CPSES-200300290, Cycle 10 Startup Report
| ML030420421 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/05/2003 |
| From: | Walker R TXU Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CPSES-200300290, TXX-03034 | |
| Download: ML030420421 (24) | |
Text
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TXU TXU Energy Comanche Peak Steam Electric Station RO Box 1002 (EO1)
Glen Rose,TX 76043 Tel 254 897 8920 Fax 254 897 6652 lance terry@txu corn C. Lance Terry Senior Vice President &
Principal Nuclear Officer CPSES-200300290 Log # TXX-03034 February 5, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NO. 50-445 UNIT 1 CYCLE 10 STARTUP REPORT Gentlemen:
TXU Generation Company LP (TXU Energy) loaded 88 Westinghouse fuel assemblies into the Unit 1 Cycle 10 reactor core. Integral Fuel Burnable Absorbers were used in 72 of the 88 Westinghouse fuel assemblies. These assemblies were loaded as the beginning of CPSES Unit 1 transition to Westinghouse fuel from Framatome.
In accordance with the FSAR Section 4.6.6, enclosed 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.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway Comanche Peak Diablo Canyon Palo Verde South Texas Project Wolf Creek
ft TXU TXX-03034 Page 2 of 2 This communication contains no new licensing basis commitments regarding CPSES Unit 1.
Sincerely, TXU Generation Company LP By:
TXU Generation Management Company LLC, Its General Partner C. L. Terry Senior Vice President and Principal Nuclear Officer By:
to le
- t.
V" Rogc*b. Walker Regulatory Affairs Manager JDS/js Enclosure c -
E. W. Merschoff, Region IV W. D. Johnson, Region IV D. H. Jaffe, NRR Resident Inspectors, CPSES
TXU Generation Company LP COMANCHE PEAK STEAM ELECTRIC STATION ENGINEERING REPORT Unit 1 Cycle 10 STARTUP REPORT ERX-03-001 Revision 0 1/8/03 Prepared By:
Reviewed By:
Approved By:
Core Te-rformance Engineering
~ Li VAAL Norman L. Terrel Core Performance Supervisor "MictXy R. Killgore f
Reactor Engineering Manager Date:______________
Date:________
Date: /-
(4-*,5
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 3
TABLE OF CONTENTS TITLE PAGE Title Page.......................................................................... I 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 10 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
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 1, Cycle 10. Cycle 10 contains 105 reload fuel assemblies supplied by Framatome ANP (FRA-ANP) (formerly Siemens Power Corporation), as well as 88 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 2 is currently in Cycle 7. Each unit utilizes a four loop Westinghouse 0V1 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 10 initial criticality occurred on November 10, 2002, and Low Power Physics Testing was completed November 11, followed by a reactor shutdown due to equipment problems. The plant was synchronized to the grid on November 18, 2002. Power ascension testing continued, and 80% RTP was reached on November 21, but the reactor was shutdown again prior to reaching 100% RTP due to equipment problems. The reactor achieved 100% RTP on November 29, but did not reach stable conditions before a dropped control rod forced another shutdown for repairs. Full power was again reached on December 15, and power ascension testing was completed with the performance of a full power flux map on December 18.
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2.0 DISCUSSION OF THE WESTINGHOUSE FUEL DESIGN The CPSES Unit 1 Cycle 10 reactor core is comprised of 193 fuel assemblies arranged in a similar core configuration as found in recent CPSES cycles. The cycle 10 core contains 105 partially spent FRA-ANP fuel assemblies (Regions 2-7A, 10A, lOB, and 11), and 88 fresh Region 12 fuel assemblies supplied by Westinghouse. The Region 12 assemblies are of the Optimized Fuel Assembly (OFA) design, similar to the design used in early CPSES cycles. Unit 2 Cycle 7 is currently using 8 lead use assemblies of a similar design. A summary of the Cycle 10 fuel inventory is provided in Table 1.
The energy content of the Cycle 10 core has been designed to accommodate a refueling interval of approximately 18 months.
