CPSES-200400132, Cycle 8 Startup Report
| ML040280475 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 01/20/2004 |
| From: | Madden F TXU Electric |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CPSES-200400132, TXX-04015 | |
| Download: ML040280475 (24) | |
Text
NAh TXU
"'441.s TXU Energy Comanche Peak Steam Electric Station P.O. Box 1002 (EO1)
Glen Rose,TX 76043 Tel: 254 897 5209 Faxc 254 897 6652 mike.blevins@txu.com Mike Blevins Senior Vice President & Principal Nuclear Officer Ref: FSAR 4.6.6 CPSES-200400132 Log# TXX-04015 January 20,2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NO. 50-446 UNIT 2, CYCLE 8 STARTUP REPORT Gentlemen:
TXU Generation Company LP (TXU Energy) used a full reload of Westinghouse fuel in Unit 2 Cycle 08 with Integral Fuel Burnable Absorbers (IFBA). In accordance with the FSAR Section 4.6.6, attached is a summary report of the unit startup and power escalation testing following installation of fuel that has a different design or has been manufactured by a different fuel supplier.
This communication contains no new licensing basis commitments regarding CPSES Units 1 and 2.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak
- Diablo Canyon
- Palo Verde
- South Texas Project Wolf Creek
TXX-0401 5 Page 2 of 2 Please contact Mr. J. D. Seawright at (254) 897-0140 (email jseawrightetxu.com) should you have any questions.
Sincerely, TXU Generation Company LP By:TXU Generation Management Company LLC, Its General Partner Mike BlevI s
By:
?rr2TL-
/Frd W. Madden Nuclear Licensing Manager JDS/js Enclosure - Unit 2 Cycle 8 Startup Report c - B. S. Mallett, Region IV W. D. Johnson, Region IV M. C. Thadani, NRR Resident Inspectors, CPSES
TXU Generation Company LP COMANCHE PEAK STEAM ELECTRIC STATION ENGINEERING REPORT Unit 2 Cycle 08 STARTUP REPORT ERX-03-007 Revision 0 12/31/03 Prepared By:
Reviewed By:
Approved By:
re Pef nance Engineering
(\\6lL L hAA4Q Date:
/2-3/-03 Date: __ S_-___
Date: / 04 Norman L. Terrel Core Performance Supervisor Mickey4A. Killgore' Y
React d Engineering Manager
DISCLAIMER This information contained in this report was prepared for the specific requirements of TXU Generation Company LP and may not be appropriate for use in situations other than those for which it was specifically prepared. TXU Generation Company LP PROVIDES NO WARRANTY HEREUNDER, EXPRESS OR IMPLIED, OR STATUTORY, OF ANY KIND OR NATURE WHATSOEVER, REGARDING THIS REPORT OR ITS
- USE, INCLUDING BUT NOT LIMITED TO ANY WARRANTIES ON MERCHANTABILITY FOR FITNESS FOR A PARTICULAR PURPOSE.
By making this report available, TXU Generation Company LP does not authorize its use by others, and any such use is forbidden except with the prior written approval of TXU Generation Company LP. Any such written approval should itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event should TXU Generation Company LP have any liability for any incidental or consequential damages of any type in connection with the use, authorized or unauthorized, of this report or the information in it.
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SECTION 1.0 2.0 2.1 2.2 3.0 3.1 3.2 3.3 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.5 3.6 3.7 4.0 TABLE OF CONTENTS TITLE PAGE Title Page....................................
Disclaimer....................................
2 Table of Contents...................................
3 List of Tables and Figures...................................
4 Introduction...................................
5 Discussion of the Westinghouse Fuel Design................................... 6 Mechanical Design...................................
8 Nuclear Design...................................
9 Discussion of the Cycle 08 Startup Tests.................................... 12 Core Loading...................................
12 Control Rod Drop Time Measurements...................................
13 Initial Criticality....................................
14 Low Power Physics Testing...................................
15 Determination of the Range for Physics Testing.............................. 15 ARO Boron Endpoint Measurement...................................
16 Moderator Temperature Coefficient Measurements.......................... 16 Reference Bank Worth Measurement....................................
17 Bank Reactivity Worth Measurements (Rod Swap).......................... 18 Flux Mapping...................................
