ML20112G406

CPSES Unit 2 Cycle 3 Startup Rept

ML20112G406
Person / Time
Site:	Comanche Peak
Issue date:	05/30/1996
From:	Bryant M, Calder R, Terrel N TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
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ML20112G401	List: ML20112G399 ML20112G406
References
NUDOCS 9606110413
Download: ML20112G406 (36)

Text

. _ _ _ _ - _ _ _.

e RXE-96-002, Rev. 0 COMANCHE PEAK STEAM ELECTRIC STATION i

UNIT 2 CYCLE 3 STARTUP REPORT MAY 1996

)

MARK B. BRYANT

]

I Reviewed: I cRmM L Teance [1nwysC,30,te: s-22.%

Norman L. Terrel Reactor Engineering Supervisor s.

R D d. id,,.

Date: 5. 30 f r.

Approved:

Richard D. Calder Engineering Analysis Manager 9606110413 960607 DR ADOCK 0500 6

DISCLAIMER The information contained in this report was prepared for the specific requirements of Texas Utilities Electric Company (TUEC), and may not be appropriate for use in situations other than those for which it was specifically prepared.

TUEC PROVIDES NO WARRANTY HEREUNDER, EXPRESSED OR IMPLIED, OR STATUTORY, OF ANY KIND OR NATURE WHATSOEVER, REGARDINO THIS REPORT OR ITS USE, INCLUDING BUT NOT LIMITED TO ANY WARRANTIES ON MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE.

By making this report available, TUEC does not authorize its use by others, and any such use is forbidden except with the prior written approval of TUEC. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall TUEC 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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TABLE OF CONTENTS SECTION TITLE PAGE Title Page... -..........................

1 i

D isc l aime r..........................................

2 i

Table o f Contents.....................................

.3 List of Tables.........................................

4 List of Figures.....................

5 1.0 Introduction.........................................

6 l

2.0 Discussion of the Siemens Power Corporation Fuel Design...........

7 i

i 2.1 Mechanical Design.....................................

9 l

l 2.2 Nuclear Design 11 l

3.0 Discussion of the Cycle 03 Startup Tests l

3.1 Co re Loading......................................... 15 1

1 3.2 Control Rod Drop Time Measurementa........................ 17 3.3 Initial Criticality......................................

20 3.4 Low Power Physics Testing..............................

22 3.4.1 Determination of the Range for Physics Testing................... 24 3.4.2 ARO Boron End Point Measurements......................... 25 3.4.s Moderator Temperature Coefficient Measurements................. 26 j

3.4.4 Bank lwactivity Worth Measurements........................

27 l

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3.5 Flux Mapping 28 i

3.6 Incore/Excore Detector Calibration........................... 33 3.7 Reactor Coolant Flow Measurements.......................... 35 3.8 Core Reactivity Balance.................................. 36 3

l 1.

LIST OF TABLES TABI E TITLE PAGE 2.0-1 Fuel Assembly Design Parameters............................ 8 L

3.2-1 Rod Drop Time Men trement..............................18 3.4-1 HZP Physics Testing Results..............................

23 3.5-1 Low Power Flux Map Results..............................

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3.5-2 Intermediate Power Flux Map Results.........................

31 3.5-3 Full Power Flux Map Results

.........32 1

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LIST OF FIGURES FIGURE ~

TITLE PAGE 2.2-1 Core Loading Pattern................

12 2.2-2 Burnable Absorber and Source Rod Locations................

13 2.2-3 Control and Shutdown Rod Locations......................... 14 3.1-1 ICRR During Core Loading 16 3.3-1 ICRR During RCS Boron Dilution........................... 21 4

5 I

1.0 ItLTRODUCTION This report presents a summary of the startup of Comanche Peak Steam Electric Station, Unit 2, Cycle 3. Cycle 3 contains the first batch of Siemens Power Corporation (SPC) fuel for Unit 2.

This report satisfies the requirements of the Comanche Peak Technical Specifications, Section 6.9.1.1, that a summary report of unit startup and power escalation testing shall be submitted following installation of fuel that has been manufactured by a different supplier. This report shall be submitted within 90 days following completion of the startup testing.

l Comanche Peak Steam Electric Station, located in North Central Texas, is a two unit nuclear power plant. Unit I completed initial startup in 1990 and was declared to be in commercial operation on August 13,1990. Unit 1 is in Cycle 5. Unit 2 completed initial startup in 1993 and was declared to be in commercial operation on August 3, 1993. Each unit utilizes a four loop Westinghouse (W) Pressurized Water Reactor as the i

Nuclear Steam Supply System. The Nuclear Steam Supply System is designed for a thermal power output of 3425 MWth (3411 MWth reactor power). The plant is operated by TU Electric Company.

Cycle 3 initial criticality occurred on May 5,1996 and Low Power Physics Testing was completed on May 6.

The plant was synchronized to the grid on May 7.

Power ascension testing was completed with the performance of a full power flux map at 96.4%

RTP on May 17. Full power testing was conducted at 96.4% RTP due to secondary problems that prevented achieving 100% RTP.

6

1 2.0,

DISCUSSION OF THE SIEMENS POWER CORPORATION FUEL DESIGN The CPSES Unit 2 Cycle 3 reactor core is comprised of 193 fuel assemblies arranged in a low leakage core configuration. During the refueling prior to operation of Cycle 3, %

fresh Region 5 SPC fuel assemblies replaced 41 Region 2 E Optimized Fuel Assembly (OFA) assemblies, and 55 Region 3 OFA assemblies. A summary of the Cycle 3 fuel inventory is provided in Table 2.01.

The nominal core design parameters used for Cycle 3 are:

Core Power (Mwt) 3411 System Pressure (psia)..............................

