ML20149D990

Cycle 3 Startup Rept

ML20149D990
Person / Time
Site:	Callaway
Issue date:	12/31/1987
From:	Bryant K, Moser T, Schnell D UNION ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
ULNRC-1711, NUDOCS 8802100143
Download: ML20149D990 (32)

Text

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UNION ELECTRIC COMPANY CALLAWAY PLANT CYCLE 3 STARTUP REPORT DECEMBER 22, 1987 1

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FACILITY OPERATING LICENSE NPF-30 l

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$$02*SOo$N $3 h P

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UNION ELECTRIC COMPANY CALLAWAY PLANT STARTUP REPORT d

Prepared by K. R. Bryant Submitted by [kf T ~

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Reviewed by 7M dbh Supervising Engineer, Reactor / Secondary Systems Approved by ^7 hi[c ~ Systems Engineering

. Fuperintende t.

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TABLE OF CONTENTS l

Page l

1.0 Introduction 1 l

l 2.0 Core Refueling 2 l 2.1 Rod Cluster Control Assembly Eddy Current Examination 12 t

3.0 Control Rod Drop Time Measurements 13 I

4.0 Initial criticality 15 5.0 Low Power Physics Testing '

5.1 Determination of Low Power Physics Testing Ranga 16 5.2 Dynamic Checkout of Reactimeter 17 5.3 Boron End Point Measurements 18 i 5.4 Moderator Temperature Coefficient Heasurements 19 5.5 Bank Reactivity Worth Measurements 20 6.0 Flux Mapping 21 -

7.0 Alignment of Nuclear Instrumentation System 25 8.0 Reactor Coolant Flow Measurement 26  ;

9.0 Core Reactivity Balance 27 l

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1.0 Ih'TRODUCTION This report presents a smemary of pisnt startup and power escalation testing for Refuel 2 at Union Electric's Callaway Plant. It satisfies the requirement of Technical Specifications that a Startup Report be submitted following installation of fuel having a different design. It also satisfies a conmitment 4 contained in the Callaway SER (NUREG-0830 Supplement 2. section 4.2.3.1) that Union Electric inspect control rods during one of the first 5 refuelings and provide the results-to the NRC.

Callaway Plant is a Westinghouse four-loop pressurized water reactor rated at 3411 MWt. It is located in Callaway County, Missouri, approximately 80 miles west of the St. Louis metropolitan area, and is owned and operated by Union Electric Company. ,

i Callaway Plant operated in Cycle 2 with a transition core  ;

consisting of 109 Westinghouse 17 x 17 low-parasitic (LOPAR) fuel assemblies and 84 Westinghouse 17 x 17 Reccnstitutable Optimized Fuel Assemblies. The unit was refueled with Westinghouse Vantage 5 fuel assemblies and the cycle 3 core now j consists of 13 LOPAR fuel assemblies 84 optimized fuel assem- l j blies, and 96 Vantage 5 fuel assemblies.

Cycle 3 fuel shuffle commenced on 9-25-87 and was completed 4

on 10/16/87. Initial criticality for the cycle occurred on 11/6/87. The plant was synchronized to the grid es. 11/15/87.

The startup testing program was completed with the flux map at full power on 11/27/87.

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2.0 CORE REFUELING (ETP-ZZ-00035)

OBJECTIVES To specify the final core configuration and to control the unloading and loading sequences.

SUMMARY

OF RESULTS Core loading was completed and verified to be in accordance with core design by video core mapping.

DISCUSSION 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. Concurrently, ultrasonic inspection of the core was performed, and 3 fuel assemblies which were to be reloaded were reconstit*'ed. In addition, an cddy current inspection was performed on the control rods.

The off-load began at 10:38 on 9/25/87, and the reload was completed at 23:43 on 10/16/87, for a total elapsed time of 21.6 days.

The Cycle 3 core configuration is shown in Figures 2.1 threugh 2.3 and Tables 2.1 through 2.6.

