ML20100G097
| ML20100G097 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/19/1985 |
| From: | Adam P, Bryant K, Fuhlage C UNION ELECTRIC CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| ULNRC-1062, NUDOCS 8504080073 | |
| Download: ML20100G097 (240) | |
Text
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UNION ELECTRIC COMPANY CALLAWAY PLANT STARTUP REPORT Prepared by 40 P. D. Adam D. T. Keating K. R. Bryant J. M. Moose C. H. Fuhlage M. R. McKinney M. A. Friedlander J. D. Schnack M. L. West
.O O Submitted by /bd SupervisifgReactor. Engineer
- O Reviewed by Superintendent, Engineering
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O Approved W L kyL l Manager, Calfaway Plant /
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8504000073 850319 PDR P ADOCK 05000403 PDR
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TABLE OF CONTENTS IC)
. 1 Page l l
1.0 INTRODUCTION
1 O- 2.0 INITIAL CORE LOADING 3 2.1 Inverse Count Rate Ratio Monitoring for Core Loading 19
. - 2.2 Nuclear Instrumentation System 35
'2.3 Core Loading Instrumentation and Neutron Source 37.
Requirements =
-2.4~ Reactor Systems Sampling for Core Load 39 C)
. 3.0 POST CORE LOAD TESTING 3.1 Incore Instrumentation Operability Test 40 l - 3.2 Control. Road Drive Mechanism Operational Test '43 I 3.3' Rod Control System. 46.
48 3.4 Control Rod Drop Time Measurements --
,C i) ' 3.5 -Digital Rod Position Indication System 55
! 3.6 RTD Bypass Flow Measurement 57 3.7 Pressurizer Heater and-Spray Capability Test .59 3.8 Reactor Coolant System Flow Measurement 62 3.9 Reactor Coolant System Flow Coastdown 64' 3.10' Pressurizer. Continuous Spray Flow Verification 70 iC) 3.11 RTD/TC Cross Calibration 73 4
4.0 LOW POWER TESTING 4.1 Initial Criticality 75 i 4.2 Inverse Count-Rate Ratio Monitoring for Approach'to 76
~ Initial Criticality' C) 4.3 Determination of Core Power' Range for Physics Testing 95 4.4 Boron Endpoint Determination 96 4.5 Isothermal _ Temperature Coefficient Measurement 99 4
- 4 '. 6 RCCA or Bank Worth Measurement at Zero Power 102 4.7. Initial Startup' Flux Mapping 120 4.8' Natural Circulation Test 169
- O 5.0 POWER ASCENSION TESTING 5.1 Dynamic Automatic Steam Dump Control Test 172 5.2 Operational Alignment of Nuclear Instrumentation 174 5.3 Shutdown.and Maintenance of Hot Standby External .183 to the Control Room 43' 5.4 -Automatic-Steam Generator > Level Control 184 5.5 Calibration of Steam and Feedwater Flow 189
. Instrumentation at Power 1
5 ~. 6 Biologica1' Shield Test 190 5.7 Thermal Power Measurement and Statepoint Data 193
, Collection
,3 4 5.8 ~ Power Coefficient Determination 195
- 5.9 ' Pseudo: Rod Ejection and Pseudo Rod Drop at Power 197 l 5.10 Automatic Reactor. Control; 199 10 L
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TABLE OF CONTENTS (Cont.)
^9 P, age, i
5.11 Operational Alignment of Process Temperature 202 Instrumentation 5.12 Startup Adjustments of Reactor Control System 204
- C) 5.13 Incore/Excore Calibration 207 5.14 Load Swing Tests 212 5.15 Large Load Reduction Tests 222 5.16- Plant Trip from 100% Power 227
-5.17 Rods Drop and Plant Trip 231 5.18 Loss of Heater Drain Pumps Test 232
- C) 5.19- Power Ascension Thermal Expansion and Dynamic Test 234 5.20 Loose Parts Monitoring System Test 236 5.21 Plant Performance Test Power Ascension 238 C) lOL jo 1
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1.0 INTRODUCTION
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' This report des.cribes the required testing at Callaway Plant from the time the first fuel assembly was loaded into the reactor until the plant was placed in operation.' It satisfies the requirement of the Callaway. Plant Technical Specifications that a Startup Report be submitted to the NRC after completion of the Startup Test Program.
- O Callaway Plant uses a 4 loop pressurized water reactor rated at 3411 MW thermal. It is located in Callaway County, Missouri, approximately 80 miles west of the St. Louis metropolitan' area, and is
-owned and operated by the Union Electric Company.
Union Electric joined with four other utilities (Kansas Gas and lO Electric, Kansas City Power and. Light, Rochester Gas and Electric, and Northern States Power) to form the SNUPPS (Standardized Nuclear Unit Power Plant System)-Project Organization to manage the design and pro-curement of 5 standardized nuclear power blocks. Subsequently, Rochester Gas and Electric and Northern States Power cancelled their units and dropped out of the SNUPPS organization. In addition, Union jC Electric cancelled'a second unit at Callaway, so that the only two units remaining-are the Callaway Plant and Wolf' Creek Generating Station, which is owned jointly by Kansas Gas and Electric and Kansas City Power and Light.
Under the SNUPPS agreement, Bechtel Power Corporation was selected
!O as the architect-engineer, westinghouse was selected to provide the Nuclear Steam Supply System (NSSS), and General Electric provided the turbine generator. At Callaway, Sverdrup & Parcel and Associates
, provided architect-engineering services which were not included in the
- . standardized portion of the plant. Daniel International Corporation
. constructed-the plant..
!O
' After receipt of a 5% operating license, fuel load was commenced on
- June 13. 1984 and' completed on June 20, 1984. Initial criticality occurred on October 2, 1984,'and initial synchronization to'the grid on-l October-24, 1984. Callaway was placed in service on December 19, 1984, 189 days after commencement of fuel load and 48 days after initial
~
- O criticality. Table 1.1 provides a summary.of-the Callaway startup program.
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TABLE 1.1 l
!O5 CALLAWAY STARTUP
SUMMARY
4 Received 5% Operating License 6/11/84 l() Commenced Fuel Load 6/13/84 Completed Fuel. Load- 6/20/84 Commenced Precritical Testing 7/11/84 Initial criticality 10/2/84 j() _ Initial Synchronization To Grid 10/24/84 4
30% Power 10/28/84 50% Power 11/5/84
$ 75%-Power 11/23/84
- ()
l 90% Power 11/28/84 100% Power 12/3/84 j Plant Released For Operation 12/19/84 Number Of Reactor Trips 16 BurnuP Accumulated on Core 2,112,099'MWHr fC)
Number of Hours Reactor' Critical 1615 Net Generation (MWe) 623,406 MWHr l
l Number of Hours Generator On Line ~910
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6 2.0 INITIAL CORE LOADING (S-070001, S-07SJ01)
Procedure ETT-ZZ-07010 OBJECTIVES Thc objectives of the subject procedure were:
1.- To specify the placement of temporary detectors during the iO- core loading sequence.
- 2. To specify the sequence of loading fuel assemblies into the core.
- 3. To specify the Westinghouse-fuel assembly identification
.O number, enrichment and type of insert for each core location.
- 4. To identify the nuclear monitoring channels to be used for each step in the core loading sequence.
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- 5. To provide a check for periodic verification of equipment
- O status and plant conditions to ensure compliance with FSAR and Technical Specifications during initial cere loading.
- 6. To verify uniform boron concentration in the Reactor Coolant r System and directly connected auxiliaries prior to fuel load.
O
SUMMARY
OF RESULTS All Acceptance Criteria for the subject procedure were met. These criteria and discussions of associated results follows:
- 1. 'A permanent record of the core configuration is obtained.
Final locations of the fuel assemblies and inserts were recorded on video tape. In addition, a written inventory was conducted by operations personnel during the video taping.
- 2. The final as-loaded core configuration is consistent with the -
>O i fuel assembly core loading plan.
I The video tape and written inventory were reviewed by Reactor Engineering. All fuel assemblies and inserts were in their
. proper locations according to the core design.
O 3. Baron concentration is between 2000 ppm and 2150 ppm upon completion of fuel loading.
l The final boron concentration, obtained from the Residual Heat Removal System A sample line, was 2119 ppm at 2000 hrs. on 6-20-84.
l0.
- 4. Prior to initiation of fuel load, the boron concentration of the reactor vessel fluid taken at four equidistant depths in the vessel are within a 30 ppm range of a sample simultaneous-
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F ly taken from the Residual Heat Removal System.
- OL- The boron: samples _obtained between 1338 and 1355 on 6-13-84 were between 2095 and 2068 ppm for the referenced locations.
f -
3 DISCUSSION.
. Prior to the initiation'of fuel load plant conditions were
- O l established in.accordance with the subject procedure. Coolant circula-
. tion was established using the Residual Heat Removal (RHR) System. The .
_RHR loop _f1 tid was sampled and analyzed for boron concentration at f.
regular intervals during core' loading and was maintained between 2000
. ppa and 2150 ppa.