2250 Core Inlet Temperature (*F)

......... 560. 8 RCS Thermal Design Flow (gpm)...................... 400,800 l

Average Linear Power Density (kw/ft).................... 5.434 The energy content of the Cycle 3 core has been designed to accommodate a refueling interval of approximately 18 months.

3 i

The CPSES Unit 2 Cycle 2 core configuration was comprised of 193 E OFA 17 x 17 fuel assemblies. The nominal outer diameter of the OFA fuel rods is 0.360". The Cycle 3 configuration includes 97 OFA fuel assemblies and 96 SPC 17 x 17 fuel assemblies.

The nominal diameter of the SPC fuel rods is 0.360".

In the CPSES Unit 2 Cycle 2 core configuration, E Wet Annular Burnable Absorber rodlets (WABAs) were used to shape the power distribution and to achieve a desirable moderator temperature coefficient. Solid burnable absorber rodlets manufactured by SPC and consisting of B C - Al 0 pellets are used in the CPSES Unit 2 Cycle 3 core design.

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TABLE 2.0-1 Fuel Assembly Design Parameters 1

CPSES Unit 2 Cycle 3 l

Region 3

4A 4B 5

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Enrichment (w/o Uus) 3.088 3.6076 4.0073 4.9435 Geometric Density (% theoretical) 96.08 95.46 95.40 94.774 i

Number of Assemblies 9

48 40 l

Expected EOC Assembly Average 30500-40700 42100 26900 Burnups (MWD /MTU) l Note:

All enrichments and densities are as-built.

Region 4 and 5 enrichments represent the value used for the central 132 inches. Regions 4 and 5 have natural uranium top and bottom axial blankets (each 6 inches in length).

]

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4 e

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l 2.1 MECHANICAL DESIGN i

The SPC 17 x 17 fuel assembly design, used for the Region 5 fuel assemblies and designated CPB-1 by SPC, contains 264 fuel rods and one instrument tube which a supported by eight bi-metallic grid spacers in the fuel assembly structure.

i The fuel assembly structure consists of an upper nonle, a lower nonle, twenty-four guide and eight spacer grids. Similar to the Region 4 E OFA fuel assemblies, the Reg SPC fuel assembly design includes a debris filter bottom nonle, a removable top and axial blankets of natural uranium.

i The major differences between the SPC fuel assembly design and the E OFA fuel i

assembly design are:

the SPC fuel rod cladding thickness is 0.025", relative to the 0.0225" thick e

cladding used in the E fuel assembly design.

The SPC grids are bi-metallic; i.e., Inconel springs imbedded in Zircaloy-4 e

grids are used to hold the fuel rods in place. The E OFA fuel assembly mid-grids are composed of Zircaloy-4; the top and bottom grids are Inconel.

In other respects, the SPC and E fuel designs are similar. Both designs include dished and chamfered pellets which are sintered to 95% of theoretical density. The SPC fuel assemblies are provided with unique serial numbers engraved on the top nonle. All locator holes in the top and bottom nonles are compatible with the upper and lower core support plates.-

Along with fuel assemblies, SPC provided 1584 B C burnable absorber rodlets 4

distributed among 76 clusters. Thimble plugs are placed in the positions in the burnable absorber assembly not taken by a burnable absorber rodlet, The physical (including geometrical) properties of the SPC fuel have been designed i

be compatible with the E OFA fuel assembly designs and with the CPSES reactor vessel internals, spent fuel racks, and fuel handling equipment. The following considerations apply:

the fuel assembly design provides adequate clearance with the core support e

plates at EOL; the fuel assembly design la dimensionally compatible with the existing rod e

control cluster assemblies; the top nonic was verified by testing to be dimensionally compatible with the e

l fuel handling tools; i

l 9

dimensional comparisons between the SPC and E fuel assemblies verified o

l' physical compatibility of the SPC fuel assembly with the new and spent storage racks; and, e

co-resident fuel dimensional compatibility was verified by evaluation of the fue assembly envelope and the centerline locations of the eight spacer grids. The assembly envelope and spacer grid locations are within design limits.

j The mechanical design criteria to which the SPC fuel rods, fuel assemblies, and absorber and thimble plug clusters have been designed are consistent w criteria used for the E OFA fuel assemblies in the CPSES Unit 2, Cycle 2 core configuration. Compliance with these mechanical design criteria has been demon through mechanical analyses of the SPC fuel rod and fuel assembly design l

methodologies which have been approved by the NRC.

1 These evaluations are valid for peak fuel assembly exposures of 54,200 MWD l

This exposure bounds the expected EOC burnup for the CPSES Unit 2, Cycl assemblies. The assumed power histories used in the mechanical design are cons with those histories expected for Cycle 3 operation. An appropriate number of tra (load changes, scrams, etc.) has been considered in the fatigue evaluations.

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2.2 NUCIPAR DESIGN The nuclear design of the CPSES ' Unit 2 Cycle 3 core was performed by TU Electric r

i accordance with methodologies approved by the NRC.

l The differences between the SPC fuel assembly design and the E OFA fuel asse designs are explicitly modeled in the neutronics codes.

The Cycle 3 core configuration is designed to meet a FTq x P ECCS limit of s 2.42 x K(z) for a flux difference (AI) bandwidth during normal operation conditions of

+3, -12% AI, where P is the reactor power normalized to rated thermal power.

The core loading pattern for Cycle 3 is shown in Figure 2.2-1.