The CPSES Unit 1 Cycle 9 core configuration was comprised of 192 FRA-ANP (formerly Siemens Power Corporation) fuel assemblies (Regions 9A, 9B, 1 A, 1 B, and 11), as well as 1 partially spent Westinghouse fuel assembly (Unit 2 Region 2). The Cycle 10 configuration includes 105 FRA-ANP fuel assemblies and 88 W OFA fuel assemblies. 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 1 Cycle 9 core, solid burnable absorbers (B4C - A120 3) 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. Cycle 10 uses two types of W fabricated burnable absorbers: Wet Annular Burnable Absorbers (WABA) and Integral Fuel Burnable Absorbers (IFBA). The WABAs consist of B4C A120 3 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 eight assemblies in Unit 2 Cycle 7. Unit 1 Cycle 10 is the first cycle to employ IFBAs at CPSES.
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TABLE 1 Fuel Assembly Design Parameters CPSES Unit 1 Cycle 10 Region 2-7A 1OA 1OB 11 12 Enrichment (w/o U235)
Central Zone 4.20 4.47 4.60 4.82 4.34 Axial Blanket Natural 2.0 2.0 2.0 2.6 Geometric Density
(% theoretical) 95.0 95.0 95.0 95.0 95.5 Number of Assemblies 1
8 4
92 88 Pellet Diameter (inches) 0.3035 0.3035 0.3035 0.3035 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 12 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 oflnconel-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 12) design and the FRA-ANP fuel assembly (Region 11) design are:
72 of the W assemblies contain IFBA as burnable absorbers, which have not been previously used at CPSES.
- 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 w/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 ZIRLOTM, while the FRA-ANP fuel uses bimetallic (Zircaloy-4/Inconel-718) grid assemblies, with Zircaloy-4 Instrumentation Thimbles and Guide Tubes.
0 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.
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.
0 The FRA-ANP fuel assemblies are equipped with the FUELGUARDTM enhanced debris filtering bottom nozzles for improved debris filtering performance. The W assemblies are equipped with the W "Small Hole" debris filtering bottom nozzle, an alternate protective grid (P-grid), and long solid end plugs.
e 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 1056 WABA rodlets distributed among 60 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 burnup for the W assemblies. The power histories used in the mechanical design are consistent with those histories expected for Cycle 10 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 1 Cycle 10 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. 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 10 core configuration is designed to meet an FQ x P / K(z) limit of< 2.42 for an axial flux difference (AI) within Technical Specification limits, where P is the reactor power normalized to rated thermal power.