{
- O! yuel' load was
- initiated with the insertion of primary source i assembly C04 into core position L-15 at 1930 hrs. on June 13, 1984. The ;
final: assembly was loaded at 1936 hrs. on June 20, 1984.- Total elapsed
- time was 7 days, 10 minutes. -Figures-2.0.1 through_-2.0.3 illustrate the j final configuration.of the core. The sequence of loading operations is
, presented by Figure 2.0.4.
- O
[- ..
Initial fuel loading at the Callaway Plant was a semi-dry operation !
with the reactor vessel fluid level maintained above the centerline of
- the. outlet nozzle.(above the fuel) but below the vessel flange. This
- provided improved visibility of the fuel assemblies during manipulations j and accessibility to the vessel flange area when positioning the tem-
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porary detectors.
, -Equipment problems caused significant delays during fuel load. The i j major _ problems follow:
1 1. On June.14 source range channel N32 began sp' king and causing a
- O spurious.High Flux at. Shutdown Alarms. The-problem was deter-P mined to be moisture on connectors in two junction boxes and'a
'; damaged connector.0-ring. This caused approximately.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
. delay.-
- . 2. On June 15 the Refueling Machine hoist motor'was replaced iO because of damaged slip-rings. This caused approximately 12 l hours delay.
l ~3. On June 17 a mechanical stop on the Fuel Transfer System bent, i creating difficulty in raising'the frame on the Fuel' Building
- side.- This caused approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> delay.
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- 4. On June 18 a shear pin on the Fuel Transfer System was re-placed. This caused'an approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> delay.
l l 5. On June 18.-the magnatorque speed controller on the Refueling Machine hoist was replaced due to a malfunction of limit
!O switches.- This: caused a delay or.approximately.ls hours. .
! Fuel assembly A47, loaded'in Step _38, was picked up for inspection
- at Step'55E'due'to-visual-' observations by the' reactivity monitoring lO _4_
s
.-._.a , . _ _ _ _ . _ , _ , _ . , , . , , - _ _ , . . . , , , , , _ . , , -,4 ____.m.,._ _ , - . ~
g crew. 'A solid' piece of grease was removed from the north side of the assembly and the affected area cleaned. The assembly was then reloaded.
., O Following completion of fuel loading, the reactor vessel and associated equipment were assembled. This was accomplished without any
. significant problems.
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.O Figure 2.0.1 Callaway Randomized Core Loading Pattern, Cycle 1
.R P N M L K J H G F E D C B A 180* ,.
O 1 C27 C43 C60 C12 C07 C50 C26 ,
2 C36 C32 C54 A38, C01 A22 C04 A16 C28 CO2 C63 3
- O C56 CSI B46 A30 B03 A40 B45 A42 B19 A33 B47 C15 C17 4 C47 B24 B48 A43 B34 A52 B43 A01 B49 B39 C05 B28_ ,B59 ,
5-- C06 C42 A49 B16 A51 B20 A32 B61 A23 B09 A14 B22 A28 C25 C45 10 6 C38 A06 B37 A05 B62 A41 B10 A26 B31 A10 B63 A46 B14 A62 C34 i
C41 C03 A13 B12 A53 BIS A15 B07 A03 B13 A63 B06 A61 C55 C40 i
C22 A37 B18 A36 B38 A60 Bil A29 B53 A35 B32 A21 B57 A24 C13 270 i C20 C61 A58 B50 A64 B33 A47 B17 a44 B02 A56 B64 A48 C11 C24 4
1
'O C09 ' A02 B56 A17 B35 A18 B23 All B01 A55 B26 A04 B05 A39 C39 11 C53 C08 A25 B04. A50 B42 A27 B58 A20 B52 A59 B40 A65 C64 C49 12 A12 B60 A07 B21 B27 B55 C18 C19 B54 B36 BS1 A31 B30
.13 C59 B29 A08 B41 A57 B44 A45 B25 AS4 B08 C33 C46 C35 i
14 C16 C23 C21 A09 C31 A34 C30 A19 C62 C57 C48 Q -
15 C44 C14 C52 C10 C37 C58 C29 lAl Region 1 (2.1 w/o) lCl Region 3 (3.1 w/o) l l
lBl Region 2-(2.6 w/o) l o r6-
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- O rigure 2.o.2 Control Rod Locations
- O R P M M L K J H G F E D C 8 A i
1800 l
1
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A 4 A E
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8 C A C B 6 S Sg 7 g
'O 900 C S E A D A SE C 2700 - 8 Sg Sg 9 8 C A C 8 10 S 3 ll D C Sg o I 12 gE A 8C S, S, S o i3
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0 SHUTDOWN NUM8ER CONTROL NUMBER BANK 0F ROOS BANK OF R005
' A 4 SA 8 8 8 Sg 8 C 8 4 5"
O n 5 S o
4
! TOTAL 25 Sg 4 TOTAL 28
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O Figure 2.0.3 Burnable Poison And Source Assembly Locations O
R P N M L K J H G F E D C B A 18 0*
I 10 10 10 2 9 12 20 ll 12 9 3 9 20 16 16 16 20 9 O 20 45 is 16 20 M
'O S
12 20 16 16 16 20 12 6 jo 20 16 16 10 16 16 20 O 7 20 20 16 20 20 16 20 8 10 16 16 20 20 16 16 10 30' 270*
8 20 20 20 16 20 20 16 10 0 16 16 20 20 16 16 10 II 20 16
, 12 16 16 20 12 I
20 20 16 16 45 20 20 g
13 3 20 16 16 16 20 g 9 12 20 ! 12 9 IS O ,o ,g gg O'
O NUMBER INDICATES NUMBER OF BURNABLE POISON RODS IS INDICATES PRIMARY SOURCE ROD 4S INDICATES FOUR SECONDARY SOURCE RODS i
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Figure 2.0.4 CORE LOADING SEQUENCE O
Legend for Core Loading figures r
,0 Assembly loaded in permanent position in previous step.
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O Asse=bly loaded into positien during leadi:g step nu=ber N.
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Location of Temporary Detac:cr A (5 and C).
g Assembly with primary source inserr.
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CORE LOADING SEQUENCE
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% Core Loading Sequence Stepe A to la i
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Figure 2.0.4 (cont.) l l
1 CORE LOADING SEQUENCE
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lO Core Loading Sequence Steps h to h ,
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Figure 2.0.4 (cont.)
CORE LOADING SEQUENCE l
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CORE LOADING SEQUENCE O
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Figure 2.0.4 (cont.)
CORE LOADING SEQUENCE O -
18 0*
O F E O C B A R P N M L K J H G I I I 8 f f f i XXXX o 2 B XXIXIX A 3 X XXX 4 , XXXX 5- X; X IX X
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Step SSE - Raise, inspect and lower assembly A47 in core location l
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Figure 2.0.4 (cont.)
3 CORE LOADING SEQUENCE TO 2
@ 18 0*
O i R P N M L K J H G F E D C 9 A s a i e i i.
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Figure 2.0.4 (cont.)
CORE LOADING SEQUENCE O
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Figure 2.0.4 (cont.)
l CORE LOADING SEQUENCE O
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in nsXXXXX 2 . a m nz XX;XXXXX O
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Figure 2.0.4(cont.) core LOADING SEQUENCE O !O C B A P N M L K J H G F E D A i .XXXXXXX O z , mie8XXXXX:X X.XX
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2.1. INVERSE COUNT RATE RATIO MONITORING.FOR CORE LOADING (S-070002) Procedure ETT-ZZ-07020 0- OBJECTIVES The' objectives of the subject procedure were: 1.~ To describe the methods for obtaining nuclear monitoring dat.a during_ initial core loading.
- 2. -To describe the sequence, frequency and core conditions for obtaining nuclear monitoring data in order to prevent criticality during core loading.
SUMMARY
OF RESULTS O-The Acceptance. Criterion for the subject procedure states that the core-is loaded without achieving criticality. This criterion was met by
. observing the trend of core neutron multiplication during the loading using the inverse count rate ratio (ICRR) method. No abnormal changes in the ICRR were noted.
O DISCUSSION-Prior to the initiation of core loading three temporary neutron-sensitive detectors (BF gion chambers) were suspended in the reactor _ vessel in locations speEified by the core loading sequence. Count rate O data from the temporary detectors were obtained at the reactivity monitoring station in containment.. Count rate data from source range channels N31 and N32 were obtained in the control room and transmitted to the reactivity monitoring station using headsets. The response from at least one responding detector was displayed continuously on a strip , chart recorder during fuel movement within the reactor vessel. Background readings were obtained for all detectors prior to loading the first assembly. Following step 10 of the core loading sequence, reference values were obtained for the responding detectors. After loading each subsequent fuel assembly, count rate data was collected and the ICRR calculated for each responding detector. The O ICRR's' vere plotted to observe the trend of core n.utron multiplication and' ensure criticality would not be achieved upon loading the next assembly. The ICRR plots are presented as Figure 2.1.1 through 2.1.5. The count rate and ICRR data are presented as Table 2.1.1. New reference values were obtained for the affected detector after O significant equipment change, after detector position change or after a significant decrease in ICRR due to building a " bridge" between a primary' source and a detector. New reference values were obtained prior to resuming core loading after a suspension of four hours or more. A chi-square statistical reliability test was performed on the data used for each reference value calculation. While obtaining background detector data, it was observed that temporary detector A was spiking. This detector 'as' w declared inoperable-O _ t, _.