19 Incore/Excore Detector Calibration....................................
20 Core Reactivity Balance...................................
21 Summary....................................
22 3
LIST OF TABLES and FIGURES TITLE PAGE Table 1 Fuel Assembly Design Parameters.................................
7 Figure 1 Core Loading Pattern................................
10 Figure 2 Burnable Absorber and Source Rod Locations................................ 11 4
1.0 INTRODUCTION
This report presents a summary of the startup of Comanche Peak Steam Electric Station (CPSES), Unit 2, Cycle 08. Cycle 08 contains 101 reload fuel assemblies supplied by Framatome ANP (FRA-ANP) (formerly Siemens Power Corporation), 8 reload fuel assemblies supplied by Westinghouse, as well as 84 fresh assemblies of Westinghouse supplied fuel.
This report satisfies the requirements of CPSES FSAR section 4.6.6, which states that a summary report of unit startup and power escalation testing shall be submitted following installation of fuel of a different design or that has been manufactured by a different supplier.
CPSES, located in North Central Texas, is a two unit nuclear power plant. Unit 1 completed initial startup in 1990 and was declared to be in commercial operation on August 13, 1990. Unit 2 completed initial startup in 1993 and was declared to be in commercial operation on August 3, 1993. Unit 1 is currently in Cycle 10. Each unit utilizes a four loop Westinghouse (X) Pressurized Water Reactor as the Nuclear Steam Supply System. Both units are rated for a thermal reactor power level of 3458 MWth.
The plant is operated by TXU Generation Company LP.
Cycle 08 initial criticality occurred on October 28, 2003, and Low Power Physics Testing was completed the same day. The plant was synchronized to the grid on October 29, 2003. Power ascension testing continued, and the plateau at approximately 80% RTP was reached on November 2. The reactor achieved approximately 100% RTP on November 5, and power ascension testing was completed with the performance of a full power flux map on November 6.
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2.0 DISCUSSION OF THE WESTINGHOUSE FUEL DESIGN The CPSES Unit 2 Cycle 08 reactor core is comprised of 193 fuel assemblies arranged in a similar core configuration as found in other recent CPSES cycles. The cycle 08 core contains 101 partially spent FRA-ANP fuel assemblies (Regions 7A, 8, and 9), 8 partially spent Westinghouse fuel assemblies (Region 9W) and 84 fresh fuel assemblies supplied by Westinghouse (Region 10). Both Region 9W and Region 10 assemblies are of the Optimized Fuel Assembly (OFA) design, similar to the design used in early CPSES cycles. Although eight Region 9W assemblies were used in Unit 2 Cycle 07 as "Lead Use" assemblies, FRA-ANP was the primary fuel supplier for that cycle. Cycle 08 is the first full reload of Westinghouse fuel for Unit 2. A summary of the Cycle 08 fuel inventory is provided in Table 1.
The energy content of the Cycle 08 core has been designed to accommodate a refueling interval of approximately 18 months.
The CPSES Unit 2 Cycle 07 core configuration was comprised of 185 FRA-ANP (formerly Siemens Power Corporation) fuel assemblies (Regions 7A, 7B, 7C, 8, and 9),
as well as 8 "Lead Use" Westinghouse fuel assemblies (Region 9W). Both the FRA-ANP and W fuel designs have a nominal outside rod diameter of 0.360 inches, and utilize a 17 x 17 lattice configuration.
In the CPSES Unit 2 Cycle 07 core (other than the "Lead Use" fuel assemblies), solid burnable absorbers (B4C - A1203) encased in a Zircaloy-4 clad and manufactured by FRA-ANP were used to shape the power distribution and to achieve a desirable moderator temperature coefficient. The eight Westinghouse supplied fresh assemblies utilized Wet Annular Burnable Absorbers (WABA). Cycle 08 uses two types of W fabricated burnable absorbers: WABAs and Integral Fuel Burnable Absorbers (IFBA).
The WABAs consist of B4C - A1203 pellets encased between inner and outer Zircaloy-4 clad. IFBAs employ a thin ZrB2 coating on the fuel pellet surface in selected fuel rods.
WABAs were previously used in early CPSES cycles, and are currently being used in Unit I Cycle 10. IFBAs are currently being used in Unit 1 Cycle 10, which was the first cycle to employ IFBAs at CPSES.