The Cycle 10 core configuration is presented in Figures 1 and 2. The core contains a total of 1056 WABA rodlets and 4704 IFBA located in the Region 12 fuel assemblies.
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FIGURE 1 CORE LOADING PATTERN CPSES Unit 1 Cycle 10 R
P N
M L
K J
H G
F E
D I I I I I I I LSG IM31 I M64[ K76 M29 M781 L77 1I d -IV K-07 12 12 N-03 12 12 F-07
/
1.49 1.03 M20 12 LOB K-O1 L90 M-07 C
I M84 L781 M301 L35 K13 12 F-01 1 12 1 G-15 M-14 M26 1.80 K5A M82 M751 L46 M79 L12 M231 L85 M88 L32 IM71 M27 K09
_-12112 12 L-02 12 G-02 12 J-02 1 12 E-02 12 12 P-12 M58 12 MS0 L83 M49 L79 M21 L56 MOl 12 K-15 12 K-11 12 F-iS 12 L57 F-09 L31 K-09 M07 12 M76 12 K66 C-13 M35 12 M52 12 B
A I
I 4-C M74 L38 IM531 L73 M56 L26 M13 L15 12 H-11 12 1 D-07 12 J-08 12 R-07
.72 M24 153 M45 L55 Ml1 L23 M141 L04 M37 L36 M48 L89 AD9 12 G-OO 12 M-09 12 H-O5 12 1D-09 12 H-07 12 R-09 K28 M51 M68 L70 M85 L21 M12 L11 M57 L63 MDA M02 K39 8-04, 12 12 L-1-4 12 G-14 12 J-1A4 12 E-1A J12 12 P-04 Quadlll j K24 L9 N44 D-o2 J-O1 Loop 4
. Unit 2 Cycle 6 Location IQ-uad I N43 2
Loop 2 3
4 5
6 7
8 9
10 11 12 13 L92 K02 Quad l1 14 G-01.M-02 oN42 coop 1 ASSEMBLY ID 15 REGION # OR Ul C9 LOCATION 0
- E]*
UNIT 2PREGION 7A
[
]
REGION 11 L9 (FRA-ANP, 4 20 w/o, Central Zone)
(FRA-ANP, A 02 w/o. Central Zone)
W REGION 1DA
[
REGION 12 (FRA-ANP, 4 47 w/o. Central Zone)
- Cvestinghouse. 4 34 w/o, Central Zone)
W REGION 108 (FRA-ANP. 4 60 w/o. Central Zone) 10 K1 7 1.39 M3A 90O L64 LOS MiS L52 M44 1.44 M36 L30 MO0 1L20 M63 L65 M65 L84 12 P-OS 12 F-11 12 K-13 12 F-13 12 L-1-0 12 B-05 12 G-06 ME1 LEA MO6 LO7 M70 L27 L16 L86 L13 L58 M32 L29 M51 L22 M22 12 P-OS 12 J-0 412 D-11 M-13 H-13 D-13 L-12 12 G-04 12 A-06 12 MiS M28 L28 M1S Li9 L51 1.71 M03 L75 LA40 L82 M10 L33 MO9 M39 12 12 P-OS 12 C-06 C-04 E-12 12 M-11 N-04 N-06 12 B-09 12 12 K69 101 M62 137 MOB L25 M66 GG02 M67 L34 M87 L68 M1i9 L01 K61 N-13 1.-OS 12 E-0B 12 C-08 12 N-03 12 N-08 12 L-08 12 E-10 C-03 MS1 M17 141 M42 L48 L02 L60 M38 L42 L62 L06 M60 L09 M33 M72 12 112 P-07 12
-10 C-12 D-05 12 L-04 N-12 N-10 12 B-07 12 12 MOE 61 M47 1.66 M59 L87 L47 L76 L10 L24 M43 1.4 M25 L69 M69 12 P-10 12 J-12 12 E-04 M-03 H-03 D-03 M-05 12 G-12 12 A-10 12 1.74 M40 117 M81
.L43 M83 L59 M86 L88 M77 L45 M73 L18 M18 L67 J-10 12 P-1 12 1E-06 12 K-03 12 F-03 12 K-05 12 B-11 12 G-1O I
I M26i 12 LBO
FIGURE 2 BURNABLE ABSORBER AND SOURCE ROD LOCATIONS CPSES Unit 1 Cycle 10 R
P N
M L
K J
H G
F E
D C
B A
I I I I I I 6S IJ Quad IVQua N41 641 1041 1041 641 N4 Loop 3 8W W
I Loop 641 641 481 641 481 641 641 8W 24W 20W _
24W 8W 641 481 481 481 481 641 24W 24W 24W 24W 8W 641 481 641 641 641 481 641 24W 20W 20W 20W 24W 4 1 81641 64!
481 24W 20W 20W 24W 1041 48!
1041 481 1041 8W 24W 8W 24W 8W 0641 641 1041 1041 641 641 20W 20W 8W 8W 20W 20W 1041 48!
1041 481 1041 8W 24W 8W 24W 8W 48!
64!
641 481 24W 20W 20W 24W 641 48!
641 641 641 481 641 24W 20W 20W 20W 24W 641 8W 64!
64!
481 1 641 1
481 6411 641 J
124W 20W 24W 8W 481 24W 641 641 6S 1041 8W 1041 8W 1
2 2
3 4
5 6
7 8
9 10 11 12 13 N 42QuadI 14 N42 Loop 15 0*
64!