The Cycle 3 core -

i configuration contains a total of 1584 burnable absorber rodlets located in the Region fuel assemblies. The locations of the burnable absorbers and source rod l

Figure 2.2-2. The locations of the control and shutdown rods are shown in Figure 2.2-3 1

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FIGURE 2.2-1 CORE LOADING PATTERN CPSES Unit 2 Cycle 3 R

P N

M L

K J

H G

F E

D C

B A

l 1

0007 EE08 EE91 DD25 EE92' EE78 0D40 l

1 (Quad IV7I

! M9 5

5 H5 5

5 D-9

' /" Quad l T i

N41 CC44 I DD55 EE42 DD02 EE85 DD52 EE12 DD34 EE56 DO64 CC29 '

N43 l

2

( bOOP 4]-

D2 l K-11 5

F-3 5

M13 5

K3 5

F-11 M2 bOOP l )

f CC45 < EE40 ! EE24 0074 EE59 D005 EE76 D006I EE95 DD70 EE71 EE27 CC06 3

P-12 5

5 R-9 5-P 11 5

B 11 5

A9 5

5 B-12 l D075l EE82 l DD38 EE81 0076 EE07 DD35 EE29 DD66 EE32 DD37 EE13 DD49 4

' E-6! 5 I G8 5

G2 5

H-13 5

J-2 5

H9 5

L-6 e

. EE05 l DD63 EE90 0044 EE45 DDO4 EE23 DD47 EE88 DD21 EE63 DD51 EE65 0D08 5

DD29!

G4 5 l G11 5 C 13 5

G10 5

J-10 5

H13 5

J-1 5

J-4 f EE54 j DD42 I EE96 0088 EE68 0086 EE64 0081 EE84 DD67 EE93 D073 EE51 0046 EE22 6

! 5 l N-10 5

P-9 5

D 13 5

L-12 5

H12 5

B-9 5

C 10 5

' EE15 l EE41 DD17 EE55 0023 EE50 DD65 EE36 0083 EE80 D018 EE10 D003 EEO9 EE39 7

5 5

E2 5

F-9 5

D-5 5

E-12 5

K-9 5

L-2 5

5 90*

DD31 DD87l EE34 D043 EE20 DD78 EE26 CC56 EE53 D068 EE87 DD48 EE19 D085 D009 8

C 12 f 5 C-8 5

D 11 5

A-8 5

M5 5

N-8 5

H4 E8 L8 lEE89 EE28 DD12 EE75 DD39 EE31 DD56 EE03 0D84 EE77 0010 EE60 DD24 EE33 EE72 9

l 5 5

E 14 5

F-7 5

L-4 5

M11 5

K-7 5

L14 5

5 EE61 DD16 EE06 0058 EE58 D072 EE74 DD71 EE67 D077 EE43 DD82 EE25 DD33 EE69 10 5

N6 5

P-7 5

C-4 5

E-4 5

M3 5

B-7 5

C-6 5

0022 EE17 DO60 EE16 D045 EE79 0001 EE38 DD19 EE57 D020 EE73 D054 EE18 DD14 11 G-12 5

G15 5

C-3 5

G6 5

16 5

H3 5

J-15 5

J 12 DD57 EE62 D041 EE70 0059 EE66 0036 EE14 DD53 EE37 0028 EE48 DD79 12 E-10 5

H7 5

G14 5

H3 5

J 14 5

J-8 5

L-10 CC49 EE47 EE46 0061 EE44 DD30 EE86 DD27 EE52 DD80 EE04 EE02 CC53 13 P-4 5

5 R-7 5

P-5 5

B-5 5

A-7 5

5 B-4 Quad lllT CC27 DD69 EE94 DD26 EE11 DD50 EE21 D011 EE01 D062 CC26 j Quad ll T 14 N44 p.14 K.5 5

F13 5

D-3 5

K-13 5

F-5 M14 N42

( Loop 3 Loop 2 >

15 i

/

0D15 EE83 EE49 0013 EE35 EE30 DD32 ASSEMBLY ID L

M7 5

5 4 11 5

5 D-7 CYCLE 2 LDCATION 0*

CC REGION 3 (3.1 w/o)

DD REGION 48 (4.0 w/o, Central Zone)

DD REGION 4A EE REGION 5 (3.6 w/o, Central Zone) 5 (4.95 w/o, Central Zone) 12

FIGURE 2.2-2 BURNABLE ABSORBER AND SOURCE ROD LOCATIONS CPSES Unit 2 Cycle 3 R

P N

M L

K J

H G

F E

D C

B A

f l

I a

i l

l

! 4S l 1

(QuadlV'Y I

l f

f iT N41

[I 88 l 208l 208' i 8B I.

Quadl 3 N43 2

(Loop 4 )!

l f

j f t(Loop 1.j

! I 20B 24B

,208l 248 l208l l

3 l

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l l

I li 208 248 248l i248l 24B l 208 4

- i i

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i

-J 8B l 248 i 24B 20Bl l 248 248 '

8B 5

I i

248 248 248

'24B' 124B 248 6

l 20B 24B 24B 20B 24B 248 208 7

90a' 208 208 208 208 208 20B 8

l 20B 24B 24B 20B 24B 24B 208 9

I 24B 248 24B 24B 24B 248 10 BB 248 24B 20B 248 248 8B 11 20B 24B 24B 248 248 20B 12 20B 24B 208 24B 20B 13 f Quad illT BB 20B 208 8B

/

N44

_ Quadil' 14 N42 Loop 3 q

j Loop 2 43 q

/

15 O*

88 8 B C RODLETS (8) 248 24 B C ROOLETS (40) 4 4

208 20 8 C RODLETS (28) 4S SECONDARY 4

SOURCES (2) i 13 l

FIGURE 2.2-3 l

CONTROL' AND SHUTDOWN ROD LOCATIONS CPSES Unit 2 Cycle 3 R

P N

M L

K J

H G

F E

D C

B A

i l

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}

j l

i f Quad IV]l f

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lfQuadI) i N41

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, SA !

