE-2. Therefore, to be concise, we will refer to the comparison of the code packages as SIMULATE-3 versus SIMULATE-2. With regard to eigenvalues, Table 5.7 summarizes the behavior of both packages benchmarked over the same five cycles. SIMULATE-3 is clearly superior. It is closer to 1.0 for the hot model eigenvalue. The latter also exhibits a much smaller standard deviation. This is because the eigenvalue drift has been cut by a factor of 3. The BOC and EOFPL average hot K-effectives for the SIMUIATE-3 model are close to one another. This again indicates the small amount of eigenvalue drift. The same is not true of SIMULATE-2. For the latter, the spread between BOC and EOFPL is almost 0.7%AK. While this amount of drift has been an inconvenience, it really does not affect the licensing applications of SIMUIATE-2. This is because of the relative tightness of the BOC and EOFPL K-effectives as shown by the standard deviations. The BOC hot excess predictions and EOFPL
cycle length predictions can still be made with SIMUIEIE-2 with a good degree of accuracy. However, the degree of accuracy will be improved upon by using SIMUIATE-3. The cold eigenvalues are roughly comparable. However, SIMUIATE 3 holds a slight advantage in standard deviation. Less uncertainty in the cold eigenv'alue produces better cold critical predictions. Also notice that the hot-to-cold bias is significantly reduced. With regard to accuracy of the power distributions, the codes are comparable with a slight 1 edge given to SIMUIATE-3. The nodal RMS errors for the five cycles for SIMUIATE-2 are 3.19%. For SIMUIAIE-3 they are 2.94%. While this improvement is marginal, it is important > to keep in mind that these results for SIMUIEIE-3 were achieved without the use of edjustment factors. In the future, the elimination of adjustment factors will considerably cimplify the building of models which have new fuel types. l
.o.
TABLE 5.1 SIMUIATE-3 Cycle Average Hot Eigenvalues l l Number of
'1TP Set Qgig Statenoints Average Eigenvalue i Standard Deviation 9 49 0.99920 ,
0.00077 10 45 0.99870 0.00075 11 .39 0.99931 0.00115 12 43 0.99956 0.00069 13 47 0.99802 0.00071 5 Cycles 223 0.99893 0.00098 l_ I
i TABLE 5.2 SIMULATE-3 Beginning of Cycle Hot Eigenvalues i Number of Average BOC l TIP Set gygig Statenoints Eigenvalue 9' 5 0.99837 10 5 0.99874 11 5 O.99807 12 5 , 0.99924 13 5 0.99811 BOC Average . 0.99851 Standard Deviation 0.00049 TABLE 5.3 SIMULATE-3 End of Full Power Life Hot Eigenvalues Number of TIP Set gygig Statenoints EOFPL Eigenvalue 9 1 0.99931 10 1 0.99972 11 1 1.00055 12 1 1.00027 13 1 0.99904 EOFPL Average 0.99978 Standard Deviation 0.00063 l f
TABLE 5.4
- VY Cold Critical Case Conditions and SIMUIATE-3 Results
- Cycle Core Control Recirc. Reactor K-effecuve Cycle Exposure Average Rod Temp. Period A4)usted
. Dats Number (mwd /St) . (mwd /St) Scauence 2)__ {3g;L. for Period 11/25/81 9 0 9,192 Local 91 226 .99258 11/26/81 9 0 9,192 A1 96 78' .99717 12/07/81 9 91 9,283 A1 180 100 .99595 01/28/82 9 976 '10.168 B2 250 45 .99635 ;
06/10/82 9 3,700 12.892 Al 200 400 .99499 08/31/82 9 5,329 14,521 B2 240 400 .99616 03/09/83 9 8,945 18,137 A2 - 100 105 .99983 03/09/83 9 8,945 18,137 B1 101 58 .99927 ; 05/28/83 10 0 10,463 Al 103 223 .99681 06/17/83 10 0 10,463 Al 189- 75 .99818
.01/06/84 10 4,085 14.548 Al 225 120 .99729 06/19/84 10 7,343 17,806 A2 126 "100 .99681 06/19/84 10 7,343 17,806 B1 127 200 .99697 07/27/84 11' O 10,418 A1 94 160 .99662 08/06/84 11 0 10,418 A1 161 197 .99634 09/29/84 11 633- 11.051 Al 227 120 .99839 06/05/86 12 0 9,820 A2 83 185 .99850 06/30/86. 12 0 9,820 B2 158 87 .99841 07/02/86 12 0 9,820 B2 165 118 .99882 08/27/87 13 0 8,613 A2 87 55 .99628 10/02/87 13 0 8,613 A2 197 72 .99498 07/02/88 13 5,386 13,999 B2 209 70 .99406 08/26/88 13 6,329' 14,942 Al 228 142 .99473 08/28/88 13 6,329 14,942 Al 234 96 99513 Average: .99680 Standard Deviation: .00168
V
< TABLE 5.5 SIMULATE-3 Nodal TIP Reading RMS Errors I
Number of j TIP Set !
,Qyglg. Statcootnts RMS Enor (%) 1 2.955 '!
9 49 10 45 2.618 ! 11 39 3.201 12 43 2.557 13 47 3.293 5 Cycles 223 2.939 1 TABLE 5.6 2 1 VY Total TIP Uncertainties at Beginning of Cycle Number of Total TIP TIP Set Uncertainty Cycle Statenoints (%) 9 1 1,46 10 1 2.12 11 1 1.98 12 1 1.52 13 1 1.52 5 Cycles 5 1.74
)
. TABLE 5.7 P anita of SIMUIATE-3 Comanred to SIMUIA'IE-2 Benchmark Results for Cycles 9-13 !
1 Benchmark New Code Package Current Code Package - j
- FWure of Merit (SIMUIATE-3) (SIMULNIE-2) l q
Hot Eigenvalue .99893 '1.00839. l t Hot Standard Deviation .00098 i.00281- ! BOC Average Hot K-eff .99851. 1.00539 ! BOC Standard Deviation .00049 - .00081. EOFPL Average Hot K-eff .39978- 1.01223
.EOFPLi Standard Deviation .00063 .00078-Hot AK BOC to EOFPL .00127 .00684 Cold Eigenvalue .99680 .99775 Cold Standard Deviation 100168' i.00262 ,
Hot to-Cold Bias '.00213 .01064 Nodal RMS Error (%) 2.939 3.185 l.
FIGURE 5.1 SIMULATE-3 Hot Eigenvalues for Cveles 9-13 vs. Core Exoosure i 1.010 1.00e- a CYCLE 9 0: CYCLE 10 AVERAGE VALUE , a CYCLE 11 } 1006- + CYCLE 12 ...._ STANDARD DEVIATION x CYCLE 13 1.004-1.002-O 4 +t kA
- 4 1.000- g -
---------------------------------g--g----G-kA
+ * * +m
, 1 4 n++ @O y O ,
0.098- -------- - - y O_ _ _ _ ____ @ _ ______________ h 0.996-0.994-0992-0990 , , , . , , , . 16.0 77.0 18.0 19.0 20.0 8.0 90 10.0 11.0 12.0 13.0 14.0 15.0 CORE AVERAGE EXPOSURE (GWD/ST) 53 - l l l
FIGURE 5.2 SIMULATE-3 Cold Eigenvalues for Oveles 9-13 vs. Core Temocrature l i i
. 1.010 AVERAGE VALUE i
----. STANDARD DEVIATION t005-i O
f 1.000- a U
$g ....g.............................g..D........................g..................
l U U y
- -r U O D D D
-~
&~~~~-~~-~~~~~~~
0.995-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-~~nD~~~~~~~~~ D D D 0.990 , , , , , 275 17 5 200 225 250 75 100 12 5 15 0 TEMPERATURE (DEGREES F) L
FIGURE 5.3 SIMUIATE-3 Cold Eigenvalues for Cveles 9-13 vs. Core Exoosure
-)
i
'1010 !