4 , A , !O. e i ! and was replaced by temporary detector D at step 22 of the core loading sequence. Temporary detector A was found to have an accumulation of moisture in it.
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2.2 - NUCLEAR INSTRUMENTATION SYSTEM (S-07SE01)
. ~ Procedure ETT-SE-07010
- O OBJECTIVES The objectives of this test were to verify that the Source and Intermediate Range Nuclear Instrumentation System performed the required
.. indication and control functions prior to core loading and that the iO Power Range Nuclear Instrumentation System-performed.the required
. indication and control functions prior to Initial Criticality.
SUMMARY
OF RESULTS 7 S It was determined that the source range level trip ocgurred at. O 1x10 cps, the Intermediate Range Trip occurred at 2.5x10-1088ps, the
; Intermediate Range P-6 setpoint was determined to be 1x10 amps and 5
the Intermediate Range High Level Rod Stop was determined to be 2x10
. amps. .Theaggeptancecriteriaforeachoccurrencewege: Source Range Trip s1.6x10 cps,IntermediateRanggrips3.53x10 amps, Intermediate Range P-6 setpoint 21x10-5 amps and the Intermediate Range
- O
- 'High Level Rod'Stop setpoint $2.1x10 amps. Therefore, all the acceptance criteria were met.
, For'each Power Range channel, the appropriate Instrumentation and Control Surveillance procedure was utilized to verify that'the following setpoints were correct: Overpower Trip Low Range, Overpower Trip High
- O
- Range, Positive and Negative Rate Trips, Power Above Permissives P-8, -
, P-9 and P-10 and Overpower Rod Stop. The acceptance criteria associated with this portion.of the test was that the setpoints were set as specified by the' appropriate I&C procedure, therefore all acceptance _ criteria were met. lO DISCUSSION l . .. Detector voltage readings were taken. Source Range Detector voltages were taken from the detector volts. meter on each drawer and the i Intermediate and Power Range detector voltages were obtained by measuring the voltage at various test points.
- O Testing of the Source and Intermediate Range Channels was conducted
- ..by placing the channel in the " Test" mode and introducing a test. signal,
- 'to simulate various. count rates. .A check of the High Flux At Shutdown l . Alarm and the Audio and Scaler Timers was-satisfactorily performed.
During this test, it was found that the Bistable reset levels on'each [O .
- Source Range Channel was greater than the given acceptable level. I&C i corrected this problem by resetting the levels. Also,'it was found that i during the test,;the channels did not1 indicate an acceptable count rate.
l This.was fcund to be due to a wrong thumbwheel setting and was corrected. The test was conducted again with satisfactory results. I l0. 1 Testing was performed on the Power. Range Channels in accordance with the appropriate I&C controlling procedure. O- _ L _ _ _ _ - _ _ _ _ _ .
F .O k The Source and Intermediate Range rate circuits were tested by utilizing the rate test switch. The circuits provided test inputs of.1 I3 and 5 dpm and the rates are then read from the appropriate instrument. A' test of the Source Range Flux Doubling Instrumentation was performed. This was accomplished by applying a voltage to the detector circuitry and.then doubling the applied voltage ten minutes later. After satisfactory completion, the voltage was returned to its normal O value. () - O O f O O O O C). -, . _ _ - - -
- O ;
2.3 CORE LOADING INSTRUMENTATION AND NEUTRON SOURCE REQUIREMENTS (S-07SC01)
.O Procedure ETT-SC-07010 OBJECTIVES The objectives of this procedure were:
,C) 1. To verify proper alignment and calibration of the temporary core
-loading instrumentation prior to the start of fuel loading,
- 2. To verify the signal to noise ratio is greater than 2, and
.3. To check the neutron response of the temporary core loading
'C) instrumentation package and the plant Nuclear Instrumentation System (NIS)' Source Range c hannels, a.- Prior to the start of fuel loading, and
- b. Prior to the resumption of fuel loading following any delay of
- C) eight hours or more.
4
SUMMARY
OF RESULTS 1 Prior to the start of fuel loading, the temporary core loading instrumentation was aligned and calibrated using a 4.9Ci Am-Be neutron , C) . source. An operating voltage of 1950 volts and a discriminator setting of 2.5 volts were determined for Detectors A, B, C, and the spare l detector. These values were identical to the pre-shipment values furnished by Westinghouse. The signal to noise ratio was satisfactory for all channels. The neutron response of the temporary core loading instrumentation and the plant NIS Source Range Channels was verified by .R C) . a positive change in count rate as the neutron level was increased before the start of' fuel loading. The neutron response was verified by performance of a statistical evaluation due to ALARA considerations prior to the resumption of fuel loading following all delays of eight hours or more. ' C) DISCUSSION The neutron source was positioned near the temporary instrumentation detectors on the operating floor of containment' to give
~
a count rate of at 1 east 10,000 CPM on each of the three channels. Starting at 1800 volts, the high voltage was varied in 50 volt
?C) increments for each channel up to 2500 volts. Voltage vs. CPM curves were then plotted.to determine the operating voltage. _ Starting ar zero volts, the discriminator voltage was varied in 0.5 volt increments for !- each_ channel up to 4.0 volts. Discriminator Voltage vs. CPM curves were
. .then~ plotted to determine'the discriminator setting. LC) Prior to the start of fuel-load, the neutron response checks were performed by moving the source to a position as close as possible to each of the detectors. Following all' delays of eight hours or more, 10 count rate observations of 400 to 5000 counts per observation were
- O , _ 37 -
.'O ' obtained for each channel. A statistical evaluation was then performed by calculating s Reliability Factor to verify that the count rate O-
~
measurements did not exhibit a greater deviation from the Poisson Distribution than was expected from chance alone. From the 6/1/84 initial checkout to the 6/13/84 start of fuel loading, high humidity inside containment required dessicant and plastic covering,to be placed over_the preamps and around all connections to O prevent moisture accumulation. This solved the moisture problem caused by high humidity. On 6/15/84 Detector A was replaced by the spare detector due to water leakage into Detector A. No additional problems were encountered during this test. O-O O O O O O . O O.
'2.4 REACTOR SYSTEMS SAMPLING FOR CORE LOAD (S-07SJ01) 0: .
.(This. procedure was combined with ETT-ZZ-07010, Initial Core Loading.)
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4 lO 13.1 INCORE' INSTRUMENTATION OPERABILITY TEST (S-07SR03 and S-07SR04) Procedure ETT-SR-07040
- O OBJECTIVES The objective of this procedure was to demonstrate the operability
-of the Flux Mapping System under hot and cold conditions, prior to physics testing.
SUMMARY
OF RESULTS The following items were verified during this test: 4
- 1. ' The Flux Mapping System (FMS) operates as designed for the
- O System initialization Mode.
- 2. The FMS operates as designed for the Manual Flux Map Mode.
- 3. The FMS operates as designed for the Set / Modify Limits Mode.
- O. 4. The FMS operates as designed for the Set / Detector Voltage i Mode.
- 5. The FMS operates as designed in the Automatic Flux Map Mode by 1- providing a full core flux map when the start action is
- entered.
- 6. The FMS operates as designed in the'Special Flux Map Mode, r- 7. All thimbles'are clear of obstruction with the exception of Thimble J01. With this path inoperable, the Flux Mapping
- System still satisfies Technical Specification requirements.
- O
- 8. Each_ drive was capable of its detector going through the full
- length of each thimble with the exception of Drives A & B not i~ being able to go through~ thimble HIS (and no detectors through J01 - See #7 above.) Auto Flux Maps normally use the C Drive for Thimble H15 and thus does not impact Technical
,O
- Specification operability.
! 9. . The FMS operated as designed in the Data Transfer Mode with
- the exception of the data link between the NSSS computer-and the FMS not functioning due to NSSS computer problems. This
, function was proven operable during the performance of ,0 S-07sR01, initial Startup Flux Mapping. 4 This. test went through all eight (8) functions that are associated with the FMS and verified that they operated correctly and that the required alarms and interlocks functioned properly, i
- 0 ._ 4o _
t DISCUSSION O This test consisted of ten separatie test sections which verified all of the~FMS capabilities: 1) System Initialization; 2) Set / Modify Limits Mode Operational. Checkout; 3) Manual Mode Operation Checkout;
- 4) Set Detector Voltage Function Checkout; 5) Thimble Checkout; 6). Data Transfer Mode Operational Checkout; 7) Peripheral Device Failure Checkout; 8) Automatic Flux Map Mode Operational Checkout; 9) Special O Flux Map Mode Operational Checkout;.and 10) Thimble. Verification.