B I C

B SA I

• l N43 2

i l I( Loop 1 j

k. LOOP 4 /

l l

i

' SD SB SB SC 3

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! D SE D !

l SA 4

i SA

{,

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l lSDl 5

l I

I l B l C

A C

B 6

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~

J I

l SB l

S8 7

l l

SE A

D A

SE C

8 90' C

l SB SB 9

B C

A C

B 10 SD SC 11 SA D

SE D

SA 12 SC SB SB SD 13

/ Quad lilt SA B

C B

SA f Quad ll T 14 N44 N42 Loop 3 j q Loop 2 >

15 q

0*

CONTROL NUMBER SHUTDOWN NUMBER BANK OF RODS BANK OF RODS j

A 4

SA 8

8 8

SB 8

C 8

SC 4

D 5

SD 4

s SE 4

l 14 l

l

3.1 CORE LOADING l

(RFO-106) l OBJECTIVES To specify the final core configuration and to control the loading sequences to ensure that the nuclear fuel assemblies are loaded in a safe and cautious manner.

TEST METHODOLOGY Refueling was performed by completely offloading the Cycle 2 core to the Spent Fuel Pool, changing out fuel inserts, and then loading the Cycle 3 core. Inmast sipping inspection was performed in the containment refueling machine mast during offload.

l Visual inspection of irradiated assemblies and UT inspection of suspect assemblies were performed in the Spent Fuel Pool. No known leaking fuel assemblies were reloaded for Cycle 3 operation.

The first assembly (one of two source assemblies) to be reloaded was la:ched on March

-22,1996 at 0734 and the last assembly to be loaded was unlatched on March 24,1996 at 1(M9. Inverse Count Rate Ratio (ICRR) plots were maintained during fuel loading (see Figure 3.1-1).

l The Cycle 3 core configuration is presented in Figures 2.2-1, 2.2-2, and 2.2-3.

SUMMARY

OF RESULTS On March 15,1996 (prior to core loading), fuel assembly insert number / type were verified in the spent fuel pool per RFO-204-2 by Reactor Engineering and Quality Control. There were no discrepancies identified.

Core loading was completed on March 24,1996 at 1049. All 193 assemblies were loaded in the core without incident.

On March 24,1996, at approximately 1500, the core loading pattern verification process I

l was completed for the Cycle 3 core loading pattern per RFO-204. RFO-204-1 was completed by Reactor Engineering and Quality Control.

Westinghouse refueling personnel, Reactor Engineering, and the Operations Fuel Handling Supervisor completed the fuel assembly alignment verifications per RFO-204.

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FIGURE 3.1-1 j

l I

ICRR During Core Loading Unit 2 Cycle 3 12 l

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..__ Q.~_w i

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~ w ~, _ %,_ e .

• e a. ' Q..

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[ Abs M+/bM hh,%g-A

-h M (f357~a,a g,,*.

09 l

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• f 't.++' y ss a g n e a

os 7

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01

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05 l

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0.3 a

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02 0i NUC 111 Revision 6l l

0 20 40 60 80 100 l20 140 160' 180 200 Number of Fuel Assemblies Loaded l6 N-31 Data + N-32 Data N-31 Trend = = N-32 Trend l j

1

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T 3.2 CONTROL ROD DROP TIME MEASUREMENTS l

l (NUC-2%)

OBJECTIVE To measure the drop time of each Rod Control Cluster Assembly (RCCA) under hot, full flow conditions in accordance with Technical Specification 3/4.1.3.4.

TEST METHODOLOGY l

The test set consists of two remote Gould strip chart recorders connected to each Digital Rod Position Indication (DRPI) cabinet inside containment via interconnecting cables at the rod control cabinets in the Safeguards Building. A rod bank is withdrawn and the reactor trip breakers are opened, causing the selected group of rods to drop. Analog signals created by the rods passing through the DRPI coils are fed to the Gould strip chart recorders, where the resultant printout is analyzed for drop times. Following completion of data transmission, the process is repeated for the remaining banks.

SUMMARY

OF RESULTS Technical Specification 3/4.1.3.4 requires that the drop time for each RCCA from the fully withdrawn position 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 l

531 F and all reactor coolant pumps operating. Under these conditions, the longest drop time was 1.49 seconds for RCCAs at locations H-8, M-2, and C-11.

l A summary of all rod drop times is presented as Table 3.2-1. The mean drop time to dashpot entry was 1.45 seconds.

The drop times for RCCAs at locations H8, M2, Cll, and L13 were greater than two standard deviations from the mean (1.42 to 1.49 seconds) but were within acceptance criteria. All four fuel assemblies are E OFA assemblies. H8 is a Region 3 assembly, and M2, C11, and L13 are Region 4B assemblies.

17

TABLE 3.2-1 ROD DROP TIME MEASUREMENT CPSES UNIT 2 CYCLE 3 Date: 5-5-96 No. of RCS Pumps Running: 4 RCS Temperature: 557"F RCS Pressure: 2240 psig i

RCCA Core Assy Fuel Guide Tube Assy Burnup Drag Force (Ibs)*

Drop Time BanivGroup Locanon ID Region Material (MWD /MTU)

(secs)

In Dashnnt Out Dashnnt CB All 11-6 DD81 4B W Improved Zirc-4 23184 15 15 1.45 CBAll H 10 DD71 4B W Improved Zirc-4 23238 15 10 1.44 CBA/2 F-8 DD68 4B E Improved Zirc4 22974 35 15 1.47 3

CBA/2 K8 DD78 4B E Improved Zirc-4 23209 15 15 1.44 j

CBB/l F2 DD34 4A E Improved Zirc4 22363 15 15 1.45 CBB!!