I AVERAGE VALUE
----- STANDARD DEVIATION t005-to O
1.000- a f D LLJ Y l ............@.......e........................................................... D O g D - DD-, g D D 0.995- -- D--~~~-~~~~---------------- Ir------------B--------------------------------- D O D 0.990 , , , , , , , , , 9 10 12 13 14 15 16 17 18 19 20 8 11 CORE AVERAGE EXPOSURE (GWD/ST) l l e I f' i
FIGURE 5.4 SIMULATE.3 Averaged TIP Integral Errors. S. D..' and RMS Errors for Oveles 913
\
i .i 44 _ _ ,,j i
.__j _ KEY (in %)
.176 e _2.124 i AVERAGE DIFF.
42 ,I I ,
. 2.002 . . 1.576 i STANDARD DEV.
I l l .RMS DIFFERENCE 40 l __t__. l 2.on l l2.643 l _ _ , . _ _ __ _ . , _ _ _ _ _ _ , _ _ _ _ _ _ _ _ . .__4__ i i g g g i I, i i i 38 , g g
@ l l * - -
zi l @ l i 3s i i i l l l- e
. _ _ . _ _ _ _ _ _p _ _ ___;.__.__q____
_p_1.n9_.__;-_. _ _l _ _ _ _ _ l_ _ _ . .__.;___
. .ug ,
34 : _. 4. . . _.3n , , , l
' ' .9o ' 2.los l ; ;
, ; lau i 2.4 5 - l ;1.96o ;, i 2.ns i n , 1.5n i 1. 5n , i , _4_ _ _ _ _ ;._ _ .. _ _ .; _ _ __.l_____4__ ___'__..___l___ ___l______+__. i i l l i ! 30 i l l ' ' ' i
, l i
zi , ; @ l, -x - l i l I i 28 - i l l l ___l__ _ _ _ ;. _ _ .. _ _ _; _ _ _ _ _ _l_ _ __ _ _ 7 _ _ .. _ _ _; _ _ _
, ___+___.___l____ _I -2.120 _ l.. __ __.;i ._ _ _
1 -1.172 g 1.969 g i .907 I 26 - i t 1 I
.344 i t t .
g
. t
.922 i I
. 1.175 j I
1.239 , , 1.497 1.958 g g g g I i I i 1.898 g g 2.788 i i 1.292 I lr__.._-.;___ 2.423 24 - l 1.283 , _ _ _l_ _ ____,.__..__.l___ _ _ _p _ _ _ _ _ p _ _ .. _ _ .l _ _ _ _ _ _;_ _ ____+______l_____ i i i i ; i i , i l ! n- i . 3 1_ t i t t, i a i , I
. * . i .
ca . 20 - l , l l l l l l l __s___ ___;.__..__ ___ _ _ _ , _ _ _ _ _ _ , . _ _ .. _ _ _ , _ _ _ _ _ , _ _ _ . _ _ . _ _ _ _ _ _ , _ _ _ ___p__.__.___
.780 . sos .936 l 18 - l a .763 l l .257 j j j l l-
- 1. m , 1.14. , 1. u. , ..u
, .6,, ,
i , 3,37, j , 3,,o, j i 3,26o 10 8 1.o29 ! ! 1.184 __.;______l___ _ _ _ y _ . . _ _ .l _ _ ___,.__..__.__. _ _ q _ _ __ _ _ ;. _ _. . _ _ .; _ _ i i , , , i i l s4 l
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8 i i 8 ' 6 1 i g g g i 1 f e 4 I i _ _ .4 __ b. ..-4 i , i S i i i i I I i i i i i 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43
- 56 l
FIGURE 5.5 i SIMULATE 3 Core Average Axial TIP Errors and S.D. for Cycles 913 4 25 I 24 - O 23- . O 22- O 21- : O l 20- l O ; 19- . O : 18- : O l 17 - : O : l SIMUIATE 3 ' SIMUIATE-3 4- O
@ 16- UNDERPREDICTS OVERPREDICTS O
o 15-O : z u- ' O ' a l R 13- 1 0 ' F3 12 - , O l 1 11- O D to- O 2i 9-O : e -- : O : 7- l 0 l 6- O 5- ; O 4- l 0 ; Underpredicts 3- l 0 l Trend at Toward 2
. - Bottom 2- EOFPL "
1l O 0
-20.0 -10.0 0.0 10.0 20.0
% PLANAR ERROR ((MODEL - PLANT)/ PLANT)*100.
l l
i I
6.0 CONCLUSION
S When the program to upgrade the Vermont Yankee physics codes began, the goal was to produce a model superior to that currently used to support and heense the plant. By the figures of merit discussed in Sedion 5 this goal has been achieved. I In terms of hot and cold eigenvalue behavior, the new code package of MICBURN-3/CASMO-3/ TABLES-3/SIMUIAIE-3 is clearly superior to its predecessor. In terms of reproducing the plant TIP traces, the new code package is slightly improved. De radial comparisons are excellent, They are very close to the accuracy limitations that TIP asymmetry imposes upon the plant measurements. Axially, there are some consistent errors; however, what errors remain, will not hamper the plant support or licensing apphcations of the codes.
-- 'Ihe trend in the axial errors leads to more conservative licensing transient results.
Finally, the investigators unanimously ag'ee that the new code package of MICBURN-3/CASMO-3/ TABLES-3/SIMUIATE-3 is easier to learn and set up. The codes provide tracking functions for QA purposes, and they run faster on both the CYBER and VAX computer systems.
7.0' REFERENCES
- 1. Edenius, M. and Ahlin, A., "MICBURN: Microscopic Burnup in Gadolinia Fuel Pins,"
Chapter 7 of ARMP Comouter Code Mannata. CCM-3, dated November 1975.
- 7. Edenius, M., et al., CASMO-2: A Fuel .*---mbly Burnuo Pronram, Studsvik/NR-81/3, dated March 1981. (Proprietary)
- 3. Ver Planck, D. M., TABLES-2 Manual. YAEC-1391P, April,1983. (Pmprietary) !
-i
- 4. VerPlanck, D. M.. SIMUIEIE-E: A Nodal Core Analysis Program for tidht-Water Reactom, EPRI NP-2792-CCM, dated March 1983. ;
- 5. Pilat, E. E., Methods for the Analysis of Boiling Water Reactors Lattice Physics. YAEC-1232, dated December 1980.
- 6. Ver Planck, D. M., Methade for the Analysis of milina Water Reactom. Steady State Core Phvsics. YAEC-1238, dated March 1981.
- 7. Ietter and SER, USNRC to J. B. Sinclair. " Acceptance for Referencing in Licensing Actions -
for the Vermont Yankee Plant of Reports: YAEC-1232. YAEC-1238, YAEC-1239P, YAEC-1299P, and YAEC-1234," NVY 82-157, dated September 15,1982.
- 8. Ietter and SER. USNRC to R. W. Capstick, " Safety Evaluation by the Office of NRR Supporting Amendment 100 to Facility Operating License No. DPR-28," NYV 87-148, dated September 18,1987.
- 9. Harris, D. R., and Hebert, M. J., Methodoloav for Incore Gamma Effects, YAEC-1456P, June 1985. (Proprietary)
- 10. Ahlin, A., et al., MICBURN Microscoolc Burnuo in Burnable Absorber Rods, Studsvik/NFA-86/26, dated November,1986. (Proprietary)
- 11. DiGiovine, A. S., et al., CASMO-3G Validation, YAEC-1363, April,1988.
- 12. Edenius, M., et al., CASMO-3. A Fuel Assembly Burnuo Pmgram, Studsvik/NFA 86/7, dated November 1986. (Proprietary)
- 13. Ver Planck, D. M., et al., TABLES-3P Library Preparation Code for SIMULATE-3P. Studsvik
/SOA-88/02, dated February,1988. (Proprietary)
- 14. DIGlovine, A. S., et al., McGuire Unit 2 SIMULATE-3 Benchmark Analysis Cycles 1 through 3, YAEC-1608, dated October 1987.
