Text

Interim Startup Test Rept
	ML20140D416
Person / Time
Site:	Fermi
Issue date:	03/12/1986
From:	; DETROIT EDISON CO.
To:	;
Shared Package
ML20140D412	List: ML20140D406; ML20140D416;
References
	NUDOCS 8603260122
	Download: ML20140D416 (95)
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{{#Wiki_filter:. _ _ _.. _ - - - - _ _ _ _ _ _ ^ i I -l e. THE DETBDIT EDISENTOMPANY TERMI 2 A'UCLEAR POWER PLAFT INTERIM STARTUP TEST REPORT ObO32601Qg g603gy DR ADOCK 05000341 p PDR .s

FERMI 2 NUCLEAR POhTR PLANT lETERIM STARTUP TEST REPDRT .. _ 2NDU Page Section '1.0 Introduction 1 1.1 Purpose 1 1.2 Test Report Format 1 1.3 Plant DeAcription 2 1.4 Test Progra::: Description 3 1.5 References 2.D General Test Prograz information 6 2.1 Chronology of Major ' Events 2.2 Matrix cf Test Completion Dates 7 3.D Test Results summary 8 31 Chemical and Radiochemical 14 32 Radiation Measurements 54 33 Fuel Loading i 66 34 Tull Ccre Shutdo+ n Mare n 68 35 Control Rod Drive System 3.6 sRM Performance and Iontrol Rod Sequence Ixchange 73 74 37 lEM Performance 3.B LPRM Calitn ation 75 76 39 APRM Calibration 77 31D Process Computer 7B 311 BCIC Syste: BD 312 HPCI System 82 3 13 Select Process Te:peratures 84 3 14 System Ixpansion 87 3 15 Feed. eater Syste: BB 3.16 FJIV Functional Test 89 317 Reller Valves 91 31B Piring system Vitration 92 319 Reacter Kater Cleanup Syste: 94 3 2D Residual Heat Removal Syste: LEP/10D/R5LE/1.0 D30.686

TERMI 2 N:ELEAR PDWER 71, ANT ~ INTERIM ETARTUP TEST REPORT s 7.D Introduction 1.1 Purpose The purpose of this Interin Startsp'Tast Repo initial fuel load through present *==M test condition This report of plant startup and power ascension fro: testing is submitted as required following receipt of the testing. Operating License in compliance with Technic l nine months after initial critica11ty are outlined herein in J This is an accordance with Technical Specification 6.9.1.3 i interim report that covers al.1 testing applicable to the test 14.2.12.2. con:Stions completed as descrlhed in TSAR Subsection Su;;1enentary reports will be issued to cover the remaining 4 testing as re:;;1 red per Technical Specification 6.9.1.3 Included in this report are 1$escriptions of the nessured values of the operating conditions and characteristics obtained during the test progra and any corrective actians that unre required to obtain satisfactory operation. 't.2 Test Report To: Tat Sections 1.0 and 2.D 1of this, w i. provi15e genera"1 inf ormation Section 3 0 about the Termi'2 plant and the testing program. provides a tasic description of the testinE we have performed along with a su-nnary of the results and analysis obtaint:d fro l Each test summary is divided into three subsections cove ~ing the purpose, test criteria, and Tesults of each test. each test. 1.3 Plar.t Descri;tien The Terri 2 Nuclear Towe" Plant is. located in FrenehtownThe Nuc To.nship,Honroe Coonty, Michigan. k nucleaa reactor ratt: Syster cens!sts of a General Electric BhTs at 3292 Wt, or 1100 we, constructed in a Mar a containmen; with a te-cical suppression pool. 7nis pa ar.1 is o.ned and operated by the Detroit Edison Cocpany (951) and the Wolverine Powe Icoperative, Incorporated (105). 4 3

,1. 4 Startup Test Program Description The Startup Test Thase began with preparation for fuel loading and will extend to the completion of the warranty demonstration. This phase.is subdivided into four parts: . 1. - Tuel Leading and Open. Vessel. 7asts . 2. Initial heatup 3 Power tests 4. Warr.anty demonstration. The Startup Test Phase and' all associated testing activities adhere closely to NRC Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Ennlad Power Reactors.' The overall objectives of the Startup Test Phase are as follows: To achieve an orderly and safe initial core loading 1. To perform all testing and sensurements necessary to 2. determine that the approach to initial criticality and the subsequent power ascension are accomplished safely and orderly 3 To conduct icw-pcwer physics tests sufficient to ensure that physics design parameters have.been Jaet li. To conduct initial heatup and hot Tanctional testing so that hot integrate;1 operation of all. systems is.shown to meet design apaM11 cations To conduct an caderly and safe ?cuer Ascension Trogram, with 5. requisite physics and system testing, to ensure that when operating at power, the plant meets design Antent 6. To conduct a successful warranty demonstration progrs=. Tests conducted du-ing the Startup Test Thase consist of Major Plant Transients and Stability Tests. The remainder of tests are t!irected toward demonstrating correct performance of the nuclea* boiler and nu~.erous auxi?lary plant systems while at power. Certain testr may be ider.tified with more than one part of the Startup Test Phase. Figure F1 shows a general view of the Startup Test Phase Program and should be considered in conjur.ction with Figure 1-2 which shows, graphically, the varicut test areas as a function of core thermal power and flow. For a more resprehensive description of the testing progra re!L to F*ference 1.5.2. -_

1.5 Peterences ~ The followinE is 2 list or documents that provide supplanentary Information of the Termi 2 Startup Test Phase Program: 1. Termi 2 Technical Specifications. Section 6. 2. Final Safety Analysis Report, Fermi 2 Nuclear Power. Plant. Section 14 i e - -...

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1 1 1 1 1 1 y 6 4 4 e a i e i i e D 10 2D .30 40 50 60 7D BD 30 1DD 110 Plotce.tege ofImre FM Notes: 1. See Tigure 1-1 fc-startup test titles 2. Power in pe*camtage of rated thermal power 3292 t%~T. 3 Core flow in percentaEe of rated core recirculation flow. 100.0 x 106 It/hr. 4. TC - test con::ition. TID":.I 1-2 1 1 ) l

2.D Genera'l Test PrcEra a. Infor1tation 4 2.1 Chronology of Major Events Date 3/20/85 Eeceived (51) Tacility Operating License No. NPF-33 3/20/B5 Started Tuel Leading 4/4/85 Completed Fuel Leading Completed Dpen Vessel Testing - 6/1/85 6721/85 2nitial criticality 7/15/B5 F.eceived (Full Power) Facnity Cperating License NPF-43 9/26/85 Ccx:pleted Initial'Tarbine Roll 1D/10/85 Fall Outage Begins 4 4 4 9 O W

2.2 Matriit of Test Cc pletion Dates Pre-Fuel Open Vessel Weatup Test No. ' rest Title 7/16/85 1 Chemical and Radiochemical 1/14/65 +*~ 1/24/85 4/19/85 7/14/85 2 Radiation measurements 4/5/95 ' 3 Tuel Loading

a-4/10/85 4

Full Core Shutdown Margin 4/5/85 9/16/B5 + 8/25/85 5 CRD 6 SRM Performance and control Rod Sequence B/30/B5 + 7 liater Level Measurements B/19/85 B IRM Performance 1/24/85 9 LPRM Calibration 8/1/85 + 10 APRM Calibration 5/30/95 11 Process Computer 7 /14/B5 "d- +" 6/31/85 12 RClc 9/25/85 13 HPCI 14 Selected Process Te=peratures 6/12/B5 9/1/85 15 Syste Expansion

16 Core Power Distribution
17 Core Performance
19 Core Power Void Mode Rasponse e--

20 Pressure Regulater 7 /9/85 21 Feed'ater-Syste: ~~' ^

22 Turbine Valve Surveillance 7/12/B5 23 MSlv 7/3/85 24 Relief Valves
25 Turbine Stop valve and Control Talv! Tast Cicoure
26 Jihutdewn frc: Dutside Contrni Roo:
27-Flow Centrol
2B Recirculation Syste:
29 Loss of Offsite Power 7/3/85 30 vibration Measure-ents
31 Recire. System Flow Calibration 7/11./E!

32 Reactor 'Wata-Cleanup Syste: 8/31/ES 33 Residual Heat Re oval Syste 9/24/E5 34 71 pane Syste Dynamic Response - Test not performed in this tam tendition ' - Tests not addressed in this su-mary repcrt i.

$1, Chemical and Radiochemical 3 121 Purpose The princip1 Turpo7es of this test are to collect 'information on the chemistry and radiochemistry of the lleactor Coolant and Support.5ystems, and to determine that the sampling equipment, procedures and analytic techniques are adequate to ensure specif.ications and process requirements are net. ~ Specific purposes of this test include evaluation of . fuel performance, evaluations of filter deelneralizer operation by direct and. indirect methods, confirmation of condenser integrity, demonstration of proper stea-separator-dryer operation, measurement and calibration of the of f* gas system and calibration of certain process instru-entation, if required. I)ata for these purposes are secured fro a variety of sources: plant operating records, reEular routine toolant analysis. radiochemical measurements of specific muclides and special chemical tests. 3.1.? criteria 1.evel 1 I hemica'1 factors befined.in the Technical Specifications and Fuel Warranty must be maintained .arithin the. limits specified. Water quality must be known ut all times and remain within the guidelines o' the Wa* er. Quality Specifications. The activity of gaseous and liquid effluents zust confcr to license. limitations. 313 Results Prior to loading fuel, app-opr.iate chemistry data was taken.. All data remained within criteria levels except for feedwater conductivity and feedwater copper concentration. These values could hav e been elevated due to low condenser vacuum, 7.inimum filter de ineralizer flow and low sample flow rates. Du-ing heatup test condition, these values were within acceptable limits. See Figuae 31'1 for specific in!cr:aticn on pre *f uel load che:1stry data.

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) turing the heatup test wndition, all chemistry data t for Control 'taken fell within applicable limits excep These levels Rod Drive (CRD) dissolved oxygen levels. are expected to decrease during.further test conditions with greater steam flow and the steen jet air ejectors 4 . Ja service rhich will more effectively purge gases.from Refer to Figure 31-2 for heatg the condenser. .che15stry data. 'M 9

2 Tre Fuel Load Chemistry Data Date 1 /1 5 /815 CTP = 0 MWe - D 7tesults Limit System Analysis EEACTOil. WATER. Conductivity (umho/cm) # 25'C D.43 < 10.0 chloride (ppb) <5 < 500 7urtidity (WrD) 0.15 J L.D. < 10 < 50.D Boron (ppb) Silica (ppm) D.023 < 5.0 6.2 5 3,< pH < B.6 pH CONDEWSATE Conductivity (ucho/cm) # 25'C 0.5 < D.5 Chloride { ppb) <5 < 10.0 j FEEDh'ATER Jnsoluble Iron (ppb) 1.8 Total Metals (ppb) 10 I Conductivity (ucho/cm) 0 25'1; D.38 j D.1 lioluble Ccpper 11. 4 i Jnsoluble Copper D.1 - C r 'The li.it of the solable and ir. solubles (Total Metals) is that 19 be < 15 ppb or which can contain no incre than 2 ppb copper. 1.1. n - 1.owest Level of Detection 1 FID'JRE 3 1-1 ^10-

Teatup Test Cendition themistry Data I) ate 7 /1 5/85 CTP = < 55 m'e - D Beactor Water Temperature 54D *F System Analysis . L11 nit 1tesults 3EACTOR WATER Conductivity (umho/cm) # 25T D.31 . ( AD 1 < 200 Ihloride (ppb) Turbidity (NTV) 3.6 l Iodine - 131 (uti/rd) <7.7 I-DB 1 - 133 (uti/ml) 2.D I-06 2 hr Gross Activity - Filtrate (cpm /Tal) 6.1 E-D2 - Crud (cpm /ml) A.D E-D1 7 dy Dross Activity - Filtrate (cpm /cI) 4.7 - Er.ud (cpm /ml) 25 silaca (ppm) 0 385 6.A 5.6 < pH < E.f pH 'The dose equivalent 2etina-131 shall net exceed a concentratien ci c.2 uCi/g=.

- Ibis analysis.is used to develop trend data.