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TABLE 1 Fuel Assembly Design Parameters CPSES Unit 2 Cycle 08 Region 7A 8
9 9W 10 Enrichment (w/o U235)
Central Zone 4.20 4.74 4.65 4.55 4.30 Axial Blanket Natural 2.0 2.0 2.0 2.6 Geometric Density
(% theoretical) 95.0 95.0 95.0 95.5 95.5 Number of Assemblies 1
16 84 8
84 Pellet Diameter (inches) 0.3035 0.3035 0.3035 0.3088 0.3088 All enrichments and densities are design values.
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2.1 MECHANICAL DESIGN The W 17 x 17 fuel assembly design, used for the Region 10 fuel assemblies, contains 264 fuel rods which are supported by eight grid spacers in the fuel assembly structure.
Mid-span grids are composed of ZIRLOTM while the top and bottom grids are composed of Inconel-718. The fuel assembly structure consists of an upper nozzle, a lower nozzle, twenty-four guide tubes, one instrument tube and eight spacer grids.
The major differences between the W fuel assembly (Region 10) design and the FRA-ANP fuel assembly (Region 9) design are:
68 of the W assemblies contain IFBA as burnable absorbers, which have not been previously used at CPSES in Unit 2.
The W fuel assemblies contain annular axial blankets to accommodate the gas volume produced in the IFBA containing fuel rods. The 2.6 w/o enriched annular axial blankets are nearly identical in reactivity characteristics to the 2.0 Nv/o enriched solid axial blankets used in the FRA-ANP fuel.
The W cladding, Guide Tube, Instrumentation Thimble, and mid-span grid assembly material is ZIRLOTN, while the FRA-ANP fuel uses bimetallic (Zircaloy-4/Inconel-718) grid assemblies, with Zircaloy-4 Instrumentation Thimbles, Guide Tubes and cladding.
The W fuel has a clad thickness of 0.0225 inches, while the FRA-ANP clad has a thickness of 0.025 inches.
The W fuel has a nominal density of 95.5 (percent of theoretical), while the FRA-ANP fuel has a nominal density of 95.0.
The W fuel pellets measure 0.370 inches in length with a 0.3088 inch diameter. FRA-ANP fuel pellets measure 0.350 inches in length with a 0.3035 inch diameter.
The FRA-ANP fuel assemblies are equipped with the FUELGUARDTm enhanced debris filtering bottom nozzles for improved debris filtering performance. The Region 10 W assemblies are equipped with the W "Small Hole" debris filtering bottom nozzle, an alternate protective grid (P-grid), and long solid end plugs. The Region 9W W assemblies do not contain the alternate P-grid, but contain the "Small Hole" debris filtering bottom nozzle.
The top nozzle design of the W fuel is incompatible with standard thimble plugs, and must use dually compatible thimble plugs. FRA-ANP fuel can use either the standard or the dually compatible thimble plugs.
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In other respects, the FRA-ANP and W fuel designs are similar. Both are provided with unique serial numbers engraved on the top nozzle. Both use removable top nozzles. All locator holes in the top and bottom nozzles are compatible with the upper and lower core support plates.
Along with the fuel assemblies, W provided 1152 WABA rodlets distributed among 56 clusters. These WABAs are similar to those used in W fuel in previous CPSES cycles.
The physical (including geometrical) properties of the W OFA fuel are compatible with the FRA-ANP fuel assembly designs and with the CPSES reactor vessel internals, spent fuel racks, and fuel handling equipment. CPSES has previously operated with mixed cores of FRA-ANP / W OFA fuel designs, and successfully demonstrated compatibility with existing rod control clusters and fuel handling equipment.
The mechanical design criteria to which the W fuel rods, fuel assemblies, and burnable absorber and thimble plug clusters have been designed are consistent with the design criteria used for the FRA-ANP fuel assemblies. Compliance with these mechanical design criteria has been demonstrated through mechanical analyses of the W fuel rod and fuel assembly designs, using W methodologies which have been approved by the NRC.