CALLA 0AY UNI 7 1 CORE LOADING PLAN CYCLE 3 180 R # N W L K J H G F t 0 C B A

^

I i da SC 4B SC 4B SC 4A Die E63 045 E71 083  !?4 024 l ,

2 SC 3 SC dB 5A da SA 4B SC 3 SC E84 C29 E85 043 til D17 t il D59 E42 C44 (93 3 BC 5A SC 48 SR 48 48 48 5B dB SC 5A SC E66 E02 E79 075 (31 058 051 041 (38 D65 E80 E04 E76 4 3 SC dB SB da li da 58 4A SB di SC 3 C49 E77 D82 Edi 022 E51 027 E52 009 E30 080 E70 C53 5 da SC 48 SB 3 SA 4A SA 4A 5A 3 58 48 SC 4A D35 E73 053 E59 C46 E20 C23 EOS 029 E26 C16 E33 1 046 E61 004 6 SC 48 58 4A 5A 4A A 1. 4A 58 4A SA 4A 58 dB SC EB3 070 E46 D21 (27 010 E47 039 E43 006 tid 020 E44 D78 EG9 7 48 ha 4B 58 4A SB dB SA dB 58 4A 55 48 54 4B 079 (28 071 E37 005 Edo 044 '210 054 (29 012 ESO D81 tis 077 90 8 BC da AB da 5A 4A 5A 3 SA da BA 4A 48 4A . 5C 270 E68 011 068 032 tii 040 (01 C36 (12 D34 E03 026 084 018 , ISO 9 dB SA dB 58 4A SP dB 5A 4B 58 da SB dB SA 48 055 E23 052 136 003 EJb 047 tot 067 E57 028 CSS 061 E24 064 to SC 48 58 da SA da 58 da 58 4A 5A 4A SB 48 BC E84 049 E48 CJO E21 D15 E60 037 E48 D25 E22 016 E45 ' 057 E82 11 da SC 48 5B 3 SA da BA 44 SA 3 55 48 SC da 002 E96 D42 (54 C63 t'7 038 ECC 033 (19 ' C56 E34 073 ttt 006 12 3 SC 4E 58 44 58 4A SB 4A 58 48 BC 3 C45 E78 CfC E58 D01 (32 D31 E39 D13 ES3 066 E87 Cos 13 BC SA SC dB 58 48 dB 4B 58 dB SC Ba SC to? E06 Es2 D69 (42 074 054 062 Els 074 E65 (CT t?2 14 BC 3 SC dB SA JA SA dB SC 3 SC E86 C26 175 D60 E13 D36 E25 072 E89 C27 E91 15 4A SC dB SC 48 SC da D19 E95 063 E44 048 (94 007 0 OtG.

Kfv AlttwSLy catfNtatiCN-R ID R = REGION NUWBER 10 = FutL ASSEMBLY IDEN71 FICA 710N Io\ 0 o e REFERENCE HOLT

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• CORE P!N HOLE O \ \

• McLDDowN BAe I

I FUEL / REGION LOCATIONS FIGURE 2.1 j

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CALLAWAY UN8T i >

CORE LOAD 2NG PLAN, 1 CYCLE 3 \

i 1801 1 R P N M L K V H G F E D C B A 1 1 PD PD Pti PD lPD PD PD e

B B B' B B B B l

2 / PD R 16A R PD R PD R 16A R PD 1 B A B A l'S A B A B A B 3 PD 8A 24A R 24A R SS4A R 24A R 24A PA PD i B B B A B 'A B A B A B B

.' Q B 1

, I 4 R 24A R 24A PD 24A R' 24A PD 24A R 24A R A B A B B B 'A B B B A B A 5 PD 16A R 24 d . PD PD PD 24A PD .PD PD 24A R 16A PD B B A S , B B B B B B B B A B B

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6 PD R t 2 3. A :PD i PD R 24A R 2/A- R PD PD 24A R PD B A Id B I'B jB A B A .B A B B B A B 7 PD PD R 24A PD 24A PD 20A PD 24A PD 24A R PD PD ,

B B, A B B B B B B B B B A B B 90 B PD R' PC R 24A R 2dA R 2OA R '24A R PD R PD 270 B A B A B A B A B A B A B A B 9 PD PD R 24A I 'd 24A PD 20A PD 24A PD C4A R PD PD B B A B B B 'S B B B B B A B B R 24A PD PD R 24A R 24A R '

PD PD 24A R PO 10JPD B A B B B A B Aa B A B B B A* B 11 PD 16A R 24A PD PD PD 24A PD PD PD 24A R 16A PD B B A B B B B B B B B B A B B 12 R 24A R 24A PD 24A R 24A D:? 24A R 24A R  !

A B A B B ,8 A B B B A B A 13 PD BA 24A R 24A R S$4A R 24A R 24A 8 .4 PD l B B B A B A B A Bs A B B B 14 PD R 16A R PD R PD R 16A R PD B A B A B A B A B A B s 15 PD PD PD PD PD .

PO PO i B B B B B B B l

1 0 DEG. f  !

l KEY: ASSEMBLY ORIENT &TIONt I

TYPE TYPE = COMPONENT TYPE PER FIGURE 4 o\ 0 o = REFERENCE HOLE '

OR OR = COM?ONENT ORIENTATION 4 \ 0 = CORE PIN HOLE L,D \ \ = HOLDDOWN BAR v ..

COMDONENT ORIENTATION- NOTE:

SEE FIGURE 4 FOR l A: NO SPECIAL ORIENTATION. COMPONENT CONFIGURATION, l B: HOLDDOWN BAR IDEN t' C APOS!TE F/A REFTRENCE HOLE. AND LOCATION OF HOLDDOWN C: HOLDDDWN BAR IDENT TOWARDS F/A REFEWENCE HOLE. BAR IDENTIFICATION.

