ML20107D684

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Cycle 11 Startup
ML20107D684
Person / Time
Site: Point Beach NextEra Energy icon.png
Issue date: 01/31/1985
From: Harris R, Kurtz P
WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20107D662 List:
References
TAC-49970, NUDOCS 8502250108
Download: ML20107D684 (51)
v • d • e


Text

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WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT UNIT 2 CYCLE 11 STARTUP JANUARY,1985 1

BY R. L. HARRIS -

P. N. KURTZ pM**SR Mfu' _. .

P,

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TABLE OF C0NTENTS Page

- LIST OF TABLES iii LIST OF FIGURES iv PREFACE v

. SECTICN 1.0, REFUELING 1 1.1 Core Unload 1 1.2 Insert Changes 1 1.3 Fuel Assembly Inspections 2 1.4 Core Reload 2 1.5 Core Design 4 SECTION 2.0, CONTROL ROD OPERATIONAL TESTING 9 ,

2.1 Rod Drop Times 9 2.2 Control Rod Mechanism Timing 9 2.3 Rod Position Calibration 9 SECTION 3.0, THERMOCOUPLE AND RTD CALIBRATION 13 SECTION 4.0, PRESSURIZER TESTS 15 4.1 Heater Capacity 15 4.2 Spray valve Effectiveness 15 4.3 Heater Effectiveness 15 SECTION 5.0, REACTOR COOLANT SYSTEM 16 5.1 RTD Manifold Flow 16 5.2 Flow Transient Times 16 SECTION 5.0, CONTROL SYSTEMS 18 SECTION 7.0, TRANSIENTS 18 SECTION 8.0, INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS 18 d.1 Initial Criticality 18 8.2 Reactivity Computer Checks 19 a SECTION 9.0, CONTROL ROD WORTH MEASURiDENT 21 9.1 Test Description 21 9.2 Data Analysis and Test Results 22 9.3 Evaluation of Test Results 22' I

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." Page SECTION 10.0, TEMPERATURE COEFFICIENT MEASUREMENTS 27 SECTION 11.0, BORON WORTH AND ENDPOINT MEASUREMENTS 28 SECTION 12.0, POWER DISTRIBUTION 30 SECTION 13.0, XENON REACTIVITY 37 SECTION 14.0, SHUTDOWN MARGIN CONSIDERATIONS 37 SECTION 15.0, EXCORE DETECTOR BEHAVIOR 37 15.1 Detector Current Versus Power Level 37 15.2 Excore Axial Offset Response 40 15.3 Channel Calibration 40 SECTION 16.0, OVERPOWER, OVERTEMPERATURE AND DELTA FLUX SETPOINTS CALCULATION 42 16.1 overpower yd Overtemperature AT Setpoints Calculation 42 16.2 Delta Flux Setpoints Calculation 42 SECTION 17.0, FUEL PERFORMANCE 45 SECTION 18.0, CONCLUSION 46 O

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LIST 0F TABLES Table Page 1-1 Changes to Core Loading Plan 3 1-2 Uranium Loading 5 3-1 RTD Calibration Check 14 4-1 Heater Group Energy Input 15 5-1 Reactor Coolant Flow Transient Times 17 8-1 Reactivity Computer Checkout 20 l 9-1 Critical Rod Configuration Data 25 9-2 Comparison of Inferred / Measured Bank Worths with Design Predictions 26 10-1 Isothermal Temperature Coefficients 27 11-1 Boron Worth and Endpoints 28 12-1 Initial Power Escalation, Flux Map Results 31 14-1 Excess Shutdc.'m Worth Available for a Full Power Trip 37 15-1 100% Currents (p amps) , 39 15-2 Excore Azial Offset Response History 41 15-3 BOL Calibration Currents 41 16-1 Overtemperature AT Constants 43 16-2 overpower AT Constants 44 17-1 Typical Isotopic composition of Primary

  • Coolant Activity 45 O

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LIST 0F FIGURES Figure Page 1-1 Core Loading 6 1-2 BOL SNH Data 7 1-3 BOL Burnup Data 8 2-1 Cold Rod Drop Times (Full-Flow) 10 2-2 Hot Rod Drop Times (No-Flow) 11 2-3 Hot Rod Drop Times (Full-Flow) 12 9-1 Control Bank A Worth 24 11-1 BOL HZP Boron Concentrations 29 12-1 Power Distribution Differences Greater than 15% 32 12-2 Power Distribution, HZP, ARO 33 12-3 Power Distribution at Power 34 12-4 Axial Power Distribution, BOL, HZP, ARO 35 12-5 Axial Power Distribution, BOL, NFP 36 15-1 Intermediate Range Detector Response to' Power Level 38 e

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PREFACE This report is intended to document in a concise format the results of

. the physics testing program and unit systems response during the startup of Unit 2 following Refueling 10. The organization of the report follows that utilized previously in startup reports.

The core loading pattern was determined by Westinghouse, the vendor for' the nuclear steam supply system. WCAP 10583, Revision 1 "The Nuclear Design Core Management of the Point Beach Unit 2 Nuclear Reactor Cycle 11,"

tabulates various parameters predicted by computer codes. All references in this report to design values pertain to WCAP 10583. Actual end of Cycle 10 burnup was 13,677 MWD /NTU. The published WCAP parameters were based on actual Cycle 10 EOL burnup. Cycle 10 was ended on September 28, 1984 with ,

a peak assembly burnup of 42,730 MWD /MTU and average assembly burnup of 24,108 MWD /NTU. Electrical power was first generated during Cycle 11 on November 20, 1984.

This report is intended primarily for the use of Wisconsin Electric Power Company personnel as a readily accessible, complete compilation of reduced data.

Copies of this report were submitted to the NRC to comply with Technical Specification 15.6.9.1.A.l.c and 15.6.9.1.A.2. A region of Westinghouse optimized fuel assemblies (OTA's) was loaded for the first time at PBNP in Unit 2 Cycle 11. The fuel design changes for 0FA's were significant enough to be classified as constituting a different fuel design.

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. Section 1.0 REFUELING Section 1.1 Core Unload The core was completely unloaded to facilitate incore thimble changeout and reactor vessel component inspections. The first fuel assembly was unloaded on October 13, 1984 at 1842 hours0.0213 days <br />0.512 hours <br />0.00305 weeks <br />7.00881e-4 months <br />. Usin~g two 10-hour shifts per day, the unload was completed without any changes to the sequence on October 15, 1984, at 1752 hours0.0203 days <br />0.487 hours <br />0.0029 weeks <br />6.66636e-4 months <br />.

All fuel was stored in the north spent fuel pit. Spent fuel receipt was suspended between core unload and core reload with one spent fuel assembly (D14) not put in the spent fuel pit.

There were no insert changes made during core unload.

One fuel assembly (MSS) sustained grid damage when being placed in i storage location SM-27. It was replaced witt. fuel assembly NO2 for the

core reload.

Section 1.2 Insert changes l 1. Eipt RCCA's were replaced because of wear found during visual

inspections performed in 1983.. All control rod transfers were made l without incident.
2. Several depleted burnable poison (BP) assemblies were removed from o:

transferred between reload fuel assemblies with no incident.

j 3. Three new BP assemblies were transferred between new fuel assemblies.

A fourth BP assembly could not be transferred because it repeatedly fell from the tool's gripper mechanism when lifted from a new fuel assembly. One of the new BP assemblies that were successfully transferred was then partially withdrawn by the tool in front of the

! periscope. It was discovered that the BP assembly crossbar was wedged in the gripper mechanism below the latching fingers. Apparently the i three transfers were made with the BP assemblies held in the tool by friction.

The fourth RF assembly was transferred to the new fuel vault so that it could be inspected at a later date. It was replaced with a new BP

. assembly left over from Unit 1.

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. 4. One plug device was damaged when a top nozzle spring clamp with orientation hole broke off and wedged between the plug device tool and top nozzle of fuel assembly K77. Several spare plug devices were available for replacement because of the changeover to optimized fuel requiring new redesigned plug devices. A spare plug device was put in fuel assembly K68 (replacement for K77).