TIGUfiE 31-2 -11*

Results Limit Analysis , System 0.08 j D.1 Conductivity (umho/cm) # 25'D CRD WATER 300 j 50 Dxygen (ppb) Completed Vff-gas Vini Sample Analysis RADIDACTIVE GASEDUS E21-P275A EFFLUENTS Off-gas Mrmitor Reading 35 .D11-K6014 (mR/tr) Dif-gas Monitor ReadinE 35 D11-K601B (rJi/nr) 1.7 E-DB

- BackEround 1

SOTS Exhaust, Div I Dorrecte:' D11-P275 (uti/cc) 1.2 I-DB

- Background SDTS Exhaust, Div 'II Corrected D11-P275 (uci/cc) 8.0 E-DB
- Background Turbine Eldg (uC1/cc)

Cerrect e:' L11-F279

- Backgrour.d Js.D5 I-D7 Reactor EldE (ci/ce)

Corrected D11-72BD 9.1 I-D7

- Background Radwaste EldE (uti/cc)

Corrected D11-P281 ~ B.1 E-tB

- BackEroar.:

Service Bldg (uti/cc) Corrected D11-P2B2

- BackEround 1.2 I-07 Site Storage Bade (uC1/cc)

Ccrrected .D11-P299 This je

Perform isotc;;ic analysis of un DII-Eas sa:ple and attach data.

for trend analyrds.

- Readout of Channel 5 on sv2ND 7)anel, Ectle Gas. Xe-133 equivalent If the SGTS is tot in This data.is for trend analysis.

(uC1/cc). se*vice enter NA.

' Tor trend analysis TIGURI 3 1-2 -- _ _.._.

, System Analysis Results Limit 1)EMIN. Conductivity lucho/c=)- @ 25C D.D72 < c.1 EFFLilEET FEIDWATER Insoluble Iron (ppb) 1 Cenductivity (umho/cm) # 254 A D.DB < D.1 3 D.1D < D.1 "Inscluble Copper (ppb) ~D.D3 Soluble topper ippb) 1 Total Metals (ppb) 2.33 CONDEN3 ATE Cer.ductivity (usho/cm) @ 25'c D.1Ji j D.5 thlwide (ppb) C2 Insoluble Iron (ppb) 3

The limit of tha solubles and insolubles (Total Metals) is that the to+M shall be < 15 ppb or which can contain no more than 2 ppb mopper. These limits nay be exceeded in this Test Condition.
- For trend analysis.

TICUEE 3 1-2 3'.2 Tiadiation Measecrents 3 2.1 Purpose ~ The purpose of this test is to determine the background radiation levels in the plant esivirons for baseline data and activity build-up dejng power ascension test 1ng to ensure the protaction af plant personnel during plant operation. 3.2.2 triteria 1.evel 1 The radiation-doses of plant origin and the occupancy tires of personnel in radiation tones shall be controlled consistent with the guidelines of the standards for protection aEninst radiation outlined it. 10CFR20. " Standards for Protection Against Radiation, and NRO General Design Criteria. 3 2.3 Results Prior to fuel load, du-ing open vessel and heatu; test conditions, radiation measurements were taken in the for of process and area radiation monitor data and This data was acceptable and : personnel site arveys. radiation protection was provided in f ull compliance with the criteria. See Figures 3 2-1 strougn 3 2-3 for appliela v.anitar and.sevey raadings.

Area Padiation Monitor Sensor Locations ' Location ICol.) T1oor-Bldg. ,, Channel Vo. '(F-1D) 2nd Fir. Reac. Bldg. (RB) Ters. Air lock 1 (B-9) 1st 71r. RB Iquip.- Air Lock (J-13) 2nd Fir. Aux. Bldg. (AB) Access Contral 2 (0-10) 2nd Fir. AB Change Area Control 3 (B-13) 3rd Fir. RB CRD Storage and Maintenance Area 4 AB Main Contral Room (CS) 5 (0-13) 3rd F17. 2 6 '(F-9) Sub'3ase. RB'S.E. Corner 7 (B-1D) Sub Base. Ita S.W. Corner 8 (B-15) Sub Base. RB N.W. Corner 9 (0-17) Sub Base. FB N.E. Corner 10 (G-11) Sub Base. RB HPCI Rs. (F-11) 1st Fir. RB Neut. Hon. Eq.1m. 11 (F-10) ist Fir. RB Neut. Hon. Control Panel. 12 13 (A-11) Sub Base.RB Supp. Pool 14 (F-15) 5th Fir. RB Fuel Stor. 7ool 15 (F-15) 4th Fir. RB New Fuel Vault 16 (F-12) 5th Far. RB Refuel Area Hear Reactor (F-13) 5th Fir. RB Refuel Area Near Reactor (High 17 18 Range) (L-12) 3rd Tir. Turbine Bldg. (TB) Turbine Anlet End 19 (R-10) Base. TB Sump 2D (N-7) 2nd Tir. TB Main Cond. Ares 21 (J-4) 1st Fir. TB Decon. Area (M-17) 1st Fir. Rad. Waste Bldg. (RWB) Contral AL 22 23 (#-17) Base. RWB Equip. Drain S. Pmp 24 (P-16) Base. RWB Floor Train S. Pump (R-17) 1st Flr. RWB Dra Eonveyor Aisle Operating Area 25 26 Spare 27 (0-11) 4th Fir. AB Vent. Equip. Ra. 2B (B-15) 4th Fir. RB Change Em. 29 (H-12) RB Basement Air Lock 30 (B-12) 1st Fir. RB Drywell Air lock labyrinth 31 (G-13) 1st Fir AB Near Blowout Pnl. 32 (C-9) 1st Fir RB South Air Lock 33 (N-2) 2nd F2r. TB Near Off Gas Equip. 34 (R-2) 1st Fir. TB Wear S.J. A.E. Area 35 (K-1) 1st Tir. TB 5.W. Corner 36 (M-2) 3rd Tir. TB South End 37 (R-14) base. RWB Scrap Cement Recovery 35 IL-13) 1st Far. RWB H.P. Lab 39 (P-16) 1st Fir. TWB Receiving Area 40 (5-17) 1st Fir. RWB Balling Roo: 41 (N-16) 1st Fir. RWB Filter Demin. Area 42 (s-17) Mezz. Ra3 Washdown Area '43 (S-12) 1st Far. Servloe Eldg. (SB) Nach. Shop. M FIGURE 3 2-1.

Area Radiation Honitor Sensor Locations Location (Col.) Floor-Bldg. . Channel No. . 45 1st T17. 7nside Drywen 1st'Tir. On Site Etg..Bles. control itamm

46 1st Fir. On Site 5tg. 31dg. Compactor Room

'47 1st Fir. On Site Stg. Bldg. Truck Unloading Station '48 is Iccated on Process Eadiation 1tonitor Panel 'The re=cte intlicat oa E11-PBS4 (Relay Roo ). t e e ) TJGURE 3 2-1 - 1 L-

f Area Radiation Monitor ' Data Pre-fuel load ' Open Vessel Heatup_i I Test Condition O <55 O CIP 0 0 0 MWe (100*F 540*F _ <100*F 1/19/85 4/16/85 7/13/85 Mod-Temp Date O.03 mr/nr <0.1 0.05 1 0.02 <0.1 0.02 2 <0.1 0.02 0.03 3 (0.1 0.02 0.02 + 4 0.08 <0.1 0.05 5 6 0.03 <0.1 0.04 0.2 0.2 0.2 7 8 0.2 0.4 0.4 9 0.3 0.3 0.3 ~~~~ 0.2 _ 0.2 0.2 10 0.3 0.3 0.3 11 4 5 12 5 0.03 <0.1 0.04 13 4 14 3 3 15 0.02 <0.1 0.02 16 0.04 <0.1 0.05 17 0.02 <0.1 2 18 300 300 300 '~ 19 0.4 0.4 0.4 0.3 0.3 ~~ 20 0, 3 21 2 3 3 22 0.03 <0.1 0.03 0.02 <0.1 0.03 23 24 03 0.3 03 25 0.3 0.3 0.3 0.2 0.2 26 0.3 27 00S 00s DOS 26 0.03 <0.1 0.03 ~~~ ~~~~~ 0.06 <0.1 i.0.06 29 l These represent remote reatings where possible NOTE 1: NOTE 2: ODE indicat es that the r.or.itCr 15 *Dut Of ServiCP'. EIGURE 3 2-1 E Area Radiation Monitor Data 4 (Cont'd) 1 Test Condition Pre-fuel load Open Vessel Heatup. O O <55 CTP i MWe 0 0 0 j <100*F (100*F 540*F_ Mod-Temp Date 1/19/65 4/16/85 7/13/85 t 1 i 30 0.03 mr/hr <0.1 0.04 l 31 0.2 0.2 0.2 32 5 4 5 33 0.02 <0.1 0.02 34 0.05 <0.1 0.07 35 0.03 <0.1 0.03 4 36 0.03 <0.1 0.03 i 37 0.02 <0.1 0.02 1 38 0.02 <0.1 0.02 39 0.05 <0.1 0.08 40 0.03 <0.1 0.03 ~~~ 41 0.03 (0.1 0.03 42 0.05 <0.1 0.04 43 0.04 <0.1 0.04 44 0.03 (0.1 0.02 l-45 oos 0.2 oos 46 0.2 0.4 0.4 47 3 3 0.3 i 48 3 4 3 i l i I i I 5 1 i NOTE 1: These represent re :te readir.E5 Where Possible EDTE 2: 005.in:11 cates that the r.cnitor is ' cut of service". j t l l T.1GUFI 3.?-1 .n- ..,,_r.r.m, _.-..v.,-,..,,y_-, .,-,- -.+.-..-,.,_--, -r ,,.y,- 7.,_.,,_,myeww.,, y.,y,..,e m - em. ..+my...,m..,,._.,w~.

Process Padiation Monitor Data Pre-fuel load' Open Vessel Heatup; i Test Condition 0 0 <55 CTP 0 0 0 MWe <100*F (100'F 540*F Mod. Temp 1/19/85 4/16/85 7/13/85 Date Off-Gas Radiation - A (D11-K601 A) 3 mr/hr 2 3.5 Off Gas Radiation -- B (D11-K6D1B) 3.5 er/hr 3 3 off Gas Radiation - Linear (D11-K602) O rr/hr SE 'B D Radaaste ET[luent (D11-K6DA) 2 eps 2 2 GSW Effluent 4 4 (D11-K605) 3.5 cps RBCCM System ( D11-K606) 3 cps 2 2 A Main Steam Line (D11-K603A) 1.2 tr/hr 1 Dos B Main Steam Line (D11-K603B) 1.h rm/hr 1 1 C Main Steam Line (D11-K603C) 1.6 w/hr 2 1.2 D Main Steam Line (D11-K6D3D) B.O tr/hr 11 1 Div-I EECW HX Inlet (D11-KB00 A ) 200 cpc 200 200 Div-l RHR S.W. (D11-K801 A) 200 cpm 200 200 Div-1 RB Vent Exh. (011-K90B) 50 c;~ 50 40 Div-I Cont. Etr. Makeup Air (D11-K809) 40 cpr 30 30 Two Min. Holdup Pape Exh. (D11-KB14) 250 cpr 200 400 Div-Il EECW HX 2nlat (D11-K800B) 200 epr 200 200 Div-Il -EME S.W. (D11-KE01B) 150 cpm 200 200 DDS dndicate.S that the' monitor is "Dut of Service". EDTE 1: EDTE 2: NA - not ava11staa or applicatie at this time. TIDURI 3 2-2, _.