A

! CORE' COMPONENT LOCATIONS l

,m g) s FIGURE 2.2 l i

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CALLAWAY UNIT 1 CDRE LOADING FLAN CYCLE 3 180 R P N M L' K V H G F E D C 8 A 1 FD MD TC 2 MD MD MD TC TC TC TC TC TC 3 MD MD MD MD 4 MD MD MD TC TC TC TC TC TC 5 MD MD MD MD 6 MD MD MD MD MD TC TC TC TC TC TC TC TC 7 MD MD MD MD 9^ 8 MD MD MD MD MD MD MD MD 270 TC TC TC TC TC TC TC TC 9 MD MD MD MD l

10  ?'O MD' MD TC TC TC TC TC TC TC .TC 11 MD MD MD MD MD 12 MD MD MD TC TO TC TC TC TC 13 MD MD MD MD 14 MD MD MD MD TC TC TC TC TC TC 15 MD MD TC 0 DEG.

KEY

• TOTAL IN CDRE FDMD FD = FIXED DETECTDR O TC MD = MDVABLE DETECTDR 58 TC = THERMDCDUPLE 50 INSTRUMENTAT!DN LDCAT!DNS i

i FIGURE 2.3

CALLAWAY. UNIT 1 CORE LOADING PLAN-

, CICLE 3

, REGION: 3 ENRICHMENT: 3.09 NUMBER REQUIRED: 13 CONDITION: RECYCLE IDENTIFICATION: AS LISTED FUEL ASSEMBLY ANSI ---CORE COMPONENT---

LOCATION IDENTIFICATION IDENTIFICATION TYPE IDENT ORIENT B-4 C53 LM0A41 R A B-12 C06 LMOA40 R A D-2 C44 LMOA48 R A D-14 C27 LMOA3J R A F-5 C16 LMOA3D PD B E-ll C6 LMOA4F PD B H-8 C36 LMOA4M R A L-5 C46 LMOA47 PD B L-ll C63 LMOA4J PD B M-2 C29 LMOA4C R A M-14 C26 LMOA4S R A P-4 C49 LMOA36 R A P-12 C45 LMOA30 R A l

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ASSEMBLY LISTING FOR REGION 3 l

TABLE 2.1 l

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CALLAWAY UNIT 1 CORE LOADING PLAN

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CYCLE 3 REGION: 4A ENRICHMENT: 3.39 NUMBER REQUIRED: 40 CONDITION: RECYCLE IDENTIFICATION: D01 - D40 .

FUEL ASSEMBLY ANSI ---CORE COMPONENT--- 1 LOCATION IDENTIFICATION IDENTIFICATION TYPE IDENT ORIENT l A-5 D04 LM0J0Q F5-- B i A-11 D08 LM0 JOE PD B B-8 D18 LMOJ0J R A D-6 D20 LM0J16 PD B D-8 D26 LM0JOC R .A D-10 D16 LM0J1C PD B E-1 D24 LMOJ0M PD B E-7 D12 LM0JO9 PD B E-9 D28 LMOJOL PD B E-15 D07 LM0J0D PD B F-4 D09 LM0J1A PD B F-6 DOS LM0JOH R A F-8 D34 LM0JOS R A F-10 D25 LM0 JOB R A F-12 D13 LM0J18 PD B G-5 D29 LM0J12 PD B G-ll D33 LM0 JOT PD B H-2 D17 LM0J0Z R A H-4 D27 LM0J19 R A H-6 D39 LM0J17 R A H-10 D37 LM0J15 R A H-12 D31 LM0J0X R A H-14 D36 LM0J0P R A J-5 D23 LM0J0G PD B J-11 D38 LM0JOR PD B K-4 D22 LMOJll PD D K-6 D10 LM0J0K R A K-8 D40 LM0J08 R A K-10 D15 LM0 JOY R A K-12 D01 LM0J13 PD B L-1 D14 LM0J0W PD B L-7 DO5 LM0JON PD B L-9 D03 LM0JOA PD B L-15 D19 LM0JOV PD B M-6 D21 LMOJ14 PD B -

M-8 D32 LM0JOU R A M-10 D30 LM0J10 PD B P-8 Dll LMOJ07 R A R-5 D35 LM0J13 PD B R-11 D02 LM0J0F PD B 1

I i ASSEMBLY LISTING FOR REGION 4A TABLE 2.2

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CALLAWAY UNIT 1 CORE LOADING PLAN 1

. CYCLE 3  !

l REGION: 4B  !

l ENRICHMENT: 3.80 NUMBER REQUIRED: 44 l CONDITION: RECYCLE IDENTIFICATION: D41 - D84 FUEL ASSEMBLY ANSI ---CORE COMPONENT---

LOCATION IDENTIFIchTION IDENTIFICATION TYPE IDENT ORIENT l A-7 D77 LM0J7M P5~~ B l A-9 D64 LM0J1H PD B  !