These evaluations are valid for peak fuel rod exposures of 60,000 MWD/MTU (for W fuel with ZIRLOTm clad). This exposure bounds the expected EOC burmup for the W assemblies. The power histories used in the mechanical design are consistent with those histories expected for Cycle 08 operation. An appropriate number of transients (load changes, trips, etc.) have been considered in the fatigue evaluations.
2.2 NUCLEAR DESIGN The nuclear design of the CPSES Unit 2 Cycle 08 core was performed by TXU in accordance with methodologies approved by the NRC.
The differences between the W OFA fuel assembly design and the FRA-ANP fuel assembly designs, including the IFBAs, are appropriately modeled in the core design and safety analysis codes. Prior to use at Comanche Peak, benchmarking was performed by using CPSES core design methodologies to analyze data from other nuclear plants which have used IFBAs. The results from this benchmarking have demonstrated that CPSES core design methodologies properly model the operating characteristics of fuel assemblies which utilize IFBAs.
The Cycle 08 core configuration is designed to meet an FQ x P / K(z) limit of < 2.42 for an axial flux difference (Al) within Technical Specification limits, where P is the reactor power normalized to rated thermal power.
The Cycle 08 core configuration is presented in Figures I and 2. The core contains a total of 152 WABA rodlets and 4864 IFBA located in the Region 10 fuel assemblies.
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FIGURE 1 CORE LOADING PATTERN CPSES Unit 2 Cycle 08 R
P N
M L
K J
H G
F E
D C
E A
1-1 HH61 KK45 KK85 JW06 KK29 KK6O HH691 QuadIV A- 09 10 1 0 M-03 10 10 R O9 Qua_
N41 HH67 JJ53 KK20 1144 KKO7 JJ79 KK84 JJ30 KK32 JJ64 I HHO I N4 Looo 4 I C-1 4 G-06 1 0 K
K-131 10 iK-1 1 0 F-1 3 1 0 J-06 M-14
- Loo, 1
adl 43 2
)1 HH40 KK62 KK75 J6II KK50 JJ29 KK23 J191 KK77 JJ32 KK71 KK27 I B-12 10 10 L-02 10 G-02 10 J-02 10 E-02 10 10 HH52 3
P13 3 JJ57 4
F-O9 w
KK58 JJ7O JJ31 K-09 KK34 10 JJB6 H-11 KK58 10 JJ70 K-01 KK64 JW07 KK53 JJ78 KK87 1JJ1 KK13 10 I M-13 10 F-01 10 L -08 10 S0-HH88 KK41 JJ61 KK1 6 JJ25 JJ90 JJ75 KK35 JJ13 JJ73 JJ92 KK63 JJ09 KK65 HH59 G-15 10 P-05 10 C-08 M-07 M-11 10 D-11 D-07 H-13 10 B-05 10 a
-15 KK54 JJ19 KK06 JJ54 JJ85 JJ52 KK26 JJ47 KK67 JJ68 JJ84 JJ39 KK51 JJ82 KK22 10 C-06 10 R-06 J-04 F-1 10 N-13 10 L-1 0 0-04 A-06 10 N-06 1 0 KK1 5 KK28 JJ28 KK46 JJ87 KK31 JJ38 JJ12 JJ26 KK57 JJ46 KK10 JJ1 8 KKO9 KK39 10 10 P-O9 10 E-04 10 H-09 H-15 J-08 10 L-04 10 B-09 10 10 JW06 JJ71 KK62 JW02 KK70 JJ65 JJ72 GG0 JJ15 JJ40 KK43 JW05 KK1 9 JJ42 JW04 N-1 2 E-10 10 C-12 10 C-13 A-0 N
P.-0 N-03 10 N-04 10 L-06 C-04 KKl1 KK17 JJ41 KK42 JJ63 KK79 JJ67 JJ49 JJ36 KK52 JJ20 KK60 JJ33 KK33 KK72 10 10 P-07 10 E-12 10 G-08 H-01 H-07 10 L-12 10 B-07 10 10 KKO5 JJ48 KK47 JJ89 JJ66 JJ43 KK66 JJ51 KK21 JJ45 JJ21 JJ69 KK25 JJ62 KK69 10 C- 0 10 R-10 J-12 E-06 10 C-03 10 K-05 G-12 A-10 10 N-10 10 HH32 KK40 JJ1 7 KK81 JJ76 JJ55 JJ24 IKK86 JJ60 JJ27 JJ34 KK73 JJ22 KK1B HH85 0-01 10 P-1 1 0 H-03 M-09 M-05 1 0 D-05 D-09 N-08 10 B-11 1 0 J-01 5