Process Radiation Monitor Data Pre-fuel load! Open Vessel Heatup' 1 Test Condition O <55 O CTP 0 0 0 MWe <100*F 540*F <100*F 1/19/85 4/16/85 7/13/85 Mod. Temp Date ~ Circ. Mater Rev. Decant Line 200 700 200 ups (D11-KB02) Div-11 RB Vent Exh. 60 40 60 cpm (D11-KB10) Div-11 Cont. Ctr. Makeup Air 60 60 60 rpm (D11-K813) Two Min. Holdup Pipe Exb. 200 400 (D11-KB15) 300 epm 1st. Fir. Inside Drywell 0.3 nr/hr 0.2 DOS (D21-K745) s 1st Fir. On Site Stg. Bldg. Control Boom ~ 0.2 mr/hr O.4 D.5 (D21-KB46) 1st Fir. On Site Stg. Eldg. Compactor Room 4 4 3 tr/hr (D21-KB47 ) 1st Fir. On Site Stg. 31dg. Truck Unicading Station 4 3 er/hr (D21-K848) Fuel Pool Vent Exh. (DIY I) 4E-2 rr/hr 2.5E-2 SE-2 (D11-K6D90/D11-R606) Fuel Pool Vent Exh. (DIV II) (D11-K6093/D11-3606) 6E-2 cr/he 6E-2 4E-2 Prl. Cont. Rad.(gamma) (T50-K003/T50-RB09) 40 cpe 45 90 Fuel Pool Vent Exh. (DIV 2) NA SE-2 NA (D11-K609 A /D11-B605) Fuel Pool Vent Exh. (DIV 71) RR AI-2 WA (D11-K6093 /011-R605) 03S indicates that the denitor is'*0ut of Servj ee". EDTE 1: NA - not availatie or applicatie at this time. W3IE 22 F200EI 3 2-2 ^ 2 D-

Process Radiation Monitor Data Test Condition Pre-fuel load Open Vessel Heatup CTP O O <55 MWe 0 0 0 Mod. Temp (100*F <100*F 510*F Date 1/19/85 4/16/85 7 /1 3/85 C.C. Emerg. Air (South), Div I i (D11-KB36A) lila WA WA E.C. Emerg. Air (North), Div 1 (D11-K837 A) NA WA NA Cont. Area Hi-Range Monitor, Div I (D11-K816 A) NA NA NA C.C. Emerg. Air (South), Div II (D11-K836B) NA NA NA C.C. Emerg. Air (North), Div II I (D11-KB37B) NA WA NA Cont. Area Hi-Range Monitor, Div II i (D11 -KB16B) NA NA NA Reactor Bldg. Vent Plenum Exh. (D11-K610) NA NA NA Radwaste Bldg. Vent Ezh. (D11-K610) NA NA NA SGTS Vent Exh. Div I (D11-K610) NA NA WA SGTS Vent Exh. Div ll (D11-K610) NA NA NA OSB Machine Shop Vent Exh. ' (D11-K610) NA NA WA Turbine Bldg. Vent Exh. (D11-K610) NA NA NA On-Site St e-age E'.::g. Vent Exh. ' (D11-K610) WA NA NA N7TE 1: DOS Andicates that the monitor ds "Out of Service'. KOTE 2: NA - not available or applicable at this time. FIGURE 3 2-2 j l

i t Survey Data I i Pre-fuel load Open Vessel Heatupj Test Condition l O O <55 CTP 0 0 ~ --4 0 (100'F (100*F 540*F _ MWe l 1/19/85 4/16/85 7/13/55 Mod. Temp + Date Location:. General Site (0.2 mr/hr <0.2 (0.2 1 ) (0.2 mr/hr (0.2 (0.2 2 <D.2 cr/br <0.2 (0.2 3 <0.2 mr/hr <0.2 _.2 4 i <0.2 mr/hr '<0.2 (0.2 5 .i 6 <0.2 mr/hr <0.2 (0.2 f <0.2 er/hr (0.2 (0.2 1 7 8 <D.2 mr/hr <0.2 <D.7 i <0.2 mr/hr <0.2 (0.2 i 1D l' <0.2 mr/hr <D.2 (0.2 11 <0.2 er/hr <0.2 <0.2 12 d, <0.2 mr/hr (0.2 <0.2 1 13 4 1A -40.2 mr/hr <0.2 <0.2 <0.2 er/hr <0.2 <0.2 l 15 16 <0.2 mr/hr <0.2 <0.2 I i, <0.2 rr/hr (0.2 <0.2 17 1B (0.2 er/hr <0.2 (0.2 i (0.2 cr/br <0.2 <0.2 19 Readings are total readings where possible EDTE: total readAnE - ga==A 4 neutron TIGURE 3 2-3 -e e i l

Survey Vata Pre-fuel load Open Vessel Heatup_ T l Test Condition O <55 ~ O CTP 0 0 0 Wie <100*F 540*F (100*F 1/19/85 4/16/85 ,7/13/85 Mod. Temp Date Locations. General Site 3 1 <0.2 mr/hr <0.2 (0.2 1 .I 20 <0.2 mr/hr <0.2 (0.2 21 <0.2 mr/hr <0.2 <0.2 22 <0.2 rr/hr <0.2 (0.2 23 <0.2 mr/hr <0.2 <0.2 24 <0.2 cr/hr <0.2 <0.2 25 <0.2 v.r/hr <0.2 <0.2 27 l 28 <0.2 mr/hr <0.2 (0.2 <0.2 mr/hr <0.2 <0.2 29 i <0.2 rr/hr (0.2 (0.2 30 L t F 7teadinEs are total readings where possible NOTE: tct al reading = ga~na * *neutm t TICTtE 3 2'3 l -2 3-l l

..O 2N A CIIMP L e ' RRE w NE M 1 i EFS 1 SR / f l E o ' ~ W v O 1 Ws nH f T = e GN W / 0 C 00 W C j 0 * ", _ - r 5 o .s a tE p v. r* e N m , t O 'r l 4M t m'- I yW ng E _mo ra* I 6 R n wt e h .~ g _ E E t,e f ) E-ge i j f K a ~p A L, Ed ~~ g8 ,a l I r - c 2 s i U ",. e" ~ F .[ t / 2 5 N"Y ~4 I rt

=

O )cs, s. t d; \\1 l '!I l .l, l

!'i;l

f Survey Data 1 Test Condition Pre-fuel load' Open Vessel Meatup' O O <55 CTP 0 0 0 E'e <100*F (100*F 540*F Mod. Temp 1/19/85 4/16/85 7/13/85 Date Location: Reactor Building - Sub Basement <0.4 l 0.4 <0.2 inr/hr 1 1 <0.2 mr/hr (0.4 D.2 2 3 <0.2 mr/hr (0.4 0.3 k <0.2 cr/hr <0.4 D.2 5 <0.2 mr/hr <0.4 0.4 1 6 <0.2 mr/tir <0.4 O.3 7 <0.2 sr/hr <0.4 O.6 B (0.2 er/hr (0.4 V.4 9 <0.2 mr/hr (0.4 O.4 1, 10- <0.2 mr/hr <0.4 0.4 J EDTE: ReadinEs a e total readings where possible total reading - Ea=re *. neutron I .s v i ~ T1DUE 3 2-3 s -25 \\

l Reactor Building --- Sub Basement ~ Elevation 540'-0" , :k'L5;',$ild ?.'W'.'.iyy *1*,'ai,t:s Sl'. Y":.":iuii"RMt.'. '.2..diR1 y *,y _1 k till HCIC Pump & Turtune 03 !d 6j 7, A, [j,.,%u g .k..h

u t n

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Survey Data i i Test Condition i 1 Pre-fuel load Open Vessel Heatup CTP O O <55 MWe 0 0 0 Mod. Temp <100*F (100'F 540*F Date 1/19/85 4/16/85 7 /13/85 Location: Reactor Building - Basement 1 <0.2 w/hr <0.4 0.3 2 <0.2 w/hr <0.2 ~<0.7 4 <0.2 w /hr <0.2 <0.2 5 <0.2 nr/hr <0.2 (0.2 <D.2 w/hr <0.4 ~0.2 6 7 <0.2 w/hr <0.4 0.2 B <0.2 mr/hr <0.4 0.2 9 <0.2 w/hr - <0.4 D.2 i f 9 NDTE: Readings are total readings where possible Actal' reading - ma

neutron TIGURE 3.2-3

~27-

o Reactor Building - Basement Elevation 562'-0" 3 a = r. M u g,v.e p w.f g.q,. a ;sJa s m :. m w,; 1 p ot.yp,wete w:wt:.sa 1.,,, ..r -l,l p,. in. .yM.i dg u }.' 'i. ,g e' w e ( 8ie i -N-i 7-m r ~ ti r n e;e m m. m...; n OI m L, g: ( 8

a l

714 i g E l.. n e 9 1.... = t hi f n c s .,f

Z MM 1 m $.

m,_ 'y. a ( l}NW ; ll's t. l ji% / % open Hatch .j. f t,. s' ,i t i:

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hm, g N e..l. < '.. y-,y; a anui

m

Survey Data i Test Condition Pre-fuel load Open Vessel Heatup CTP O O <55 MWe 0 0 0 Mod. Temp (100'F (100*F 540'F Date 1/19/85 4/16/85 7 /1 3/85 a ' location: Reactor Building - First Floor 1 <0.?'er/hr <0.2 l <0.2 2 CD.2 mr/hr <0.2 (0.2 4 <0.2 tr/hr <0.k D.2 5 (0.2 mr/hr <0.4 0.2 7 <0.2 dr/hr <0.4 0.2 B <0.2 tr/hr <D.* D.2 9 <0.2 tr/hr <0.4 0.2 10 (0.2 mr/hr <D.4 D.2 11 <0.2 mr/hr <0.4 V.2 12 <D.2 mr/hr <D.4 0.2 13 <0.2 mr/hr <0.4 0.2 15 <0.2 nr/hr <0.4 O.2 16 <0.2 tr/hr <0.4 0.2 17 <0.2 tr/hr (0.4 0.2 <0.4 0.4 1B (0.2.tr/hr 19 (0.2 rr/hr <0.4 0.2 in i <D.2 tr/hr CD.4 D.? 23 <0.2 cr/hr <0.4 0.2 .XJTE: Readings are total readings where possible total rea:11nE - ga =a + Teutron 73EDEE 3.2-3 -2 9-

Survey Data Pre-fuel load Open Vessel

Heatup, Test Ccndition O

(55 ~ O CTP 0 0 0 MWe <1DO*F 540*F <100*F 4/16/85 7/1 3/85 Mod. Temp 11;i9/65 Date Resctor Building Qirst Floor ', Location: <0.2 <0.2 .<0.2 arlhr 25 <0.2 mr/hr ~<0.4 0.4 26 <0.2 mr/br <0.4 D.2 27 <0.2 mr/hr <0.4 0.4 2B <0.2 mr/hr <0.4 0.2 29 <0.2 cr/hr <0.4 D.2 31 <0.2 cr/hr <0.4 0.4 32 <0.2 mr/hr <0.4 D.4 13 <0.2 mr/hr <0.4 D.2 34 <0.2 rr/br <0.4 0.2 35 <0.2 mr/hr <0.2 <0.2 37 <0.2 mr/hr <0.2 <0.2 3B <0.2 mr/hr CC.2 CD.2 39 <0.2 mr/hr <0.2 (0.2 4D Readings a-e total readings'where possible ROTE: total restng = ga-:r.a

neutrDn TIGUEE 3 2-3.... -

1 r n -ze e p-a g i y Ic i,il]i fi z...._-. L. -.e-c_- =:- n

gy c_.

1 1 bet c ? ( 65 k M2 al h n*~ u f ( .g q . i psw.zw.--w.r.w.-#a 8 3 5, 4 9 f c a j.g 5 3 gtnr ag hb t. En "1 .......n. -,l.n *- 1 ~ s v ....=.-.-,..c }. I I n I .c SL l Si P.s E i =. m @N. # J,i 1 5 Y J-Ig t w to s m. ,bg- -. g ,e .f p g. ..m.. - -- -. ~. _. i,. s ..y*. [ , 7

$:.*L T,

6 e 1g [. P C, b e-W

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i t 4L: 'I I.I a)$ j5p 4. w\\ g u,.. <p ~-v. 4 g, te l ~ C c g. -.4s v. 1.,a. g g 1 v p,- -- ) eZ 9 w

==. ,3 V

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D s1 C D '5 F' ~ ' ~ E }qJ p -e.- 7

O 6.T h }=

a yr i o u.qngs l %n" __n. E 2.4 .,/ b ( l TIGrE 3.2-3.. -... ~

\\ Survey Data I Test Condition ' Pre-fuel load Open Vessel Heatup CTP O O <5% MWe 0 0 0 Mod. Temp <100*F <100'F 540*F 1/19/85 4/16/85 7/13/85 Date 1,ocation: Reactor Building - Second Floor 1 <0.7 w/hr <0.4 0.4 2 (0.2 w/hr <0.4 0.4 3 <0.2 w /hr <0.4 0.6 4 <D.2 mr/hr <0.4 0.2 5 <D.2 w /hr <D.2 <D.2 6 <0.2 mr/hr <0.4 0.6

/

<0.2 mr/hr <0.4 0.6 B <0.2 mr/hr <D.4 D.6 10 <0.2 cr/hr <0.4 0.8 I D.6 11 <0.2 w/hr <0.4 12 <O.2 mr/hr (0.5 D.k 13 (0.2 mr/hr <0.4 0.5 14 <D.2 1cr /h-- <D.4 1.4 15 Cs.2 mr/hr <0.t 0.8 16 <0.2 mr/hr <0.4 0.2 1B <0.2 mr/hr <0.2 <D.2 19 <0.2 mr/hr <0.h 0.3 23 <',2 r.-/hr <0.4 0.3 WDTE: Readings are total readimEs where possible total reading - garra

neutron Time 3.2-3 - -.