B-10 DD33 4A E Improved Zirc4 22264 20 15 1.45 CBB/l K 14 DD26 4A E 1mproved Zirc-4 22209 15 15 1.45 CBB/l P4 DD42 4A E Improved Zirc4 21828 30 20 1.46 CBB/2 B4 DD46 4A E Improved Zirc-4 22305 20 10 1.44 CBB/2 F-14 DDil 4A E 1mproved Zirc4 22068 10 15 1.45 CBB/2 P-10 DD16 4A E 1mproved Zirc-4 21897 30 20 1.44 CBB/2 K-2 DD02 4A E lmproved Zirc4 21931 15 10 1.45 CBC/1 H2 DD52 4B E 1mproved Zirc-4 20895 10 10 1.47 CBC/l B-8 DD85 4B E 1mproved Zirc-4 20375 15 10 1.44 CBCll H 14 DD50 4B E 1mproved Zirc-4 20456 15 15 1.47 CBC/l P-8 DD87 4B E 1mproved Zirc-4 21281 20 15 1.44 CBC/2 F-6 DD67 4B E 1mproved Zirc 4 20667 20 20 1.46 CBC/2 F-10 DD77 4B E 1mproved Zirc-4 21020 10 10 1.44 CBC/2 K 10 DD72 4B E Improved Zirc4 20990 15 15 1.45 CBC/2 K4 DD86 4B E Improved Zirc4 21034 15 15 1.45 CBD/l D-4 DD37 4A E 1mproved Zirc-4 22690 10 15 1.45 CBD/l M 12 DD41 4A E Improved Zirc-4 22816 15 15 1.46 CBD/2 D-12 DD28 4A E 1mproved Zirc-4 22578 15 15 1.44 CBD/2 M4 DD38 4A E 1mproved Zirc-4 22814 15 10 1.45 CBD/2 H-8 CC56 3

E starxiard Zirc-4 21219 10 10 1.49 i

1 18

o TABLE 3.2-1 ROD DROP TIME MEASUREMENT RCCA Core Assy Fuel Guide Tube Assy Burnup Drag Force (lbs)*

Drop Time Bank / Group Locanon ID Region Matenal (MWD /MTU)

(secs) in Dashoot Out Dashret SBA/1 D2 DD64 4B E Improved Zirc4 22484 10 10 1.48 SDAll B 12 DD79 4B E 1mproved Zirc-4 22635 15 20 1.45 SBAll M 14 DD69 4B E 1mproved Zirc-4 22900 25 25 1.48 SBA/1 P-4 DD75 4B E 1mproved Zirc4 22686 10 10 1.45 SBA/2 B4 DD49 4B E 1mproved Zirc4 22694 10 10 1.46 SBA/2 D-14 DD62 4B E 1mproved Zirc-4 22951 10 15 1.45 SBA/2 P 12 DD57 4B E 1mproved Zirca 22713 20 15 1.43 SBA/2 M2 DD5$

4B E lmproved Zirc4 22633 10 10 1,49 SBB/l G3 DD06 4A E Improved Zirc-4 18645 10 15 1.44 SBB/1 C-9 DD24 4A E Improved Zirc-4 18065 10 10 1.45 SBB/1 J-13 DD30 4A E Improved Zirc4 17855 15 15 1.46 SBB/1 N-7 DD17 4A E 1mproved Zirc-4 17748 15 15 1.46 SBB/2 C-7 DD03 4A E improved Zirca 17765 10 15 1.47 SBB/2 G-13 DD27 4A E Improved Zirc4 17710 10 10 1.46 SBB/2 N-9 DD12 4A E Improved Zirc-4 18076 15 10 1.47 SBB/2 J3 DDOS 4A E 1mproved Zirc-4 18235 15 15 1.44 SBC/l E3 DD70 4B E 1mproved Zirc-4 13933 15 15 1.45 l

SBC/1 C 11 DD54 4B E improved Zirca 13740 10 10 1.49 l

SBC/l L 13 DD61 4B E Improved Zirc4 14355 15 15 1.42 SBC/1 N-5 DD63 4B E improved Zirc.4 14308 15 10 1.45 SBD/1 C-5 DD51 4B E lmproved Zirc-4 14229 15 20 1.46 SBD/l E-13 DD80 4B E lmproved Zirc-4 14134 10 15 1.45 SBD/l N il DD60 4B E lmproved Zirc-4 14312 15 15 1.44 l

SBD/l L-3 DD74 4B E Improved Zirc4 14072 15 15 1.45 I

l SBE/l H4 DD35 4A E 1mproved Zirc4 21928 15 10 1.45 I

SBEll D-8 DD48 4A E Impmved Zirc-4 21891 10 10 1.43 1

SBE/1 H-12 DD36 4A E Improved Zirc-4 21940 15 15 1.45 SBE/1 M8 DD43 4A E Improved Zirc-4 21751 15 15 1.46 1

RCCA drag force measurements taken during latching of control rod drive Shafts per RFO-403. Drag force limits are as followS:

Rods completely out of " dashpot"......

40 lbS Tips of Rods in " dashpot" i 100 lbS 19

3.3 INITIAL CRITICALITY 1

'(NUC-101/ IPO-002B)

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 2028 ppm, the Shutdown and Control Banks were withdrawn to a desired critical position of Control l

Bank D at 187 steps in proper overlap sequence. Inverse Count Rate Ratio (ICRR) plots were maintained during bank withdrawal.

Reactor Coolant System dilution was initiated at 60 gpm. During dilution, the ICRR was plotted. When the ICRR decreased to 0.184, dilution was stopped; subsequent mixing with Control Bank D rod pulls to 202 steps brought the reactor critical.

SUMMARY

OF RESULTS Cycle 3 initial criticality was achieved on May 5,1996 at 2146. Criticality was achieved with Control Bank D at 202 steps and a boron concentration of 1911 ppm. The design estimated critical boron concentration with Control Bank D at 187 steps was 1921 ppm.