All other plug device changes were made without incident.

Section 1.3 Fuel Assembly Inspections 0FA demonstration assemblies ZD1, ZD2, and ZD4 were inspected by Westinghouse. These assemblies have removable rods and have had 3 cycles

.of burnup. The inspection program included general visual examinations of the fuel assemblies and high magnification visual examination of several individual fuel rods. No abnormalities were found.

Section 1.4 Core Reload Change's were made to the original core loading plan for Cycle 10 because of damage to the following fuel assemblies:

1. M55 - Replaced with NO2 after sustaining grid damage from spent fuel pit storage rack at location SM-27.
2. K77 - Replaced with K68 after the top nozzle spring clamp with orientation hole broke off.

As a result of the above replacements, changes were made to the original core loading sequence as described in Table 1-1.

Numerous changes had to be made to the core loading sequence because assemblies were bowed. This problem is expected to occur during full core l reloads.

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1 TABLE 1-1 CHANGES TO CORE LOADING PLAN Original Final Core Location F/A Insert Q Insert E-3 M55 RCCA M77 RCCA D-4 M77 RCCA NO2 RCCA G-1 ZD2 ZPD K68 PD G-13 K77 8P50 ZD2 ZPD D-3 N81 2P105Z N81 PDZ 6

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[ ,. Section 1.5 Core Design l

1. Optimized Fuel Assemblies l

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A region of 32 new optimized fuel assemblies were used for the first time at PBNP in Cycle 11. Their distribution in core is typical of the low leakage concept in which new fuel assemblies are loaded between the center area and extreme periphery of the core.

Three demonstration optimized fuel assemblies with removable fuel rods were loaded for a fourth cycle of operation at peripheral locations A-7, G-13, and M-7. These assemblies were found to be in good condition when inspected prior to core load.

The optimized fuel assembly employs a slightly reduced fuel rod clad OD (0.400 inch) compared to the standard fuel rod clad OD (0.422 inch) while retaining the same fuel rod pitch. This increases the water to uranium ratio which improves rieutron moderation and, efficiency eventually lowering fuel cycle costs. The fuel pellets are enriched j to 3.4% in U-235.

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Another feature of the optimized fuel assembly design -is the use of l zircaloy spacer grids for all but the top and bottom spacer grids.
The top and bottom spacer grids are Inconel, the same material used in standard fuel assembly spacer grids.

l Slight reductions in the guide thimble and instrument thimble diameters I

were also made. Standard control rods and burnable . poison rods are l compatible with optimized fuel assemblies. Standard plug devices,

all having thicker plugging rods are not compatible however, and new plug devices were provided for use in optimized fuel assemblies.

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2. Inserts i

New burnable poison assemblies were provided in Cycle 11 to control..

radial power distribution. Seven 2P and eight 14P asymmetric burnable

- poison assemblies were loaded in optimized fuel-assemblies. Four 4P burnable poison assemblies were loaded in once-burned . fuel near the core's center.

Eight control rods -were replaced in a continuing -program -leading eventually to total' replacement. A total of 10 original control rods i have been replaced since the program started in 1983.

I The two secondary sources were returned to their normal locations at G-2 and G-12.

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. 3. Fuel Loading Table 1-2 lists the uranium weight by region. Figure 1-1 shows the final core load pattern, Figure 1-2 BOL SNM data, and Figure 1-3 BOL burnup data.

TABLE 1-2 URANIUM LOADING Number of U Weight (NTU) Current

-Region Assemblies Original Current Enrichment (%U235) i .

0.74 9A 1 0.40 0.38 10 5 2.00 1.92 1.10 10A 3 1.06 1.01 0.75 11 23 11.24 10.86 1.23 13A. 4 1.60 1.55 1.36 13B - 1 0.40- 0.39 1.65 12 7 2.81 2.76 2.05 12A 40 16.13 15.80 1.88 13* 32 11.40 11.40 3.40 TOTAL 47.04 46.07 -2.02

  • New Assemblies i

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FIGURE l-1 CORE LOADING PSNP S M DATA - UNIT 2 CTCLI 11 - START OF CTCLE 48 0F 10/24/84 11/14/04, t 2 3 4 5 6 7 8 f IG 11 12 13

. Lee Z31 Le2 4 LM W O Lae3PK LM WL P3 SPS PS L32 N53 No M45 H3 N73 L77 9 LacMT LME35 LHE3M L309VI Ln0E2X LMt38 LaoMC PI GPS R 44 PS R I4 0P8 PS L72 M74 22 Ne7 L&3 L38 NSS MS2 L35 C LacW 8 Lace 37 Lae m9 Lac 9T3 LacM N LNe w T LaeM 6 La0E3T LA N FU PS 2P 107 2 34 PS 2 114 PS R 65 2P tee P3 Lee NI N2 N79 255 E39 Me M7 NG2 49 L&7 3 Lac W R Lace 30 La097J Lae M E Lae U S LM 3 m Lace 31 Laegue Lassu1 Lao c3 LatW M PS OPS E 74 PS I W102 Pt 1W1H P3 R 74 2P 102 PS MS N77 Ne5 M4 N2 Ne8 M1 Mot NOT M57 No E LaeE32 Lae9N Lae858 LMSUF La0004 LactuN Laogus LaoguZ Lac 8UN LaemF LaeE34 OPS A B3 P9 R 37 PS PS P9 R 68 P9 8 72 GPS L39 M59 L34 N62 M7 N76 Net NG A43 at Ne8 N72 Lal F LaeWS LaeE3L La6WI LaeF3M LMCM LacNe'LaetuFlaeW7 LM84 LaeE3M Lae97C LaeE35 LaeWJ PS 2 55 Pt 1#105 PS t M W 102 2 69 PS I W194 PS R 107 PS us mi u3 Ret Nee m m we m ut vi m . zt2 l 5 LM3PE Lae8vE LaoM T L303Pe L3039 LatSv3 LM757 LMput LM957 Lae3P9 LacePM LactuN L303PL l P9 SPS 32 113 PS P9 W IM I 77 4F 101 PS PS I 112 SPS 4 SPS j L51 NSB NS NEF Me MT 594 NO3 N94 w9 L33 M77 L70 N tm.m taas tmm u.aP Laws Laws LassE Lauw Lums Lama tanz tau 3C tun, PC R 61 P9 14P'03 P9 R 79 W 103 R 43 PS IW1e7 P9 R 110 PS 556 N06 N74 Ref N75 nee W3 N04 agt NE2 M71 I Last2Y LacME Las3E LacN4 LN485J La0054 Laogur LacIv2 La0IUT LacAM LacGF OPS I 75 P9 R 42 PS PS P9 R 81 P9 R M GPS l

us m w3 m2 ur m- m et m me v4 '

J LatWS LaeE34 LaoguB LacMJ LaeE27 L303PC Lace 39 L30N1 La005C LNGE3a LaeWE PS 2P le3 3 73 PS 1 #1e4 PS IWitt P9 R 92 2P 1e4 P9 u, Nee me u7 N. . .4 u.