Section 10 was performed at cold shutdown and sections 8 and 9 at hot standby. After performance.of this test, a large number of deficiencies were.
'identified, for which Westin' house g assistance was requested for O. resolution. Westinghouse determined that the majority of the problems were procedural errors carried over from the original, Westinghouse test procedures. These_ changes had not been incorporated by. Westinghouse since the system software was still being modified. These procedural errors do not affect test acceptance. The problems encountered are 8u888ri2*d in the f 11 win 8. paragraphs.
O The System Initialization test checked the ability of the FMS to properly initialize the system for use from either CRT, transfer control between the CRT's, and signal appropriate error messages. This section was performed satisfactorily. The only problem encountered in this section was an inadvertent "15-Path Transfer Failure - A and B" which O was due to power being off to the A and B drives.
'The. Set / Modify Limits Mode Operational Checkout section tested the ability of the FMS to enter new drive path limits, obtain a printout of the values and load the new values onto a cassette tape. A problem was
~
encountered with the printouts not matching up to'the entered values, O. which was caused by the write protect < tabs not being in the correct position on the cassette tapes. This section.was performed satisfactorily after this problem was resolved. The Manual-Mode Operational Checkout section tested the ability of the FMS to perform manual manipulations of~the drives / detectors to O preset locations and verify that the system performed as designed and/or signalled appropriate error messages. Several problems were encountered during performance of this section, which were resolved by Westinghouse. Final detector positions were required to be within .1 "of the input-
. position - nonconformance was determined to be a procedural error since g slow speed tolerance is t.5" and fast speed tolerance is 12". Other problems were encountered due to unexpected or incorrect. error messages, which were-due to procedural errors. This section was completed successfully after correction of the' procedural deficiencies.-
The Set Detector Voltage Function Checkout section tested the. ability of the-FMS to set detector voltages and obtain a printout of the i voltage plot.- The detector voltages are set. prior to running any auto or special. flux map. This-section was performed satisfactorily with the-l- caly problem being a procedural deficiency due to an error message not ' being received.
'O
_ 4g _
G IOi The Thimble Checkout section tested the ability of each detector to be inserted into every thimble. . The major problem encountered in this
- O, section was the message "s0 DETECTOR MOTION" caused by the detector.
running'into the end of the thimble. This was due to the difference in the number of turns on the drive cable which drives the resolver and the inaccuracy in the way the first path insertion lengths were measured with the manual control box. Revised insert. ion limits were developed . and input each time this problem was observed. Final insertion limits
- O . vere eventually verified in.a separate procedure. Thimble J01 was found to be blocked during performance of this section which caused several deficiencies since the thimble must be deleted to run an auto flux map.
Generic problems were encountered in this section and others with the overtorque' function of the drive units. The overtorques could not be ; set properly, however this is a redundant function since a "NO DETECTOR O MOTION". message will be received if the drives become stalled. Another , overtorque problem was traced to a. faulty triggering card in the motor ; control box.- A final problem was encountered with Drive B not being able to enter Thimble H15. ;
. The Data Transfer Mode Operational Checkout section tested the l O ability of the FMS to transfer data from a floppy disk to the printer, a second floppy disk, or to the NSSS computer. This section was performed satisfactorily with the exception of the data link between the FMS' and the NSSS computer not functioning due to NSSS computer problems. This problem was resolved and the link verified during later. testing (S -07SR01) .
. . The Peripheral Device Failure Checkout section tested the ability i of the FMS to signal appropriate warnings to the console operator due to any equipment failures, e.g. printer, floppy disk unit, etc. This section was performed satisfactorily following resolution of some procedural problems ~(conflicting error messages received).
The Automatic Flux Map Mode Operational Checkout section tested the ability of the FMS to perform an automatic flux map using all or any combination of drives and signal appropriate errors.to the console operator. In addition to miscellaneous procedural problems, the problem with "NO DETECTOR MOTION", as discussed previously, was encountered. O Drive A also would not go through Thimble H15.- With Drives A and B not being able to go through H15 and no drives through J01, the system is still capable of meeting Technical Specification operability
-requirements.
[The Special. Flux Map Mode Operational Checkout section tested the O- ability of the FMS to perform a special flux map using selected thimbles
,and drives and signal appropriate error messages. No additional
' deficiencies were discovered during this section but several similar problems _(procedural errors,' Thimble HIS stuck for Drives A and B "NO DETECTOR MOTION") were encountered.
O The Thimble Verification section (performed at cold shutdown) l tested each thimble using a dummy detector. This section was performed r satisfactorily after resolving the initial problem with path insertion limits <
- O :
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r1
.O 3.2 L ~ CONTROL' ROD DRIVE MECHANISM (CRDM) OPERATIONAL TEST (S-07SF01)
Procedures ETT-SF-07014 and ETT-SF-07015 OBJECTIVES The objectives of this test were to verify proper slave cycler timing, to demonstrate the proper operation of each CRDM under both cold O- shutdown _and hot standby conditions, and to demonstrate that all Digital
~
Rod Position Indicators (DRPI) operate satisfactorily over the entire range of_ rod travel;
SUMMARY
OF RESULTS O The operability of each CRDM was verified by a demonstration of
. drive line insertion and withdrawal during which CRDM coil timing traces were obtained. .These traces showed conformance to the requirements for proper CRDM operation and timing. This criteria was verified during both, cold..and_ hot conditions. The DRPI System control board display indication and NSSS computer readout of demand position (not necessarily O the. actual position)-for each rod agreed with its associated group step counter within 14 steps. All rod bottom LEDs actuated within 4 steps of the zero indication of-their associated group step counters. For control banks, the pulse-to-analog converter bank position readout agreed with its associated group step counter within 11 step.' All rods operated satisfactorily when being withdrawn and-inserted over their O
entire range of travel. All traces.were normal during both cold and' hot mechanism timing tests. DISCUSSION
-These tests were performed by pulling control rods and stopping at
.O various points to reccrd data. .This data verified DRPI display and gray code (at Data Cabinets A and B in containment) data versus control room
. step' counters. For control banks, pulse-to-analog converter data was ;
verified versus the step. counters. At 48 steps, rods were withdrawn and i reinserted 10 steps while operating a visicorder that was connected to - the stationary gripper, movable gripper and lift-coil _of each mechanism 0 to obtain a trace of the associated coil currents. In addition, sound signals from a carbon microphone affixed to the' top of each rod travel thousing were selected for the appropriate rod and recorded as a fourth
~'
visicorder trace. The resultant traces were examined by comparison with
.a correct. trace using a~ clear plastic overlay. A normal CRDM stepping trace is illustrated in Figure 3.2.1.
Several problems were encountered during the performance of the , test under cold conditions. NSSS computer readout of demand position l
/was' unavailable due to a' design discrepancy between the DRPI wiring and. '
NSSS computer. -This discrepancy was resolved and testing was successfully completed. .During the testing of Control Bank B (CBB), Rod 0 - F-14 DRPI indicacea_66 step. while all other rods in CBB and the step counters for CBB read-72 steps. . (This was still within 112 steps'as required by Tech Specs.) Testing for CBB was continued until at--76 steps (on step counters) F-14 showed General Warning-(GW) and Rod Bottom
~
O_- . _ 43 _. i
h
.O .
(RB) LEDs. All other rods indicated normal. CBB was reinserted. Upon investigation it was discovered that two pins were touching in the DRPI O connector for.' rod F-14. The pins were straightened out and CBB was suc'cessfully-retested. During Control Bank D (CBD) testing, Rod _D-4 DRPI indicated 72 while the step couaters for CBD indicated 79 steps. i Rods were reinserted after some troubleshooting was performed. The '
-problem was a bad-DRPI connector for 7od D-4. The connector was ,
replaced and CBD wns successfully res.ested. During Control Bank A , (CBA),L testing rod K-8 would not withdraw. Upon investigation it was
.O-discovered.that the blown _ fuse indicator (BFI) was indicating that the stationary gripper coil for K-8 had a blown fuse. The fuse and.BFI were replaced and CBA was successfully tested. During CBC testing the timing trace of rod B-8 did not show a sound trace. Upon investigation it was found that a wiring mismatch was present. A jumper was installed to O- allow completion of testing. The wiring problem was corrected subsequently.-
During testing at hot standby conditions several problems were encountered. An attempt was made to withdraw Shutdown Bank A. There
.was no movement due to a Power. Range Overpower Rod Stop. Power Range O Channel N44 Rod Stop was placed in bypass and testing resumed. While withdrawing Shutdown-Bank A DRPI showed no indication of rod movement
~
with the step counters showing 38' steps. Shutdown Bank A was inserted to 0 steps and testing suspended. Troubleshooting revealed that DRPI ' Data Cabinet B had been placed in " Test". DRPI Data Cabinet B was returned to normal.' While attempting to withdraw Shutdown Bank C, a Rod O Control Urgent Failure Alarm was received. The same alarm was received while attempting to withdraw Shutdown Bank D. Following replacement of a card in Power Cabinet SCDE, an attempt was made to withdraw Shutdown Bank C and the Rod Control Urgent Failure Alarm occurred. An attempt to-withdraw Shutdown Bank E was made with the same result. Testing was suspended to determine the problem, which turned out to be a slave O cycler card in the Logic Cabinet. Testing of Shutdown Banks C, D, and E was then completed.- During hot testing of CBC'a sound signal was not obtained from microphone B-8. Troubleshooting revealed incorrect wiring inside the head area junction box. It was determined a retest was not necessary because all coil' traces were satisfactory, and satisfactory sound and coil traces were obtained under cold conditions. O O l0 i _ 44 _
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1 O'
!3.3 -ROD CONTROL SYSTEM (S-07SF02)'
Procedure ETT-SF-07020 OBJECTIVES To demonstrate and document that the hod Control System satisfactorily performs the required control, inhibit and indication functions in order to verify it is ready for use prior to initial O' criticality.