B-6 D78 LM0J22 R A ,

B-10 D57 LM0J7Q R A l C-5 D46 LM0J26 R A l C-7 D81 LM0J29 R A l C-8 D84 LM0J1R PD B C-9 D61 LM0J1D R A C-11 D73 LM0J1P R A i D-4 D80 LM0J7V R A D-12 D66 LM0J2A R A E-3 D65 LM0J7N R A E-lD D76 LM0J1N R A F-2 DS9 LM0J1Q R A F-14 D72 LM0J21 R A l G-1 L83 LM0J7R PD B G-3 D41 LM0J7W R A G-7 D54 LM0J1F FD B G-9 D67 LM0J7T ?D B G-13 D62 LM0J23 R A G-15 D48 LM0J1U PD B H-3 D51 LM0J1L SS4A B H-13 D56 LM0J1F SS4A B J-l D45 LM0J1V PD B J-3 DS8 LM0J7S R A J-7 D44 LM0J24 PD B J-9 D47 LM0J27 PD B J-13 D74 LM0J20 R A J-15 D63 LM0J1M PD B K-2 D43 LM0J7U R A K-14 D60 LM0J1X R A I L-3 D75 LM0J1W R A L-13 D69 LM0J1T R A M-4 D82 LM0J7Y R A M-12 D50 LM0J7X R A N-5 D53 LM0J28 R A )

N-7 D71 LM0J25 R A N-8 D68 LM0J15 PD B N-9 D52 LM0J1J R A i N-12 D42 LM0J1Y R A l P-6 D70 LM0J7P R A P-10 D49 LM0J1K R A R-7 D79 LM0J1G PD B R-9 D55 LM0J12 PD B l

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ASSEMBLY LISTING FOR REGION 4B l

TABLE 2.3 j l

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CALLAWAY UNIT 1 CORE LOADING PLAN CYCLE 3 .

REGION: SA ,

ENRICHMENT: 3.60 NUMBER REQUIRED: 28 CONDITION: NEW IDENTIFICATION: E01 THRU E28 FUEL ASSEMBLY ANSI ---CORE COMPONENT---

LOCATION IDENTIFICATION IDENTIFICATION TYPE IDENT ORIENT B-7 E16 LMOLJV PD B B-9 E24 LMOLK2 PD B C-3 E04 LMOLJF 8A B C-13 E07 LMOLJH 8A B E-6 E14 LMOLK4 PD B E-8 E03 LMOLJQ 24A B E-10 E22 LMOLK1 PD B F-5 E26 LMOLK8 PD B F-11 E19 *MOLJW

. PD B G-2 E18 LMOLK3 PD B G-8 E12 LMOLJK 20A B G-14 E25 LMOLK5 PD B H-5 EOS LMOLJG 24A B H-7 E10 LMOLJN 20A B H-9 EOS LMOLJL 20A B H-11 E09 LMOLJS '24A B J-2 E15 LMOLJU FD B J-8 E01 LMOLJR 20A B J-14 E13 LMOLKO PD B K-5 E20 LMOLJT PD B K-ll E17 LMOLJZ PD B L-6 E27 LMOLJY PD B L-8 Ell LMOLJM 24A B L-10 E21 LMOLK7 PD B N-3 E02 LMOLJJ 8A B N-lI E06 LMOLJP 8A B P-7 E2B LMOLJX PD B P-9 E23 LMOLK6 PD B FUEL ASSEMBLIES IDENTIFIED E13 THRU E28 CONTAIN lliTEGRAL FUEL BURNABLE ABSORBER RODS l

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I ASSEMBLY LISTING FOR REGION SA TABLE 2.4 9 -

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l CALLAWAY UNIT 1 CORE LOADING PLAN i

, CYCLE 3 j REGION: 5B ENRICHMENT: 3.80 NUMBER REQUIRED: 32 CONDITION: NEW IDENTIFICATION: E29 THRU E60 FUEL ASSEMBLY ANSI ---CORE COMPONENT---

LOCATION IDENTIFICATION IDENTIFICATION TYPE IDENT ORIENT C-6 E44 LMOLKJ F B C-10 E45 LMOLKX 24A B D-5 E33 LMOLKC 24A B D-7 E50 LMOLKV 24A B D-9 E55 LMOLKY 24A B D-11 E34 LMOLKE 24A B E-4 E30 LMOLKD 24A B E-12 E53 LMOLL5 24A B F-3 E38' LMOLKP 24A B F-7 E29 LMOLKB 24A B F-9 E57 LMOLL3 24A B F-13 E56 LMOLL4 24A B G-4 E52 LMOLKX 24A B G-6 E43 LMOLKK 24A B G-10 E48 LMOLKZ 24A B G-12 E39 LMOLKN 24A B J-4 E51 LMOLLO 24A B '

J-6 E47 LMOLL6 24A. B J-10 E60 LMOLKL 24A B J-12 E32 LMOLKF 24A B K-3 E31 LMOLKG 24A B K-7 E40 LMOLKT 24A B K-9 E35 LMOLKA 24A B K-13 E42 LMOLKH 24A B L-4 E41 LMOLKR 24A B L-12 E58 LMOLL1 24A B M-5 E59 LMOLKS 24A B M-7 E37 LMOLKQ 24A B M-9 E36 LMOLK9 24A B M-11 E54 LMOLKW 24A B N-6 E46 LMOLL2 24A B N-10 E49 LMOLKU 24A B l