1 Sur.'cy Data i Pre-Tuel load-Open Vessel Heatup ' T Test Condition O O <55 CTP 0 0 0 MWe <100*F (100*F 540*F 1/19/85 4/16/85 7/1375 ^ Mod. Temp Date locationi Reactor Building - Second Y1oor 21 <0.2 1er/hr i <0.4 0.4 1 \\ <D.2 w/hr (0.4 0.2 22 23 <D.2 v.r/hr <D.4 0.4 24 <0.2 w /hr <0.4 0.3 l 25 <0.2 mr/hr <0.4 0.2 26 <0.2 mr/hr <0.4 0.2 25 <0.2 mr/hr <0.4 <0.2 29 <0.2 mr/hr (0.2 <D.2 30 <0.2 mr/hr <0.2-2.5 31 <0.2 w /hr <0.2 <0.2 32 <0.2 er/hr <0.2 <0.2 33 <0.2 w /hr <0.2 <0.2 i Readings are total readinEs where possible NUTE: tctal reading - ga=a + Tieutron TIGUE 3.2-3 1 j

d L Reactor Building - 2nd Floor Elevation 613'-6" ,gg,_ p hyn a,g3g,7g.;, ] t' i L.M. UQy,:.L*."1; id '.}!s;,c q,izild 2ig.,;,fim h ~~L 1 A 5E f" I- ;[ k.) ) Tj# Wa MSe a C. Heald 5' l$'f.M"N [;' "$"g T UHR ) DELAY N00M E**"8 IQ b 8!2 ~i Eactu:ngers I il ~ -~ ~ ' A+-- - lQ. d EECW ,ht'!%Com m w;., [TA ~"s,,; .g% -;,; v h g /, - lif$$', f555$$j_ ggy ' ? h 43 ~1 y e u ?< e, ,h"

IE",

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anar Pumps 3 ch,;-a.

,, - opeu O, t L: !. AREA Personnes a if !? tb j .yh s;;- ~ o Cm / / ;i j:t to .I %,,, n Aooin ~ E ") " 2' b.. b. s w D i . k.'.ht;.gpfgj> e Personnel an%,,G w a ga. Emchanger ($ d f 4 shy 2.4' i.- F AI' LOCH 1 L g ./ N X~A k ~~ Equipment ' q at tachatw[st E u +hHR Heal 4,n-t. SWMCHGEAM DOCW /, y ['l-h m ". ' 7 " jd / T i nomo s d 31 jL~6 d ~ fenma , / .wg - ,,,,,,.uc.f,.n.....u,. ~ a.., -

l Survey Data i s Pre-ruel load Open Vessel Heatup 1 Test Condition CTP O -0 <55 1 MWe 0 0 0 Mod. Temp <100*F (100'F 540'F Date 1/19/85 4/16/85 7 /13/85 i ' Location: Reactor Building - Third Floor I 1 <0.2 w/hr (0.4 O.3 2 (0.2 w/hr (0.k D.3 3 <0.2 mr/hr <0.4 0.3 4 <0.2 w /hr <0.4 0.2 '5 <0.2 mr/hr (0.4 O.3 6 l <0.2 w/hr 1 (0.2 <0.2 7 <0.2 cr/hr <0.4 0.4 J 9 (D.2 w/hr-(0.4 1 0.4 10 <0.2 ter/hr (0.4 0.4 11 (0.2 w/hr <D.4 0.4 12 <0.2 w/hr <0.2 <0.2 1 13 <0.2 w/hr <0.2 <0.2 14 (0.2 w/hr (0.2 (0.2 15 <0.2 w/hr <D.2 <0.2 k t NUTE: Readings are total readings where possible total rea:iing = ga:::ma

neutron 1

TIGUE 3.2-3 , ll 1

Reactor Building - 3rd Floor u Elevation 643'-6" $@,,6

,7d..,. g' ) 1

_m 4 !,n O 9 ' g e. contam g t.p 3 ( $.}$$""lI@ pated .Rf f 9 L,,_.s?, ? O '~ l L .=.*- %? a q) .c CONTROL ~ t .'?. 4 , s; f:.- I -N-f - g:* FUEL f,. StonAGE I.b J I I " -' ' " l. ~ N ,l j I l.f;{ll 4 ?, poot h._._. IM.f, ,I rncco n.3.. li: 7 's 'i u

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3

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u .)'1..,---v .,r.v ' I.i (a....tr s m f. 3 ..'. ri s,s. r T1,; /.I Y h e* c,. d,2 j=={rw

s

{.L 3 W s WERY i >, 2-)- gg j M.' 8!: s v'ywt nows i(s 1 h Y.fi.h [. l f_ n

=. m Y' k.

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s: CL.

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Survey Data ~ Test Condition Pre-fuel load Open Vessel Heatup CTP O O <55 MWe 0 0 0 Mod. Temp <100*F <100*F S40'F Date 1/19/85 4/16/85 7/13/85 ' location: Reactor Building - Fourth Flev ? <0.2 ar/hr <0.2 <0.2 4 <D.2 mr/hr <0.4 0.3 5 - <0.2 r:r/hr. <0.4 0.3 6 <0.2 mr/hr' <0.4 0.4 7 <D.2 mr/hr <D.4 0.3 8 <D.2 r.r/hr <0.5 0.5 i 9 <D.2 rr/hr <0.4 03 10 <0.2 mr/ fir <0.4 03 11 (0.21cr/hr <0.4 0.2 i 12 (0.2 mr/br- <0.k D.h 13 <0.2 mr/hr <0.4 0.2 a = NDTE: ReadinEs are total readings where possible total reading - Fwa

neutron F1GJEE 3.2-3.-

I I e Reactor Building - 4th Floor s Elevation 659'L6" 3 9.wr==.u w.w. m. 9 p:r. . -L_ ~pb H. g.w.; ;.g.c.>.s,1, s.ru g : Eg4 ' l 'T !, *)

Hr

_3 _._ ,a

l nwcu cena

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l Survey Tata i Pre-fuel load Open Vessel T Heatup Test Condition O <55 O CTP O O O <100*F <100*F 540*F MWe 1/19/85 4/16/85 7/13/85 Mod. Temp Date Reactor Buildirig - Tifth Floor _i Location: <0.4 0.2 <0.71ser/hr 1 <0.4 D.2 <0.2 mr/hr 2 <0.2 mr/br <0.k D.2 3 <0.2 nr/hr <0.4 0.3 4 <0.2 r.r/hr (0.4 0.3 5 <0.2 mr/hr <0.4 0.3 6 <0.2 v.r/hr <0.k D.2 7 <0.2 mr/hr (0.5 0.3 8 <0.2 nr/hr <D.2 (0.2 9 <D.2 rr/hr <0.2 <0.2 1D <0.2 ter/hr <0.7 (0.2 11 <D.2 1er/hr <0.2 (0.2 12 <0.2 mr/hr <0.2 <0.2 13 <0.2 er/hr ' <0.1 0.3 la ReadinEs are total readinES Where possible' FJTE: toteC reading = gar'a

neutron E1 DURE.5 2-3. _.

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Survey Data . Test Condition l Pre-fuel loadi Open Vfssel Heatup O O <51 CTP 0 0 0 MWe (100*F 540*F <100*F Mod. Temp 1/19/85 4/16/85 7/13/B5 l Date Location: Turbine Building - Basement (0.2 lur/hr <0.7 (0.2 1 (0.2 w /hr <0.7 <D.2 2 3 <0.2 tr/hr <0.2 (D.2 4 <0.2 r:r/hr <0.2 <0.2 i 5 <0.2 v:r/hr <0.2 (D.2 i <0.2 mr/hr <0.2 <0.2 B 9 <0.2 r:r/hr <0.2 <0.2 4 10 (0.2 w /hr (0.2 <0.2 11 <0.2 w /hr <0.2 (0.7 12 (0.2 nr/hr <D.2 <0.2 13 <0.2 r:r/hr <0.2 <0.2 la- <0.2 tur/hr <D.2 (0.2 15 <0.2 r:r/tr <0.7 (0.2 \\ 16 <0.2 cr/hr <0.2 <0.2 17 <0.2 mr/hr (0.2 <0.2 ) i Readings are total readings where possible NUTE: total reading = ga:mna + neutron T2tTJFtE 3.2-3 -

i i 4 i ) 1 Turbine Building i Basement d. Elevation 564'-0" ..u N.,. e 3 I .,s. ., w q -c ". p .. :.n n. ......,.c', .i .. e- ,,3.< i 'l g A 4y ..I, 8 4) p-g "y e. 0,9esrq &,.,.g ogn,. 4 p,yqex e,. IO _p.1,i o, i s, i[g i l Z &A y Q l ~w e m'm o e v.. ' g .g, A.m' '.*03 ' y; G.. u%<.,

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.=. t Survey Data Test Condition Pre-fuel load Open Vessel Heatup i CTP O O <55 MWe 0 0 0 Mod. Temp (100*F (100*F 540'F Date 1/19/85 4/16/85 7 /13/85 --Location: Turbine Building - First' Floor 1 <0.2 ar/hr <0.2 <D.2 2 (0.2 mr/hr <0.2 (0.2 3 <0.2 r.r/hr <0.2 (0.2 4 4 <0.2 nr/hr <0.2 <D.2 i 5 <0.2 mr/hr (0.2 (0.2 d 6 <0.2 ar/hr (0.2 <0.2 t 7 <0.2 mr/hr <0.2 <0.2 <0.2 B (0.2 mr/hr <0.2 9 <0.2 mr/hr CD.2 (0.2 10 <D.2 mr/hr <0.2 <0.2 11 <0.2 mr/hr (0.2 (0.2 12 <0.2 mr/hr (0.2 (0.2 i 13 (0.2 m.'/hr CD.2 (D.2 i 14 <0.2 Tr/hr <0.2 <0.2 15 (0.2 nr/hr <0.2 <0.2 16 <0.2 mr/hr <0.2 <0.2 17 <0.2 mr/hr <0.2 <0.2 16 <0.2 mr/tr <0.2 (0.2 NDTE: Readings are total readings where possible i l total reading - gamma + neutron l TIG'JF.I 3 2-3 l I f ~43' I