6 7
9 10 11 JJ50 KK24 JJ137 KK08 JJ83 KK1l JW03 KK35 JJ56 KK37 JJ23 KK48 JJ77 K-07 10 E-08 10 K-1 5 10 D -03 10 F-1i 10 H-05 10 F-07 12 13 HH31 KK55 KK68 J1158 KK44 JJ35 KK12 J111 KK30 JJ10 KK041 KK02 HH65 B-03 10 10 L-14 10 G-14 10 J-14 10 E-14 10 10 P-04 r Quad III I HH46 J 1N441
[D-02 I I Loop 3 C D
E
- 'Unit 2 Cycle 6 Location JJ14 KK59 JJ9I KK49 J1B0 KK78 JJ88 KK01 3-10 10 K-03 10 F-05 10 F-03 10 JJ74 HH6 Q
Quad 11 1 4 J-10 J N-02]
N42 ASSEMBLY ID 1 5 REGION # OR U2C7 LOCATION HH54 KK83 I KK761 WMl KKBB KK56 HH39 A-07 10 10 D -13 10 10 R-07 0O
[ia]
REGION 7A
[i] REGION 9W l
(FRA-ANP. 4.20 w/o, Central Zone) l l Westinghouse. 4.55 w/o. Central Zone)
Rj REGION 8 K
REGION 10 (FRA-ANP. 4.74 w/o. Central Zone) [l lW'estinghouse. 4.30 w/o. Central Zone) jm j REGION 9 m
I (FRA-ANP. 4.65 w/o. Central Zone) 10
FIGURE 2 BURNABLE ABSORBER AND SOURCE ROD LOCATIONS CPSES Unit 2 Cycle 08 R
P N
M L
K J
H G
F E
D C
B A
ua
~
~
~~~IS I II Ia a
l l
~65
_________1 Lo 4 641 641 641 641 No 1 641 641 801 641 8a0 641 641 3
8W 24W 24W 24W 8W 641 80!
801 E01 801 641 4
8W 24W 24W 24W 24W 8W 90-641 801 641 801 641 24W 24W 224W 801 80!
801 8
24W 24W 24W 24W 641 801 801 801 801 641 16W 24W 24W 24W 24W 16W 641 641 641 641 24W 24W 24W 24W
.4 641 801 801 801 80!
641 16W 24W 24W
- 24W, 24W 16W 801 8!
801 801 24W 24W 24W 24W 641 801 641 801 641 24W 24W 24W 5
6 7
B 9
10 11 641 63W 8a1 24W 801 24W Sol 24W 801 24W 641 8W 641 641 801 641 1
801 1
641 641 8W 24W 24W I24W SW_
12 13 Quadll 14 N42 Loop 2 15 1N441 641 641 116W 641 16W 641 1
-c 6S O-ES SECONDARY
[
- of IF8A rods (64, or 601 SOURCES (2)
[L J# of WABA rods (8.16. or 24) 11
3.0 DISCUSSION OF THE CYCLE 08 STARTUP TESTS The objectives, methods, and results of each startup test is described in the following sections. The purpose of the overall test program is to ensure the new cycle reactor core behaves in a manner consistent with the design and safety analyses.
3.1 CORE LOADING OBJECTIVES Control the loading sequences to ensure that the nuclear fuel assemblies are loaded in a safe and cautious manner, and that the final core configuration is in agreement with the specified design.
TEST METHODOLOGY Refueling was performed by completely offloading the Cycle 07 core to the Spent Fuel Pool, changing out fuel inserts, and then loading the Cycle 08 core. Cycle 07 had no indications of leaking fuel assemblies.
The first assembly (one of two source assemblies) to be reloaded was latched on October 17, 2003 and the last assembly to be loaded was unlatched on October 20. Inverse Count Rate Ratio (ICRR) was monitored during fuel loading.
The Cycle 08 core configuration is presented in Figure 1.
SUMMARY
OF RESULTS Prior to reload, fuel assembly insert number/type were verified in the spent fuel pool by Core Performance Engineering and Quality Control. There were no discrepancies identified. Fuel assembly identifications were again verified via underwater camera for each assembly as it was loaded into the core.