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1 ASSEMBLY LISTING FOR REGION SB TABLE 2.5 l

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CALLAWAY' UNIT 1 CORE LOADING PLAN CYCLE 3 REGION: SC ENRICHMENT: 4.20 NUMBER REQUIRED: 36 CONDITION: NEW IDENTIFICATION: E61 THRU E96 FUEL ASSEMBLY ANSI ---CORE COMPONENT---

LOCATION IDENTIFICATION IDENTIFICATION TYPE IDENT ORIENT A-6 E69 LMOLLX PD B A-8 E90 LMOLLZ PD B A-10 E82 LMOLLK PD B B-3 E76 LM0LLV PD B B-5 E61 LMOLL8 16A B B-ll E81 LMOLLE 16A B B-13 E72 LMOLLN PD B C-2 E93 LMOLM2 PD B C-4 E70 LMOLLH 24A B C-12 E87 LMOLLM 24A B C-14 E91 LMOLM8 PD B D-3 E80 LMOLLG 24A B D-13 E65 LMOLLB 24A B E-2 E62 LMOLLL 16A B E-14 E89 LMOLM1 16A B F-1 E74 LMOLLY PD B F-15 E94 LMOLM6 PD B H-1 E71 LMOL'.W PD B H-15 E64 LMOLLC PD B K-1 E63 LMOLL9 PD B K-15 E95 LMOLM7 PD B L-2 E85 LMOLLD 16A B L-14 E75 LMOLLQ 16A B M-3 E79 LMOLLF 24A B M-13 E92 LMOLLR 24A B N-2 E84 LMOLM5 PD B N-4 E77 LMOLMO 24A B N-12 E78 LMOLM4 24A B N-14 E86 LMOLM3 PD B P-3 E66 LMOLLA PD B P-5 E73 LMOLLP 16A B P-ll E96 LMOLL7 16A B P-13 E67 LMOLLT PD B R-6 E83 LMOLLJ PD B R-8 E68 LMOLLU PD B R-10 E88 LMOLLS PD B i

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i ASSEMBLY LISTING FOR REGION SC l TABLE 2.6 I l

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2.1 ROD CLUSTER CONTROL ASSEMBLY (RCCA) EDDY CURRENT EX^.MINATION (ETP-ZZ-00044)

OBJECTIVES To perform eddy current examinations of the Callaway RCCA's.

SUMMARY

OF RESULTS An RCCA is considered acceptable for continued use if at least 50% of the cladding cross sectional area for each rodlet remains at all axial locations. All RCCA's examined by eddy current testing met this acceptance criterion.

DISCUSSION RCCA wear has been observed at a number of plants. This wear is believed to be due to fretting caused by flow induced vibrations between the RCCA rodlets and upper internals guide cards. Based on hafnium levels in the Reactor Coolant during Cycle 2, it was believed that the cladding of a number of rodlets at Callaway had worn through.

Westinghouse was therefore contracted to perform eddy current testing on the RCCA's to determine the extent of wear.

The Westinghouse eddy current fixture used encircling coils to take data on 6 rodlets while the RCCA was being inserted into the fixture. Each encircling coil provided data on the percent of eroded cladding cross-sectional area. Using this information, the operator selected 1 of the 6 rods, and used a pancake coil array to obtain sequential rod radii during withdrawal of the RCCA. The rod radii were used to plot wear scar profiles and estimate scar depths. The RCCA was rotated and the process repeated until encircling coil data had been obtained on all 24 rodlets of the RCCA.

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3.0 CD';fROL R0D DROP TIME MEASUREMENTS (ESP-ZZ-00016)

OBJECTIVES 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.

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 2700 msee from the beginning of decay of stationary gripper coil voltage to dashpot entry with Tavg greater than or equal to 551*F and all reactor coolant pumps operating. Under these conditions, the longest drop time was 1466 msec for rod E03.

A summary of all rod drop times is presented as Figure 3.1.

The mean drop time to dashpot entry was 1425 msec.

DISCUSSION The Westinghouse Automatic Rod Drop Test Set was used to measure rod drop times.

The test set consists of two Remote Units (one connected to each Digital Red Position Indication (DRPI) cabinet inside containment) and a Central Unit (connected to the Rod Control Logic cabinet). A rod bank is withdrawn and the control circuit for one group of tads is interrupted from the Central Unit, causing the selected group of rods to drop. Analog signals created by the rods passing through the DRPI coils are fed into the Remote Units, where they are converted to digital signals and stored. Upon command, the digital data is transmitted to the Central Unit for display (or printout) of a graphic representation of the rod drop signals and associated drop times. Following completion of data transmission the process is repeated for the remaining withdrawn group and, subsequently, for each bank.