- 15. VerPlanck, D. M., et al., SIMULATE-3P Advanced Three-Dimensional Two-Group Reactor Analysts Code, Studsvik/SOA - 88/01, dated February,1988. (Proprietary)
- 16. D1Giovine A. S., et al., SIMULATE-3 Validation and Verification, YAEC-1659, dated September 1988.
- 17. Delp, D. L., et al., FLARE A Three-Dimensional Boiling Water Reactor Simulator, GEAP-4598 (1964).
- 18. Goldstein, L., et al., " Calculation of Fuel-Cycle Burnup and Power Distribution of Dresden-I Reactor with the TRILUX Fuel Management Program," Trans. Am. Nuc. Soc.,10, 300 (1967).
- 19. Borrensen, S., "A Simplified, Coarse Mesh, Three-Dimensional Diffusion Scheme for Calculating the Gross Power Distribution in a Boiling Water Reactor," Nuclear Science and Engineering, d4, 37-43 (1971).
- 20. Smith, K. S., Rempe, K. R., ' Testing and Applications of the QPANDA Nodal Model" Proceedings International Meeting on Advances in Reactor Physics and Computation.
l Volume 2, p. 861, Paris, France, dated April,1987. I
- 21. General Electric Standard Application for Reactor Fuel (GESTARII), NEDE-24011-P-A-9, GE Company Proprietary, February 1988, as amended.
l
- 22. Schultz, S. P. and St. John, K. E., Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY1 Code /Model DeWotion Manual, YAEC-1249P, April 1981.
- 23. Schultz S. P. and St. John, K. E., Methads for the Analysis of oride Fuel Rod Steady-State )
Thermal Effects (FROSSTEY1 Code Qualification and ADDlication. YAEC-1265P, June 1981.
'24.14tter and SER, USNRC to R. W. Capstick, " Approval of Use of Fuel Performance Code FROSSTEY," NVY 85 205, September 27,1985.
- 25. Ansari, A. A. F., Methods for the Analysis of Boiling Water Reactors: Steady-State Core Flow Distribution Code (FIBWR), YAEC-1234, December 1980.
- 26. Ansari, A. A. F., et al., FIBWR: A Steady State Core Flow Distribution Code for Boiling
- i. Water Reactors - Code Verification and Qualification ReDort. EPRI NP-1923, Project 1754-1 Final Report, July 1981.
- 27. Lellouche, G. S. and Zolotar, B. A., Mechanistic Model for Predicting Wro-Phase Vold Fraction for Water in Vertical Tubes Channels, and Rod Bundles EPRI NP-2246 SR, i February,1982.
- 28. Letter E. W. Jackson to USNRC (Region 1), " Cycle 9 Startup Test Report," FVY82-21, dated February 25,1982.
- 29. letter W. P. Murphy to USMRC (Region I), " Cycle 10 Startup Test Report," FVY83-100, dated September 15,1983.
30, letter, W. P. Murphy to USNRC (Region I), " Cycle 11 Startup Test Report," FVY84-132, dated November 6,1984.
- 31. letter R. W. Capstick to USNRC (Region I), " Cycle 12 Startup Test Report," FVY86-92, dated October 6,1986.
- 32. letter, R. W. Capstick to USNRC (Region I), " Cycle 13 Startup Test Report " FVY88-1, dated January 4,1988.
I h
'q q
- 33. Personal Communication (1988) with D. M. VerPlanck Studsvik of America,1087 Beacon Street, Newton, Massachusetts 02159, i
- 34. 'Iwo papers, W. A. Golub. " Effects of Channel Bow " and T. Rausch, " Effects of Channel i Bow," Station Nuclear Engineer's Conference, June 20-24, 1988, San Jose, Cahfronia. I 1
I
- 35. Gorman, J A. and Lipsey, G. W., An A-e.amment of BWR Fuel Channel Lifetimes, EPRI NP-2483, July 1982.
- 36. Gonnan, J. A., et al., Fuel Channel Lifetimes: Statistical Predictive Models, EPRI NP-3937 and NP-3938, February 1984.
- 37. Gorman, J. A., et al., Fuel Channel Lifetimes: Exoansion of Statistical Predictive Models, Final Report, EPRI NP-4225M. September 1985.
- 38. Gorman, J. A., and Turner A. P. L., Inveaflaation of Large Bows in Reused BWR Fuel Channels. EPRI NP-5718P, April 1988.
- 39. Telecon, J. W. Heard with D. T. Weiss (GE), February 1,1989.
APPENDIX A HOT DEPLERON STA'IEPOINTS During the five cycles modeled here, some 630 complete sets of MP data were taken at the plant. To reduce the number of comparisons to a manageable number, most non-equilibrium xenon TIP sea were eliminated from the benchmark. What follows is a l description of each cycle and the depletion steps used to modelit. Generally, the depletion prograd from 'I1P set to TIP set, using the plant conditions ' of the concluding TIP set. 'Ihe preferred depletion interval between TIP sets was two weeks, but it often varied from one to three weeks. The length of the interval roughly coincided with the frequency of rod moves. When the TIP machine was broken for an extended period, or a TIP set was not taken near the end of a sequence, a deoletion step was inserted at suitable plant conditions. In these instances, no comparison to plant data was made in the benchmark. Figure A.1 shows the reload design for the baginning of Cycle 9. 'Ihe darker the chading, the older the batch of fuel. Figure A.2 shows the plant power versus exposure for Cycle 9. 'Ihe statepoints where MP data set comparisons were made are indicated. The plant ran close to full capacity at all times, so most statepoints are near full power, except during - coastdown. Figure A.3 shows the md inventory for the cycle with the modeled statepoints indicated. The statepoints are evenly distributed among A and B sequences. Finally, Figure A.4 provides the reactor conditions at the comparison statepoints. Also shown are the additional depletion steps where no comparisons were made. Figures A.5-A.8 provide similar information for Cycle 10. Figures A.9 A.12 for Cycle 11. Figures A.13-A.16 for Cycle 12. Figures A.17-A.20 for Cycle 13. l A-1 l
FIGURE A.1 i I Reload Design of Cycle 9 44 P[ h 42 F M MM 4o M M @ ^N PJB M M E M 2e TcF#fs FA&MM M RES.i as
%dB kn *M MW M T@lf MM ll $ adwYimY 3.i[$
MTdMNBF #_ m M K4 . D M . M 2e n u m m a em a a n n F42rd^ W MiE#AM 24 {@f rds 2 M iWeFAWJ A ;; 2 ,s lq M & M M E8i e
; :: "g ,8 - sy W- -
W$}$sc
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xe ggggg e JiE42 N2 R M t@s lG RUMM MMt W*S M. I R M l '; ?Q, g age g eg g gM L > Vd f
== m - a mm.