I Lacert LaeE3s Lasest ueWR Lasaru Laetit Laseet Laena LasaFP l Pt 2P tot t 34 P9 R 114 PS I 34 2P IN PS H

us mi es w2 me et us L* LateF4 LaeE3J Lac G 4 La00 H L30E30 LaeE3B LasAP7 P9 OPS 3 111 Pg R 109 99 PS L34 234 L74 ,

N LaodPU LNOWW LasAF9 P9 SP9 PS l

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FIGURE l-2

> BOL SNM DATA

  • START QF CYCLE A3 0F 10/26/84 11/14/44 POMP SNM DA7a ! UN!? 2 CYCLE 11 6 7 8 9 to 11 12 13 1 2 3 4 5 L60 101 L64 A 4S44 2509 4S17 2434 2020 2444 LS2 NS3 N60 M6S N63 Nf3 L77 9327 18430 12139 4524 5 4574 12216- 13031 0 2444 2447 0 0 1414 0 MS2 N07 L63 (S4 MS4 N52 LSS L72 N74 i

C 4947 12034- 4044- S361 4446 SISS 0126 12066 4941 0 1744 3333 2345 43S1 1724 0 3375 2371 '

M79 NSS MS9 N64 M9F M42 N64 L67 L64 N41 NO2 4547 6446 7390 13040 4543 13041 7344 7326 11994 l

.0 4600 }4116 3493 0 1959 1941 0 4434 i 3442 0 2019 1949 0 M44 M41 M47 MS7 N64 N65 M77 MSS M64 #93 M91 7402 7335 6737 6690 7337 6743 6713 7354 4070 12055 E 11943 3043 1951 1730 0 0 1923 1954 3074 3095 1934 3046-M67 476 M49 M64 M63 NS1 NOS Nf3 L61 LS9 NS9 LS4 N64 5314 13039 4605 4SS4 12034 S185 12034 6445 6754 8947' 6779 6445 120a4 F

3073 2047 0 2440 0 3437 t

t 3439 0 3353 0 3097- 3070 154%

J76 M49 M90 N61 L71 M95 202 K64 M41 L73 nSt M40 M70 3554 7419 4/34 3414 4443 - 7341 4S09 4444 9144 4 3131 93g7 4474 4546 J 3014 1934 1554 3674 1943 ,1940 3503 3391 1441 4644 g'414 3396 3493 M64 rag M94 M43 M94 N79 LS3 N77 L70 f LS1 NS4 NOS NS7 4410 M 4S43 12004 SaS3 13044 4733. 6703 4934 4799 6447 13031 5154 18134 atte, 4041 1531 4040 3047 0 3343 4 4440 3441 0- 3%44 0 M74 MS9 M75 M60 M73 M44 MS1 Nea M71 M56 Met 4039 13003 l

I 13045 4125- 7313 6734 6673 7334 6712 6449 7331 0 1789 1970 3071 3108 1940 4090 3044 1999 1743 0 N.4 N67 R70 N76- M71 M74 NS6 ~ L74 L75 N70 M93 4440 J 4543 )2029 7344 7365 13049 449E 13040 7399 ~7441 12036 1997' 8 4544- 0 1994 1945 0 4441 3444 0 1943 649 NGO MS4 L64 L57 Net MS3 NS4 L69 R 5044- 12049 4014 $194 S083 534S 4114 12044 5040 .

0 1754- 1337 4394- 3449 1733 0 3343 l 2376 L74 . N75 M7 N L.4

4. 1,.,1..

1 ii9 7 93, i. 03. 1.707

-0 49 7 3

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4 0 4 16 0 n':3 M

3459 .it!0 Str t.it.

47 CON?tNTS OF gACM C948 LOCATION FUEL IDENTIFICATION 8 CURREN7 U.23S Sw&M4 CURRENT FISSILE Py GRAMS l 7 L

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1 FICUP2 1-3

' BO! BURNUP DATA 4

PSNP UNIT 2 CTCLE 11 - START OF CTCLE EUNMUP D4TA - 11/14/84

{

J 2 3 4 5 6 7 8 9 to 11 12 13 l 1 4

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Lee Z31 L&2

': 4 15452 14770 15433 j- 26248 34823 24545 L52 N53 N64 N3 H3 N73 L77

0 15554
3 15444 0 0 8895 0 24244 0 0 8895 0 0 26499 j

i L72 N74 N52 M7 L43 L58 N58 N52 L55 l

l C 14915 0 12274 15849 16140 14382 12131 0 14491 .

24493 0 12274 22435 24729 23044 12131 0 24184 l

LM Mt M2 N79 N55 K59 NS. N97 NO2 M9 Le7 3 15473 0 8948 14937 0 4374 0 14954 144M 0 15493 24I22 0 17592 14937 0 20144 0 14954 144M 0 24344 NS N77 N05 N64 N2 NOS N91 M1 N87 N57 He I, 0 14544 14937 14944 17254 14764 16984 17034 15045 12094 0 0

0 14544 14937 16944 17254 14744 14944 17034 15445 12094 L59 185 9 L54 H2 N7 N74 M9 M8 Ne3 M1 N00 N72 Let F 15424 0 14415 0 17233 14877 9787 14833 17242 0 15774 0 15444 24341 0 232M e 17233 14477 9787 16833 17242 0 22522 0 24078 Kee El L73 K31 NM N70 - J74 Me N99 Ret L71 M5 Z32 9 '14847 .9004 15983 4419 14597 1 M26 . 3499 1M14 14774 e535 15917 8943 14201 34943 3004 24711 28243 14597 1M24 39134 1M14 14774 23354 24849 8943 34428 N50 <- NOS NET No Mf ; N94 N83 M94 N79 L53 187 7 L70 j L51 N i335: e i5u3 e in?5 im4 fase iM n i7143 e tan 4 is7u j 24309 , 0 22745 6 17075 17224 9004 164N 17143 4 23314 4 24145 N74 N59 NFS Neo N73 NS4 M1 - N02 l L M71 454 Nee '

0 1 0 12171 15232 17949 172M 14001 17139 17421 15124 12414 0

- 0 12171 15232 17969 172M 14001 17139 17421 15124 12414 1 L75 579 W3 H2 N7 E70 274 N71 N79 N64 L74 J. 15738 0 14844 14918 0 4352 0- 15137 14474 0 13495 24117 4 14044 14918 0 28435 0 15137 14474 0 26475 Lef NGO 884 L44 - L57 Nee NO3 N54 -L&S

~T' 14884 4~ 12513' r4555 155!T ~15811 ' !!194 ~ 0 14844 s 23802 0 12513 231N 24168 22905 12194 0 23871 L7S El NF5 572 M4 N78 L&4 L 15863 0 0 ONS e 0 15689 24387 0 0 ONE 0 0 242M L34 ZB4 L74 M

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~,.y N 184E2 ' tee 84 15474 24347 34414 24204 g:; {

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CONTDTS OF E4CN CONE LOCATION l +

FML 19tNTIFICATION I CTCLE ASSO ET IUNNUP TOTAL ASSENELT DUNNUP P9MP UNIT 2 CTCLE 11 -ITANTOFChCLEBUENUPBATA- 11/14/84 9

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Section 2.0 CONTROL ROD OPERATIONAL TESTING Cold control rod testing was conducted on November 15, 1984, just prior to initial cycle heatup.

Hot control rod testing was conducted shortly after primary system heatup on November 17, 1984.

Section 2.1 Rod Drop Times 1 Rod drop times to dashpot in the cold full-flow condition ranged from 1.27 seconds to 1.50 seconds with several rod drop times near each end of the range.

I Rod drop timas to dashpot in the hot zero flow condition ranged from 1.08 seconds to 1.18 seconds with several rod drop times near each end of the range. .

L Rod drop times to dashpot in the hot full-flow condition ranged from 1.23 seconds to 1.36 seconds with several rod drop times near each end of the range.

See Figures 2-1 through 2-3 for rod drop times and core parameters.

Locations containing optimized fuel assemblies are marked because the narrower thimble tubes increase rod drop times slightly in the dashpot area. Locations with new control rods are also shown.

All rod drop times to dashpot were well within the Technical Specifi-cation limit of 2.2 seconds (15.3.10,5).

Section 2.2 Control Rod Mechanism Timinq control-rod mechanism timing was. conducted in cold plant conditions on November 15, 1984. -The visicorder traces of the ' lift, ' movable and stationary gripper coil voltages. of each rod mechanism were - reviewed by

-plant personnel. No rod misstepping occurred.

Section 2.3 Rod Position Calibration During hot rod testing, LVDT voltages ' were read at 20 steps and 200

. steps to determine if any voltages were abnormal.

"Zeroa adjustments 1 were made with rods at 20 steps under hot zero

. power full. flow conditions.

" Span" adjustments were made at full power l after- rods were' verified to be at 228 steps using WMTP 9.19.