Survey Data Pre-fuel load Open vessel Heatup Test Condition O O <55 CTP 0 0 ~ 0 ~ We (100'F (100'F 540*F Mod. Temp 1/19/85 4/16/85 7 /13/85 Date ~. location: Turbine Building - Tirst Floor i <0.2 mr/hr <0.2 (0.2 19 (0.2 mr/hr <0.2 <0.2 20 <0.2 cr/hr (0.2 <0.2 21 <0.2 nrIhr <0.2 <0.2 22 <0.2 mr/hr <0.2 <0.2 23 24 <0.2 mr/hr <0.2 <0.2 25 <0.2 er/hr <0.2 (0.2 26 <0.2 mr/hr <0.2 (D.2 27 <0.2 tr/hr <0.2 <0.2 28 <0.2 rtr/hr <0.2 '<0.2 29 <0.2 mr/hr <0.2 <0.2 <0.2 mr/hr <0.2 (0.2 30 31 <0.2 w/hr <0.2 <0.2 32 <0.2 vr/hr <0.2 (0.2 33 <0.2 cr/hr <0.2 <0.2 34 (0.2 rr/hr <0.2 <0.2 35 <0.2 mr/hr <0.2 <0.2 <0.2 <0.2 36 <0.2 r-/he Readings are total readinEs where possible 31DTE: total reading - Easna

neutron TIGURE 3.2-3

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Survey Data Pre-fuel load Open Vessel' Heatup-Test Condition i O O (55 CTP 0 0 0 MWe <100*F <100 *F 540*F _ Mod. Temp 1/19/85 4/16/85 7/13/85 i Date ) l - Location: Turbine Building - First Floor 37 (0.2 w/hr (0.2 D.2 3B <D.2 w/hr <D.2 0.2 l 39 <0.2 mr/hr <0.2 0.2 40 <D.2 tr/hr <0.2 0.2 (D.2 w /hr <D.2 0.2 41 42 <D.2 w/hr <D.2 0.2 43 (0.2 mr/hr <0.2 0.2 44 (D.2 w/hr <D.? D.2 NOTE: Readings are total readings where ; usible total reading - sama

neutron j

FI L T2 3 2-3 ^4rf

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Survey Data . Pre-fuel load Open Vessel Hestup_ '! Test Condition 0 <55 O CTP 0 0 0 ' MWe i (100*F (100*F 540*F Mod. Temp Date - 1/19/85-4/16/85 7 /13/85 i , Location: Turbine BuiltiiTIE - Second Floor i (0.2 w/hr <D.2 <D.2 1 <0.2 w/hr CD.2 (0.2-f 2 l <0.2 w /hr <0.2 <D.2 3 4 <0.2 tr/hr (0.2 (0.2 I I' <0.2 sr/hr (0.2 <0.2 i 5 i I 6 <0.2 w/hr <0.2 (0.2 <0.2 mr/hr <0.2 <0.2 ) 7 B <D.2 w/hr (0.2 (D.2 (D.2 er/hr <D.2 <D.2 9 <0.2 w/hr <0.2 (D.2 10 1 (D.2 w/hr <0.2 <D.2 11 - (D.2 w/hr <0.2 <D.2 12 (0.2 w/hr <0.2 (0.2 13 1h <0.2 tr/hr <0.2 <0.2 0.2 t.r/hr <0.2 (0.2 15 16 <0.2 rr/hr <0.2 <0.2_ __ l <0.2 tr/hr <0.2 <0.2 17 1B <0.2 tr/tr <0.2 <0.2 l ReadiTIEs are total readings where possible EDTE: total readinE = Ea a + neutron l I n F. EE 3 2-3 Z.

Survey Data i Test Condition Pre-fuel load Open Vessel Heatup CTP O O <55 MWe 0 0 0 Mod. Temp <100*F (100*F 540*F Date 1/19/85 4/16/85 7/13/65 Location: Turbine Building - Second Floor 19 CD.i w/hr <0.2 <D.? 20 <D.2 tr/hr (0.2 CD.2 1 <0.2 tr/hr <0.2 <D.2 21 22 <D.2 tr/hr <0.2 (0.2 23 <0.2 En /hr CD.2 (D.2 <D.2 w/hr <0.2 <D.2 24 25 <0.2 r.r/hr <0.2 <0.2 26 <0.2 w/hr <0.2 <D.2 27 <0.2 tr/hr <0.2 <0.2 26 <0.2 tr/hr <0.2 <D.2 29 <0.2 mr/hr <0.2 <0.2 30 <0.2 w /hr <D.2 (0.2 31 <0.2 tr/hr <0.2 '<0.2 32 <0.2 tr/hr <0.2 <D.2 .NDTE: F.eadings are total readings where possible 1.otal reading - ga==a + neutron T1D'Jhi 3 2-3 -2 3-

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, Survey Data .) Test Condition Pre-fuel load! Open Vessel Heatup-CTP O O <55 We 0 0 0 Mod. Temp <100*F (100*F-540'F 1/19/65 4/16/85 7 /1 3/85 Date Location: Turbine Building - Third Floor 1 <0.2 er/hr <0.7 (0.2 2 (0.2 nr/hr <0.2 <0.2 1 '<0.2 tr/hr <0.2 <0.2 1 3 4 <0.2 mr/hr <0.2 <0.2 5 <0.2' er/hr <0.2 <0.2 6 <0.2 rr/hr <0.2 <0.2 7 <D.2 r.r/hr <0.2 <0.2 B <0.2 mr/hr <0.2 <0.2 9 <0.2 tr/hr <0.2 (0.2 10 <0.2 er/hr <0.2 <0.2 11 <0.2 er/hr <0.2 <0.2 12 <0.2 tr/hr <0.2 <0.2 13 <0.2 tr/hr <0.2 (0.2 14 <0.2 tr/hr <0.2 <0.2 15 <0.: tr/hr <0.2 <0.2 16 <0.2 er/hr (0.2 <0.2 17 (0.2 tr/hr (0.2 <0.2 1E <0.7 tr/tr <0.2 <0.2 N3TE: P.eadinEs are total readings where possible total reatinE = Eansa + neutron FIGUFI 3 2-3 ~ i Survey Data Pre-fuel load. Open Vessel Heatup' 1 Test Condition O 0 <55 CTP 0 0 0 MWe (100*F (100*F 540*F Mod. Temp -1/19/85 4/16/85 7/13/85 Date Location: Turbine Building - Third Floor 19 <0.2 mr/hr <0.2 (D.2 20 <0.2 mr/hr <0.2 <D.2 21 ~ <0.2 er/hr <0.2 <0.2 22 (0.2 nr/br <0.2 (0.2 23 (0.2 nr/hr <0.2 __[ <0.2 24 9 <0.2 wr/hr <D.2 (0.2 J 25 <0.2 nr/hr <0.2 <0.2 <0.2 <0.2 I <0.2 nr/hr l 26 (0.2 <0.2 <0.2 mr/hr 1 27-28 <0.2 rn /hr <0.2 <0.2 l 29 (0.2 er/hr <0.2 (0.2 /~ 30 <0.2 * /hr <0.2 <0.2 31 <0.2 mr/hr <0.2 <0.2 32 l <0.2 tr/hr <0.2 <0.2 33 <0.2 Er/hr (O'. 2 (6.2 3t (0.2 er/hr <0.2 <0.2 35 <0.2 nr/hr <0.2- <0.2 \\ r' l 3E (0.2 n /hr <0.2 <0.2[ l ETE: Readings are total readings where possible 4 total reading - gamma + nemron TFRI 3 2-3

Survey Data Pre-fuel load ! Open Vessel ' Heatupj ' ' Test Condition O O <55 CTP 0 0 0 1 We <100'F (100*F 540'F Mod. Temp 1/19/85 4/16/85 7/13/85 Date Turbine Building - Third Floor ~ Location: <0.2 w/hr <D.2 (0.2 37 38 (0.2 w/hr <0.2 (0.2 <0.2 mr/hr <0.2 <0.2 39 <0.2 r:r/hr <0.2 (0.2 to 21 <0.2 er/hr <0.2 <0.2 ) 52 <0.2 mr/hr <0.2 <0.2 43 <0.2 nr/hr <0.2 <0.2 Readings a*e total readings where possible EDTE: tctal reading - ga=::a

neutron FIGUEE 3 2-3

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l ] 33 Fuel 1.cading 331 Purpose The purpose of this test. aan to load fuel asafely and - efficiently to the Sull more size 1764 assamh11es). 332 Criteria . Level 1 ' The partially Ioaded core isustine 'suberitical tuy at least 0 38 percent delta k/k with the analytically determined strongest rod fully withdrawn. There rust' be a neutron signal wout-tenoise count ratio 'of at least 2:1 on the required operable SRMs or fuel loading chambers.(FLC). The minimum count rate, as defined by the Technical Specifications, must be r.et on the required operable SRMs or fuel leading chambers. 3.3 3 Results Prior to fuel loading, all fuel assemblies were inspected and then stored in the fuel pool in such a way that no rotation of fuel assemblies would be required during their transfer to the reactor wessel and also that no assembly would pasa. over any other The assembly in th'e fuel pool during fuel loading. only exception to this asas handle LJK 954 mhich was oriented 'SW instead of SE in the fuel pool, but was Before verified to be properly oriented in the core. the start of fuel load, all control rods were fully inserted, all blade guides were positioned as shown on F.igure 3 3-1. Seven sb-Be neutron sources were installed at locations shown mn Tigure 3 3*1. All applicable initial conditions were verified prior Four times during the to the start of fuel loading. fuel loading process, fuel loading was suspended for greater than eight hours, and all applicable initial conditions were reverified before fuel loading was resuned. The Bottom head drain te=perature indication was used to obtain the Reactor Coolant Temperature at least once every eight hours (3 15 minutes) during the fuel loading process. ~55 l

Detailed fuel loading sheets, approved by the 1teactor Engineer, provided the instructions on each' Individual fuel assembly to be moved from a specific location in the Yuel-pool to a pre-aaalgned location in the core. It also provided the. instructions on what control rods vere to be exercised for functionalmd muttwiticality checks for pre-defired tore configurations. '1LC 1 moves to be made during the fuel loading were also included. Most of the changes required to the fuel loading.aheets ~ during fuel loading were to inove the TLCs earlier due to high count rates experienced when fuel assemblies and/ar the neutron sonroes were too misse to tJte FLCs. The only other change involved using contral rod 10-27 (instead of 06-27) for a subacriticality check due to an accumulator problem with rod D6a27. Four FLCs (one per quadrant) were used to 1 monitor the count rate from the start of fuel loading up to the In point when 532 bundles were loaded in the core. order to keep the TLC count rate within a desirable range and to accommodate an increasing core size, it was necessary to move the FLCs outward by approxittately one cell routinely as fuel loading progressed. The location of FLCs was selected to ensure that each quadrant of the core was adequately monitored. (See Figure 3 3-Ji) ~ The upscale alarm setpoint was set at 1 x 105 cps and the upscale trip a=+Fnt was met at 2x 105 cps for each TLC. The downscale rod block metpoint was 3 mps. i The TLCs were checked for flux response either by control rod pulls during scheduled nutr criticality checks or by lifting the FLCs partially out of the core. These flux response checks.were made at least once every eight hours during fuel loading and prlor to the resumption of fuel loading when fuel loading was delayed for eight hours crincre. In addition, the Signal-te-Noise ratio was calculated for each FLC paict to start of fuel load, durinE any required reverification of plant initial conditions and every time the FLCs were moved to a mu location. (See Figure 3.3-2) Four SRMs (one per quad ant) were used to scnitor the neutron count rate starting from the point when 532 bundles were loaded in the core to the completion of fuel load (764 bundles). With the SRM detectors connected to the SEM Anstrument channels, the rod block 4 and'the upscale trip setpoints were set down to 1x10

and 7x104 respectively, since no previous saturation The down test was performed on the SRM detectors. The SRM flur scale rod block setpoint-was 3 cps. response check was ' performed at 12ast ~cnce every eight hours.iiuring the fuel loading process by partially withdrawing each SRN. Tuel-loading co=nenced on Marc. 2D 1935 with, the loading of four fuel assemblies around the central The loading continued in control cell neutron source. units that sequentially completed each face of an increasing square core, loading in-a clockwise direction until a 12 x 12 square was completed with The thirteen contro3 symmetry about the' center source. cells (52 bundles) needed to form a 14 x 14 square array of bundles around the center control rod (30-31) were loaded next. The remaining control cells were loaded, one on each face at a time, in a clockwise manner, such that the core was rotationally syt:metrie (See after every four control cells had been loaded. Figure 3 3-3) Control rod functional and sub-criticelity checks were performed either after levery cell (first 4 cells in the core), or af ter every two or four cells as dictated by the detailed fuel loading sheets. The purpose of the sub-criticality checks-was to ensure that it was safe to load the next control cell (s). For each bundle a visum1 verification uns m fm ad to ensure that the bundle was properly grappled before the bundle was lifted frcxn the fuel pool racks, that there was adequate clearance on all sides while the bundle was being moved to the reactor cavity and that it was loaded in the core in the proper location with the proper orientation. Also, physical verification was made of the fact that the bundle was ungreppled before the hoist was raised. Similar verifications were made 1"or the blade Euldes lifted out of the core and the C - toves made. during the fuel loading process. f A t!ay-by-day account of the fuel load progress is given in Figure 3 3-5. Most of the problems that caused i delays were related to the refueling bridge (limit switch, power loss, grapple indication, air hose ' break, etc.). Tuel Icading was halted on Sundays in order tc perform required weekly surveillances on ELE /SRMs, IEP.s, APRMs and the refueling bridge. 4 6-

Dur.ing the fuel loading process, TLC /SRM count rates were wonitored periodically and 1/M calculations were performed and plotted for each FLC/SRM and for the average of the four FLC/SRMs (See Figure 3 3*6). The-average 1M plot was used to project the est,1 mated number of bundles for criticality. 'If criticality was Trojected during the next loading increment then the .~1ncrement size was reduced between 1/M calculations. i Strong geometric effects were seen, particularly during f the first few bundles loaded in the core and also when These geometric the tundles were loaded near an FLC. effects resulted 'In erronious (but highly conservative) projections which often resulted in very small incre=ent sizes (1 ^ 2 bundles) between 1M j After eighty bundles were Iceded in the j calculations. core, the maximum increment size between 1/M calculations 'was reduced to one cell (4 bundles except i for the peripheral locations where a maximum of.five bundles were loaded between 1/M calculations). Bundle LJK 677 was identified to have a rusted channel Some debris was fastener that had to be replaced. identified in the core on bundles LKJ 398, LJK 506 and Af ter fuel loading was completed, these LJK 957. bundles were pulled out of the core to correct the respective problems and reinserted back into the core. After the 12 x 12 squaae array of bundles was co=pleted, a partial core shutdown margin (SDM) demonstration was performed by withdrawing the analytically determined strongest rod (26 - 27) and a diagonally adjacent rod (22 ^ 23) out of the core. ..Sub-criticality with these two rods withdrawn demonstrated that there was at.least a D.385 delta K/K shutdown margin for the existing core configuration. 'Because the calculated Keff for the 12 x 12 array with the two rods withdaawn was 0.9758, and the calculated f 41 c6re with only the strongest rod Keff fcr tr.e withdrawn is 0.97, sub-criticality for the partial core demonstrated that the shutdown margin would be cet l throughout the re=aining fuel loading process. The fuel loading was completed after fifteen days on Apall 4, 1955. All criteria were satisfied.