i W l A-2
FIGURE A.2 j Power History of Cvele 9 Showing TIP Statenoints w
"g 68
! 'b g
-g ,
j -4 8 ,; I d'
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- e ~
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FIGURE A.3 Control Rod Inventory of Ovele 9 Showing TIP Statenoints
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l
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............................ .......................................+.......__.. ;
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009 006 ODE Out 001 0 - 03183SNI S3H010N 008 1081N00 A-4
FIGURE A.4 Reactor Conditions for Cycle 9 Depletion Steps i Step TIP No. Date Rod Pattern Power Flow Exposure 4 0 No TIP 12/1/81 BOC Startup NA NA 9192 1 631 12/14/81 Al-1, Deep @ 14 1589 45.6 9324 2 640 1/5/82 Al-1, Deep @ 14 ' 1593 45.6 9742 3 641 1/25/82 Al-1, Deep @ 10 1591 47.3 10168 4 654 2/4/82 B2-1, Deep @ 10 1589 45.2 10302 5 655 2/10/82 B2-1, Deep @ 10 1592 46.6 10434 6 656 2/24/82 B2-1, Deep @ 08 1593 46.7 10731 7 658 3/2/82 B2-1, Deep @ 08 1590 45.9 10863 8 No TIP 3/13/82 B2-1, Deep @ 08 1592 44.7 11077 9 664 3/17/82 A2-1, Deep @ 04 1592 45.5 11177 10 671 4/6/82 A2-1, Deep @ 04 1592 46.9 11579 11 672 4/15/82 A2-1 Deep @ 00 1590 46.7 11772 12 673 4/22/82 A2-1 Deep @ 00 1593 47.4 11918 13 679 4/29/82 B1-1, Deep @ 08 1584 45.8 12037 14 681 5/4/82 B1-1 Deep @ 06 1592 46.7 12148 15 682 5/18/82 B11 Deep @ 04 1591 47.1 12444 16 684 5/27/82 B1-1. Deep @ 04 1591 46.3 12640 17 685 6/8/82 B1-1 Deep ' 04 1591 46.2 12892 18 692 6/15/82 Al-2, Deep 4. 06 1591 45.9 12980 19 693 6/25/82 Al-2, Deep @ 06 1592 47.2 13197 20 695 7/1/89 Al-2, Deep @ 06 1592 47.4 13S2S 21 696 7/7/82 Al-2, Deep @ 06 1591 47.4 13450 f 22 697 7/16/82 Al-2, Deep @ 06 1591 46.4 13647 23 698 7/22/82 Al-2 Deep @ 06 1591 46.6 13774 24 701 7/28/82 B2-2, Deep @ 04 1593 47.6 13894 25 702 8/4/82 B2-2, Deep @ 06 1592 47.1 14043 26 703 8/4/82 B2-2, Deep @ 06 1591 47.1 14046 ) 27 704 8/11/82 B2-2, Deep @ 06 1591 47.2 14198 A-5 }
FIGURE A.4 (Continued) i q l Reactor Conditions for Cycle 9 Deoletion Steos - ] l l i p gigg TIP No. Dag. . Rod Pattem Egw_g E]gE Entgagg 28 709 8/19/82 B2 2 Deep @ 04 1590, 45.8 14339 l 29 716. 9/9/82 B2-2. Deep @ 06 1589' 47.5 14896 30 719 9/13/82 A2 2, Deep @ 04 1585 47.9 14784 { 31 720 9/17/82 A2-2, Deep @ 04 1589 47.6 14862 l 32 722 9/21/82 A2-2, Deep @ 04 1588 46.8 14943 ; 33 723 9/28/82 A2-2. Deep @ 04 1587 46.0 15093 34 725 10/5/82 A2 2, Deep @ 06 1589 46.2 15245 35 727 10/12/82 A2-2. Deep @ 06 1589- 46.1 15394 36 734 10/20/82 A2-2. Deep @ 06 1591 46.3 15490 37 735 10/26/82- A2-2, Deep @ 06 1588 47.5 15623 38 736 10/28/82 A2-2, Deep @ 06 1589 47.3' 15666 39 739 11/2/82 B1-2, Deep @ 10 1589 47.2 15758 40 741 11/9/82 B1-2 Deep @ 12 1589 45.8 15915 45.6 41 743 11/17/82 B1-2, Deep @ 12 1589 16084 42 745 11/23/82 B1-2, Deep @ 20 1588 46.9 16211
'43 747 12/1/82 B1-2. Deep @ 24 1590 46.1 16380 44 749 10/7/82 B1-2, Deep @ 30 1589 46.7 16506 45 750 12/15/82 B1-2, Deep @ 30 1590 47.8 16676 46 751 12/28/82 B1-2, Deep @ 30 1524 47.8 16944 47 753 1/4/83 B1-2, Deep @ 30 1493 47.9 17092 48 762 1/17/83 ARO 1470 47.9 17291 49 765 2/1/83 ARO 1397 47.7 17581 50 766 2/16/83 ARO 1327 47.7 17853 51 No TIP 3/1/83 ARO 1277 47.6 18077 52 No TP 3/5/83 ARO to EOC 1263 48.0 18137 l
A-6 l
i l FIGURE A.5 ) Reload Deslan of Cycle 10 h 42
% M MM MNd$r@M M^M MMM 4o >8 PPRWa M M M MM se IMP} M M* M 'E sss ist M 32 3h & M >o maammm gr g gg a 'jggyg n o*gs g-g 2ejgg 2eMsn E M @ MR sss REem 24 M9 s m "' W / R f B MWMM" a t#M 22 MM E23 E t QTd EB M EMMM saa sssg 2o MM M
- M sfa EM M N 19 mm E
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FIGURE A.7 Control Rod Inventory of Cvele 10 Showing 'ITP Statenoints a 3 .. k% lc.g- 5
+.. '
+ a
+ A,. -
t n - g
+
+ -
+ .'l' 5
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FIGURE A.8 i Reactor Conditions for Cvele 10 Deolclion Steos ) 2112 11P No, Dalt Rod Pattern Egy.g E]gg Exposure l 0 No TIP 6/17/83 BOC Startup NA NA 10463 l 1 777 6/29/83 Al-1, Deep @ 18 1590 46.7 10002 1 2 785 7/12/83 Al-1, Deep @ 18' 1591 47.8 10850 3 786 7/26/83 Al-1 Deep @ 18 1592 46.9 11150 ; 4 788 8/2/83 Al-1. Deep @ 18 1591 48.5 11304 5 789 8/11/83 Al-1 Deep @ 16 1592 47.3 11496 l 6 793 8/16/83 B2-1. Deep @ 12 1591 . 44.5 11592 7 794 8/24/83 B2-1 Deep @ 10 1592 46.6 11765 i 8 803 9/6/83 B21 Deep @ 08 1590 46.0- 12002 9 804 9/14/83 B2-1 Deep @ 08 1588 46.5 12170 10 805 9/21/83 B2-1 Deep @ 08 1590 46.0 12320 11 806 9/29/83 B2-1 Deep @ 06 1590 46.3 12490 12 810 10/4/83 A2-1. Deep @ 04 1591 47.1 12574 13 812 10/5/83- A21 Deep @ 04 1589 46.6 12615 14 813 10/14/83 A2-1. Deep @ 04 1591 46.5 12807 15 814 10/18/83 A21 Deep @ 08 1592 46.6 12890 16' 815 10/25/83 A2-1. Deep @ 04 1590 46.9 13040 17 817 11/1/83 A2-1 Deep @ 00 1590 46.6 13191 18 821 11/9/83 B1-1, Deep @ 10 1591 46.4 13354 19 822 11/17/83 B1-1. Deep @ 08 1591 47.2 13524 20 823 11/22/83 B1-1, Deep @ 08 1591 47.0 13631 21 824 12/5/83 B1-1. Deep @ 06 1591 46.9 13914 22 826 12/8/83 B1-1, Deep @ 06 1591 46.8 14002 23 827 12/12/83 B1-1, Deep @ 06 1592 46.6 14064 Al-2 Deep @ 10 46.9 24 831 12/20/83 1590 14222 25 832 12/28/83 Al-2, Deep @ 10 1593 47.3 14394 26 834 1/4/84 Al 2, Deep @ 10 1590 47.4 14548 27 840 1/10/84 Al-2, Deep @ 10 1587 45.5 14626 A 10 1