Core loading was completed on October 20, 2003. All 193 assemblies were loaded into the core without incident.
Following reload, the core loading pattern verification process was completed for the Cycle 08 loading pattern by Core Performance Engineering and Quality Control.
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3.2 CONTROL ROD DROP TIME MEASUREMENTS OBJECTIVE To determine the drop time of each Rod Control Cluster Assembly (RCCA) under hot, full flow conditions in accordance with Technical Specification SR 3.1.4.3.
TEST METHODOLOGY The Plant Process Computer (PPC) method was used to determine the rod drop times for Unit 2 Cycle 08. This involves withdrawing each rod bank and opening the reactor trip breakers. The difference between the time the reactor trip breakers open and the time a RCCA has entered the dashpot (according to PPC DRPI indications) is used to determine the rod drop time. This process is repeated for the remaining banks.
SUMMARY
OF RESULTS Technical Specification SR 3.1.4.3 requires the drop time for each RCCA from the fully withdrawn position to be less than or equal to 2.4 seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry with Tavg greater than or equal to 500'F and all reactor coolant pumps running. Under these conditions, the longest drop time measurement was 2.13 seconds for RCCAs at core locations B12 and M14.
All rod banks satisfied review and acceptance criteria.
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3.3 INITIAL CRITICALITY OBJECTIVE To achieve initial criticality following refueling in a deliberate and controlled manner.
TEST METHODOLOGY From an initial condition of all rods in and a boron concentration of 1866 ppm, the Shutdown and Control Banks were withdrawn to the full out position (FOP) in proper overlap and sequence. Inverse Count Rate Ratio (ICRR) was monitored during Control Bank withdrawal.
Reactor Coolant System (RCS) dilution was initiated. During dilution, ICRR was monitored. Criticality was declared on October 28, 2003, and dilution was terminated.
Control Bank D (CBD) motion was used to stabilize flux level.
SUMMARY
OF RESULTS Cycle 08 initial criticality was achieved in a controlled manner on October 28, 2003 at 0959 hours0.0111 days <br />0.266 hours <br />0.00159 weeks <br />3.648995e-4 months <br />.
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3.4 LOW POWER PHYSICS TESTING Low Power Physics Testing (LPPT) verifies the design of the reactor by performing a series of selected measurements including control/shutdown bank worths, moderator temperature coefficient and boron worth. These measurements are performed by using the Digital Reactivity Computer (DRC) resident on the Plant Process Computer (PPC) to indicate reactivity changes below the point of adding heat.
The individual tests completed during the initial criticality and the low power test sequences are discussed in the following sections of this report. All required tests were satisfactorily completed.
3.4.1 DETERMINATION OF THE RANGE FOR PHYSICS TESING OBJECTIVE To determine the neutron flux level at which detectable reactivity feedback from fuel heating occurs and to establish the flux range for low power physics testing.
TEST METHODOLOGY With the reactor critical at a power level of approximately 1.0 E-8 amps (as indicated by the primary IR channel), approximately +40 pcm of positive reactivity was added by withdrawal of Control Bank D. Flux was allowed to increase until fuel temperature feedback effects were observed by a decrease in the indicated core reactivity, as indicated on strip chart recorders.
The physics testing range upper limit was set at 30% of the flux level at which the point of adding heat was observed. The LPPT lower limit is 3% of this point, giving a one decade range in which to perform LPPT.
SUMMARY
OF RESULTS Fuel temperature reactivity feedback was observed at flux levels similar to past CPSES cycles. The LPPT range was set appropriately. There are no review or acceptance criteria for this test.
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3.4.2 ARO BORON ENDPOINT MEASUREMENT OBJECTIVES To measure the critical boron concentration at the All Rods Out configuration.
TEST METHODOLOGY Conditions were established with Control Bank D within 30-50 pcm of its full out position configuration with the reactor critical in the low power physics testing range.
The control bank was withdrawn to the full out position while monitoring reactivity. The changes in reactivity due to bank movement and Tavg deviation from Tref were converted to equivalent boron concentration units and used to correct the initial boron concentration, yielding the endpoint boron concentration.