SUMMARY
OF RESULTS The control, inhibit and indicating functions of the Rod Control
. System were satisfactorily demonstrated and verified, with the exception O: 'of speed indication for Shutdown Banks C, D and E.
DISCUSSION The urgent and non-urgent alarm functions were tested in the Power cabinets.. The urgent alarm resulted for each test case where a rod O control inhibit function was tested. The rod inhibit functions which were tested are listed below: A. Regulation Failure (Stationary and Movable) B. Phase Failure
- C. Multiplexing Error O D. Logic Error For each inhibit function listed above, a failure symptom was physically simulated for each power cabinet and the Reactor Operator was requested to attempt to withdraw rods. The net result was that rods didn't withdraw and the urgent alarm was received at the Main Control i O Board..(MCB) and locally at the power cabinet being tested, thus satisfying the objective of the test.. The non-urgent alarm circuits were also tested at each power cabinet by removing' fuses and thus simulating a Power Supply (P/S) failure. The non-urgent alarm was received at. the MCB and the power supply failure was received at - the
~
power cabinet, thus satisfying the objective of the test. O.' The urgent and non-urgent alarms were tested in the logic cabinet-in the same manner as the power cabinets.- An actual logic circuit defect was inserted into the-logic cabinet and the Reactor Operator was !
; requested to attempt to withdraw control rods. The urgent alarm was '
verified at the MCB and locally at the logic cabinet and there was no 0: 1 outward motion of>the control rods. soch the logic circuits and the l pulse failure test were. items tested for-the urgent alarm inhibit
' function. 'The non-urgent' alarm function was physically tested by..
- de-energizing a Power Supply (P/S) by pulling fuses-in the logic cabinet and= verifying the non-urgent alarm was~ received at the MCB and at the logic cabinet.: Thus,'all objectives were met. l
.O The control and indication functions'were tested by actual control rod withdrawal, insertion, DRPI reading, bank position indication
-readings, rod speed verification, and: rod group transfer to and from the O: -
- -~46~-
^
F O DC Hold Bus. Both the shutdown and Control Rod banks underwent this form of testing. The Control Bank rods also underwent a sequenced bank (3 - overlap check by control rod withdrawal and insertion. The control system was also checked by inserting a manual rod withdrawal block condition and then having the Reactor Operator attempt to withdrawn control rods. The rods didn't withdrawn and the test was satisfactory. Thus, all objectives concerning control functions were satisfied. () A Design Review Criterion was not met for Shutdown Banks C, D, and E speed indication. Rod speed was ac expected when measured independently, but an indication _prcblem exists. Westinghouse has been requested to review the design for this indication and initiate the necessary design change to correct _the problem. This deficiency has no impact on nuclear safety and was evaluated by the Onsite Review () Committee as posing no restraint to continued power operation. O O
;O
- C) o iO (O
O _ 47 _ i
Q. l l 3.4 CONTROL ROD DROP TIME HEASUREMENTS (S-07SF03) ; Procedures ETT-SF-07033, ETT-SF-07034 ; OBJECTIVES The objectives of the subject procedures were as follows:
- 1. To measure the drop time of each Rod Cluster Control Assembly O. (RCCA) with the Reactor Coolant System (RCS) in a hot, no flow condition (HNF).-
- 2. To measure the drop time of each RCCA with the RCS in a hot, full flow condition (HFF).
O 3. To retest those RCCA's whose drop time deviated from the mean drop time by more than two standard deviations a sufficient number of times (2 3) to reasonably ensure proper' performance during subsequent plant operations.
- 4. To monitor the ?ffectiveness of the thimbles dashpot region O for decelerating the RCCA's during each rod drop time measurement.
SUMMARY
OF RESULTS O All acceptance criteria for the subject procedures were met. These criteria and discussions of associated results follows:
- 1. The drop time for each RCCA is less than or equal to 2.2 seconds (HFF).
O The longest drop time as measured from the beginning of decay of stationary gripper coil voltage to dashpot entry (HFF) was 1.40 seconds at core location J-03.
- 2. The rod drop traces are ' consistent and exhibit free fall with no abnormalities or evidence of binding.
v Review of the rod drop traces indicated satisfactory results.
- 3. The rod deceleration through the dashpot region must be similar for all rods being dropped.
'O The. deceleration time (time from dashpot entry to bottom of dashpot) for the rods. ranged from 0.39 to 0.45 seconds for the HNF case and from 0.45 to 0.50 seconds for the HFF case.
- 4. Each RCCA whose drop time ~ deviated from the mean drop time-by more than two standard deviations, when retested, exhibits a
'O time consistency with 0.02 seconds. All rod drop times were within the allowed band. No rods were required to be retested. O I O DISCUSSION'
~
- O ' The.Callaway Plant utilizes a Digital Rod Position Indication (DRPI) System. A Rod Position Detector is supplied for each rod; this iincludes the rods for the control banks and the shutdown banks. Each standard detector is in essence a tube with 42 assemblies of coiled wire slipped over the tube and spaced out along its length at 3.75 inch intervals. Each shutdown detector has a reduced number of coil O: assemblies (20) and a space without coils in the middle section of the detector. .The tube-fits'over the rod travel housing. AC power from a constant voltage source is applied to each coil of wire on the tube. A low frequency;is used so that-the electromagnetic flux generated by the coils penetrates the thickness of the rod' travel housing. Since the rod within the housing is ferromagnetic, a concentration of flux is realized
.O_ as the rod moves through a coil. The concentration of flux is reflected as ar. increased impedance to the AC flow.
To prevent total loss of information due to a single coil failure,
-the outputs of every other coil on the tube are connected as inputs to one Data Cabinet, while the outputs of the remaining coils are connected O to the other. Data Cabinet.. Within each Data Cabinet there is a detector / encoder card for each Rod Position Detector. A detector /
encoder card samples the voltages across terminating resistors, each of
~
which is effectively in series with one of the detector coils. When a rod is.out of a coil, the voltage sample level is high; when the rod moves into'a coil, the voltage sample level drops. At the detector / O encoder card, samples from adjacent coils are monitored by differential amplifiers. When the rod end_is between two. adjacent coils, the corresponding differential amplifier output is high; otherwise, the output'is low.- Thus, the position of a particular rod is reflected in the ~ output = states of 'the associated detector / encoder card dif ferential amplifiers.
-A test point on each detector / encoder card in each' Data Cabinet provides a convenient means of' recording the profile of a rod drop test.
LUnder normal operations, the test point measures the ac excitation applied to the particular A or-B channel of coils. When'this voltage is removed, the residual magnetism of the rod causes a voltage to be i
- O momentarily' induced in each coil as the rod passes through. These l _ voltages sustain a pulse of current in the conmion-line, which can be
!' recorded using a visicorder. f _ Figure 3.4.1-is an example of.a recording of such a current' pulse-monitored at the test point for a standard detector. The notches in the jo pulse indicate the rod top passing through a coil.. Therefore, if a particular coil is shorted or_open, the profile identifies it. Also, if I. the rod temporarily binds at any point, the approximate. location of the bind can be identified in the profile. 'For shutdown detectors, since rod _ travel _is not monitored to the same accuracy in the mid-range as for a standard ~ detector, visicorder trace sensitivity is not_as sharp in O- that area. nowever, all critical parameters,-such as initiation of drop,1 dashpot entry, and snubber contact are present at full sensitivity;.and the rod drop for shutdown detectors.can be measured, as with the standard detectors. l lO- _ 49 _
.i
r-O' A' block diagram of the test arrangement is presented as Figure 3.4.2. Note in Figure 3.4.1 that the Data A and Data B traces are out () of phase, due to the alternating arrangement of the A and B coils. By sumr'_g the A and B signals a smooth. trace is output which allows
. precise definition of dashpot entry. A 60-Hz trace was recorded for use
,as a time standard. Sound traces were obtained from carbon microphones mounted in housings attached to each CRDM rod travel housing.
I)I Figures 3.4.3 and 3.4.4 presents the rod drop times. .O O O O O
- O C) .