~ l l l TEUTEDN SOURCE LOCAT1DN AND BLADE CUIDE ORIENTATION j , PRIOR TO FUEL LDADING .m l 7l [ l ~~~ ~ ~ ~ . sis ~ ~ 7" ) ] f - f ~ ~ 2 7F7F 7F 7'# # 7F 7FM N< ~ ? 71 7F & K&N 7.'m r 92 M7(ii! 7'- 7 "--MM N 2' ztzt ME f rf M NE#A g Ng z$# # zfr #4 6ff#d .~ 7t ~ ~ ~~~ v_7,4,l8 w v v v-mz,?S x v v % ll l fJ fi L l* ll /s /, ft - $ ?+$ $ Yh 7f Ysfi$A? $ $ $k7 22 #z f 7f # 7 # # 26f77j K M E M,P& M E& f f E#;Mh F N# f ff/H @eef (7,jf '> -- + 7 h t e# r # fff u b -F cg: 7Fzf F f z6zf # Ed = > -+t',, ', #7(t# # # # #H i c ,\\ \\ l l \\ \\ l I l l 1 02 D6 10 14 IB 22 26 30 34 38 42 46 50 S

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_ Signal to Weise Meastrrement C D

OF BUNDLES y DATE

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~70 T74 AB (D34D) 03-22"B5 11t' E.B 16 3.B 9.8 6.5 64 6.D 5 96 (2005) 95 ~ -+ 7.0 3.4 03-22-85 TLC' (2227) 144 + + D3-23-B5 FLC* 5 4 12 11 (2110) 03-25-85 EC 10 19.0 11 14.7 12 19.0 12 14.0 156 (1420) 195 D3-26-B5 TLC

1D 49.0 20 69.9 (002D)

D3-26-B5 TLt* 3B 1 69 32 759 AD 159 4.B 15 -260 (1915) D3-28-85 TLC

30 99 4

39 35 116 2.5 7 3 3BB (1116) D3-29-65 FLC 300 9?? 100 999 150 3 14 90 299 440 (0937) D4-01-B5 SF.M 16 15" 12 119 AD 399 .15 149 532 (1528) 'S/N Ratics tttained dring TLC toves -TLC not :Oved Fig r e 3 3-2 ~59-

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yally Tual Loading Progress B"NULTS LDarED CDMMENTS DATE DAY TO DATE Tuea inad at.arted at 2130. E 85 4 4 Rod 31ock 2131t switen inalrmetion. ~D3-21-55 32 36 53-22-85 62 98 V3-23'B5 58 156 ' Weekly surveillance on SRMs, 2RMs, AFF" 03-24-B5 D 155 Refueling Bridge. Tuel Acad resu:: red at 1500. D3-25-E5 3B 196 .03-26-B5 B2 27B D3-27-E5 B4 362 U3-2B-B5 76 43B Transformer #64 lost due to initiation c. D3-29-B5 66 504 its deluge Ifire protection) systen. 0400 rarme2 bridge power cable proble:. D3-3D-B5 2B 532 Cable cut and re-termed to restore the systum. weekly surveillance. YLD to SRM switcho'. D3-31-B5 0 532 Tuel load resmed at 2000. D4-D1-B5 14 546 04-D2-85 74 620 Air hose damaged when stuck center sert:< D4-D3-B5 45 665 of the 7tast was released and dropped. Fuel load completed at 2350 'OhD4-B5 96 764 Tigure 3.3-5 ^ b2-

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FUEL LOCATION VERIFICATION Wi s %ggggy-4/Mt4RRMF+ vrMMM M *S MF- ~ i tstfRMM$bl&- 1MRR MMt4bt> iIMMMMMMM &P lMRMMM!M MF- ,i MRMMMMM&; MEMMRAAMP MRM14MMMF-4 m$ Y' $ aaar iif7MRRP .:. + + + e > ~+ -: nse 3 2-7 - ]

FUEL LOCATION VERIFICATION y = ?- M 7 - T '.*c .n 3-e i , a M MRR T i. ~ 3:M NMMR&i 4RMMMMR d ne a agy:aaug 4tM MMMMNR et & MQ4555 g MMMM M M M R' 'dRteMMRMW -itM MSM4W:2 d RMMMMMR! --QR RN5?gMM 4%MMMM* "MIRMMMt i wa+ y -t e n gsk. . ka = 1 1 F1rre 3 3-7 1 --

3.4 Full Core Shutdown Margin 3.4.1 Purpose The purpose of this test is to assure that the reactor will be suberitical throughout the first cycle with an,- single control rod fully withdrawn and all sther rods fully inserted with the more 'in its maximum reactivity . state. 3.4.2 . Criteria Level 1 The shutdown margin of the fully loaded core with the analytically determined strongest rod withdrawn must be at least D.38 percent delta k/g plus R (an additional margin for exposure) where R - U.S percent delta k/g. Level 2 Criticality should occur within + 1.0 percent delta k/g of the predicted critical. 3.4.3 Pesults The fully loaded core was made critical by withdrawing control rods following the B sequence, using the Reduced Notch Worth Procedure. This sequence contained the analytically strongest Tod D6-39, which was fully withdrawn before reaching criticality. Prior to performing the shutdown margin demonstration, as Tequired by Technical Specifications, the shorting links were removed to put the Reactor Protection Syster in the non-coincidence Jcra= mode. The point of criticality was demonstrated by withdrawing control rods following the order given in f the rod pull sheets until an (approximate) 300 second period was observed with Group 3 rod 18-51 withdrawn t notch position DB. Moderator temperature was recorded at 96*F. Later, with moderator temperature still at 96*T, the reactor was then made supercritical by withdrawing control rod 10-43 to position 08. SRM A,B,C and D measurements were taken every 30 seconds for three and one half minutes. Period analysis was perforced by fitting the data linearly on a semi-loE l plot and measuring time to increase one decade free which period was calculated. The average period was . determined to be ~l6.5 seconds. _

ihe shutdown margin nf the fully loaded core at 6B'F with the analytically strongest rod withdrawn was zietermined to be 2.725 delta k/k. 1,evel 1 crlieria were satisfied since the lseasred shutdown margin was 0 385 - 0.885 a k/k where R la defined larger than R + here as the analytical difference in shutdown margin (cold) at the most limiting point.in the. cycle and Beginning of Life - af the core. The difference in keff between the theoretical critical configration and the actual measured witloal This configuration was found to be 0.285 A k/k. satisfies Level 2 criteria since criticality occurred within 15 h k/k of the theoretical.witical sigenvalue.

35 Control Rod I)tive system 351 Pur;:ese Tach control Tod drive (CND) was' tested to 1ceasure insert / withdraw and scram times an1 friction dP levels This was done to in the CRD hydraulic system. demonstrate that the CRD system operates properly over the full range of Trlmary coolant temperatures and pressures. 3 5.2 ._ Critersa Level 1 Each IRD aust have a normal withdrawal speed less than or equal to' 3 6 inches per second, indicated by a full 12 foot stroke in greater than or equal to 40 seconds. time of all the operable CRD's with The 7nean scra functioning accumulators must not exceed the following times (scr'am time is measured from the time the pilot scram valve solenolds are de-energized). Position Inserted Scram Tire TSeconds) Trc-Fully Withdrawn 0 358 46 1.0% 36 1.560 26 3 419 6 ine mean scram time dY the three fastemt IRD's in a two-by-two array must not exceed the following times (scra. tire is censured from the time the pilot scra valve solenoids are de-energized). Position Inserted !!econds) _Frc, Fully _ Withdrawn ,Scra-T i-a 0.379 46 1.161 36 1.971 26 3.642 6 Level 2 Each CRD nust have a normal withdrawa! speed less tha:. c-equal to 3.6 inches per second Indicated by a full 12 foot stroke in 40 to 60 seconds. f ^68-

If the differential pressure variation exceeds 15 psid for a continuous drive-in, a settling test must be ~ In this case the differential settling pressure should not be.less than 30 Tsid, nor shou 115 it performed. vary by more than' 10 psid over a ' full stroka. 352 'Results Insert / withdraw timing, friction testing, und scram timing were performed on the CRDs at the conditions specified.in Figure 3.'F'1. All of the individual contral rods were~3aram time tested, friction tested and insert / withdrew timed Adjustments tc during the Open Vessel test condition. some CRDs had to be done in some cases to bring Dur.ing insert / withdraw times into acceptance limits. the friction testing, no pressure differential measurements exceeded the criteria of 15 psid and me The four slowest settling tests had te be performed. rods in each sequence were also scrammed at reduced All test criteria were accumulator pressure. satisfied. Heatup, the four slowest rods in each sequence UPon' Durir.E were scram timed at 600 poig and at 800 psig. reaching rated temperature and pressure conditions, all The eight slowest rods CRDs were scram timed. determined during Open Vessel.nd Heatup testing were then insert / withdraw timed, friction tasted, andYigure 3 5-2 scrar.:med at reduced accumulator pressure. shows the average scram time of the eight slowest rods, four in each sequence, at various reactor pressures compared to the maximum per:Issible..__ _ +y-<

Tha specific results from our rated pressure testing are as follows: Mean Ecram Times Rod Position 46 g 26 06 Mean Scram Time'for ' O.302 D.852 1.398 2.501 all BB ' Seq. B reds (sec) Mean Scram Time for D.2BB 0.802 1.340 2.436 all 97 Seq. A rods (sec) Mean scram Time for D.295 0.826 1.368 2.467 ALL rods, Seq. A and Seq. E (sec) (core average) Mean Scram Time of the 0 325 0.900 1.481 2.655 l 3 fastest CRDs in a two-by-two array for ALL rods, Seq. A and Seq. 3 (core average) 1 I ) ^7D-

CDNTEDL-PDDDRIVE SYSTEM TESTS Reactor Pressure With Core Loaded Test Accumulator Preop, psig .DescrJption Pressure Tests _ 0 600 800, Rated All All Position Indication 4(a) Ecrimal Stroke Times All All 2neert/ Withdraw A1.1 All Ioupling %(a) All Friction Scra: Worcal All All 4(a) 4(a) g13 4(a) 4(a) Scra: Minimu 4(a) Scram 2ero Scram (scram 5creal discharge volu-e hiEh level)(c) 4(b) Scram Normal Refers to four CRDs selected for continuous monitoring based on slow a. normal. accumulator pressure scran times, or unusual operating characteristics, at zero reactor pressure. The four selected URDs zust be co.patitle with rod worth minimi.zer, RSCS systems, and CEL sequence requirements, t. Scram times of the for slowest' ERDs will be determined at Test Conditions 1 and 6 during planned reactor scrams. c. T r.a scre-discharEe volu-e fill tir.e will be determined at Test Conditions 1 and 6 during planned reactor scrams. Single CRD scra.s should be performed with the charging valve Note: riosed (do not ride the charging pump head). T1gure 3 5-1 ~71-