Figure 3.3-1 is a plot of ICRR for initial criticality plotted versus Reactor Makeup Water (RMUW) added.

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l FIGURE 3.3-1 ICRR During RCS Boron Dilution on 5/5/96 Unit 2 Cycle 3 12

- "3 ii i

1 Iw i

0.9 0:

I 07 o6 l

u o5

. i 4553 total gallons addedb 04 03

'N 0.2 v

0l lNUC-111 Revision 6l 0

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Gallons RMUW Added l-e-N-31 Data -+-N-32 Does l l

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3,4 LOW POWER PHYSICS TESTING (NUC-301)

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 an ABB/CE Reactivity Monitoring System to indicate reactivity changes below the point of ::dding heat.

LPPT also ensures the surveillance requirements for Technical Specification Special Test Exception 3.10.3 are met.

Results of individual tests completed during the initial criticality and low power test sequence are discussed in the following sections of this report. A tabulation of key physics measurement results is also included in Table 3.4-1. All required tests were performed. Initial criticality was achieved without incident on May 5,1996. A low power physics testing power range was determined and the reactivity monitoring system was verified to be operating properly. Boron endpoint concentration measurements and associated core reactivity balance calculations were performed.

The Isothermal Temperature Coefficient was then measured in order to calculate the Moderator Temperature Coefficient. The Moderator Temperature Coefficient, while slightly positive, was less positive than the Technical Specification limit and rod withdrawal limits were not required. Control rod worths were inferred using the bank excharige method (rod swap method). All required testing was satisfactorily completed.

Upon completion of LPPT, the plant was aligned as directed by the Shift Manager for power operations and additional power ascension testing.

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TABLE 3.4-1 4

HZP PHYSICS TESTING RESULTS d

i REVIEW CRITERIA TEST PARAMETER Design Actual Review Criteria All Rods Out Critical Boron (ppm) 1928 1894 1878 to 1978 Isothermal Temperature

-0.44

-1.02

-2.44 to + 1.56 Coefficient (pcm/*F) j Differential Boron Worth (pcm/ ppm)

-7.25

-7.44

-8.34 to -6.16 Reactivity Computer Error NA

• 1.52 %

4%

Reference Rod Bank Worth Error NA

-0.24 %

-10% to +10%

4 All Other Banks Worth Error (max.)

NA 7.94 %

-15 % to + 15 %

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Total Rod Bank Worth Error NA 1.18 %

s +10%

4

• The average of the absolute values of the difference between the reactivity indicated by the reactivity compute i

the corresponding reactivity calculated from the reactor period or doubling time for three separate checkouts of the I

ABB/CE Reactivity Monitoring System.

ACCEPTANCE CRITERIA TEST PARAMETER Design Actual Acceptance Criteria j

All Rods Out Critical Boron NA 246

< 1000 Difference (pcm) 4 Moderator Temperature Coefficient

+ 1.10

+0.52

< +5 l

q (pcm/*F) l Reference Rod Bank Worth (pcm) 819 821 696 to 942 Total Rod Bank Worth (pcm) 4652 4597 2 4187 t

NOTE: pcm means percent millitho, equivalent to a reactivity value of 105 AK/K 23

t 33.1

• NETERMINATION OF THE RANGE FOR PHYSICS TESTING

)

i 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.

j TEST METHODOLOGY With the reactor critical at a power level of approximately 1.0 E-8 amps (as indicated by

)

the Intermediate Range channels), approximately 40 pcm of positive reactivity was added by withdrawal of Control Bank D. Flux was allowed to increase until reactivity feedback effects were observed by a decrease in the indicated core reactivity, as indicated on strip chart recorders.

The lower physics testing range limit was set at the low end of the decade which was -

below the decade in which the point of adding heat was observed.

SUMMARY

OF RESULTS Detectable reactivity feedback was observed at approximately 9.5 E-7 amps as indicated i

by the Intermediate Range channels. The low power physics testing range was set at 2.0

)

E-8 to 2.0 E-7 amps.

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i 1.4.2.

'ARO BORON ENDPOINT MEASUREhtIUiTS OBJECTIVES To measure the critical boron concentration at various Control Bank configurations, TEST METHODOLOGY Conditions were established with Control Bank D greater than or equal to 190 steps (within 30-50 pcm of its endpoint configuration) with the reactor critical in the low power physics testing range. The control bank was inserted / withdrawn as applicable to its endpoint configuration 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 All boron endpoint measurements met the review criteria, as summarized below, Review Acceptance Bank Design Measured Difference Deria Criteria Position (ppm)

(ppm)

(ppm /pcm) 4 (pcm)

All Rods 1928 1894 34 / 253 50

< 1000 Out 1815 1784 31/ NA i 272 NA B

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l 3.4.3 ' MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS OBJECTIVE l

To measure the Isothermal Temperature Coefficient (ITC) and calculate the Moderator l

Temperature Coefficient (MTC).

l TEST METHODOLOGY The ITC measurement was performed by first decreasing, then increasing Tavg using 1

I steam dump control and measuring the resulting reactivity changes. 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.

i

SUMMARY

OF RESULTS

)

As required by Technical Specification 3/4.1.1.3.a. the BOL/ARO/HZP-MTC must be less positive than +5 pcm/"F. The subject MTC was determined to be +0.52 pcm/ F.

All measurements met the review criteria, as summarized below.

Review Acceptance Design Measured Difference Criteria Criteria (pcm/*R (pcm/ B (pcm/*R (pcm/*R (pcm/ F)

ITC

-0.44

-1.02 0.58 2.0 NA MTC

+ 1.10

+ 0.52 0.58 NA

< +5.0 f

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3:4.4 BANK REACTIVITY WORTH MEASUREMENTS OBJECTIVE i

To measure integral reactivity worth of each Control and Shutdown Bank.