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{\ ! 1 THIS TRACE FORMAT IDENTIFIES:
- 1- DROP TIME VALUE, TO DASHPOT
- 4- DASHPOT ENTRY
- 7- S.G. COIL SIGNAL
- 2- DROP TIME VALUE, TOTAL
- 5- BOTTOM OF DASHPOT
- 8-INITIATION OF MOTION
- 3- INITI ATION OF EVENT
- 6- SOUND (SIGN AL
- 9- 60~ TIMING SIGNAL RECOMMENDED OPTIMUM " ROD DROP" TEST TRACE FORMAT ARRANGEMENT
O FIGURE 3.4.2 ROD DROP TESTS BLOCK DIAGRAM O l
. CONTAINMENT CRDM POWER CA3DET RCCM v 1 O
a
- O Data Cabinet A l l A+B ,
^
O Testbex 60 Es for Timing Trace ! l i Data l Cabiner. 3 lO Visicorder E_ s ~ E oscillograph c. r ~O - [ Patch l Sound E Box c-- -O RDM Microphone Signals g i lo i O l l l O
,_, FIGURE 3.4.3 ROD DROP TIME TABULATION
~
HOT, NO FLOW K l- M N P 5 O. D E F G H J s C 0* 0 15 14 1.18 1.20 1.19 1.19 1.19 1.58 1.62 ,1.63\ 1.62 1.61 1.18 1.19 1.20 1.1G 0 13 1.61 1.61 1.62 1.60 1.20 1.20 1.18 1.18 1.18
'12 1.63 1.61 1.60 1.62 1.61 '
1.17 11 1.20 1.59 O 1.63 . _ _ 1.18 1.19 1.17 41.20 1.19 10 1.62 1.61 1.60 l1.64 1.63 ! c, . 1.19 1.19 l 1.62 1.61 I- i o , 7 9 , ,. _ 1.19 1.19 1.17 1.19 1.17 1.19 19;
*[_ 90*
, -j 1.58 1.62 1.60 1.62 l'.61 1 ,
1.62 1.61 1.20 1.19 7 1.63 1.62 1.19 I 1.18 1.20 1.19 'O 6 1.18 1.59 1.63 .1.62 1.62 1.62 1.20 5 1.20 1.60 1.62 1 19 I9 I 19
, 4 1.20 1.18 l'.63 1 :1.60 ' 1.62 1.61 1.63 1.17 1.19 1.19 1.20 3
1.60 1.61 1 1,62 1.59 1.18 1.19 1.19 I 1.19 2 1.19 1.62 1.60 1.61 1.61 1.61
.O 1
183* O X.XX TIME TO DASHPOT ENTRY { SECONDS} - AVERAGE = 1.19 X.XX- TIME TO DASHPOT BOTTOM { SECONDS} - AVERAGE = 1.61
'O L
- FIC'TRE 3.4.4 O '
ROD DROP TIME TABULATION HOT, FULL FLOW L M N P R O 3 c D E F G H J K
-3 O' 'O 15 1.38 1.38 1.36 1.36 1.39 14- ~~
1.83- 1.85 1.85 1.84 1.86 l'.35 1.37 1.37 1.34
.O 13 1.82 1.85 1.84 1.82 -- - - - - - - -
. .n - .- . . . . .
1.36 1.36 1.35- 1.35 1.36 12 1.83 1.83 1.83 1.84 1.84 , 1.39 1.34 11 _ . . - 1.80 0 1.86 10- 1.38 1.37 1.35 1.38 1.36 1.86 1.85 1.82 1.85 1.84 1.38 1.38 9 1.86 1.86 j, n 270._ 1.36- 1.35 i 1.38 1.38 1.37 1.38 L 90' "3 1.34 1.81 1.85 1.85- 1.85 1.85 1.82 1.80 1.38 7 1.35 1.87 i 1.83 1.38 1.38 1.37 O6 _ 1.37 1.86 1.36 1.83 1.84 1.86 1.85
~
.5 1.37 1.38
- 1.85- 1.84 1.38 1.33 l.36 1.35 1.36-n.4
's 1.85 1.82 1.85 1.84 1.83 1.37 1.37 1.40 1.37 3
1.84 1.85 1.89 1.83 1.39 1.37 1.35 1.39' 1.39 l 2 1.85 1.89 1.87 '1.84 1.~ 88
.g -
f
~1 180*
- O X.XX . TIME TO' DASHPOT ENTRY { SECONDS} - AVERAGE = 1537-X.XX TIME T0 DASHPOT BOTTOM (SECONDS}'- AVERAGE = 1.84 r
V - 54 - 1 g !
01
.-3.5 DIGITAL ROD POSITION INDICATION SYSTEM (S-07SF04)
Procedure ETT-SF-07040 OBJECTIVES.
- The object of this test was to verify.that the Digital Rod Position g Indication System (DRPI)' satisfactorily performs the required indication v- .and alarm functions prior to the control rods being operated.
SUMMARY
OF RESULTS e This test was performed satisfactorily with all alarm and
.. indication functions working as required. Acceptance criteria for- .O verifying-control rod position vs.-DRPI output during rod movement was transferred to procedure ETT-SF-07014, CRDM Timing Test (ambient).- -
DISCUSSION Performance of-this test involved five,(5) separat'e test sections; O (1) nelf Accuracy Test; (2) Rod Bottom Test; (3) Rod Error Test; (4) Simulated Data Transmission-(Normal Codes); and (5) Sinulated Data Transmission (Error. Codes) Test. The Half. Accuracy Test checks the ability'of DRPI to signal applicable warnings when either A data or B data from a Data Cabinet is
- O: rejected (simulating Failure). This is performed.by using the Accuracy Mode switch on the rear of the DRPI display in the control room. This section was performed satisfactorily.
The' Rod Bottom Test checks the ability.ofJthe DRPI to signal the applicable ~ warnings when rod' bottom codes are manually, inputted to the
.O DRPI aisplay by means of the RBT-1 and RBT-2 test push-buttons. A problem was encountered with the RPI Rod DEV' annunciator not working which was due to an annunciator input wire being lifted per a temporary modification., The temporary modification was restored and the annunciator worked satisfactorily after which this section was performed
-satisfactorily.- In addition,'it was discovered at the end of this O section.that the RPI: ROD DEV annunciator'would not clear; however, this was determined to be a normal condition'whenever the shutdown banks are less~than 210' steps and the control banks are not greater than 12 steps.
~
The= Rod Error Test checks the ability oflthe DRPI to. signal
> appropriate warnings when rod error codes are manually inputted to the 10" DRPI display by means of the ETA and ETB test push-buttons. . This test
! section_was. performed satisfactorily. The: Simulated Data Transmission (Normal Codes) Test checks the H L ability'of the DRPI-to display the appropriate rod positions by setting the switches'on-the test / monitor cards in the Data Cabinets to the (O. appropriate position.c-The results are viewed at the-DRPI display in the
~
control room to verify ~the correct codes have been transmitted without-error'andithe' correct rod positions are displayed. . This test section was! performed = satisfactorily. O .- 55' -- s
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~
The Simulated Data Transmission (Error Codes) Test check the ability of the DRPI.to signal the appropriate warnings whea rod position.
;O- codes contain errors or differ by an excessive amount. This is done by inputting test signals to the test / monitor cards and checking the DRPI display-to verify that the error'have been correctly identified and the proper alarms displayed. This test section was performed satisfactorily.
.:O 4 ,iO f-1 10 i l O O 10 i. 10' 4 i + (0 1 ?- !O -
6
' 3.6 RTD BYPASS FLOW MEASUREMENT (S-07BB01)
Procedure ETT-BB-07010 0.- 0BJECTIVES The objectives of this procedure were to determine the flow rate necessary to achieve the design reactor coolant transport time in each () resistance temperature detector (RTD) bypass loop, to measure the flow rate in'each RTD bypass loop to ensure that the transport times are *
. acceptable, and to set and verify the RTD bypass loop low flow alarm setpoint.
SUMMARY
OF RESULTS O The flow rate in the hot leg RTD bypass loop of reactor coolant loops 1, 2, 3 and 4,'with flow in both the hot and cold bypass loops was greater than or equal to the minimum required flow for the design transport time of one second. () The results are tabulated below: MINIMUM REQUIRED BYPASS ACTUAL BYPASS LOOP FLOW (GPM) FLOW (GPM) 1 100.22 115 2 99.72 153
- C) 3 100.47 157 4 111.03 129 The flow rate in the cold leg RTD bypass loop of reactor coolant loops 1, 2, 3 and 4 with flow in both the hot and cold leg bypass loops C) was greater than or equal to the minimum required flow for the deLign transport time of'one second.