Seram Testing Restdts 'l - 9 I*i $! l..;...d..._.M ! serk ita. s!sa.e vs.du=

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.4 . -.. :..= .....i -::-:..! ? _-.. - - -,....:.3. -. i-....: .u c.2.[.. etwel.....;i._. 3; 4.. . 6, r,, g.:. .=_ 5 _e .. a.. ;_.: ..A,. u F 4D . J. 4.. ... y . _.m - _ =- _.3. _2 _ 3, e .. g = _ _ _ __.....E .....0. 20D 400.....600 SK .. 2. 6(v se,c. ten _f.I+mn bsul I i . ~.. .i. I. I, I I i 'E1GURE 3.5-2 3.6 Source Range Monitor Performance and Control Rod Sequence Exchange i 3.6.1 Purpose The purpose UT this prence was to itemonstrate that .Sourne B unge Monitor (SRN) Jumat-traoise ratio, ininimum count r. te, and saturation requirements are satisfied and to wonitor Intermediate Range Monitor (IRM) readinEs dur.ing heatup. This procedure was performed successfully. j 3.6.2 criteria There must be a neutron signal countdo noise count ratio DT at least 2:1 Dn Lbe required Dperable SRMs. ' There must be a minitra count rate as defined by Technical Specification on the required operable 3RMs. 3.6.3 Results Prior to the initial criticality in sequence 'B, the count-to noise ratio for SRM ( A, B, C and 'D) were 43, 149,199 and 49 respectively. These ratios were well above the Level 1 criteria of 2:1. The minimum counts on the 3RMs (1, B, C and D) were 2D.15, 40 and 15 ces Taspectively. These were well above the minimum 1,evel 1 crite-ia ret;uire.1 of D.7 cps. 3RM readings we-e also taken periodically durinE initial celticality in both sequences and 7E" readir.p Altre obtained Cring the initial heatup.in sequence h. All test criteria were satisfied. l ~73*

I 37 IRM Perfername 3 7.1 Purpose The purpose Vf this test is to verify that the Intermediate Range Monitor (IRM) System is functional try increasims the neutron fluz in the reactor following initial criticality, until all eigttt thannels come on scale with significant everlap to the SRM's and to subsequently ensure adequate overlap with the APRM syste. 3.T.2 Criteria level 1 Each IRM thannel nust be on scale before the JSRMs exceed their rcn h1Dck setpoint. EDTE: Tech Spec table 4.3.1.1-1 requires 1/2 decade overlap between the SRMs and IRMs. Each APRM must be on scale before the IRMs exceed their rod block setpoint. NOTE: Tech Spec Table 4.3.3.1-1 requires 1/2 decado of Dyerinp 21uring a controlled shutdown. 3.7.3 Fesults During the initial criticality, all IMM's wami TRM D 4 showed response prior to the SRM's reaching 5 x 10 cps. 3RM D was repaired and tested satisfactorily at a later date. Range 6/7 overlap calibration was also co:tpleted for each IRM, except JRM G which was reading errati cally. This IRM was replaced and retested successf ully. The second reiteria could not be verified but will ta done upon reaching higher Pcwe* levels in Test Condition 1. ~74-

3.'S IFRM Calibration 3 8.1 Purpose The purpose of this test in to verify LPRM response to flux changes and proper LPEM connection to neutron monitcring electronics and to malibrate the.LPRM's.to their calculated unives. 3.B.2 . Criteria level '2 Each LPRM reading will be within 1D percent nf its calculated value. 3.B.3 Tesults This test was performed while the 1teactor was at rated pressure in the heatup. test condition, in conjunction with serem time testing. Specific control rods were selected to be used for flux response checks based on The withdrawal of their proximity to the LPRM strings. these rods from position 00 (FULL IN') to position 48 (FULL OUT) was observed in terms of the LPRM flux respcnse as the rod was withdrawn past each of the fou-All 172 LPRMs LPRMs for the associated LPRM string. (43 LPRM strings with JI LPRMs per atring) were observed, using Brush recorders and STARTREC System for flux response. lnitially, as fina response asas observed on 25 of the 172 LPRMs. For the LPRMs that showed flux response, the proper nrder nf the LPRM response (D, C,'B, A) was observed. 1)uring supplemental testing, it was found that some LPRM detectors were connected in reverse order and these were corrected. One detector was found damaged and had to ba repaired. At the present time four LPRMS located on the par.iphery of the core have failed to respond. These peripheral detectors are expected to respond during' further testing at higher power levels im' Test Condition 1. Also at that time, the LPRMs will be calibrated to the Level 2 criteria. l.

Average Tower Range Monitor Calibration 39 3 9.1 Purpose The purpose of this test is % ca13trate the A?ftM syr; tem. 3.1.2 'L'riteria level 1 In the startup mode, all APRM channels must produce a scram at less than er equal to 15 percent at rated l thermal power. The APRM channels must be calltrated to read equal to, or greater than the actual cerc the 1erl pwer. Recalibration of the ATAM system is not necessary frw. a safety standpoint if at least two APftM channels per P.PS trip circuit have readings greater than or equal to core power. l Leval 2 Jf the above criteria are satisfied, then the A?RM channels will be considered to be reading accurately if they agree with the heat balance to within (47, eD) percent of rated power. 3 9.3 Results During heatup, each APRM channel was me11brated to read greater than or equal to a manual calculation of Core Thermal Power based upon a constant heatup rate The APEM scram trip setpoints were also analysis. adjusted to produce a scram at less than 155 of rated 1 The level 1 criteria was satisfied. power. The Lovel 2 criteria could not be proven as reacter power and heatup rate decreased af ter completion of tr.e Since the adjusted power readings wert heat balance. Iar below 75, this criteria is not a major concern at j l Itu power levels, and will be applied to APRM testing for Test Condition 1 and beyond.

79

3'10 Process Corputer ~ 3 10.1 Purpose ' The purpose of this test was to perform the Transversing 2n-Core Probe (TIP) acid aligtanent and Static System Test Case (SSTC) af ter anstallatlon of the 72P System. 3 10.7 Criteria Level 2 The told TIP alignment matting and the Static System Test Case (SSTC) testing is completed by the satisfartory performance nf this test procedure. 3 10.3 Results The TIP Syste J consists of five identical probes used to 1teasure and record the axial neutron flux profile at 43 radial core locations. The recorded information As used by the Process Coraputer to calibrate the fixed in core Local ' Power Range 91onitors. Emeh probe da driven into and withdrawn from the core by its associated drive techanism. In order to operate muttamation11y, the TIP drive control units must be programmed with the probe position at top and bottom of the more. These top and .tottom limits are programmed and serified du the T2P cnid alignment. This portion of the test uma performed successf ully by hand cranking the TIPS to the top of the core and sett1DE the tore limits based on the resulting position readings. In Drder to follow and read data from the TIP sachines, the Process Computer must receive position information and flux siEnals fror. the TIP Systen. Tnis interfact is tested in the SSI", ty running the TIP machines ir. various conf 1Eurations and verif ying the proper responses on the Process Computer. The Static Syste-Test Case had twn objectives: verification of the proE"am logic and checkout of the T2P interface. The first objective was successfully achieved, but the TIr interface checkout was unsuccessful due to a desigt, error in the TIP Syste. that resulted in the loss of TIP position indication. The TIP interface probler will be corrected and the cr.iterla u111 be verified it, the Test Condition 1 testing. ~~l T* .I

3 11 RUIC System 3 11.1 Purpose The purpose of this test war-to set up the TCIC control system and to verify the proper operation of the Reactor Core Isolation Cooling (RCIC) System adtile :he reactor As at, Jrated prASSre S0nditinea. 3 11.2 triteria i.evel 1 The averaEv pump discharge flow launt be equal to or greater than the 100-percent-rated value after 50 seconde have elapsed from initiation an all auto starts at any reactor pressure between 150 paig and rated. - With pump discharge at any pressure between 250 psig and 100 psi above rated pressure, the required flow is 600 gpm. (The 100 psi is a conservatively high value The measured value may be used if for line losses. l available). I The RCIC turbine shall not trip or isolate during auto l or nanual starts. Level 2 To provide a cargin on the overspeed trip and isolation, the first and subsequent speed peaks un ttee transient start shall Tsot emoeed ttie rated speed af the TICIC turbine by more than 5 percent. ) For small speed or flow changes in either manual or automatic mode, the decay ratic of each recorded RCIC system variable must he less than 4.25. The turbine gland seal condenser system shall te capable of preventing steam leakaEe to the atmosphere. The delta P switch for the RCIC stes:n supply line high-flow. isolation trip shall te adjusted to actuate at 300 percent of the maximum raquired steady state flow, with the Reactor assumed to be near the in essure 1 J f or main relief valve actuation. l '7 6"'

O 3 11.3 Results J .? l I)uring the Heatup Test Condition, the RCC pump suction ) and discharEe was.11ned-up in a closed loop with the condensate storage tank. The system was subjected to j negative tnd. positive 101 step changes in flow at aystera floas nf 600 sps and E10 gym 4aming both a step Minimum flow Benerator und the RC1C flow montroller. data was (1so taken at a speed of 200D. rpm and a RCIE i quickstaM was performed. The BQC system was able to supply 600 gym at a . discharge pressure of 1140 palg in 35 seconds when autmatically started using 91101 mig steau from the 10 vessel. The 172 time delay relay was set down froc: Sec to 5 see to prevent the RUC turbine fras coasting i i down excessively before the opkning of the Steam Admission Valve, thus reducing the experienced transient. The RCIC turbine siid not isolate or' trip during the auto and manual starts. In additio. there i were no RCIC turbine speed peaks or oscillatioc la RCIC system' varlables in the transient testing. The ROIC system was lilso subjected to an extended run at rated flow conditions. RCIC performed satisfactorily with all system te=peratures stabilized r f below alarm levels and a negative pressure maintained on the gland seal condenser system. All Level 1 and Level 2 criteria were satisfied except the 7tGC stean supply taigh flow ist!1rtion trip setting. This will te verified later after engineering nodifications to the instraent lines are completed. l i I l ,e -7 9- . _ _ _ _ _. '. L',, _ _

3 12 WPCI System 3 12.1 Purpose The purpose of this test was to verify proper 19eration of the HIEh Pressure toolant Injection (HPCI) system with the. reactor operating at rated presswe. 3 12.2 Criteria Level 1 'The averaEe pump discharge flow must 1e equal-to ar greater than the 100" percent-rated value after 25 seconds have elapsed from initiation on all auto starts lit any reactor pressure betmeen 150 peig and rated. With pump discharEe at any pressure between 250 psig and 100 psi above rated pressure, the flow should be at least 5000 gpm. (The 100 psi is a conservatively high value for line losses. The measured value may be used .if available). The HPCI turbine shan not trip or isolate during auto or manual starts. Level 2 The turbine gland seal condenser system shall be capable of preventing stea= leakage to the atmosphere. The delta P switch for the RPCI stesa supply line hiEh flow isolation trip shall be adjusted to actuate at 300 percent of the maximum required steady-state flow with the reactor assu~.ed to be near main relief valve actuation pressure. Tor small speed or flow changes in either manual nr autor.atic mode, the decay ratio of each recorded HPCI syntes variable must be less than D.25. Io provide a targin on the overspeed trip and isolation, the.rans.ent start first speed peak shall not come closer to the overspoed trip than 15 percent of rated speed, and subsequent speed peaks shall nct be Feater than 5 percent above the rated turbine speed. B0- ,,._ss_.