FIGURE A.8 (Continued) Reactor Conditions for Ovele 10 Deoletion Steos 31g2 TIP No. Rod Pattern E2Eg Egg Exposum Dats ] 28 841 1/19/84 Al-2, Deep @ 10 1588 46.0 14822 i 29 849 2/3/84 B2-2. Deep @ 06 1591 46.1 15048 30 851 2/7/84 B2-2 Deep @ 06 1592. 46.6 15132 31 853 2/14/84 B2-2, Deep @ 08 1588 46.0 15282 32 855 2/28/84 B2-2, Deep @ 10 1593 45.8 15580 33 859 3/6/84 A2-2 Deep @ 06 1588 47.1 15702 34 861 3/15/84 A2-2 Deep @ 06. 1591 46.5 15909 35 863 3/28/84 A2-2, Deep @ 10 1592 46.4 16167 ,
~
36 864 4/4/84 A2-2, Deep @ 12 1591 46.8 16333 37 866 4/10/84 A2-2, Deep @ 14 1592 46.7 16466 38 872 4/24/84 B1-2, Deep @ 12 1591 46.7 16722 39 874 5/1/84 B1-2, Deep @ 18 1590 4L 2 16870 40 875 5/4/84 B1-2, Deep @ 18 1588 . 46.9 16936 I 41 877 5/8/84 B1-2, Deep @ 24 1590 46.2 17016 42 879 5/15/84 B1-2 Deep @ 30 1587 47.2 17141 43 881 5/23/84 B1-2, Deep @ 42 1573 48.0 17333 44 882 5/30/84 B1-2 Deep @ 42 1538 47.7 17478 45 884 6/6/84 ARO 1502 47.7 17623 46 No TIP 6/15/84 ARO to EOC 1457 48.0 17806 f l l [ A-11 L
i l FIGURE A.9 Reload Desina of Cycle 11 44
$ MMM Ed^ h MMM
>. M M M a M MM se MMfB E *M MBr M E5* MMM L 24 M M P/AR RM M M 32
% %i % MMPN&i% Bl' M M l >e w 74 m ag MMM 2e gg gggg gst g glm gm e M M e 2e 4 M fd MM MMEM EM M RM 2442 ER' M s MMMlMMiiE"' N W.'6 M :
22 f M M M 6 MffefMM MMtR H E E 2o WA E 1 x MURAM MM M MM
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e Yh gc g 0%g* g WMM MMM M MMM e hh 4 TAL%M MMM l a MEMEMF#4 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 A-12 f l
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FIGURE A.11' Control Rod Inventory of Cvele 11 Showing TYP Statenoints
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FIGURE A.12 Remetor Con @tiana for Cycle '11 Deoletion Steos jgg, TIP No. 233 - Rod Pattern Egw.ggr & ExDO6Ure O N/A 8/6/84 BOC Startup NA NA 10418 1 894 8/23/84- Power Ascent 1274 44.1 10592 2 895 9/4/84 Power Ascent 1272 44.3 10780 3 897 9/4/84 Power Ascent 1381 33.3 10832 4' 898 9/14/84 Al-1, Deep @ 14 1519 46.8 10965 5 904 .10/9/84 Al-1. Deep @ 14 1593 46.8 11192 6 907 10/18/84 Al-1. Deep @ 14 1590 46.6 11379 7 917 10/29/84 B2-1, Deep @ 06 1584 44.0 11576 8 918 11/2/84 B2-1, Deep @ 06 1591 46.7 11641 9 920 11/7/84 B2-1, Deep @ 06 1592 46.5 11773 10 921 11/15/84 B2-1. Deep @ 04 1590 47.2 11943 11 922 11/26/84 B2-1 Deep @ 04 1591 46.2 12176 12 924 12/04/84 B2-1. Deep @ 00 1592 46.8 12348 13 925 12/13/84 B2-1 Deep @ 00 1591 46.2 12540 14 927 12/21/84 A2-1, Deep @ 04 1591 .47.7 12724 15 929 1/3/85 A2-1 Deep @ 06 1592 47.0 12977 16 931 1/23/85 A2-1. Deep @ 04 1593 46.4 13402 17 932 1/29/85 A2-1, Deep @ 04 1592 46.1 13530 18 937 2/5/85 B1-1, Deep @ 10 1590 46.4 13681 19 943 02/12/85 B1-1, Deep @ 10 1589 46.2 13799 ) 20 944 02/26/85 B1-1, Deep @ 10 1590 46.4 14101 21 947 03/07/85 B1-1 Deep @ 10 1591 46.4 14295 22 948 03/19/85 B1-1 Deep @ 10 1592 46.7 14551 23 951 03/26/85 Al-2 Deep @ 12 1593 46.2 14692 24 952 04/04/85 Al-2, Deep @ 12. 1593 46.9 14884 i 25 955 04/15/85 Al-2, Deep @ 12 1585 45.5 15118 26 956 04/18/85 Al-2. Deep @ 12' 1593 46.9 15179 A-15 )
FIGURE A.12 (Continued) Reactor Conditions for Cvele 11 Deoletion Steos I gig ' IIP No. 231g. Rod Pattern Egw.E E]gy EKDo8ure , 27 957 04/23/85 Al 2. Deep @ 14 1590 45.6 15289 28 959 05/07/85 Al-2, Deep @ 16 1590 46.0 15588 29 -960 05/14/85 Al-2 Deep @ 18 1590 45.5 15732-30 963 05/21/85 B2-2, Deep @ 10 1590 44.9 15871 31 966 06/06/85 B2-2, Deep @ 14 1590 46.5 16210
'32 968 06/12/85 B2-2, Deep @ 24 1589 44.4 16338 33 971 06/19/85 B2-2, Deep @ 32 1592 ,
45.4 16481 34 974- 07/02/85 ARO* Deep @ 46 1583 47.6 16760 35 975 07/17/85 ARO*, Deep @ 46 1529 47.7 17069 36 976 08/08/85 ARO*, Deep @ 46 1430 47.7 17523 37 978 08/21/85 ARO*, Deep @ 46 1372 47.6 17697 38 980 09/03/85 ARO*, Deep @ 46 1326 47.8 17989 39 981 09/19/85 ARO*, Deep @ 46 1261 47.7 18260 40 No TIP 09/21/85 ARO* to EOC 1200 48.0 18283
- In Cycle 11 VY had a single rod " impeded" at 46; that is, it gave no settle indication at 48.