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l FIGURE 3.1 i RCD DROP TDS TABUULTION Hof,11.',11 TIW; CYCLE 3 A B C D E F G H J K L M N P R is _. l 14 1414 1432 1440 1424 1454  !

2044 2052 2050 2044 2094 _

13 1430 1428 1410 1404 2040 2058 2020 2014 ~

12 1424 1436 1412 1400 l396 2044 20 % 2012 20:0_ 2026 11 1428 1440 2048 2056 10 1416 1438 1412 1428 1422 2052 I 2046 20_58 2022 2038 9 1412 1408  !

20'2 2038 l 2M, 8 1420 1422 1428 1406 1426 1422 1424 -90' 2030 2012 2048 2016 2036 2012 2034 7 1422 1414 2042 2024 6 1436 1416 1420 1424 1410 2056 2026 2040 2014 - 2040 s 1434 1428 2044 2048 6 1404 1418 1418 1430 1438 2024 203R 202R ,2050 2038 3 1466 1428' 1430 1414 7076 705R 2040 2024 2 1464 1426 I1434 1446' *420 2094 2076 2054 2066 2050 1

180' xxzz TDs To DAS100T E M Y (HSEC)

,xxxx TDE TO DASUPOT BOTTOM (MSEC) l l

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4.0 INITIAL CRITICALITY ,

l (ETP-ZZ-ST002)

OBJECTIVES To achieve initial criticality following refueling in a.

cautious and controlled manner.

t

SUMMARY

OF RESULTS Initial criticality was achieved on 11/6/87 at 1126.

Following stabilization, Control Bank D was at 172 steps with a boron concentration of 1520 ppm. The design critical boron concentration with Control Bank D at 160 steps was 1507 ppm.

DISCUSSION i

From an initial condition of all rods in and a boron concentration of 2076 ppm, the Shutdown and Control Banks were ,

withdrawn to a final position of Control Bank D at 160 steps with all other banks fully withdrawn. Inverse Count Rate Ratio (ICRR) l plots were maintained during bank withdrawal.

Reactor Coolant System dilution was initiated at 60 gpm.

During dilution, the ICRR was plotted. When-the ICRR reached 0.1, the dilution rate was reduced to 30 gpm and the ICRR was normalized to a value of 1.0. When the ICRR decreased below 0.2, dilution was ,

stopped; subsequent mixing brought the reactor critical.

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5.1 DETERMINATION OF LOW POWER PHYSICS TESTING RANGE (ETP-ZZ-ST003)

OBJECTIVES To determine the power level at which detectable reactivity

feedback from fuel heating occurs ..nd to establish the power range for low power physics testing.

SUMMARY

OF RESULTS, Detectable reactivity feedback was observed at 1.2 E-6 amps as indicated by the Reactimeter. The low power. physics testing range was set at 5.0 E-8 to 4.0 E-7 amps.

DISCUSSION With the reactor critical at a powar icvel of approximacely 1 E-8 amps (as indicated by the Intermediate Range channels),

approximately 30 pcm of positive reactivity was added by withdrawal of Control Bank D. Powar was allowed to increase until reactivity feedback effects were observed as indicated on strip chart recorders by an incresce in Tavg and a decrease in reactivity.

The upper physics testing range limit was set at one-third of the current at which fuel heating occurred. The lower physics testing range limit was set at .125 of the upper limit.

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5.2 DYNAMIC CHECK 0UT OF REACTIMETER (ETP-SE-ST002)

OBJECTIVES To verify accuracy _of reactivity measurements obtained from.

the Reactimeter. .

SUMMARY

OF RESULTS The absolute deviation between indicated and theoretical reactivity was determined to be 2.7%. This met the acceptance criteria of 4.0%.

DISCUSSION The Reactimeter is a digital computer-based system for the on-line computation of dynamic reactivity.during low power physics ,

testing.

4 For dynamic checkout of the Reactimeter, the controlling bank

• was withdrawn approximately 25 pcm while monitoring reactivity using the Reactimeter. The resulting reactor period was determined from the increasing flux level. The measured reactor period was -

used to calculate a theoretical reactivity which was compared with the reactivity measured by the Reactimeter. This process was repeated for positive reactivity additions of approximately 50 and 15 pcm. The average absolute deviation between indicated and theoretical reactivity was then calculated for comparison with the +

acceptance criteria.

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5.3 BORON ENDPOINT MEASUREMENTS (ETP-ZZ-SI004)

OBJECTIVES To measure the critical boron concentration at various Control Bank configurations.

SUMMARY

OF RESULTS All boron er.dpoint measurements met the design criteria, as summarized below.

Bank Measured Boron Acceptance Configuration Concentration (ppm) Criteria (ppm)

All Rods Out 1550 1523 1 50 Control Bank 1486 1459 1 219 D In Control Banks 1379 1340 2 201 C and D In DISCUSSION Conditions were established with the controlling rod bank within 50 pcm of its endpoint configuration with the reactor critical in the low power physics testing range. The controlling 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.