SUMMARY
OF RESULTS The ARO boron endpoint measurement satisfied the review criteria of+/- 50 ppm (RCS boron) of predicted, as well as the acceptance criteria of+ 1000 pcm (reactivity) of predicted. The difference between the measured value and design value was similar to past CPSES cycles.
3.4.3 MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS OBJECTIVE To measure the Isothermal Temperature Coefficient (ITC) and calculate the Moderator Temperature Coefficient (MTC).
TEST METHODOLOGY The ITC measurement was performed by first decreasing, then increasing Tavg using Steam Generator blowdown flow and increasing Auxiliary Feedwater Flow to compensate. The resulting reactivity changes were measured and used to calculate the ITC. The ITC is the change in reactivity divided by the associated change in temperature.
The MTC was determined by subtracting the design Doppler Temperature Coefficient from the ITC.
SUMMARY
OF RESULTS The measurement of ITC met the review criteria of being within + 2 pcm/IF of the design value. The difference between the measured value and design value was similar to past CPSES cycles. MTC met the acceptance criteria of < +5.0 pcm/IF.
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3.4.4 REFERENCE BANK WORTH MEASUREMENT OBJECTIVE To measure the Integral Rod Worth (IRW) of the Reference Bank using the standard boron dilution technique.
TEST METHODOLOGY The Reference Bank is the RCCA bank with the highest predicted IRW. For Unit 2 Cycle 08, the Reference Bank was Control Bank C.
Conditions were established with Control Bank D within 30-50 pcm of its full out position configuration with the reactor critical in the low power physics testing range.
CBD is withdrawn in MANUAL to the FOP. Following a short wait for a reactivity measurement, the Reference Bank is selected in individual bank select and inserted to establish reactivity indication on the DRC near zero.
A RCS dilution is then initiated. The Reference Bank is inserted in incremental reactivity steps sufficient to maintain flux and reactivity in the LPPT range as the dilution continues. Reactivity measurements are registered for each incremental insertion. A Target Rod Position is selected for the Reference Bank that corresponds to approximately 60 pcm. of remaining worth which indicates when to secure the dilution. After the dilution is terminated and RCS mixing is complete, the Reference Bank will have a small amount of remaining worth at the critical position. The Reference Bank is then fully inserted for the final reactivity measurement and withdrawn back to the critical position.
The incremental reactivity steps are summed to obtain the total worth for the Reference Bank.
SUMMARY
OF RESULTS The Review Criteria states that the absolute value of the percent difference between measured and predicted IRW for the Reference Bank is < 10%. This criteria was satisfied.
The Acceptance Criteria states that the absolute value of the percent difference between measured and predicted IRW for the Reference Bank is < 15%. This criteria was also satisfied.
The differences between the measured values and design values were similar to past CPSES cycles.
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3.4.5 BANK REACTIVITY WORTH MEASUREMENTS (ROD SWAP)
OBJECTIVE To infer the integral reactivity worth of each Control and Shutdown Bank based on the known IRW of the Reference Bank measurement.
TEST METHODOLOGY Integral bank worths were measured using the rod swap method. The subject bank was inserted then compensated for by pulling the reference bank in response to the change in reactivity caused by the insertion of the measured bank. Each bank's worth was determined by comparison to the Reference Bank's measured worth.
SUMMARY
OF RESULTS The following review and acceptance criteria were satisfied.
Review Criteria:
Individual Banks within 15% or within 100 pcm of design worths, whichever is greater.
Total Worth is
- 110% of design.
Acceptance Criteria:
Sum of measured bank worths shall be no less than 90% of the design sum of bank worths.
The differences between the measured values and design values were similar to past CPSES cycles.
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3.5 FLUX MAPPING OBJECTIVE To verify adequate flux symmetry and power distribution during initial startup following refueling.
TEST METHODOLOGY Flux maps were taken at the 28%, 80%, and 100% RTP plateaus to monitor flux symmetry and power distribution.
SUMMARY
OF TEST RESULTS A flux map was taken at the 28% plateau. The maximum allowable power level extrapolated above 80% (the next target plateau) based on peaking factors. A check of the core loading pattern was performed by comparing the Relative Power Densities (RPD) from the flux map to design predicted values. All RPD values satisfied review criteria limits.