'Ihe mirror symmetric rods were also ddven to 46. Therefore, during coastdown the model
) {: was nearly ARO. A.ie )
i FIGURE A.13 Feload Design of Cycle 12
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) FIGURE A.16 Reactor Conditions for Cvele 12 Deoletion Steps , j{gp TIP No. Date Rod Pattern Power Flow Exposure O No TIP 06/30/86 BOC Startup NA NA 9820 ..g 1 995 07/15/86 A2-1, Deep @ 12 1588 46.1 9983 2 996 07/24/86 A2-1, Deep @ 14 1581 45.6 10170 3 997 07/81/86 A2-1 Deep @ 12 1594 46.9 10320 4 999 08/12/86 A2-1. Deep @ 12 1590 45.6 10580 l 1 5 1002 09/05/86 A2-1, Deep @ 12 1591 45.2 11082 6 1005 09/09/86 B1-1, Deep @ 10 1589 46.4 11162 . L 7 1006 09/16/86 B1-1, Deep @ 10 1590 46.0 11307 8 1007 09/17/86 B1-1, Deep @ 10 1592 46.0 11341 l 9 1008 09/30/86 B1-1. Deep @ 10 1592 45.9 11606 10 1017 10/10/86 B1-1, Deep @ 10 1591 45.4 11751 - 11 1018 10/16/86 B1-1. Deep @ 10 1590 45.0 11875 12 1020 10/21/86 B1-1, Deep @ 10 1589 46.2 11983 13 1021 10/30/86 B1-1. Deep @ 10 1588 47.0 12173 14 1024 11/05/86 Al-1, Deep @ 12 1593 46.1 12298 Al-1, Deep @ 12 46.8 15 1026 11/13/86 1593 12465 16 1027 11/25/86 Al-1. Deep @ 12 1593 46.2 12721 ' 17 1029 12/04/86 Al-1, Deep @ 12 1592 46.7 12916 . 18 1030 12/18/86 Al-1, Deep @ 12 1593 45.6 13213 , 19 1033 12/23/86 B21 Deep @ 06 1589 46.2 13312 20 1034 01/02/87 B2-1, Deep @ 06 1592 45.7 13525 21 1036 01/06/87 B2-1. Deep @ 04 1591 47.1 13615 22 1037 01/20/87 B2-1, Deep @ 04 1591 46.8 13913 - 23 1038 02/04/87 B2-1 Deep @ 04 1593 46.8 14230 i 24 1043 02/10/87 A2-2, Deep @ 06 1591 45.9 14358 l 25 1044 02/24/87 A2-2, Deep @ 06 1593 47.2 14653 26 1046 03/03/87 A2 2 Deep @ 06 1591 45.6 14804 A-20
FIGURE A.16 (Continued) Reactor Conditions for Cycle 12 Depletion Steps I Step TTP No. Date Rod Pattern Power Flow ExDosure I l 27 1047 03/11/87 A2-2, Deep @ 06 1592 . 46.7 14970 28 1048 03/17/87 A2-2, Deep @ 08 1591 46.9 15098 29 1050 03/25/87 A2-2, Deep @ 08 1591 45.7 15267 30 1052 04/01/87 A2-2, Deep @ 08 1591 46.8 15421 l 31 1063 04/14/87 B1-2, Deep @ 10 1593 46.8 15669 I 15818 31 1064 04/21/87 B1-2. Deep @ 14 1590 46.4 j 33 1066 04/28/87 B1-2, Deep @ 18 1591 46.1 15970 34 1068 05/06/87 B1-2, Deep @ 24 1594 46.4 I,6137 , 35 1072 05/13/87 B1-2, Deep @ 40 2590 46.7 16273 36 1073 05/19/87 B1-2, Deep @ 40 1583 47.4 16402 37 1074 05/26/87 B1-2, Deep @ 40 1557 47.4 16555 l 38 1076 06/04/87 ARO 1526 48.0 16740 39 1078 06/18/87 ARO 1473 47.6 17021 40 1079 07/01/87 ARO 1430 47.8 17265 41 1081 07/09/87 ARO 1398 48.0 17426 42 1082 07/28/87 ARO 1324 47.8 17763 43 1083 08/06/87 ARO 1285 47.7 17926 44 No TIP 08/07/87 ARO to EOC 1280 48.0 17949 l l ) A-21 I f I
- - - - - ~ - - -- __. _
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FIGURE A.20 l 1 I Reactor Conditions for Cvele 13 Deoletion Steos j l l M TIP No. Date Rod Pattern f.2Efz Dmg Exposure O No Tip N/A BOC Startup N/A N/A 8613 1 1103 10/21/87 A2-1. Deep @ 18 1592 47.6 8897 l 2 1104 10/28/87 A2-1 Deep @ 18 1590 47.5 9047 3 1105 11/04/87 A2-1. Deep @ 18 1591 46.8 9195 4 1116 11/17/87 A2-1, Deep @ 16 1592 45.8 9406 l L. 5 1118 12/02/87 A2-1, Deep @ 16 1592 45.2 9725 l 6 1123 12/09/87 B1-1 Deep @ 12 1591 45.7 9866 7 1127 12/29/87 B1-1 Deep @ 10 1591 45.9 10288 8 1130 01/06/88 B1-1 Deep @ 08 1591 46.3 10463 ) 9 1131 01/12/88 B1-1, Deep @ 08 1593 45.8 10588 l 10 1132 01/19/88 B1-1, Deep @ 08 1590 46.3 10737 I 11 1134 01/28/88 B1-1, Deep @ 08 1589 44.9 10929 12 1138 02/04/88 Al-1 Deep @ 12 1592 46.2 11068 13 1140 02/09/88 Al-1, Deep @ 12 1592 45.4 11173 14 No TIP 02/16/88 Al-1, Deep @ 12. 1592 46.3 11322 > 15 1142 02/25/88 Al-1, Deep @ 10 1593 46.3 11514 16 1144 03/08/88 Al-1. Deep @ 10 1592 46.6 11772 17 1146 03/18/88 Al-1. Deep @ 10 1593 .45.6 11980 18 1151 03/25/88 B2-1, Deep @ 06 1592 46.4 12118 19 1153 04/06/88 B21, Deep @ 04 1593 46.8 12370 20 1155 04/19/88 B2-1 Deep @ 04 1592 46.0 12649 j 21 1156 04/26/88 B2-1. Deep @ 00 1591 46.9 12796 l 22 1161 05/10/88 A2-2. Deep @ 10 1593 46.6 13085 23 1163 05/13/88 A2-2, D:ep @ 10 1591 46.5 13154 24 1164 05/20/88 A2-2, Deep @ 10 1591 46.2 13299 25 1165 05/26/88 A2-2, Deep @ 10 1593 46.1 13425 26 1166 06/03/88 A2 2, Deep @ 10 1592 46.6 13596 A-25
f FIGURE A.20 (Continued) Reactor Conditions for Cycle 13 Deoletion Steps l l Step TTP No. Date Rod Pattern Power Flow Exposure 27 1174 06/23/88 A2-2, Deep @ 10 1590 46.0 13999 28 NoTIP 07/18/88 B1-2, Deep @ 04 1588 46.3 14186 l 29 1193 07/22/88 ill-2, Deep @ 04 1592 46.7 14268 30 1194 05/26/88 B1-2, Deep @ 06 1591 46.9 14415 l 31 1197 08/02/88 B1-2. Deep @ 06 1586 46.0 14481 ) 32 1198 08/05/88 B1-2, Deep @ 06 1695 46.6 14564 i 33 1201 08/17/88 B1-2, Deep @ 08 1592 46.1 14814 I f 34 1202 08/23/88 B1-2, Deep @ 08 1592 46.5 14942 35 1213 09/02/88 Al-2. Deep @ 08 1589 45.9 15054 l 36 No TIP 09/16/88 Al 2, Deep @ 08 1593 47.9 15352 37 1219 09/19/88 Al-2, Deep @ 10 1595 46.1 15413 > 38 1220 09/27/88 Al-2 Deep @ 12 1590 45.4 15578 39 1223 10/07/88 Al-2, Deep @ 12 1592 47.4 15787 40 1225 10/11/88 Al-2, Deep @ 14 1591 46.3 15871 41 1227 10/18/88 Al-2 Deep @ 16 1590 46.3 16020 42 1228 10/25/88 Al-2, Deep @ 18 1591 45.9 16169 ) 43 1232 11/02/88 B2-2, Deep @ 16 1591 47.5 16327 44 1234 11/08/88 B2-2 Deep @ 18 1590 46.3 16454 45 1237 11/15/88 B2-2, Deep @ 24 1591 46.1 16602 46 1239 11/22/88 B2-2, Deep @ 38. 1592 46.4 16748 l 47 1240 11/29/88 B2 2 Deep @ 38 1584 48.0 16898 } 48 1241 12/06/88 B2-2. Deep @ 40 1554 47.8 17045 7 49 1242 12/13/88 B2-2, Deep @ 40 1525 47.8 17187 l l 50 1244 01/03/89 B2-2, Deep @ 40 1442 47.8 17606 Note: The Cycle 13 Benchmark was concluded in January, prior to shutdown. VY continued to operate until 2/11/89 and shutdown at 18,307 mwd /St. A-26 l l
APPENDIX B COLD CRmCAL STATEPOINTS f 1 i
'Ihe following figures give the control rod patterns for the cold criticals performed at Vermont Yankee during the period of the benchmark. All but one of the criticals are !