The results are tabulated below: MINIMUM REQUIRED BYPASS. ACTUAL BYPASS () LOOP FLOW (GPM) FLOW (GPM) 1 64.70 147 2 63.38 109 3 64.98 137-4 65.47 133 O Low flow alarm setpoints were determined and the instruments were adjusted within tolerance. The low flow alarms were verified. DISCUSSION () Flow in the hot leg bypass loops was. measured with no flow in the cold leg bypass 'ioops; likewise, flow in the cold leg bypass loops was measured with no flow in the~ hot leg bypass loops. To verify.that Acceptance Criteria were met.-the~following equations were used. 'O - !
,, _ _ ~ . _ , - . - .__ , _ . - _ . , -,-. _
7-; O: 1 F F* h" O: (1+ F c h} F I =--F -F 1 t h O-
~ WHERE:1 P =. Calculated hot leg RTD bypass flow with flow in both the !
h - hot and-cold leg bypass loops. F1 = Calculated cold leg RTD bypass flow with flow in both the hot and cold leg bypass loops. . O- - F = Measured hot leg bypass flow. h-F = Measured cold leg bypass flow. F .= Measured total ~ bypass flow.- ; i O O.
}
5 O i O
- O 3
0 _
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U O 3.7J PRESSURIZER HEATER AND SPRAY CAPABILITY TEST (S-07BB02)
- Procedure ETT-BB-07020
- 0BJECTIVES The objectives of thin procedure were to determine the rate of pressure reduction caused by fully opening the pressurizer spray valves and the rate of; pressure increase from the operation of all pressurizer O heaters.
SUMMARY
OF RESULTS-The pressurizer was depressurized to 1985 psig in 124 seconds. It O was also pressurized to 22F5 psig in 156 seconds. This met the acceptance criteria. The depressurization and pressurization are Ldepicted in Figures 3.7.1 and 3.7.2, respectively. DISCUSSION The depressurization was performed by placing the pressurizer level controller in manual, .de-energizing the pressurizer heaters and then opening spray valves PCV-455B and PCV-455C. The time that it took the Reactor Coolant System (RCS) to depressurize to 1985 psig was then recorded. The pressurization was performed by placing the pressurizer level controller in manual, placing the spray valves in manual, placing a power signal to the control group heaters so they_wouldn't cycle off,
~ placing the-PORV's in the closed position, and then energizing the g pressurizer heaters. The time that it took the RCS to pressurize to V 2285 psig was then recorded.
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f-O 3.8 REACTOR COOLANT SYSTEM FLOW MEASUREMENT (S-07BB03) Procedure ETT-BB-07030 0 OBJECTIVES-Tha *jectives of this test were:
- 1. To confirm that Reactor Coolant System (RCS) Flow with the
.O- core installed and at hot standby conditions, but before initia1' criticality, is equal to or greater than 90% of the thermal design value by measuring loop elbow differential pressures and using them to compute RCS Flow.
- 2. To confirm using calorimetric data taken at 50, 75, 90 and O- 100% power, that RCS Flow is equal to or greater than the thermal design value.
SUMMARY
OF RESULTS Based on calculations performed at the various plant conditions - O hot standby and 50, 75, 90 and 100% power - all acceptance criteria were met and'the test was performed satisfactorily. A compilation of the required flow rates versus the actual flow rates for each power level is provided below: Power Level Required Flow (GPM) Actual Flow (GPM) Hot Standby 344,520 432,682 50% 382,800 412,823 75% 390,838 397,100 90% 390,838 411,862 100%- 390,838 .411,100 DISCUSSION-The method for determining RCS Flow during precritical testing (at
~
hot standby) was by measuring loop elbow differential pressure (AP); This was done by measuring the output voltage from each loop elbow tap O differential pressure transmitter-(3 per-loop), correcting the readings for' instrument drift and then converting the voltages to AP (in. H O). 2 The three transmitter flows were then averaged to determine each loop flow rate. The total RCS Flow was determined by adding up the flow for , each loop'and was. calculated to be 432,682 G*M which met the acceptance criteria of - 344,520 GPM. This flow calculation is not as precise as O using. calorimetric data obtained during power operations, but provides sufficient accuracy for plant startup prior to power ascension.
'For determining the RCS Flow during power ascension, hot leg temperature'(T and. steam generator thermal power ),coldlegtemperature(T'kkn)edfromtherespective
)dataforeachloop'wasohi O Thermal Power Measurement and Statepoint Data Collection procedure for
- O i
, , . _ _ _ _ . - . _ _ . _ __ _ _ ,.. . _. - .;_.._-, _._ _ . ._.w... , -
~
F C) each power level. From this data and using values obtained from the ASME Steam Tables, flow was calculated by the following formula:
- C) 0 (Vc (Qg - 12 x 10 ) )
FLOW (GPIO = (hHot .02
-hCold) where h is the enthalpy for each loop T 3 and T C1 and V is the L> specific volume for each cold leg. Flowsforeac0koopwerethanadded -
together to obtain a total RCS Flow Rate. The calculated flows for each power level met'the specified' acceptance criteria as delineated in the previous section.
'O
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- 63'-
f. 0: 3.9 REACTOR COOLANT SYSTEM FLOW C0ASTDOWN (S-07BB04) Procedure ETT-BB-07040 ' O . OBJECTIVES The objectives of this procedure were to measure the rate of which reactor coolant flow changes subsequent to various reactor. coolant pump : trips, to compare the measured flow coastdown profiles assumed in the O FSAR Safety Analysis, to verify that the reactor coolant system low-flow j trip delay times are less than or equal to the icw-flow trip delay times assumed in the FSAR Safety Analysis, and to verify that the reactor ,
, coolant pump undervoltage trip and underfrequency trip delay times are
-within design specifications.
O .
SUMMARY
OF RESULTS I The: slope of the inverse of flow versus time curve, for the 4/4 case Figure 3.9.1 was 0.090, compared with the slope of the inverse of Figure 3.9.2 of.0.078. Because the measured slope was greater than that assumed in the FSAR Safety Analysis, Westinghouse was requested to O' evaluate the impact. In ensuing response. Westinghouse stated that following a re-analysis of FSAR Chapter 15.3.2, all safety criteria were met. _ The re-analysis will be included in the next revision of the FSAR. Figure 3.9.2 shows the data that was taken during the performance of this procedure superimposed on FSAR Figure 15.3-9. The best estimate , O actual core flow is not greater than or equal to that assumed in the FSAR Safety Analysis.(FSAR Figure 15.3-9). Westinghouse re-analyzed FSAR Chapter 15.3-2,'and determined'that all safety criteria were met. Figure 3.9.3 shows that data that was taken during the performance- I of this procedure superimposed on FSAR Figure 15.3-1. The best estimate O actual core flow for the 2/4 case was not greater than that assumed in the FSAR Safety Analysis and hence, assistance was requested from Westinghouse to evaluate the impact. The Westinghouse response states-that following a re-analysis of.FSAR Chapter 15.3.2,'all safety criteria were met. O Table 3.9.1 shows the data that was used to verify that the trip delay times were acceptable. At the time the test was performed, it was assumed that the gripper response times were the same as the acceptance criteria of procedure ESP-SA-02415, (Response Time Test For Steam Generator Level High-High Feedwater Isolation and Turbine Trip) namely 0.15 seconds. Since that time, successful completion of procedure O ESP-SA-02415 indicates that this was'a conservative assumption'and hence, all safety criteria and design specifications were met. DISCUSSION i '
- O The procedure for this test involved tripping two or four reactor i coolant pumps (depending upon the portion of the' test being conducted) l from the full-flow condition to produce a flow coastdown.- A strip chart l .
recorder was employed'to record loop. flows and reactor coolant pump-i l0 a-
,-4m--w-.,.,,,--+ - - . , , -# -. - __4 ,_. , _ , , _ , , , . , . , , ,#,, , , , , , , , , , , , , , . , _ , , , ,, ,,._,, ,y_
F
- 7-l breaker p>sition. The Reactor Trip Time and Longest Flow Sensor Re-sponse Times were obtained from procedure CS-03SB01 (Reactor Protection O ' Logic System Pre-Operational Test). The gripper release times used were the acceptance criteria of ESP-SA-02415, as the gripper release times had not been measured at the time of the test.