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FIGURE 2-1 COLD ROD DROP TIMES (FULL FLOW) 1 2 3 4 5 6 7 8 9 10 11 12 13 27e*

A 1.33* 1.41*

O 1.99 2.12 1.35 1.48 1.39 C 1.91 2.09 1.94 N #

/ ~

4 1.34 1.40 D 1.88 1.93 1.36 1.36 1.36 1.35 E 1.93 1.89 1.89 1.86 1.42* 1.27 1.33 1.49*

p 2.12 1.81 1.86 2.21 1.40 1.30 1.40 G 1.99 1.85 2.01 **

1.39* 1.32 1.27 1.45*

H 2.04 1.85- 1.81 2.22 .

1.36 1.40 1.33 1.33

! 1.86 1.93 1.85 1.87 1.38 1.34 3 1.91 1.91 1.41 1.45 1.32 K 1.96 2.04 1.88 1.50* 1.40*

L 2.20 2.11 M 2 UmIT so-

, Optimized Fuel Assembly DATE 11-15-84 TDE TO DASHPOT (SE)

> TDtB TO BOT 1DM (SEC) TEMP. 295 'F New Control-Rod FIDW 100 g

- PRESSURE 340 psia POINT BEACH NUCLEAR PLANT CONTROL ROD TESTING ROD DROP TIMES RE-D6 10 (10-78)

FIGURE 2-2 HOT ROD DROP TIMES (NO FIrW) 1 2 3 4 5 6 7 8 9 10 11 12 13 270

  • A 1.17* 1.17*

O 1.68 1.70 1.11 1.14 1.13 C 1.s4 1.s9 1.57

\ 1.10 1.12 #

D 1.s3 1.s3 1.10 1.10 1.09 1.11 E 1.s1 1.50 1.s0 1.s4 1.18* 1.10 1.11 1.15*

F 1.71 1.52 1.s2 1.69 1.12 1.12 1.13 G *~

1.58 1.s4 1.s8 1.13* 1.12 1.12 1.17*

H 1.67 1.53 1.54 1.71 1.12 1.09 1.08 1.12 *

-l 1.53 1.50 1.48 1.54 1.13 1.12 J 1.52 1.s3 1.13 1.13 1.11 K 1.s6 1.s7 1.56 1.16* 1.14*

l 1.69 1.68 l .

M 2 UNIT se*

,7 Optimized Fuel Assembly DATE 11-17-84 TIME 10 DASEPOT (SEC)

TIME 10 BOT 1DM (SEC) TEMP. 4s20 '

  • F-New Control Rod-FIDW O t PRESSURE %1990 psia POINT-BEACH NUCLEAR PLANT CONTROL ROD TESTING-ROD DROP TIMES RE-D6 i (10-78) 11 .

FIGURE 2-3 hor ROD DROP TIMES (FULL FLOW) 1 2 3 4 5 6 7 8 9 10 11 12 13 270*

A 1.29$ 1.32*

O 1.85 1.93 1.23 1.34 1.32 C 1.72 1.84 1.82 N #

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1.26 1.29 b 1.73 1.74 1.28 1.25 1.26 1.27 E 1.76 1.71 1.72 1.75 1.32* 1.28 1.28 1.36*

F 1.93 1.75 1.74 1.98 1.28 1.32 1.30 G 1,79 1.80 e-1.79 ,

1.26* 1.30 1.29 1.33*

H 1.85 1.78 1.77 1.97 1.29 1.20 1.27 1.25 l 1.75 1.74 1.72 1.75 1.30 1.26 3 1.76 1.73 1.31 1.32 1.24 K 1.81 1.81 1.74 1.36' 1.30*

L 1.97 1.90 M

UNIT 2 se*

Optimized Fuel Assembly DATE 11-17-84 TIls TO DASHPOT (SEC)

  • : TIME TO BOT 1DM (SEC) TEMP. N 535 -

- PRESSURg %2000 psia-POINT BEACH NUCLEAR PLANT CONTROL RCD TESTING ROD DROP TIMES RE-D6 2 (10-78)

. Section 3.0 THERMOCOUPLE AND RTD CALIBRATION During initial cycle heatup, thermocouple and loop RTD signals were l recorded at different temperature levels under partial and full-flow condi-tions. See Table 3-1 for the results for full flow conditions. The RTD resistance readings were obtained at the protection racks in the control l room using a digital multimeter that subtracted lead resistance. l Thermocouple temperatures were read at the toggle readout panel. l Since the core was producing very little heat, the hot and cold leg RTD's were at about the same temperature. Thus both hot leg and cold leg readings were averaged into one temperature for the RTD's. The RTD resistances were converted to degrees Fahrenheit by using the vendor's calibration curves.

Due to the use of optimized fuel assemblies, the Improved Thermal Design Procedure (ITDP) was implemented. The ITDP has a requirement of l 0.9% of span accuracy for the bypass manifold RTD's. To obtain the required )

accuracy, the existing Sostman RTD's were removed from the bypass manifolds for recalibration at PBNP. Because of poor calibration results, the Sostman RTD's were replaced with four Rosemount Model 176 and eight Model 189 RTD's.

It was found, however, that the yellow channel hot and cold leg RTD's

  • being used were still not accurate enough as indicated in Table 3-1 -and readings during initial power escalation. The spare RTD's (407A and 407B) were wired in place of 404A and 404B to obtain the required accuracy.

The T/C readout panel indicated that thermocouples at I-10, K-3, L-7, .

E-4, I-4, and M-6 were not functioning properly.

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TABLE 3-1 RTD CALIBRATION CHECK 7

RTD Elements RTD Temperatures from Measured Resistances ('F)

Loop A - Cold Leg R 401B 413.11 466.76 475.10 515.39 539.25 R 405B 413.48 466.67 475.42 515.56 539.05 W 402B 413.56 467.11 475.61 515.90 539.77 Loop A - Hot Leg R 401A 412.81 466.02 474.63 514.80 538.38 R 405A 413.98 467.09 475.75 515.84 539.30 W 402A 413.37 466.88 475.27 515.50 538.96

' Loop B - Cold Leg 4 B 403B 414.23 467.77 476.30 516.62 539.99 B 407B 414.02 467.13 475.70 515.77 538.94 Y 404B 410.55 463.53 472.10 512.27 534.83 Loop B - Hot Leg B 403A 413.27 466.66 475.07 515.26 538.46 B 407A 414.43 468.66 476.22 516.62 539.15 Y 404A 414.70 468.56 476.88 517.42~ 540.57 l

RTD Average 413 467 475 516 539

. T/C Average 421 479 --

.524 545 l

Saturation Temp. 414 '455 --

520 535-s

'o 1

  • i i-l 14 t .-

Section 4.0 PRESSURIZIR TESTS Section 4.1 Heater Capacity Pressurizer heater capacity was calculated using volt and ampere readings for each group of heaters. Table 4-1 shows that heater capacity is above Technical Specification requirements of 100 KW minimum total.

TABLE 4-1 i

HEATER GROUP ENERGY INPUT i

Group I-Current V-Voltage KW-Energy Input f

(amps) (volts) KW = 6 x V x I/1000 A 271 480 225 l B 227 480 189

[ C 226 480 188 D '

209 480 174 j

E 225 480 187 TOTAL 963 l .

Section 4.2 Spray Valve Effectiveness Spray valve effectiveness is determined by measuring how fast each-spray valve decreases pressurizer pressure when fully opened with the other valve closed and heaters off. For the test, spray valve "A"' decreased l_ pressure at the rate of 116 psi / min. Spray valve "B" decreased pressure at i -the rate of 113 psi / min. These are typical values and indicate _ that mass / flow through each valve is greater than design. It can be shown that

.given normal heat balance characteristics of the pressurizer, 200 gun design spray flow decreases pressure by about 70 psi / min well below - the results achieved above.

~

Section 4.3 Heater Effectiveness-l

! Heater. effectiveness is determined by measuring how fast pressuriser pressure increases with all _ heaters on and spray flow only through :the

~

-bypass valves. . For - the_ test, pressurizer pressure increased at an average.

rate of 15.6 psi / min between 1840 and 2150 psia _using all heaters. This.is well above design heater capacity of 14.0 psi / min.