in-sequence criticals. There is only one local critical. The latter is, a critical performed in a localized region of the core by pulling two adjacent control rods. VY Technical Specifications do not allow the pulling of " face-adjacent" control rods in either shutdown margin or cold critical demonstrations. The beginning of Cycle 9 was the only time, in 'recent history, that criticality could be achieved using diagonally adjacent rods. Figure B.1 provides the cold critical patterns for Cycle 9. Figures B.2-B.5 provide the same for Cycles 10 through 13, respecuvely. Also provided are the approximate hours of xenon depletion at zero power which preceeded the cold critical. This information is important to the quality of the data. It is not sufficient to run these cases at zero xenon. In some instances, f the time between scram and recovery was sufIlciently short that failure to model the xenon depletion shifts the eigenvalue significantly. 1 l I I t l B-1
FIGURE B.1 Ovele 9 Cold Critical Patterns CRmCAL PATTERN CRmCAL PATTERN 43 4: 39 39 48 48 48 48
]
i 35 48 35- 'l 31 16 31 48 48 48 r- 27 - l 23- 23- 48 48 48 48 48 48 19 - 19 - 15 15 48 12 48 11 11 07 I 07 48 48 48 48 l 03 0* 1 I I I I I I 02 06 to 14 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 i' DATE: 11/26/81 PERIOD: 78 SEC. E FRS: >1000 f DATE: 11/25/81 PERIOD: 726SEC. E FRS: C000 CRmCAL PATTERN CRmCAL PATTERN 43 43 6 6 48 48 39 48 48 48 39 f48 l48
- 35 4 4 35 6 6 31 48- 48 48 48 31 48 48 48 48 48 27 - 6 6 8 6 27 - 4 23 48 48 48 48 48 23- 48 48 48 48 48 4 4 19 - 6 8 6 6 15 48 48 48 48 15 48 48 48 48 48 11 11 6 6 J
1 1 07 48 48 48 48 07 48 48 48 03 4 03 6 6 I I I I i 1 L 02 06 10 14 18 22 26 30 34 38 42 02 06 to 14 18 22 26 30 34 38 42 DATE: 12/07/81 PERICL-. 100 SEC. E 2: 46 DAX: 01/28/82 PERIOD: 45 SEC. E PRS: 48 B-2 'rm-lu 1i- ri 9 iii - . . . . .
FIGURE B.1 (Continued) Ovele 9 Cold Critkal Pattems CRmCAL PATTERN CRITICAL PATTERN 00 40 39 48 48 48 48 39 48 48 48 , 35 35 ____ 48 48 48 48 31 48 48 48 48 31 27 - 27 - 48 48 48 48 48 23 - 48 48 6 48 33- 48 19 - 19 - I 48 48 48 15 48 48 48 48 19 11 11 07 48 48 48 48 07 48 48 48 0: 0: I I I I I I i 02 06 10 14 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38'42-DATE. 06/10/82 PERIOD: 400 SEC. E 2: 47 DATE: 08/31/82 PERIOD: 400 SEC._ XE 2 : 88 CRITICAL PATTERN CRITICAL PATTERN j 1. 43 43 l l l 39 48 48 48 48 39 48 48 48 1 4 35 4 4 35 S? 48 48 48 48 31 48 48 48 48 48 27 - 4 27 - 4 33- 48 48 48 48 48 48 23 - 48 48 48 48 48 9- 4 4 19 - 4 15 48 48 48 48 15 48 48 48 48 48 11 11 l 07 48 48 48 48 07 48 48 48 03-- 4 0; i l I i I I I 02 06 to 14 18 22 26 30 34 38 42 02 06 10 14 18 22 16 30 34 38 42 DATE. 03/09/83 PERIOD: 105 SEC. E 2: 108 DATE. 03/09/83 PERIOD: 58 SEC. XE 2: 110 , B-3
FIGURE B.2 Cvele 10 Cold Critical Patigna j d CRmCAL PATTERN CRmCAL PATTERN 43 4; 4 39 48 48 48 48 39 48 48 48 48
'35 35 4 4 31 48 12 48 48 31 48 48 48 48 y_ D- 4 8 4 33- 48 48 48 48 48 48 23- 48 48 48 48 48 48 19 - 19 - 4 4 4 15 48 48 48 15 48 48 48 48 11 11 4 4 07 48 48 48 07 48 48 48 48
,48 0: 0: 4 l l I i Ti 1 02 06 to 14 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 DATE.05/28/g PERIOD: 223 SEC XE efts: >1000 DATE: 06/17/83 PERIOD: 75 SEC- XE HRS: >1000 CRmCAL PATTERN CRmCAL PATTERN 03 10 43 39 48 48 48 ' 39 48 48 48 48 48 l 35 to 10 35 31 48 t.8 48 48 31 48 12 48 48 27 - 10 12 to 27 -
23- 48 48 40 48 48 48 23- 48 48 48 48 48 48 19 - 10 10 12 19 - 15 48 48 48 48 15 48 48 48 11 10 10 11 07 48 48 48 48 07 48 48 48 48 0; 12 03 i l ) l l l 02 06 to 14 18 22 26 30 34 38 42 02 06 to 14 18 22 26 30 34 38 42 DATE: 01/06/84 PERIOD: 120 SEC. XE 2 : 32 DATE: 06/19/84 PERIOD: 100 SEC XE 2 : 75 B-4
FIGURE B.2 (Continued)
.J Cvele 10 Cold Critical Patterns CRmCAL PATTERN CRmCAL PATTERN 4: 4:
39 48 48 48 39 35 4 35 31 48 48 48 48 48 31 i n 27 - 4 27 - 1 23- 48 48 48 48 48 23 - l' 19 - 4 19 - 15 48 48 48 48 48 15 11 11 07 48 48 48 07 0 03 l l 1 l l I I 02 06 10 14 18 22 26 30 34 38 .42 02 06 to 14 18 22 26 30 34 38 42 CATE: 06/19/84 PERICD; 200 TC. E MtS: 75 DATE: PERIOD: SEC. XE NtS: CRmCAL PATTERN CRmCAL PATTERN 43 4 39 39 35 35 31 31 27 - 27 - 23- 23-19 - 19 - l 15 15 11 11 07 07 C: 0: l l I I I I I I 02 06 10 14 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 30 42 OATE: PERfCD: SEC. E MtS: DAIE: PERIOD: SEC._ XE MtS: B-5
i FIGURE B.3 Cvele 11 Cold Crttical Patterns . r CRmCAL PATTERN CRmCAL FATTE".N I 4: 4* 39 48 48 48 48 39 48 48 48 48 35 35 31 48 16 48 31 48 6 48 48 , 17 - n- l 23 - 48 48 48 48 48 48 23 - 48 48 48 48 48 48 19 - 19 - 15 46 48 15 48 48 48 11 11 I 48 48 48 07 48 48 48 48 07 48 0: 0: l l l l l l l l 02 08 10 14 18 22 26 30 34 IB 42 02 06 to 14 18 22 26 30 -34 38 42 DATE; 07/27/84 PERIOD: 160 SEC. E rRS: 51000 DATE: 08/06/84 PERIOD: 197 SEC. E HRS: >1000 CRmCAL PATTERN CRmCAL PATTERN 43 43 39 48 48 48 39 l l48 35 4 4 35 31 48 48 48 48 31 27 - 4 27 - 23 - 48 48 48 48 48 48 23-19 - 4 4 19 - 15 48 4? 48 48 15 11 4 4 11 07 ,48 48 48 48 07 . C 4 0; 1 I I I i 1 02 06 10 14 18 22 26 30 34 38 42 02 08 10 14 18 22 26 30 34 38 42 DATE: A/29/84 PERICD: *20 SEC. E 2: 252 DATE: PERICD: SEC. E FRS: B-6
FIGURE B.4 Cycle 12 Cold Critical Patiggg l CRmCAL PATIERN CRmCAL PATTERN l 43 40 l 39 48 48 48 48 39 48 48 48 35 . 35 31 48 22 48 31 48 48 48 48 48 27 - 27 - 23-- 48 48 48 48 48 48 23-- 48 48 48 48 48 19 - 19 - 15 *e 48 15 48 48 8 48 48 if 11 , 07 48 48 48 07 48 48 48 48 l 0; 0; l
- 1. I I I I I 02 06 10 14 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 DATE: 06/05/86 PERIOD: 185 SEC_ XE d6: >1000 DATE: 06/30/86 PERIOD: 87 SEC. XE tRS: >1000 CRIT. CAL FAT 7ERN CRIT CAL PATTERN 43 43 >
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19 - 19 - 15 '48 j 48 12 48 48 15 11 11 07 48 48 48 07 00 0; i l 1 l l l T 02 C6 to 14 16 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 PERICO: 118 SEC. XE 2 : >1000 DATE: PERICD: SEC. XE E : J2/86 DATE: 07 B-7
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I FIGURE B.5 I l 1 l l Je 13 Cold Critical Pattems s CRmCAL PATTERN CRmCAL PATTERN 43 43 l 39 48 48 48 48 39 48 48 48 48 j 35 35 31 -- 48 48 31 48 10 48 27 - 27 - 48 48 48 12 48 23 - 48 48 48 48 48 48 33-19 - 19 - 15 48 48 15 48 48 11 11 07 48 48 48 48 07 48 ;48 48 48 l 0: , 0: y l l- 1 I I I l C2 08 10 14 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 DATE. 08/27/87 PERIOD; 55 SEC. E %: >1000 DATE *0/02/87 PERIOD: 72 C .