5.4 MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS ,

(ESP-ZZ-00009)

OBJECTIVES To measure the Isothermal. Temperature Coefficients (ITC) and determine the Moderator Temperature Coefficients (MTC) under various Control Bank configurations.

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SUMMARY

OF RESULTS As required by Te:hnical Specification 3/4.1.1.3, the MTC for the all rods out (AR0) configuration must be less positive than 0 pcm/*F. The subject MTC was determined to be 0.37 pcm/*F.

All ITC measurements met the design criteria, as summarized below.

Bank- Measured Boron Acceptance-Configuration ITC (pcm/*F) Criteria (pcm/*F)

All Rods Out -1.97 -2.79 2 3 l

Control Bank -2.92 -3.68 2 3 D In i i

Control BankJ -6.06 -6.92 2 3  :

C and D In DISCUSSION The ITC measurement was performed by first decreasing, then j increasing Tavg using steam dumps and measuring the resulting I reactivity changes. The'ITC is the change in reactivity divided by l the associated change in temperature, j The MTC was determined by subtracting the design Doppler Temperature Coefficient from the ITC.

Doe to measurement of a positive MTC for the ARO cor. figuration, rod withdrawal limits were established (presented as Figure 5.4.1) and a Special Report (ULNRC-1679) was submitted to the Commission.

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5.5 BANK REACTIVITY WORTH MEASURCMENTS' -

(ETP-ZZ-ST005)  ;

OBJECTIVES To measure integral reactivity worth of each Control Bank.

SUMMARY

OF RESULTS All design criteria were met, as summarized below.

Control Integral Acceptance l Bank Worth (pcm) Criteria (pcm) +

D 551 533 ! 53 C 896 986 1 99 l l

B 1327 1391 139 A 394 384 1 38 l

DISCUSSION i

Integral bank worths were measured using the boron exchange method. Reactor Coolant System dilution was initiated at a l l

constant rate. The subject bank was periodically inserted in response to the change'in reactivity caused by dilution. The reactivity resulting from each incremental bank movement was measured using the Reactimeter and summed to yield integral bank worth.

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6.0 FLUX MAPPING (ETP-SR-ST001)

OBJECTIVES To verify adequate flux symmetry and power distribution during initial startup following refueling.

SUMMARY

OF TEST RESULTS Low power flux map results are presentett as Table'6.1. The Safety Review Criteria that incore tilt is .i.04-was met. .The Design Review Criteria that incore tilt is st.02 and the error between measured and predicted normalized en:halpy rise for each instrumented assembly is s 10% was also met. ,

Intermediate power flux map results are presented as Table 6.2. The Design Review Criteria that the error between measured and predicted normalized enthalpy rise for each instrumented assembly is s 10% was met.

Full power flux map results are presented as Table 6.3. The

Design Review Criteria that the error between measured and predicted normalized enthalpy rise for each instrumented assembly is s 10% was met. ,

DISCUSSION A full core flux map was taken prior to criticality to check the operability of the Flux Mapping System.

At approximately 30 percent power a full core flux was taken. '

Safety review criteria for incore tilt s1.04 was met, as well as the design review criteria for incore tilt s1.02 and s10% error  ;

between measured and predicted normalized enthalpy rise for each instrumented assembly.

At approximately 48% power the nuclear enthalpy rise hot channel factor and the heat flux hot channel factor were verified to be within the limits prescribed by Technical Specifications.

At approximately 100% power the following surveillances were  ;

performed: l

1. The absolute incore versus indicated axial flux difference was verified to be less than 3%.

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2. The nuclear enthalpy rise hot channel factor and the heat flux hot channel factor were verified to be within the limits prescribed by Technical Specifications. )

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TABLE 6.1 LOW POWER FLUX MAP RESULTS Map ID: 87-026 Date Performed: 11-17-87 Power Level: 31%

Cycle Burnup: 24.7 MWD /MTU Boron Concentration: 1250 ppm Rod Position: CBD at 200 steps Limiting Measured F : 2.0623 Axial Location 32 Core Location N12 T.S. Limit:

9 4.0623 Limiting Measured F I

• AH T.S. Limit: 1.8002 Total Core Axial Offset: 12.717%

Maximum F AH Error:: 8.1% Core Location N12 Incore Tilts:

QUADRANT UPPER CORE TILT LOWER CORE TILT 1 1.012 1.014 2 0.992 0.993 3 1.003 0.999 4 0.994 0.994

TABLE 6.2

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INTERMEDIATE POWER FLUX MAP RESULTS Map ID: C7-027 Date Performed: 11-18-87 Power Level: 48%

Cycle Burnup: 51.6 MWD /MTU Boron Concentration: 1181 ppm Rod Position: CBD at 199 steps Limiting Measured F :

q 2.0182 Axial Location 49 Core Location D03 T.S. Limit: 3.0160 Limiting Measured F H: 1.4132 Core Location P09 T.S. Limit: 1.7195 Total Core Axial Offset: 2.274%