1 .

m 15

s

. Section 5.0 REACTOR COOLANT SYSTEM Section 5.1 RTD Manifold Flow After the initial cycle heatup, the reactor coolant bypass flow through the RTD manifold was checked and found to be adequate for both loops. The i flows were 215 gpm through Loop "A" and 190 gpm through Loop "B".

I Section 5.2 Flow Transient Times Table 5-1 gives the times to reach certain percentages of full-flow i from the time a reactor coolant pump is tripped or started. The times are I consistent with those obtained in previous measurements.

i b

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e 16 1

TABLE 5-1 s

l REACTOR COOLANT FLOW TRANSIENT TIMES t

Flow Flow Through Time to Reach Time to Reach Through Inactive

! 90% Flow 50% Flow Active Loop Loop i Condition (Sec.) (Sec.) (%) (%)

! A Tripped 2.1 14.0' ----

0 B Tripped 2.1 14.5 ---- 0 E

A Not Running ---- ----

-13.8 B Started 17.5* ---- --- 51)

>108.3 ----

1 A Started 18.0*- ----

'100 ----

B Running ---- ----

100 -----

A' Running ---- ----

>107.9 II) ----

.B Tripped 1.8 11.12 .---- '- 17.3 l A Running ---- . ----

100 ----

19.6*

B Started ----

100 ' ----

l-A Tripped 1.8 10.6 ----

-14.3 l '

B Running ---- ----

107.7 ----

  • Time to reach 100% flow.

(1)' signal was off-scale high. Values given are for the highest scale reading.

'a S:

4

h I

. Section 6.0 CONTROL SYSTEMS There were no difficulties encountered during heatup or testing in the control syst. ems of pressurizer level, pressurizer pressure, or the rod control system.

?

Section 7.0 TRANSIENTS There were no significant transients during the startup or approach to full power. There were no violations of the fuel conditioning restrictions on power and rod stepping change rates.

Section 8.0 INITIAL CRITICALITY AND REACTIVITY COMPUTER CHECKS Section 8.1 Initial Criticality The approach to criticality was made in two phases. The first step, which began at 2020 hours0.0234 days <br />0.561 hours <br />0.00334 weeks <br />7.6861e-4 months <br /> on November 17, 1984, was the normal withdrawal of control rods until Bank D reached 180 steps at 2111 hours0.0244 days <br />0.586 hours <br />0.00349 weeks <br />8.032355e-4 months <br />. Then the reactor coolant boron concentration was decreased by dilution until criticality was

, achieved. The dilution began at 2114 hours0.0245 days <br />0.587 hours <br />0.0035 weeks <br />8.04377e-4 months <br />. The initial boron

concentration was 1978 ppe. 10,400 gallons of water were used to reduce boron concentration by 594 ppe until criticality was achieved.

ICRR plots were maintained during each phase of the approach to criticality.

!' The reactor conditions at the time of criticality were determined to be as follows:

Date November 18, 1984 Time 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> E RCS Temperature 530*F RCS Pressure 1985 psig Rod Position Bank D at 180 steps Boron Concentration 1384 ppe i

18

. Section 8.2 Reactivity Computer Checks

1. Following criticality, acceptable zero power physics testing flux levels were determined. The flux level at which nuclear heat appeared was at 5 x 10~8 amps on N-35, 6 x 106 on N-36 and 3 x 10 s amps on the Keithley picoammeter. Normal flux levels for physics testing are one-third of these values.
2. A check of the reactivity computer was made by comparing the computer's calculated reactivity for a certain doubling time versus the reactivity obtained from Figure A.1 of the WCAP.

Reactor coolant system temperature was near 535*F. Table 8-1 shows the results of this check.

4 1

9 9

4 5

19

TABLE 8-1 REACTIVITY COMPUTER' OECKOUT Bank D Steps Measured Measured Calculated Doubling Reactivity Reactivity From T J Time (Sec.) (pca) (pca) .

173 185 75.89 48 48 172 187 57.04 59 60 172 194 40.33 78 78 t

9 e

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I

, Section 9.0 CONTROL ROD WORTH MEASUREMENT

,- l 1

l, Section 9.1 Test Description l The rod worth verification utilizing rod exchange (" rod swap") was divided into two parts. In the first part, the reactivity worth of the i reference bank was obtained from reactivity computer measurements and boron endpoint' data during RCS boron dilution. In the second part, the critical height of the reference bank was measured after exchange with each remaining .

bank. l i  !

} In the rod exchange technique, the reference bank is defined as that

bank which has the highest worth of all banks, control or shutdown, when l inserted into the core alone. For Cycle 11 the reference bank was Control i Bank A (CA) as was the case in all prior rod swap tests. (

Using the analog reactivity computer, reactivity measurements were made i during the insertion of control Bank A from the fully-withdrawn to the i I felf.y-inserted position. The average current (flux level) during the 7

meauurement was approximately 5 x 10 amps'. Critical boron concentration measurements (boron endpoints) were made before and after the insertion of i control Bank A (see section 11.0). Figure 9-1 shows the results of the j differential worth measurements. .

Starting at a critical position with the reference bank fully inserted and control Bank C at 212 steps, 2 new critical configuration at constant R(d boron concentration was established with control Bank c fully inserted ar.d control Bank A at 141 steps. Control Bank c was then withdrawn and *

! control Bank A inserted to one step to establish the initial conditions for

64 next exchange. This sequence was repeated until a critical position was
established for the reference bank with each of the other banks individually inserted. Criticality determinations before and after each exchange' were j ,

made with the reactivity conputer.

The sequence of events during the rod enchange and a summary of the rod e 4 data is presented in Table 9-1.

e e

.9 r 21

.. Section 9.2 Data Analysis and Test Results The integral reactivity worth of the measured bank is inferred from the swapped portion of Control Bank A by the following equation:

Wf = - apt - (a ) (OE ) + where:

x 2 X Wf=TheinferredworthofBankX,pcm E = The measured worth of the reference bank, control A, from R

fully withdrawn to fully inserted with no other bank in the core.

aX= A design correction factor taking into account the fact that the presence of another control rod bank is affecting the worth of the reference bank.

Ap2 = The measured worth of the reference bank from the elevation at which the reactor is just critical with Bank X in the core to the reference bank fully withdrawn condition. This worth was measured with no other bank in the core.

Apg = ne measured worth of the reference bank fmn the hlly berted condition to the elevation at which the reactor was just critical prior to the worth measurement of Bank X. In this test ap is sero.

t I

W X

= The worth of Bank X from the initial position (before the start of the eschange) to 228 steps. This worth is measured by the normal en4oint worth method.

Final values for the integral worth of control and shutdown banks inferred from the measurement data are tabulated in' Table 9-2. Values for a, were obtained from the design predictions . are also listed in Table 9-2.'

section 9.3 Evaluation of Test Results , ,

& comparison of the measured / inferred bank worths with desiqpi predic-tions is presented in Table 9-2.

. ~In evaluating the test results, the standard review and acceptance -

-' . criteria were used. . . .

r 9

22-i

Review Criteria

a. The measured worth of the reference bank agrees with design predictions within 110%.
b. The inferred individual worth of each remaining bank agrees with design predictions within 115% or i100 pcm whichever is greater.
c. The sum of the measured and inferred worths of all control and shutdown banks is less than 1.1 times the predicted sum.

Acceptance Criteria

a. The sum of the measured / inferred worths of all control and shut-down banks is greater than 0.9 times the predicted sum.

As shown on Table 9-2, all review and acceptance criteria were met.

8-

'O e

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e 23

FIGURE 9-1 CONTROL BANK A WORTH PENP UNIT 2 CYCLE 11 BOL HZP O - Measured Data All Other Rods Fully Uithdrawn Solid Lir.e - Design 15 - _ _ . . _ _ _ _ _ ,

I I E 14 -

'N$

_: 1 1.;

"1 .

7; 'l 13 -  : - ;- =r
-.n

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= -t

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=- . ,r - . .

11 - . _. .

. f%

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w 10 - .: T.
:_,=.

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m = v; s 9- .. - --_

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-~,

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- 7-

.=_ ..

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'L

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= . _

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1

. -wa=, . _-

1-t

.-1-----

_.___.r. . _.