. E rftS: >1000 CRmCAL PAT 7ERN CRmCAL PATXRN 43 43 l
48 39 48 48 48 39 l 48 l 48 48 l 35 35 31 48 48 48 48 31 48 8 48 48 27 27 - 23 - 48 48 14 48 23 - 48 48 48 48 48 48 19 - 19 - 15 j48 48 48 48 15 48 48 48 11 11 07 48 48 48 07 48 48 48 48 0' 0; - j l l I I I I I I 02 06 10 14 18 22 26 30 34 38 42 02 06 to 14 18 22 26 30 34 38 42 DATE: 07/02/88 PERICD: 70 SEC E %: ~82 DATE: 08/26/88 PERIOD: 142 SEC. E %: 61 B-8
FIGURE B.5 (Continued) Ovele 13 Cold Critical Pattems CRITICAL PATTERN CRmCAL PATTERN 03 43 39 48 l 48 48 48 39 35 35 31 48 48 48 31 27 - 27 - 23 - 48 48 48 48 48 48 23 - 19 -
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02 08 10 w to 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 DATE: 08/28/Be PERICO: 96 SEC XE rRS: 100 DATE: PERIOD: SEC. AE r1RS: CRmCAL PATRRN CRIT' CAL PATERN 43 l 43 l 39 lt 39 l l i 35 33 l I 31 J1 37 - 27 - 33- 23 - 19 - 19 - ~ 15 15 11 11 07 07 0: 0: I I I I I I 02 06 10 M 18 22 26 30 34 38 42 02 06 10 14 18 22 26 30 34 38 42 DATE: PERIOD: SEC XE W6: DAT: PERIOD: SEC. XE MtS: B-9
APPENDIX C k CHANNEL BOWING AT D-IATTICE PIJMI'S Early in the benchmarking process, the code vendor, Studsvik of America (SOA), advised us that the eigenvalues appeared too high. They had performed numerous benchmarks for other foreign BWR's which produced consistently lower eigenvalues. When reviewing this data with SOA'** it came to Itght that all the units in question were C-lattice plants. Subsequent investigation by SOA of a D-lattice reactor revealed high eigenvalues similar to Vermont Yankee's results.
'Ihis led to a search for the possible differences between D lattice anr' C-lattice plants.
The most fruitful discovery was that of pxtferential channel bowing in D bttice plants. This issue first came to our attention at the 1988 Station Nuclear Engineers Corderence'**. Subsequent investigation of EPRI channel measurement studiestas. 4 a7. sa revealet that the charmel bowing effect is unique to D lattices and is quite prevalent across the industry. Reference 35 provides a good overview of the mechanism causing the bow and the range of possible channel deflections. The primary mechanism is the fast flux gradient across the channel. In the interior of the core, the wide water gaps significantly soften the spectrum. Thus, the gradient results in a higher fast flux on the narrcw sides. The higher fast flux results in a higher rate of differential growth on the narrow sides of the channel, bowing the channel away from the wide side as shown in Figure C.I. There is also some contribution from differential oxide growth on the channel, but the impact is in the same direction: Channels deflect away from the wide wide corner (i.e., the control blade location) as the burnup on the channel (fluence) increases. l The amount of deflection, with burnup, is a function of the initial channel materials and the initial channel deflection. For typical channel material, Commonwealth Edisont *" calculated ) deflecuons away from the control blade as a function of initial deflection on the as-built channel. This is presented in Table C.I. From Table C.1 it is obvious that channel bowing exhibits positive feedback. That is, a wider than usual wide water gap results in greater fast flux gradients, further increasing the rate of bowing. l C-1
'As far as Vermont Yankee is concerned, it is difficult to determine what the average amount of channel bow in the reactor is. We do know that the majority channel vendor"" has . j pmoriented channda away from the control blade (wide-wide corner) since 1974. Bis is to prevent channel interferer.ce with blade scram times. Thus, even fresh channels start with some average bow grcater than sem.
Vermont Yankee has also been reusing channels during the period of the benchmark. We very process of selecting channels for reuse eliminates any channels which might be bowed 1 in the direction of the blade. Thus, it is relatively certain that average bowing on the order of 40 mils, or more exists at VY. < i A value of 40 mils was arbitrarily selected by the investigators. It represents half of a channel thickness and is considered to be quite reasonable, since it lies in the middle of the vendor's channel pre-selection criterion. While not such a big deflection, the impact it has on the D-lattice assembly is significant. The reason for this is that the lattice gets its rigidity from the channel being in contact with the spacers. As the channel deflects (see Figure C.2) so do all the pins in the lattice. As shown in Figure C.2, the enrichment distribution in a D-lattice is not symmetrical. It is tuned for the size of the water gaps. When the narrow gap is narrowed further, the medium enrichment pins on the narrow side suffer imm a reduction in thermal utilization. 'lhe increase in thermal utlhzation in the low enrichment pins along the wide side do not take up the slack. The net result is a loss of assembly reactivity. For C-lattice plants bowing is not a problem. The water gaps in a C-lattice are nearly the same. This reduces the fast flux gradient across the channel to practically zero, eliminating the feedback mechanism that causes bowing. Moreover, even with pmorientation of new l- channels, the reactivity impact is negligible because the C-lattice has a symmetrical enrichment distribution. L C-2
TABLE C.1 u Channel Bowinst vs. Burnuo as a Function of Initial Bowing Initial Deflection (mils) Deflection (mils) - . Change (mils) in Away from Control Rod Away from Control Rod Deflection After (Wide wide Corner) After 25.000 mwd /St 25.000 mwd /St 0.0 56 56 50 130 80 100 204 104 150 278 128 I h l l C.3
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l APPENDIX D HCYT MODEL-TO-PIMP DETECTOR COMPARISONS ne five cyclea modeled in this benchmark produced comparisons between the hot model and the plant samma detector readings at 223 EP set statepoints. With 20 instrument stdags ) per set and 24 data points per string, the number of comparison points exceeds 100,000. To reduce this to a manteable, yet presentable form, the data was normalized and summed radially and axially as explained in Section 5.3. De summations at each radial location are called the "11P integrals." he summations for each planar slice, when plotted axially, produce the core average axial'I1Ps." This appendix presents the plant to model comparisons for each cycle. For Cycle 9, Figure D.1 shows the average error and standard deviation in the TIP integrals at each location. Note: percent error is always defined as 100*(Model Plant)/ Plant. Figure D.1 also shows the RMS crmr in the 'I1P integrals at each location. Figure D.2 shows the core average axial TIPS for the model (lines) compared to the plant (+). De axially averaged TIPS are shown for each TIP set modeled in the Cych 9 depletion. The order of the TIP sets can be identified by either the TIP set number or the cycle exposure. As the cycle progresses, the model exhibits the f:llowing axial trend: The model begins to underpredict the power in the bottom of the core towards EOFPL. This same trend can be observed in all the cycles benchmarked. Figures D.3 and D.4 present similar information for Cycle 10. Figures D.5 and D.6 for Cycle 11. Figures D.7 and D.8 for Cycle 12. Figures D.9 and D.10 for Cycle 13. l l L D-1
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