TEST METHODOLOGY Integral bank worths were measured using the rod swap method. The subject bank was inserted then countered 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 worth of 821.0 pcm.

SUMMARY

OF RESULTS The review criteria was met, as summarized below.

Review Criterla-Individual Banks s 15 % of design or s 100 pcm, whichever is greater, Total Worth s 110% of design.

Acceptance Criteria:

Sum of measured bank worths shall be no less than 90% of the design sum of bank worths.

DIFFERENCE DESIGN MEASURED

[d-m]

[(d-m)/d]

BANK (oem)

(nem)

(ocm)

(%)

sBB Ref 819.00 821.00 2.00

-0.24 CDA 315.74 290.67 25.07 7.94 SBA 314.53 304.63 9.91 3.15 SBE 313.02 312.02 0.99 0.32 SBD 439.73 439.50 0.23 0.05 SBC 440.79 438.00 2.79 0.63 CBD 531.29 512.50 18.79 3.54 CBB 679.49 687.58

-8.10

-1.19 CBC 798.56 791.50 7.06 0.88 TOTAL 4.652.14 4.597.40 54.74 1.18 27

3 /5. FLUX MAPPING (NUC-101) l OBJECTIVE l

To verify adequate flux symmetry and power distribution during initial startup following r

' refueling.

I TEST METHODOLOGY i

[

Flux maps were taken at 27.87%, 78.42% and %.4% RTP. All acceptance criteria were met for the flux maps.

SUMMARY

OF TEST RFRULTS t

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t l

Low power flux map results are presented as Table 3.5-1. Incore tilts for lower and i

upper quadrant 2 exceeded 1.02. NUC-208 states "If the largest quadrant incore tilt is t

l greater than 1.02 but less than or equal to 1.04 the incore tilt is acceptable and reactor power can be increased with normal plant procedures. However, the core designers shall be notitled of the tilt. This notification is not a prerequisite for power escalation." The excore tilts matched the quadrants showing incore tilts.

Discussions with Reactor Physics (core designers) and Nuclear Engineering Management led to performing intercept current alignments on all four power range Nuclear Instrumentation channels per NUC-203 to zero the excore indicated tilts. New 120% currents were transmitted to I&C for the intercept current alignments.

Management expected that further contingency monitoring of tilts may be required up to the 80% plateau. FQ(Z) and l

FDHN were less than applicable limits. The extrapolated maximum allowable power l

level based on FQ(Z) was 97.43% RTP, and based on FDHN it was 97.02% RTP.

l Intermediate power flux map results are presented as Table 3.5-2. Upper quadrant 2 incore tilt still slightly exceeded 1.02. Further discussions with Reactor Physics and Nuclear Engineering Management resulted in an administrative power level limit of 95%

RTP until the core load verification tape could be reviewed for the possibility of a mistoaded assembly. Subsequent review of the tape confirmed a properly loaded core.

No technical concerns remained once the tape had been reviewed and power ascension to 100% RTP was authorized. NUC-203 base case map and quarter core maps were taken for the Incore/Excore Detector Calibration. See Section 3.6 for Incore/Excore Detector Calibration.

FQ(Z) and FDHN were less than applicable limits. The extrapolated maximum allowable power level based on FQ(Z) was 102.78% RTP, and based on FDHN it was 101.38% RTP.

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' Full power flux map results are presented as Table 3.5-3. Power level was stabilized at

%.4% RTP due to valve problems in the secondary plant. This power level satisfied the NUC-101 full power flux map prerequisite of 91-100% RTP. Upper quadrant 2 incore tilt still slightly exceeded 1.02. New 120% currents were transmitted to I&C for Nuclear Instrumentation alignment. See Section 3.6 for Incore/Excore Detector Calibration.

i FQ(Z) and FDHN were less than applicable limits. The extrapolated maximum allowable power level based on FQ(Z) was 105.06% RTP, and based on FDHN it was 103.33 %

RTP.

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l TABLE 3.5-1 LOW POWER FLUX MAP RESULTS i

(NUC-201, NUC-208)

Map ID:

U2C03M06 Date Performed:

05/09/1996 Power Level:

27.99 %

Cycle Burnup:

18.6 MWD /MTU or 0.4 EFPD l

Boron Concentration:

1633 ppm @ 0010 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> on 05/09/1996 Rod Position:

CBD at 180 Core Axial Offset:

4.04 %

I Measured Peak FQ(Z):

2.2098 FQ(Z) Limit:

4.7432 Maximum Measured FDHN: 1.5638 4

FDHN Limit:

1.8848 Incore Tilts: Review Criteria s 1.02. If :t 1.02 but s 1.04 the tilt is acceptable, however, the core designer is notified.

OUADRANT UPPER CORE TILT LOWER CORE TILT 1

1.00440 1.00357 2

1.02498 1.02276 3

0.97508 0.98527 4

0.99555 0.98839 I

NOTE 1:

Nuclear Peaking Factors include required uncertainties.

NOTE 2:

FQ limit includes the K(Z) axial penalty factor.

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TABLE 3.5-2 INTERMEDIATE POWER FLUX MAP RESULTS (NUC-201, NUC-203)

Map ID:

U'.C03M07 Date Performed:

05/12/1996 Power Level:

78.42 %

Cycle Burnup:

95.2 MWD /MTU or 2.2 EFPD Boron Concentration:

1408 ppm @ 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on 05/12/1996 Rod Position:

CBD at 183 Core Axial Offset:

-1.310 %

Measured Peak FQ(Z):

2.1266 FQ(Z) Limit:

3.0859 Maximum Measured FDHN: 1.5436 FDHN Limit:

1.6503 Incore Tilts: Review Criteria s 1.02. If 21.02 but s 1.04 the tilt is acceptable, however, the core designer is notified.