$.2--i--- -: -

C..

. _-3 0- e==__.3 ---

- - - - + - - ' -- 2 F- -

0 20 40 60 80 100 120 140 160 180 200 220 Steps Withdrawn *

~24

TABLE 9-1 CRITICAL ROD CONFIGURATION DATA 11-19-84 Measured RCS CA Bank Bank Tavg Position Position

, Measured Time (*F) (Steps) (Steps)

CC 1721 530 1 212 CC 1732 530 141 1 CC 1744 531 1 214 SB 1747 531 1 217 SB 1757 531 131- 1 .

SB 1808 531 1 217 .

1 SA 1819 531 1 213 l

SA 1827 531 127 1 SA 1839 531 1 214 CD 1849 531 1 213 CD 1856 531 83 1 CD 1905 531 1 220 CB 1906 531 1 218-CB 1912 530 101 1 CB 1920 530 1 220 Boron concentration was 1190 ppe.

6-

-e 4

9 i

qq-

TABLE 9-2 COMPARISON OF INFERRED / MEASURED BANK WORTHS WITH DESIGN PREDICTIONS op2 EX W I

X

/X (

I~'

P

) x 100 Bank X M h M M M _

(%)

CC 530 0.956 37 1125 1149 - 2.1 SB 602 1.009 23 1011 996 + 1.5 SA 631 0.953 23 1017 993 + 2.4 CD 1014 0.991 24 . 607 580 + 5.9 CB 837 1.077 27 714 690 + 3.7 CA ---- ---- ---- 1595 1650 - 3.3 TOTAL 6076 6058 + 0.3 O

e G

26 r

4 1

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i l

Section 10.0 TEMPERATURE COEFFICIENT MEASUREMENTS  !

I Isothermal temperature coefficient measurements were taken during zero i- power physics testing. The measurement test conditions and results are i given in Table 10-1. The measured values are the average of the recorded 1 reactor coolant system heatups and cooldowns. Reactivity from the reactivity computer and reactor coolant system temperature were recorded on I

an X-Y plotter and two-pen recorder.

The measured temperature coefficients are within the review criteria of 13 pcm/*F.

i TABLE 10-1 l ISOTHERMAL TEMPERATURE COEFFICIENTS i

Control Boron Avg.

Bank conc., Temp. Measured Design

  • Difference t

configuration _gg_, 'F pcm/*F pcm/*F pcm/*F (M-D)

ARO 1351 534 -3.1 -3.1 0.0 i A in 1189 530 -7.3 -6.3 -1.0

  • WCAP Figure 5.1 and Figure 5.8 i

4

@ g .

'9 1 D.

27 -

.- p

. . - - _ _ = _ _ _ _ _ _ _ _ - . _ _ _ . . . - - - _ _ _ _ _ _ _ - - . . . _ _ - __ - -. _ .- - .-__

1

+

1 Section 11.0 BORON WORTH AND ENDPOINT MEASUREMENTS Figure 11-1 shows RCS boron concentration during zero power physics testing. Table 11-1 shows results of the endpoint measurements. Design j 4 values are for 530'F testing temperature. The measured boron worth was i obtained by dividing bank worth (pcm) into change in boron concentration between the endpoints.

Review criterion was not met (10.5 pcm/ ppm). This is a typical problem with boron endpoint measurements where measured boron endpoints are not close to design.

TABLE 11-1 BORON WORTH AND ENDPOINTS Endpoint Bank Worth Boron Worth Bank Design I1) Measured Design Measured Design I) Measured configuration (ope) (pon) (oca) (oca) (ocm/pos) (ocm/ ppm)

ARO 1373 1355 --- ---

-9.19 ---

CA in 1189 1192 1650 1595 -9.15 -9.8 (1) Figure 5.1 Table A.2 -

(2) Figure 2 - Supplement to WCAP, Letter 84WE-G-080 Table A.2 6

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29 i

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

! .' Section 12.0 POWER DISTRIBUTION Table 12-1 illustrates the lowering of maximum hot channel factors during initial power increase to full load. More flux maps were required because allowed power levels based on maximum hot channel factors were less than 100% for the HZP flux map. Allowed power levels were calculated using the relationships for FoH and FQ versus power level in Technical Specification 15.3.10.B.1.a. The relationships have been changed due to the use of optimized fuel assemblies in Cycle 11. Zero power flux map results typically do not show that full power operation is permitted due to hot channel factor limitations.

i Measured power sharing factors (FoH) for each fuel assembly were compared to predicted values. Differences of more than 5% were listed in Figure 12-1 for the ARO HZP flux map and for a full power map (No. 9) taken after a mon,th of operation.

Figures 12-2 and 12-3 show the actual power sharing factors at each ,

location for the same flux maps.

Measured axial power distribution compared t,o design is shown in ,

Figures 12-4 and 12-5 for the same flux maps.

e O

4

'e 6

i-l i , .

l

? .

(

i TABLE 12-1 INITIAL POWER ESCALATION rLUX MAP RESULTS Flux Map Power Thimbles Allowed Number Date M Missing reos Power (%)

r_g!!

1 11-18-84 0 5 100 76 2 11-20-84 20 1 112 104 3 11-26-84 50 1 117 113

  • 4 11-27-84 50 2 114 109
  • 5 11-27-84 50 1 116 107 6 12-03-84 100 1 116 118
  • 7 12-04-84 100 1 116 115
  • 8 12-04-84 -

100 1 ,

116 115 9 12-19-84 100 1 117 119

  • Q40 flux naps taken when delta flux or control rods were not near their normal operating positions.

O e

a 0

31

FIGURE 12-1 INIT 2 CYCLE 11 STARTUP

[g DIFFERENCES GREATER THAN 25%

1 2 3 4 5 6 7 8 9 10 11 12 13 2ro*

79 14 3 A 5.1 8.5 7.1 8.9 8.8 3.9 g

\

8.5 6.1 l

\

%% ./

/

0 '

I

-*- -*- " 53 E

F G a

-5.8 -7.3

-0.2 ~4.2 -0.2 J

s.3 s.2 -lo.7 h

K

-18.4 -18.4 L .

-9.o M 11.'9 11.9 5.8 se*

  • -- Flus Map 61 Difference (t) - 557
  • - - Flus Map #9 Difference (t) - Full Power 4 Diff. = #2 POINT BEACE NUCLEAR PLAarf

~

FIGURE 12-2 POWER DISTRIB17 TION, HZP, ARO PDFM 211-I4 HZP ARG .I1/10/04 I 2 3 4 3 4 7 0  ? to 11 12 13 0.309 0.204 0.312 4 0.313 0.347 0.337 1.2 7.? 14.3 0.400 0.f34 1.099 0.f49 1.111 0.f71 0.434 1 3 0.443 0.f47 1.112 1.016 1.719 1.034 0.433 l 0.3 1.2 1.2 7.1 0.9 0.0 3.9 i

0.470 1.003 1.132 0.ff6 0.093 0.790 1.214 1.183 0.319

. C 4.310 1.122 1.131 0.ff0 0.914 1.043 1.290 1.229 0.327

  • l 0.3 3.4 -0.2 0.1 2.4 4.4 4.1 3.f 1.3 0.403 1.119 1.113 1.101 f.134 0.950 1.134 1.276 1.313 1.213 0.434 3 0.403 1.123 1.131 1.103 1.142 0.704 1.204 1.339 f.338 1.230 0.463

-0.1 0.3 3.4 0.3 0.3 2.9 4.3 3.0 3.3 2.0 1.3 0.MS 1.113 1.143 1.199 1.219 1.244 1.231 1.294 1.311 1.204 1.034 E 4.030 1.021 1.150 1.199 1.222 .1.2M 1.307 f.343 1.346 1.317 1.043

-0.3 -0.3 -1.3 -0.0 0.3 3.7 4.3 3.3 2.4 2.4 2.4 0.307 1.074 4.753 1.t97 I.I99 1.213 1.203 I.234 1.2N I.222 I.004 1.200 0.340 F" W2N7se, v. m i. sue i.4 4 i.44e i.4ei i.4 = i.4ma i.4 4 i.ias i.4a4 s.353