The loop flow values were normalized to the loop flows existing with all reactor coolant pumps in operation and were adjusted for sensor O time delay. These normalized flows were used to obtain a second order least squares fit for each curve and these fitted curves averaged to obtain the total core flow ratio for each scenario. These corrected total core flow ratio curves were compared to the FSAR curves. O O O O O O-O O l
- o. '
TABLE 3.9.1 TRIP DELAY TINES O Reactor Trip ( ) Longest Sensor ( Gripper ( Total Acceptance Release Tim Time Criteria Breaker Trip Time Ruponse Time O Reactor Coolant 0.15 1.245 4.1. 5 0.08 1.150 Pump ' Undervoltage Reace r coolant 0.15 0.385 / 0.6 O Pump 0.08 0.155 Underfrequency 0.15 0.315 f;1.0 0.08 0.085 Low Flow o (1) All times in seconds. (2) From Prs-Operational Test CS-03SB02, Data Sheet 8.4.1, and Supplemental Record (JTG Punchlist Item #181) . O (3) usins Acceptance criteria from ESP-SA-02415. O l l O l 0-O l l u_____._______._____ _ . _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ . - _ _ _ _ _ _ _ _ _ _ . . _ . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ . _ . _ _ - _ . _ _ _ _ _
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0 2 '4 6 8 10 TIME (SECONOS) 'o 'O $ uATA P0lNT 4 CALLAWAY NUCLEAR PLANT FSAR FIGURE 15.3-9 lO CORE FLOW COASTOOWN FOR FOUR l LOOPS IN OPER ATION, FOUR PUMPS COASTING DOWN 1 fC
O FIGURE 3,9,3 1.2 g g l l O 3* 1.0 t~ 0.8 - O 5 4 5 6 0.6 - 5 O $ 0.4 - m LAJ 8 0 0.2 - 0 0 1.2 O l.0 dC I50.8 - O zw
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O FSAR FIGURE 15.3-1 FLOW TRANSIENTS FOR FOUR LOOPS IN OPERATION, TWO PUMPS COASTING DOWN O i
O 3.10 PRESSURIZER CONTINUOUS SPRAY FLOW VERIFICATION (S-07BB05) Procedure ETT-BB-07050 OBJECTIVES The objectives of this test were to establish a setting for the pressurizer continuous spray throttle valves to obtain an optimum continuous spray flow and to establish the setpoint for the pressurizer U, spray line low temperature alarms.
SUMMARY
OF RESULTS The continuous spray flow valves BB-V-082 and BB-V-083 were .O . throttled to Sis and 5 turns open, respectively. This established a break-point temperature of 539'F for both pressurizer spray loops. The setpoints for the pressurizer spray line low temperature alarms were established at 504*F. O DISCUSSION This test throttled pressurizer continuous spray flow valves BB-V-0082 and BB-V-0083 and established a " break-point". The
" break-point" is defined as the point where a change in valve position produces an abrupt change in spray line equilibrium temperature. The O valve is left in this position in order to maintain an optimum continuous spray flow. This optimum spray flow is intended to avoid thermal shock to the spray lines in the event of initiation of spray flow. This test also establishes the setpoint for the pressurizer spray line. low temperature alarms as the equilibrium temperature at the
" break-point" minus 35215'F. The data obtained by throttling these O valves are presented in Figures 3.10.1 and 3.10.2.
O o 0- 7o _ c _ _ - - _ _ _ _ _ _ _ - - - _ _ _ _ _ - - _ _ _ - _ _ - - _ _ - _ - _ _ - _ _ - - _ _ _ _ _ _ _ - _ - - - _ _ _ _ _ - _ _ _ _ _ - - _ _ _ - - _ _ - _ - _ _ - - _ _ -
~
'O O O 9 O O O
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Spray Line Equilibrium Temperature - Vs. 88-V-0083 Position 580
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() -- t 510 0 1 2 3 4 5 6 7 8 9 10 88-V-0083 Valve Position (Turns Open)
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Q 3.11 RTD/TC CROSS CALIBRATION (S-07BB06) Procedure ETP-ZZ-00008 O OBJECTIVES The objectives of this test were to verify the expected resistance versus temperature characteristics of installed RTD's at the 435'F and 557'F plateaus, to verify the incore thermocouple response, and to O determine installation corrections for individual RTD's, as necessary.
SUMMARY
OF RESULTS During the test, it was determined that RTD TE-411B had the leads crossed and RTD TE-433A had a bad card. All other RTD's exhibited a O hehavior such that the measured resistance corresponded to temperature that was within 12.0*F of the average indicated reactor coolant system temperature. No acceptance criteria was associated with the data taken at this plateau. After correcting the problems found with RTD's TE-411B and TE-433A, the test was performed at the 557'F plateau. The resistance measured for each RTD corresponded to a temperature within 2*F of the average indicated RCS temperature. It was determined that 10 RTD's indicated a temperature that was greater than 0.3*F different than the average of .O the RTD readings. For the RTD's that indicated a temperature greater than 0.3'F different than the RCS average indicated temperature, the difference was forwarded to I6C for input as a correction to the appropriate loop calibration procedure. Because the acceptance criteria for this procedure was that at the O 557 F plateau, all RTD's must indicate a temperature within 2*F of the RCS calculated average indicated temperature and that any RTD indicating a temperature different from the average of the RTD temperature indications by more than 10.3*F must be corrected, all acceptance criteria for this test were met. O At the 435'F plateau, the steam pressure saturation temperature was not within 12*F of each of the thermocouple readings. This was attrib-uted to steam line pressure losses at a lower steam pressure. At the 557'F plateau, the thermocouple map average temperature was not within 12*F of each of the individual thermocouple indications. O This was due to three thermocouples that were giving an indication approximately 30*F below the average. When the three thermocouple readings that are questionable are removed from the average, each individual thermocouple reading is within 12*F of the average. DISCUSSION This procedure was performed at two temperature plateaus during the heatup phase, 435'F and 557'F. The data obtained at the 557'F plateau was used for development of the individual RTD correction factors. O _ 73 _
O l Individual RTD readings were obtained by pulling the appropriate card j
.from the Reactor Protection cabinets and reading the RTD resistance J . utilizing a card edge connector and a digital voltmeter. Two RTD's were
-O-read off.each card' edge connector, the second being a spare. Each RTD was read four different times, reversing the data taking sequence each time. The four RTD resistance readings were then averaged and converted i to temperature using the manufacturer's data. Each indication was then compared with the average indicated RCS temperature to determine the O installation correction factor. Before and after all the data was collected at.each plateau, a thermocouple map was taken. The average RCS temperature was compared with the average of the temperatures obtained from the incore thermocouple map and the temperature calculated from the steam pressures recorded-for each loop. O - T.stins at the 557'F plateau was performed satisfactorily after some difficulty was encountered in establishing stable RCS temperature. Review of the data determined that the mean temperature deviation was 0.50*F, excluding RTD TE-421B, having a deviation of 2.35'F. An analysis performed by I&C indicates that erroneous data was obtained for TE-421B during the test performed at the 557'F plateau. This was O. further substantiated in that data taken at the 435'F plateau did not
-indicate a problem and additional data taken at a later date did not indicate a problem. Hence, no corrective action was taken for TE-4218.
During the comparison of the calculated RCS average temperature and the incore thermocouple map, two problems were encountered. At the O 435'F plateau, the steam pressure saturacion temperature was not within 12'F of each thermocouple readings. It was determined that this is due to line losses at a lower steam pressure. At'the 557'F plateau, the thermocouple map average temperature was not within 12*F of each of the thermocouple readings.. This was because three thermocouples, (E-02, G-12 'and L-08) were reading approximately 30*F lower than the other O thermocouples, when the three erroneous thermocouples were eliminated from the average, the average thermocouple temperature was within 12*F of each of the thermocouple readings. O O O O _ 74 _
O 4.1 INITIAL' CRITICALITY (S-070004) Procedure ETT-ZZ-07040 0-OBJECTIVES The objective of this procedure was to achieve initial reactor criticality in a cautious and controlled manner. O
SUMMARY
OF RESULTS Initial criticality was achieved at 0422 on October 2, 1984. After power was leveled out above the source range, Control Bank D was at 147 steps with a boron concentration of 1311 ppm. The design critical position was Control Bank D at 160 steps with a boron concentration of O 1292 ppm. DISCUSSION This procedure provided for a slow approach to initial criticality. The approach to criticality was initiated by withdrawing Shutdown Banks O A, B, C, D and E, stopping at 50 step intervals to obtain inverse count rate ratio (ICRR) data. Control Banks were then withdrawn in 50 step increments to a final position with Bank D at 160 steps. The reactor coolant system was then diluted at 60 gpm to reduce the boron concentration. During the dilution, the reactor coolant system was sampled every 15 minutes.and the ICRR plotted. When the ICRR reached O .1, the dilution rate was reduced to 30 gpm. When criticality was achieved, reactor power was leveled off above the source range and critical data was obtained. O O O O iO - 9 4.2 INVERSE COUNT RATE RATIO MONITORING FOR APPROACH TO INITIAL CRITICALITY (S-070003) GD Procedure ETT-ZZ-07030 OBJECTIVES This procedure was used in conjunction with the initial criticality procedure to predict when criticality would occur by plotting Inverse GD Count Rate Ratios (ICRR).
SUMMARY
OF RESULTS This procedure was completed satisfactorily.
? DISCUSSION During rod withdrawal, the ICRR was computed at 50 step increments.
During Reactor Coolant System (RCS) dilution, the ICRR was calculated every 15 minutes (concurrent with boron concentration sampling) and plotted as a function of time, boron concentration, and gallons of water GD added to the RCS. When the ICRR reached 0.1, the ICRR's were renormalized to 1.0, the dilution rate was reduced, and the ICRR's were plotted at 5 minute intervals. This was continued until the reactor was critical. The ICRR plots are shown in Figures 4.2.1 to 4.2.18. .) J m/ 7
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