481 24W 481 24W 481 24W 6S1 SECONDARY
_ SOURCES (2)
IOFIFBARODS {48,64or104]
- tOF WABA RODS (8.20 or 24) 11 9'
Quad III N44 Loop 4
--4-4-t-11tT
-- -(
3.0 DISCUSSION OF THE CYCLE 10 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 9 core to the Spent Fuel Pool, changing out fuel inserts, and then loading the Cycle 10 core. Cycle 9 had indications of one leaking fuel pin in a high bumup assembly. Inmast sipping inspections and UT inspections were performed and positively identified the leaking assembly, which was a discharge fuel assembly. No leaking fuel assemblies were reloaded into the Cycle 10 core.
The first assembly (one of two source assemblies) to be reloaded was latched on October 17, 2002 and the last assembly to be loaded was unlatched on October 19. Inverse Count Rate Ratio (ICRR) was monitored during fuel loading.
The Cycle 10 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 assemblies identifications were again verified via underwater camera for each assembly as it was loaded into the core.
Core loading was completed on October 19, 2002. All 193 assemblies were loaded into the core without incident.
Following reload, the core loading pattern verification process was completed for the Cycle 10 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 1 Cycle 10. 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 5007F and all reactor coolant pumps running. Under these conditions, the longest drop time was 2.11 seconds for RCCAs at locations D02, B04, D14, P12, and M02.
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 1953 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 bank withdrawal.
Reactor Coolant System (RCS) dilution was initiated. During dilution, ICRR was monitored. Criticality was declared on November 10, 2002, and dilution was terminated.
Control Bank D (CBD) motion was used to stabilize flux level.
SUMMARY
OF RESULTS Cycle 10 initial criticality was achieved in a controlled manner on November 10, 2002 at 2322 hours0.0269 days <br />0.645 hours <br />0.00384 weeks <br />8.83521e-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.
Upon completion of LPPT, the plant was shutdown as directed by the Shift Manager to implement repairs due to equipment problems unrelated to Physics Testing.
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 and acceptance criteria.
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/°F 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/°F.
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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 1 Cycle 10, the Reference Bank was Shutdown Bank B.
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 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 100% power and a full core flux map was performed on December 18. 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-1 6 function.
TEST METHODOLOGY AND RESULTS Pre-critical adjustment ratios from the Unit 1 Cycle 10 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. AFD Monitor Check calculations passed acceptance criteria, but did not pass review criteria. Therefore, excore detector calibrations were required. Power ascension was allowed to continue as excore detectors were calibrated.
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 22, 2002. It was determined that AFD indications exceeded the acceptance criteria, therefore excore calibrations were performed prior to starting the Multipoint Measurement.
Six Quarter Core flux maps were performed on November 23, 2002 to be used in the Confirmation of the Calibration Standard. The flux maps were measured over a total change of 18% 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 1 Cycle 10 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 18, 2002 with the reactor at 100%
RTP. The AFD Monitor check satisfied acceptance criteria, but did not satisfy review criteria. Therefore, both the Intercept Current and Delta Q alignments for each excore NIS channel were 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 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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4.0
SUMMARY
This report is submitted as required following installation of fuel of a different design.
Cycle 10 contains 88 fresh assemblies supplied by W, 72 of which contain Integral Fuel Burnable Absorbers. Comanche Peak has not previously loaded fuel containing this type of burnable absorber.
Comanche Peak has previously used fuel of the Westinghouse OFA design. Since 1993, however, Siemens Power Corporation (now FRA-ANP) has been the primary fuel supplier. Although the Unit 2 Cycle 7 core contains eight W "lead use" assemblies of a similar fuel design, Unit 1 Cycle 10 is the first cycle in recent years in which the full reload has been supplied by Westinghouse. The design of this Westinghouse fuel, including the WABA burnable absorbers, is similar to the previous fuel used at CPSES; however, it uses ZIRLOTh materials to replace Zircaloy and contains IFBAs.
Unit 1 Cycle 10 reload, startup, and physics tests were performed 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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