-3.7 -3.7 -3.4 -2.2 04 1.7 4.0 -0.0 -0.3 -l.4 3.1 3.7 4.1 0.313 0.fte 0.094 0.954 1.234 1.201 0.003 1.233 1.293 1.014 4.743 1.029 0.314 0 0.317 0.724 0.042. 0.913 f.231 1.219 0.M4 J.208 1.244 0.949 0.954 1.0M 0.327 0.7 -3.3 -3.4 -4.1 -0.2 f.3 2.4 -2.4 -3.0 -4.7 -1.0 3.f 4.1 0.321 1.144 f.035 1.174 1.232 f.249 f.241 1.203 1.290 1.214 1.000 f. tee 0.334 N 0.324 1.101 0.794 f.tN 1.254 1.273 1.240 1.220 f.197 1.139 1.05 f.202 0.300 0.7 -3.7 -3.9 -S.0 0.1 2.0 -0.1 -4.7 -F.3 -4.3 -2.1 1.4 4.1 8.794 1.237 I.200 t.281 1.2M t .2M i.310 t.337 1.335 1.273 1.423

! 9.913 1.134 f.173 1.284 1.2M 1.309 f.333 f.308 f.244 1.217 0.707

-0.2 -0.2 -0.2 9.4 0.4 1.1 1.2 al.7 ,-4.4 -4.4 -3.3 0.442 I.IN f.277 I.278 i.tet t.004 1.229 1.330 t.348 i.224 4.405 J 0.440 t.234 1.302 I.332 1.214 I.000 t.284 I.334 I.299 1.I84 0.439 4.2 4.2 4.2 4.2 2.0 4.3 4.3 1.7 -3.4 -3. 4 -3.3 0.313 1.tet i.230 t.439 4.f42 I.477 I.233 1.223 0.333 E 6.325 I.210 t .2M i.003 I.000 t.000 1.210 t.097 4.31 2.4 2.3 2.4 3.3 6.2 al.7 -S.I -10.7 -3.3 0.439 0.790 1.130 1.002 f.107 1.431 4.400

t. 0.449 f.013 f.179 f.030 1.131 0.041 0.374 2.3 2.3 2.4 2.4 -4.7 -10.4 -10.4 0.333 4.J00 0.J33 0 0.334 0.300 0.203

-0.t =0.I -9.0 P9PIl 211-t0 NIP 400 t1/18/04 0.000 POE01CTES Fells 0.000 fl0A00003 PENN 0.0 3 IIFF. (N=Pl/P 4.0 1 FZ e 1.408 Fenn a 1.320 Af Ft23E F0 e 2.340 Af Ft 2M

e IUOME-~rz-3 POWER DISTRIBUTION AT POWER PSFN 211-9 1300.f WT ROUTINE 12/1t/94

(

t 2 3 4 3 4 7 I f to 11 12 13 0.334 0.334 0.333 4 0.334 0.343 0.373

-0.3 2.4 3.1

. 0.453 0.794 1.149 1.0N 1.144 0.994 0.42 3 0.43 0.700 1.145 1.030 1.185 1.024 0.472 2.1 -4.3 -4.3 2.4 3.4 2.4 2.0 l 0.317 1.125 1.183 1.041 0.f37 1.024 1.201 1.144 0.337 C 0.328 1.138 1.182 t.043 0.f47 1.030 1.235 1.184 0.342 2.1 1.2 -4.0 0.4 1.1 2.3 2.8 2.0 0.7

! 0.444 1.144 1.142 1.201 f. tee 0.991 1.144 1.241 1.238 f.178 0.473

, B 0.484 1.174 1.135 1.198 1.148 1.041 1.191 1.242 1.270 1.100 0.478 l.3 1.4 1.2 -4.3 0.2 1.0 2.3 1.7 1.0 0.1 0.7 i 0.f54 1.!47 1.188 1.209 1.219 1.233 1.227 1.231 1.258 1.235 1.02t i E 0.f42 1.154 1.190 1.207 1.220 1.234 1.233 1.245 1.249 1.218 f.000 0.4 0.4 0.2 -4.2 0.1 0.1 0.3 -4.3 -4.7 -1.4 -2.1 0.354 1.125 1.002 1.142 1.203 1.207 1.109 1.221 1.241 1.197 1.072 f.184 0.349

  • i F 0.355 f.144 1.014 1.142 1.243 1.297 1.189 1.214 1.235 1.182 1.053 1.lM 0.344 i I.4 ' t.4 8.T ~-4.1 ~ ;6;4 0;0' ~-4.1 -" -4.T -4.T- ':1.3 -1.T *-1.7 -t:3 0.344 1.004 8.727 0.f00 1.228 1.199 0.897 1.291 1.247 1.011 0.943 1.435 0.352 0 0.342 f .M4 0.735 0.781 1.234 1.195 0.894 f.lft 1.232 0.ff2 4.941 1.017 0.347 0.7 -4.1 -4.2 -4.7 0.2 0.3 -4.3 - -4.9 -l.2 -1.8 -2.3 -l.1 -1.3 0.341 1.148 t.484, t.181 1.230 t.221 I.200 1.234 1.237 1.184 1.048 1.Iof 0.348 N 0.343' t.I48 1.484 I.168 1.234 I.227 1 294 1.224 1.217 f.!M t.024 1.154 0.343 0.4 -4.2 -4.2 -t.t 0.3 0.3 -4.3 -4.8 -1.4 -t.4 -t.8 -t.2 -4.4 I.000 t.219 1.240 t.2M I.227 1.23I I.253 I.239 1.254 1.217 I .0N 1 0.793 1.177 1.226 1.240 1.229.1 242 1.243 1.240 1.230 1.199 9.791

! -l.4 -1.8 -1.8 9.2 0.2 -4.7 -0.4 *l.3 -1.4 -1.4 -1.3 0.M8 1.149 1.252 1.241 1.171 1.004 1.190 1.239 1.244 f.175 0.449 J 9.440 1.154 1.239 1.238 f.159 0.777 1.187 1.2M 1.231 1.150 0.441

-1.7 -1.1 -l.1 -0.3 -1.0 -4.7 -4.f -1.0 -1.0 -1.4 l.4 4.335 1.Iet t.294 1.434 t'.ftt I.443 1.224 I.I74 0.344 E 4.328 1.140 f.192 1.011 0.73e 1.001 1.215 1.145 0.330

  • t.2 *l.2 -1.0 *2.2 -l.4 -1.1 -4.7 -l.4 -I.4 0.M3 4.794 1.153 1.019 f.172 1.01 2 9.471 -

L 4.400 0.704 f.214 1.049 f.200 1.004 0.444

-1.0 -l.3 4.9 4.7 3.0 -4.4 -4.e 0.385 4.339 4.343 a .6.401 9.379 0.384 11.9 11.7 - 5.8 f

PIPW 211-9 1900.9 NET RENT!III 12/19/04 e

CONTSEL 000 P981710B0 t$7EPS) . 0.000. PREBICTta FINE Ingue segue CafEL CafRL CafRL CITAL 9.000 #EASURED F3NN 4 8 4 I C 3 0.0 2 B2FF. la-ft/P

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l 1

. Section 13.0 XENON REACTIVITY j 1

Xenon reactivity behavior data for Unit 2 Cycle 11 was supplied by l Westinghouse as part of the WATCH data package. Point Beach code XENALG will be run with a TDF1 of 0.95 and a TDF2 of 1.2 to remain consistent with the Xenon Tables. Tables are supplied for BOL, MOL and EOL conditions.

Section 14.0 SHUTDOWN MARGIN CONSIDERATIONS Rod swap results were within acceptance criteria and were accepted as valid proof of rod worth for shutdown margin determination. See Section 9.0 for rod swap details. Thus WCAP Table 6.3 was accepted as a valid shutdown margin determination. Table 14-1 calculates the excess worth available to Unit 2 Cycle 11.