At approximately 80% RTP, a base case flux map and six quarter-core flux maps were taken for the Confirmation of the Calibration Standard. Peaking factor extrapolation resulted in a most limiting allowable power level in excess of 100% RTP.
Xenon equilibrium was established at approximately 100% power and a full core flux map was performed on November 6. Power distribution factors and flux symmetry satisfied all requirements. Target AFD was established based on the measured axial offset.
The differences between the measured values and design values were similar to past CPSES cycles. All flux maps taken during power ascension displayed adequate flux symmetry and power distributions, and all acceptance criteria were met.
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3.6 INCORE/EXCORE DETECTOR CALIBRATION OBJECTIVES The objective of this surveillance is to check the validity of the current incore/excore detector calibration equations. The incore axial flux difference (AFD) is measured with a full core flux map and compared to the AFD indicated by the control board indicators, the plant process computer, and the NIS power range excore detector currents. This procedure satisfies Technical Specifications Surveillance Requirements 3.3.1.3.6 and 3.3.1.6.6 for Overtemperature N-16 function.
TEST METHODOLOGY AND RESULTS Pre-critical adjustment ratios from the Unit 2 Cycle 08 Startup and Operations Report were used to adjust the latest calibration currents from the previous cycle.
A full core flux map was taken at 28% power. The results of the AFD Monitor Check calculations exceeded the acceptance criteria, which required that excore detector calibrations be performed prior to the continuation of power ascension.
At the next calibration plateau, power was held near 80% for a sufficient amount of time to reach xenon stability. A full core flux map was performed on November 3, 2003. The results of the AFD Monitor Check calculations exceeded the acceptance criteria, which required excore calibrations to be performed prior to starting the Multipoint Measurement.
Six Quarter Core flux maps were performed, starting on November 3, to be used in the Confirmation of the Calibration Standard. The flux maps were measured over a total change of 13.4% in incore axial offset. The measurements confirmed that the Calibration Standard could be used in place of multipoint measurements for the calibration of the power range NIS throughout Unit 2 Cycle 08 operation.
Neutron Streaming Gains were determined and transmitted to I&C for calibration of the N16 system.
A full core flux map was performed on November 6, 2003 with the reactor at approximately 100% RTP. The results of the AFD Monitor Check calculations exceeded the review criteria, which required that Intercept Current and Delta Q alignments for each excore NIS channel be performed.
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3.7 CORE REACTIVITY BALANCE OBJECTIVE To compare the overall core reactivity balance with predicted values at hot full power (HFP), all rods out (ARO), equilibrium Xenon/Samarium boron concentration.
TEST METHODOLOGY Under equilibrium conditions at approximately 100% RTP, the Reactor Coolant System measured boron concentration was corrected to yield the Hot Full Power, All Rods Out, Equilibrium Xenon/Samarium boron concentration for comparison with the predicted boron concentration.
SUMMARY
OF RESULTS The equivalent reactivity difference between measured and predicted boron concentration was within the acceptance criteria of 1000 pcm, as required by Technical Specification SR 3.1.2.1. The difference between the measured value and design value was similar to past CPSES cycles.
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SUMMARY
This report is submitted as required following installation of fuel of a different design.
Cycle 08 contains 84 fresh assemblies supplied by _, 68 of which contain Integral Fuel Burnable Absorbers. Comanche Peak has not previously loaded fuel in Unit 2 containing this type of burnable absorber.
Comanche Peak has previously used fuel of the Westinghouse OFA design. From 1993 to 2001, however, Siemens Power Corporation (now FRA-ANP) was the primary fuel supplier. The current Unit 1 cycle (Cycle 10) is the first cycle in recent years in which the full reload has been supplied by Westinghouse. The design of the Unit 2 Cycle 08 Westinghouse fuel is similar to the fuel loaded into Unit I Cycle 10, including ZIRLOTh materials and use of IFBAs.
Unit 2 Cycle 08 reload, startup, and physics tests were completed without incident. All required testing was performed, and all acceptance criteria were satisfied. The differences between the measured values and design values were similar to past CPSES cycles. Based on the results, the Westinghouse OFA assemblies and Integral Fuel Burnable Absorbers were properly modeled in the design of the core, and there was no need to perform further testing.
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