Maximum F rr r: .9 re cation G12 AH Incore Tilts:

QUADRANT UPPER CORE TILT LOWER CORE TILT 1 1.009 1.011 2 0.992 0.993 3 1.001 0.999 4 0.998 0.997 TABLE 6.3 FULL POWER FLUX MAP RESULTS Map ID: 87-033 '

Date Performed: 11-27-87 Power Level: 100%

Cycle Burnup: 249 MWD /MTU Boron Concentration: 1033 ppm Rod Position: CBD at 224 steps Limiting Measured F : 1.8284 Axial Location 19 Core Location N08 T.S . Limit : 2.0163 Limiting Measured F 1.3742 Core Location N08 H

T.S. Limit : 1.4874 Total Core Axial Offset: -4.123% .

I Maximum F Error: 8.01% Core Location H06 aH Incore Tilts:

QUADRANT UPPER CORE TILT LOWER CORE TILT i l

1 1 1.001 1.000 l 2 1.000 1.000 '

3 0.999 0.999 4 1.000 1,000

. 7.0 ALIGNMENT OF NUCLEAR INSTRUMENTATION SYSTEM (ETP-SE-ST001)

OBJECTIVES i

To generate alignment data for the Intermediate and Power Range detectors prior to criticality.

4 To verify adequate overlap between the Source and Intermediate Range channels and between the Intermediate and Power Range channels.

To verify that initial calibration of the Intermediate and Power Range channels v. sing calculated alignment data did not result in non-conservative trip setpoints.

To normalize Power Range detectors during power ascension.

SUMMARY

OF RESULTS The acceptance criteria that the overlap between the Source and Intermediate Range channels is 2 1) decades was met.

$ DISCUSSION

• 1 l A preliminary alignment of the Intermediate and Power Range 4

channels was performed prior to criticality in anticipation of reduced currents and a mismatch between the Intermediate and Power Range channels (based on Cycle 2 alignment).

The preliminary alignment was based on calculations which multiplied Cycle 2 currents by the ratio of the sum of selected

-l assembly powers (predicted) of Cycle 3 to the same (actual) from Cycle 2. Intermediate and Power Range trip setpoints were i

monitored during power ascension to ensure that the initial alignment did not result in non-conservative setpoints. The percent power indications of the Power Range channels were adjusted during power ascension based on secondary calorimetrics.

An incore/excore calibration was performed at approximately 77% RTP. An incore/excore comparison was performed at approximately 100% RTP, which verified agreement.

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> L 8.0 REACTOR COOLANT FLOW MEASUREMENT (ESP-BB-03015)

OBJECTIVES To determine the Reactor Coolant System-(RCS) flow rate by precision heat balance measurements.  :

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SUMMARY

OF RESULTS i

The RCS flow rate was determined to be 410,750 gpm.. This met the acceptance criteria that the RCS flow rate _be a 382,630 gpm.

4 DISCUSSION The RCS flow rate measurement was performed at approximately '

100% RTP. The instrumentation used for determination of steam pressure, feedwater temperature, and feedwater venturi delta-P was  ;

calibrated within seven days of performing the calorimetric.

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9.0 CORE REACTIVITY BALANCE, (ESP-ZZ-00013)

OBJECTIVES To compare the overall core reactivity balance with predicted ,

values'at full power.

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SUMMARY

OF RESULTS The equivalent reactivity difference between measured and

~

predicted boron concentration was 284 pcm (21 ppm) which met the acceptance criteria of 1000 pcm, as required by Technical Specification 4.1.1.1.2. i DISCUSSION Under equilibrium conditions at approximately 100% RTP, the Reactor Coolant System 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.

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___-________________1

__ _--- _. . . . . ~ . . .

, Umov ,

Etscraic >

a CMaway Flant January 22, 1988 I

i U.S. Nuclear Regulatory Commission Document Control Desk '

Washington, DC 20555 ULNRC-1711 DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 FACILITY OPERATING LICENSE NPF-30 START _UP REPORT The enclosed Startup Report is submitted pursuant to Sections 6.9.1.1, 6.9.1.2, and 6.9.1.3 of the callaway Unit 1 Technical Specifications. .I Very truly yours,

}

sAh ,

Donald F. Schnell  !

DFS/J / crc Encloeure  !

cc: Distribution attached 1

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Manng hadh!ss: PO, Su 62G FJKn MO 65251

U.S. Nuclear Regulatory Commission Page 2 January 22, 1988 cc: Mr. A. Bert Davis, Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road i Glen Ellyn, IL 60137 1 Mr. J. M. Hinds >

l Chief, Proj ect Section 1A U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Mr. T. W. Alexion Licensing Proj ect Manager, Callaway Plant Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop P-316 7920 Norfolk Avenue Bethesda, MD 20014 Mr. B. H. Little Resident NRC Inspector, Callaway Plant Callaway Resident Office U.S. Nuclear Regulatory Commission RR #1 Steedman, MO 65077 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 I

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