TABLE 14-1 EXCESS SHUTDOWN WORTH AVAILABLE FOR A FULL POWER TRIP

~

BOL (pcm) EOL (pcm)

Shutdown Margin From WCAP Table 6.3 -4230 -3500

- Required Shutdown -1000 -2770

= Excess Worth -3230 .730 Section 15.0 EXCORE DETECTOR BEHAVIOR Section 15.1 Excore Detector Current Versus-Power Level The upper and lower excore detector currents for each power . range .

channel were recorded and calorimetrics were - performed at various power i levels. - The upper and ' lower detector currents were stansed for each channel

~

and then normalized to obtain predicted currents for 100% power. These 100%

currents are listed in Table 15-1.

~

.I - Intermediate range detector - currents versus power level' are- shown in Figure 15-1. The intermediate rang detector trip signals activated at

'about 2.9 x 10 4 amps - and - 3.2 x 10 amps for N35 ' and N36 respectively.

~

-- From Figure 15-1, the trip signals occurred between 28% and. 30% power :as:

.~~

expected.-

~

37

FIGURE 15-1 INTERMEDIATE RANGE DETECTOR

, RESPONSE TO POWER LEVEL 1200 -

1100 - ,e N36 1000  ;

/ N35 900-_.. .;-.: ,

o, . :  :

g  : .

E E 2 800 -

m- m

.f s

. M 700 - /

D U

m .:  :-

o 600 - -

M -

y M.

,e c

500 - .-, .

m  : . . - --

a  :

y  ; .-

E 400 '/ .

W .:

5 s ,-

300 1^

J i

200 .

100 ' ."'

L' O-

' -- 0 10 20 30- 40- 50 60 70' 80 90 .100 POWER LEVEL (%)

~[.

}.

38

9 TABLE 15-1 100% CURRENTS (p AMPS)

H 42 43. - M Cycle 11 576 626 379 573 Cycle 10 625 641 403 574 Cycle 9 615 651 415 610 9,

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139. ~ '

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, Section 15.2 Excore Axial Offset Response

~

Excore axial offset responds to actual incore axial offset (calculated from flux map data) in a linear fashion but not with a one-to-one correspondence. Excore axial offset " sees" only about 70% of the actual

axial offset. Table 15-2 shows the historic response of the detectors. -

Section 15.3 Channel Calibration The currents measured during flux maps at three different axial offsets were first corrected for quadrant tilt by dividing each channel current by the quadrant tilt factor calculated by PBCORE for the associated quadrant.

Then because each flux map was taken at a slightly different power level, the currents from two maps were ratioed up or down by the percent the power level had changed from the reference map.

Straight line fits of the " corrected" currents for each channel versus incore axial offset as determined by the flux maps were obtained. The intersection of the line with zero axial offset was the calibration current at the ' power level of the reference map.

Power range quadrant tilt alarms were meant for rapidly developing
tilts. Natural core tilts were washed out to prevent a bias in alarming of rapidly occurring tilts. This was accomplished by multiplying the calibration currents for each channel by the quadrant tilt factors calculated by PBCORE, to obtain " tilt free" calibration . currents. Thus, after the " tilt ' free" calibration currents were entered, the computer and the Hagan recorders ~ indicated the same power (voltage) on'~ all upper half
quadrants and the same power on all lower half quadrants.

In actual practice, - the. " tilt free" calibration currents are ratioed to a power level slightly above normal . operating full power . to - make' it possible ; for the currents to be set in ~ at power. Table 15-3 ~ lists the actual " tilt free" calibration currents at 100% power at BOL'.-

t a

f e

e 3

5 -

g , y

~.

TABLE 15-2 EXCORE AXIAL OFFSET RESPONSE HISTORY Slope (Incore vs. Excore)

S S S S Cycle 11 1.45 1.37 1.17 1.38 Cycle 10 1.56 1.55 1.27 1.50 Cycle 9 1.49 1.58 1.27 1.66 TABLE 15-3 BOL CALIBRATION CURRENTS (100%)

4.1 E M M T 304 321 201 290 -

B 265 291 168 271

  • S e

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41-

I i

i 4 . .

. Section 16.0 OVERPOWER, OVERTEMPERATURE AND DELTA FLUX SETPOINTS CALCULATION Section 16.1 Overpower and Overtemperature AT Setpoints Calculation Discussion of the setpoints and equations has been sufficiently covered in previous reports.

l l

The equations are:

Overpower AT ( y,1Tg) 3 I8 5 1 1 SAT [K4 - K5( I S + 1)( 1+t 4S )

~

6 b ( 1+T4 S } ~ l~ (

5 1

overtemperature AT( 1+t S )

3 1+T S SAT. (K y } ~ '(O 

                           -K((1+4S 2
                                         )~ }( 1+I2 S }
  • b ( ~

See Tables 16-1 and 16-2 for. the constants associated with this cycle of operation. Section 16.2 Delta Flux Setpoints Calculation The overpower and overtemperature AT setpoints are reduced when . the [ excore detectors sense a power mismatch between the top and bottom of the L core. The dead band is +5% and -17% before the setpoints are reduced. .For l each percent (more than 5%) the top detector output exceeds ~ the botton . j . detector, the setpoints are reduced an equivalent of 2% of- the rated power. l For each percent (more than -17%) the bottom detector exceeds the ' top I detector, the setpoints are reduced an equivalent of 2% of rated Lpower. l. , t. O e 

                                                          'f.                                  5 42'                                  ._ , ,

I l I 

  .                                 TABLE 16-1 l

OVERTEMPERATURE AT CONSTANTS i AT, = Indicated AT at rated power, F T = Average temperature, *F T1 = 574.2*F P = Pressurizer pressure, psig P1 = 2235 psig , Kg 51.117 for operation at 2250 psia' primary system pressure 51.30 for operation at 2000 psia primary system pressure K2 = 0.0150 

                               ~

K3 = 0.00079'1 . Il = 25 seconds l I2 = 3 seconds T3 = 2 seconds for Rosemount or equivalent RTD 

                 = 0 seconds for Sostman or equivalent RTD 1

l I4 = 2 seconds for Rosemount or equivalent RTD L 

                = 0 seconds for Sostman or equivalent RTD
                                                                               -- s
 *9 a

g M r 44 s S 43'

s TABLE 16-2 OVERPOWER AT CONSTANTS AT,= Indicated AT at rated power, *F T = Average temperature, *F T' = 574.2*F K4 $ 1.089 of rated power , K5 = 0.0262 for increasing T 

                 = 0.0 for decreasing T                        .

Ks = 0.00123 for T 2 T 

                 = 0.0 for T < T' T5    = 10 seconds f(AI) as defined in Section 16.2 13    = 2 seconds for Rosemount or equivalent RTD
                 = 0 seconds for Sostman or equivalent RTD I4 = 2 seconds for Rosemount or equivalent RTD
                 = 0 seconds for Sostman or equivalent RTD 4

-4 s 9.: 

                                     ;44-
                                                                 ._..J

4 4 

     ~

Section 17.0 FUEL PERFORMANCE Reactor coolant activity is summarized in Table 17-1 and indicates good fuel integrity. 1' Because of low Cycle 10 activity, no fuel assembly failures were expected. Thus, there were no fuel assembly inspections scheduled other than the demonstration optimized fuel assemblies. Those assemblies were in good condition. TABLE 17-1 TYPICAL ISOTOPIC COMPOSITION OF PRIMARY COOLANT ACTIVITY End of Cycle 10 Start of Cycle 11 Isotope Half Life pC/cc x 10 1 pC/cc x 10 1 I-131 8.05 days 0.1 0.0 I-132 2.3 hours 1.5 0.7-I-133 21 hours 1.0 G.5 I-134 53 minutes 2.5 1.2 I-135 6.7 hours 2.0' 1.0 TOTAL 7.1- 3.5 Gross Activity (pCi/cc) 30 minute decay 0.6 0.3 i e 

  ,e
                                                '45 .-                      -
                                                                                           .. i

1: 4 Section

18.0 CONCLUSION

i. The use of optimized fuel assemblies produced no unusual physics
  • testing results. The use of. optimized fuel assemblies had no significant effects on other phases of startup testing.

x a 

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