QLTADRANT UPPER CORE TILT LOWER CORE TILT 1

1.00545 1.00328 2

1.02081 1.01925 3

0.97876 0.98708 4

0.99498 0.99039 NOTE 1:

Nuclear Peaking Factors include required uncertainties.

NOTE 2:

FQ limit includes the K(Z) axial penalty factor.

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l TABLE 3.5-3 FULL POWER FLUX MAP RESULTS (NUC-201, NUC-203)

J Map ID:

U2C03M17 Date Performed:

05/15/1996 Power Level:

96.4 %

Cycle Burnup:

206.4 MWD /MTU or 4.9 EFPD Boron Concentration:

1340 ppm @ 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on 05/15/1996 Rod Position:

CBD at 200 Core Axial Offset:

-2.235 %

Measured Peak FQ(Z):

2. % 03 FQ(Z) Limit:

2.5104 Maximum Measured F1)HN: 1.5345 FDHN Limit:

1.5667 Incore Tilts: Review Criteria s 1.02. If 21.02 but s 1.04 the tilt is acceptable, however, the core designer is notified.

OUADRANT UPPER CORE TILT LOWER CORE TILT 1

1.00433 1.00327 2

1.02052 1.01945 3

0.97806 0.98711 4

0.99709 0.99017 NOTE 1:

Nuclear Peaking Factors include required uncertainties.

NOTE 2:

FQ limit includes the K(Z) axial penalty factor.

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o

3. 6,

INCORE/EXCORE DETECTOR CAI IRRATION

{

(NUC-203)

OBJECTIVES The object of this test is to measure the Nuclear Instrumentation System (NIS) res to a varying core flux distribution for determination of the normalized detector currents

.and the excore NIS and plant computer constants for the Incore vs. Excore detector calibration of CPSES Unit 2, Cycle 3. This procedure satisfies Technical Specification 4

surveillance requirements 4.3.1.1.2a and 4.3.1.1.7 for the NIS and N-16 systems.

TEST METHODOLOGY s

An incore/excore calibration was performed at 96.4% RTP.

The base case flux map, U2C03M07, was taken with the AFD near the target AFD at approximately 78.42% RTP, The transient portion of the test was performed at 78.42%

RTP by using bank D motion to vary the AFD between the lower and upper AFD limits.

The RCS boron dilution was initiated at approximately 1115 on May 12,1996. The AFD was reduced to and maintained near the lower target band limit for approxima I

two hours to allow the xenon concentration to redistribute. After this time, the rods wbre borated out and the first quarter core flux map (QCFM), U2C03M08, was taken with Control Bank D at 180 steps. A three-pass QCFM sequence was used. Additional three-2 pass QCFMs were taken at changes in the AFD of approximately 2-3%. A total of nine i

quarter core maps were obtained during the AFD transition. The axial flux difference approached the upper target bad limit and rod motion and dilution were used to return

}

the AFD to within the target band.

a l

A total of ten flux maps were taken at the 78.42% RTP plateau and the data we i

processed in parallel with the testing. The data reduction was completed. Based on the AFD Monitor Check from map U2CO3M07, no adjustments to the NI system were.

required to be performed at this power plateau. The confirmation of the calibration standard was performed per NUC-203 in accordance with methodology detailed in RXE-94-02. The confhmation resulted in a high degree of confidence that the standard is still valid.

Following confirmation of the calibration standard, reactor power was increased to about 96.4% RTP and stabilized with Xenon equilibrium. A higher power level was precluded due to valve problems in the secondary plant. U2C03M17 was taken at this power plateau and used to perform the fical testing required per NUC-101. An AFD Monitor Check was performed per NUC-203 and revealed the need for an intercept current alignment and delta-Q alignment. New 120% currents were transmitted to I&C for the final power ascension Nuclear Instrumentation alignment.

33

SUMMARY

OF RESULTS The incore/excore detector calibration sheets were forwarded to I&C for calibration of the NIS channels on May 16,1996 after they were reviewed and approved. Instrument and Control started and completed the calibrations of the NIS channels on May 17,1996.

This concluded the power ascension program outlined in NUC-101.

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3.7

• REACTOR COOLANT FLOW MEASUREMENTS (PPT-S2-7018)

OBJECTIVE l

To determine the total Reactor Coolant System (RCS) flow rate by precision heat balance l

measurements, with the RCS at s: 592*F and a 2219 psig, prior to exceeding 85% RTP, TEST METHODOLOGY The total RCS flow rate measurement was performed at 78.4% RTP.

The instrumentation used for determination of steam pressure, feedwater temperature, feedwater pressure, and feedwater venturi delta-P was calibrated within ninety days of performing the calorimetric flow measurement.

The RCS N-16 Transit Time Flow Meter (TTFM) is also utilized to measure the RCS flow rate. The TTFM measures N-16 produced in the reactor by activation of O-16.

l

SUMMARY

OF RESULTS l

The measured total RCS flow rate was 419,550 gpm. This met the acceptance criteria that the total RCS flow rate be a 408,000 gpm, per Technical Specification 3/4.2.5, which includes a flow measurement uncertainty of 1.8%.

For comparison, the measured total RCS flow rate for Unit 2, Cycle 2 was 419,080 gpm.

This flow rate was also well above the minimum required Technical Specification limit stated above.

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i 3.8

' CORE REACTIVITY BALANCE (NUC-205)

OBJECTIVE 1

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 96.4% 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.

l

SUMMARY

OF RESULTS The equivalent reactivity difference between measured and predicted boron concentration was 445.3 pcm (64.05 ppm) which met the acceptance criteria of 1000 pcm, as required by Technical Specification 4.1.1.1.2.

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.