'3 72.3 1Results TollowinE setup of the control system, initial coupled ' turbine performance runs were per of rmed on the HPCI systes.- Eynamic stability checks were conducted with the HPCI pump suction and discharge lined-up in a closed Inop with _the EST ii. which 500 sps flow step changes were inanually and autr=atically introduced by the flow controller with HFCI system ficus at 5000 spe .and 2700 gpm. Iha*ing the automatic initiation testing of 'EPCI, a discharge flow of 50D0 gym was reached la 23 4 seconds. Twenty-five seconds af ter the automatic initiation HPCJ flow had reached 5310 sps at a discharge pressure of 1140 psig.193 psig greater than reactor pressure. HPCI did not trip or isolate durinE any manual or automatic starts. ' There was also adegaate margin on turbine speed peaks and oscillations of system An extended run was also performed in which variables.. syste:n temperatures stabilized at acceptable levels and the gland seal system performed satisfactorily. All Level 1 and Level 2 criteria are satisfied except the steam supply ' isolation trip setpoint. This will be verified later af ter engineering modifications to the .instrinnent lines are completed. 4 r f a 1 s -

3'13 select Process Teeperatures 3.13.1 Purpose The purpose of this test is to russe the seasred bottom head drain temperatre morresponds to the tottom head coolant temperatre and to verify that temperatre stratification in the bottom head region is avoided by p*oper setting of the Recirculation low speed airiter. In addition, the calibration of the reactor water level indicators was checked as well as the temperatre of the reference legs. 3 13 2 Criteria Level 1 The reactor recirculation pu=ps shall Tiot te staMed nor flow increased unless the coolant temperatures between the stea dome and bottom head drain are within 145'F. Level 2 Tra ca ricence between the actual reference lee temperatu c and the value assumed dring initial calitration shall be Isss than that amount which would result in a scale end point error nf 15 of the .1nstrument. span. 3 13.3 Results The coolant temperatres Treasred at 305 Recirculation pr.; speed satisfied the Level.1 criteria. The instability of the recirc. speed controller that ecorred delng this test precluded an effective invest 1Eation of the stratification phenomenon at low flows. The test also allowed setting of the low speed 11riter based on flow contrc11er variations of

25 of rated speed. Flow controller variations of
SI were experieneet prior to stratification so the test was ter=inated. Further investigation of the 106. flo -

region will be conducted dr.ing the completion of remaininE heatup testinE. ~ B ?-

Testing of the level instrumentation accuracy showed that scale end point errors when actual drywell temp.ratures and assumed calibration temperatures were i compared were 0.7085 D.5545,1.05075 and D.3205 fr wide range (Div. I). ad14e range (Div. II), narrow range The (Div. J) and narrow range (Div. II). Tespectively. slight Level 2 criteria violation for Tiv.1 marrow ranee level instrumentation was found ac However, subsequent to the performance of this test, a drywell cooler that was out of service at ~ support beams has been added which should improve on As such, this data will be the previous results.re-taken during the remaining heatu appropedate analysis redone. -B3+

3 14 System Expansion 3 14.1 Purpose The purpose of this test as to verify that. selected plant piping systems are i aee and unrestrained with regard t31heranal'.==Pa=4&n. 3 14.2 criteria 1 Level 1 7he seasured damp 1mewments ut the $nstroented locations shall be within the greater _cf the specified 0.25 allowable tolerance of the calculated values, or 3 inches fw the specific points. There shall be no obstruction which w111' interfere with the expected thermal expansion tf the piping system. Ilectrical' cables shall te able to accommodate expected thermal expansion of the piping system. h=L unentation and tranch piping can accomodate expected thermal expansion of. the piping system. The constant hanEer shall not be bottmed or topped out. The sprinE hanger shall not be bottmed Dr topped cut. The snubber shall not be bottmed or topped nut. Leve', 2 The measured disp 1meements at the instrumented locations should be within the greater of the speciflet expected tolerance of the calculated values, or ; D.25 inches for the specific points. 1 The installed cold position of the constant hange-ru: de within 1 55 of the design cold load. The installed cold position of the spring tanger must be withan 2 5% of the design mold. load. f Tr2e snutter r.ay deviate from its design cold positlen ) 1/2*, providing the position is not less thar. setting 3 1/2" from tottming out.._. _.

3103 ~iesults Tiping Inspection Selected piping. systems were malked llown at various plant conditions to identify possible restraints to pojected thermal espinalon. These wa14 downs accurred at actient temperature, 25D*

and rated temperatre.

a Hanger and snubber settings were recorded and thermal expansion (PVDET) sensors were wrified to be Antact. No Testraints to projected thermal expansion 4 sere identified. One-hundred and forty-three (143) supports were identified as being out of tolerance or topped w bottomed out. FollowinE re verification and I engineering evaluation, sixteen (16). supports were ad. justed or modified and the remainder. accepted as is. Systen Expansion Selected points on the piping systems were wired with remote senso'rs to monitor the thermally induced pipinF zovements during system operation. The monitored points were expected to undergo large luovewnts w experience large thermal stresses. After establishing initial Tsadings for the sensors at ambient conditions, the mensors were 1 monitored during the. initial heatup of the plant. Data was recorded at SD*F intervals until the Teactor reached nyerating temperature. The evaluations found several criteris exceedances, but upon engineering evaluation of the exceedances, all were found acceptable. In addition, initial ambient sensor readings taken before Heatup were etz: pared to anbient sensor readings after the Heatup and cooldown cycle esas completed. No appaeciable difference in the before and after readinF5 were noted, indicatinE piping movement was not l restrained. Thermal Expansion esta was again taken at 5D*F intervals at moderator temperatures beginning at 1DD*F during the subsequent heatup cycle following initial j l l heatup. The data was evaluated at each temperatu*e plateau before proceeding to the next level. Upon reaching rated temperature, four Level 2 criteria violations existed, tut these were very minor and accepted as-As. y-

1 1 3.15 reedwater system 3151 Purpose The purpose of this test, was to verify the operi bility of the Feedwater control System in the manual and aut . modes of oper.ation aasing the Startup J.evel Contrn11er. 3 15.2 Criteria Level 1 '7he response of luny level related variable to any test input chanBe. or disturbance, laust not diverge dur.inE the setpoint changes. Level 2 7ne level control systerelated.variatles 1 cay contain oscillatcry modes of response. In these cases. the decay ' ratio f or each controlled mode of response must be less than or equal to 0.25. as a result of the setpoint change testing. 3 15.3 Results Duaing the initial heatup, the Teedwate systu performed satisfactorily in toth the inanual and automatic modes. All level-related variables did not diverge during testing and all system related variables did mot exceed a D.25 decay ratio for their esci11 story responses in the level setpoint changes. All criteria were satisfied. -se-

f 316 MSIV Tunctional Test 3 16.1 Purpese 7he purpose of this test was 1.oTunctionally check.the main steam line isolation valves (MSIVs) for proper operation by performing a slow riosure from the ~305 open position to approximately 905 open position followed by a return to 1005 open position on each valve. 3 1b.2 Criteria None 3 16.3 Results DurinE the Heatup Test Condition, with the RPV at rated temperature and pressee conditions, each of the inboard and outboard isolation valves were successfully closed slowly to the approximately 905 open position and then TQlly reopened without any noticeable change .in reactor pressee, APAM readings or reactor water level. 1 .e g ~~

3.17 Belief valves 3 17.1 Purpose The purposes of this test wert to verifythat -the Safety Relief Valves (SRV) function properly (can be npened and closed *2==11y), reset properly after operation, and there are molnajor blockages in thz relief valve J11scharEe pip 1ng. 3.17.2 criteria level 1 There should be a Tositive indication of steam discharge dur.ing the==>ml actuation of each valve. level 2 Ya*iatles related to the pressure control system may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than.or equal to D.25. The temperature seasu*ed by thermocouples on the discharEe side of the valves shall return to within 10*F of the te:peratre recorded before the valve was opened. If pressure sensors'are available, they shall return to their initial state upon valve closure. 'Dring the 25D psig functional test,the steam flow through each relief valve as measu*ed by the initial and final bypass valve (BPV) position shall not differ ty more than 10 percent from the average relief valve steam flo.: as zeasu*ed by bypass valve position. 3173 Results k*ing the hes* up testing, all 15 SR'."s wre v. anus 11y actuated. There was positive indication of stea. discharge upon' actuation of each SRV. As each SRV u s op rated there was a sudden te=perztu e r.ise on the SRV discharEe tailpipe, the appropriate pressure switch j responded, and BPV position decreased to control reactor pressu*e as indicated on the CETARS off _line plots. The Level 1 triteria was satisfied. All pe-tanent varlables related to pressure control dic not extitit any oscillatory responses with secay ratios g' eater than D.25

W l

The SRV discharge line temperirtures for Tive 3RV's 'did not veturn to within 1D'F af the temperatre recorded p*ior to actuation as soon as the others, however, they did tool down sufficiently to. indicate 'that they were not.lenkinE by. Shortly after the perTarzance of this test a. reactor aram accurred.and on the subsequent startup, the SEV tailpipe temperatures remained low, farther verifyingM the 3RV's did wayraciose. Three SRV's trad steam flow values,.as measred by BPV position change, that differed frcrc the average relier valve steam flow by greater than 105. The bypass valve 1msition was. inadequate to get a Troper value of stem flow from BPV position change. Upon the actuation of each SEV the BPV closed completely. Had there been more bypass steam flow, the BPV would not futve close:: cocpletely and there would te a more uccrate value c! SRV steam flow. This steam flow variance will be reevaluated dwing the TC2 25RV 1esting. -B P

3'.18 Piping syste= vidration 3 18.1 Purpose The' purpose t6 This test 13 to deterzine the flow dnduced transient vibration 1 responses of the Main Steam lines inside the drywall and Safety Relief Valve (SRV) lines duriiig planned 3RV nyeration for dynanic transient v11 ration tsardwired tasting. 3 1B.2 criteria Level 1 The reasuaed vitration levels of the pipinE shall Tiet exceed the A11 m ahie apecified walnes. Level 2 The reasu ed vitration levels of the piping must not exceed the expected specified valmes. 3 18.3 Results All vitaation data taken in the conduction of the Heatup SRV testing was well below all specified Level 1 and Level 2 criteria. -9 0-

319 Reseto-water cleanup system 3 19.1 Turpose The purpose of this test was to wify the proper loperation of the RWCU system in the Blowdown and Normal Included in this was a calltration liarif.ication modes. of the RWCU bottzug heafPfl0w tranSMittWr. 3.17.2 triteria level 2 The temperature at the tube side otftlet af the T.on-regenerative heat exchanEers (NRHX) shall not exceed 130*F in the blowdown mode and shall mot exceet 120*F in the normal rode. The coolitE water supplied to the his *Eenerative -heat exchanEers shall be less than 6 ' percent above the flow corresponding to the heat exchangers capacity (as determined f. rom the process diagram) and the existing temperature differential across the heat exchangers. The. outlet temperature shall not exceed 1BD'E. The bottom head flow indicator will be recalitrated against the RWCU flow.inzlicator.if the deviation is greater than 25 Epm. Pump vibration shall be less than or equal to 2 mils (in any direction) as measuaed on the peak-to peak bearing housing and 2 mils peak *to peak shaft vibration as seasured on the 2.oup11ng end. 3 19.3 Results During the Heatup test condition, the RW",U system was placed in a configu-ation so that flow was taken fro. the Lottoc drain and directly fed back to the vesse!, bypassinE the demi 2ntralizers. In this confiEucatic:. U33-610 tottoc drain flow, should read the same as D33-609.. system inlet flow. Oua data show d a.aximu-deviation of.62 apm. Bottom drain flow was recalibrated such that the Level 2 criterla could be satisfied.

91*

The W.TU syste's was Vperated in teth the inormal and A10wdown itodes with the reactor at rated temperature and pressure. Process variables were Tecorded in orde* to demonstrate the proper performance of the RWCU system in each of these modes. The Junengenerattime beat exchange tube side nutist temperaturas for. the normal and blowdown mode were 112*F and 1224 respectively. These values were within the Level 2 crateria limits of 12D*E and 130*F for each mode. Using temperature meascements from the RBCCW side of the non* regenerative heat exche.ngers (WRHX) the cooling water flow was calenistad to be less t!an 65 shoes the NRMX capacity. The non regenerative heat exchanger coolinE water outlet temperatures were well within or Level 2 criteria c? 1BD*T. All Level 2 criteria uere satisfied. l

92*

f I i '-*'-7 r- --3

f 3.20 liesidual Heat Removal system 3 2D.1 Purpose The purpase of this test is to demonstrate the ability .ar the Residual Jient Removal (RHR) System to rmove Deat Jileposited Jn.1he suppramirm pool. 3 2D.2 Criteria Level 2 ' The ilHR System is rapatie nr. operating in the suppaession wol cooling mode at the flow rates and temperatuee dirTerentials.imlicated on the vocess 111a gra=s. 3 20.3 Results Each tivision of the RHR syste= was placed in the Suppression Pool Cooling Mode and process data was taken for a 30 minute time period. The extrapolated heat capacity for both' heat exchangers had an excess capacity of 67.51. This was expected since in early heat exchanger life the heat transfer coefficient is larger and capacity was determined to accommodate some neteriovaticn. 4 l '93* . - ---.}}

ML20140D416

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