ML20204H007
| ML20204H007 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 03/10/1987 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20204G974 | List: |
| References | |
| NUDOCS 8703260523 | |
| Download: ML20204H007 (135) | |
Text
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es THE DETROIT EDISON COMPANY PERMI 2 EUCLEAR POWER PLAllT INTERIM STARTUP TEST REPORT SUPPLEMENT E0 3 March 10, 1986 i
I 1
8703260523 870320 PDR ADOCK 05000341 l P PDR
Supples;nt 3 FERMI 2 EUCIJtAR POWER PLANT INTERIM STARTUP TEST REPORT
% INDEI 1.0 Introduction 1.1 Purpose 1.2 Test Report Format 1.3 Plant Description 1.4 Startup Test Program Description 1.5 References 2.0 General Test Program Information 2.1 Chronology of Major Events 2.2 Matrix of Test Completion Dates 30 Test Results Summary 31 Chemical and Radiochemical 32 Radiation Measurements l 3.3 Fuel Loading 34 Full Core Shutdown Margin 35 Control Rod Drive System 36 3RM Performance and Control Rod Sequence l
37 Water IAvel Measurements l 38 IRN Performance l 39 LPRM Calibration 3 10 APRN Calibration 3 11 Process Computer 3 12 RCIC 3 13 HPCI 3 14 selected Process Temperatures 3 15 Systen Expansion 3 16 (Deleted) 3 17 Core Performance 3 18 (Deleted) 3 19 (Deleted) 3 20 Pressure Regulator 3 21 Feedwater system 3 22 Turbine valve surveillance 3 23 MsIV 3 24 Relief Valves 3 25 Turbine stop valve and Control valve Fast Closure
Supplement 3 PERNI 2 NUCLEAR POWER PLANT INTERIM STARTUP TEST REPORT INDEI 30 Test Results Summary (Continued) 3 26 Shutdown from Outside Control Room 3 27 Flow Control 3 28 Recirculation System 3 29 Loss of Turbine-Generator and Offsite Power 3 30 Steady-State vibration 3 31 Recirc. Systen Flow Calibration 3 32 Reactor water Cleanup Systea _
3 33 Residual Heat Removal System 3 34 Piping System Dynamic Response Testing l .
l l
Supplement 3 FOREWARD !
This Supplementary Startup Test Report includes the testing performed i since the previous interia summary report dated December 8, 1986. t This report was transmitted to the NRC via VP-86-0177 dated December l 17, 1986. Since that report was issued Fermi 2 has completed all of I the tests required for Test Condition One, and Test Condition Two '
, testing is in progress.
In this supplement we are transmitting an updated copy of the entire l test report. Revision bars have been added to show where changes have been made, except for changes which are only cosmetic in nature or which only involve renumbering sections or pages.
The results sections of this report will be filled in as the tests are completed in the future.
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i Supplesint 3 Page 1-1
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FERMI 2 NUCLEAR POWER PLANT INTERIM STARTUP TEST EPORT 1.0 Introduction 1.1 Purpose The purpose of this Interim Startup Test Report and its associated supplements is to provide a suanary of the test results obtained in startup testing completed from initial fuel load to the present. This report of plant startti and power ascension testing is submitted as required per Technical Specification 6.9.1.1. This interia report plus
, its supplements cover all testing applicable to the test -
conditions completed as described in FSAR Subsection 14.1.4.8. Supplements will be issued as the remaining l testing is completed, at the intervals specified per Technical Specification 6.9.1 3 Included in this report are descriptions of the esasured values of the operating conditions and characteristics obtained during the test program and any corrective actions that were required to obtain satisfactory operation.
1.2 Test Report Format Sections 1.0 and 2.0 of this report provide general information about the Fermi 2 plant and the testing progras. Section 3 0 provides a basic description of the testing we have performed along with a summary of the i
results and analysis obtained from each test. Each test
- summary is divided into three subsections covering the l
purpose, test criteria, and results of each test.
1 3 Plant nascription The Persi 2 Nuclear Power Plant is located in Frenchtown Township, Monroe County, Michigan. The Nuclear Steam Supply System consists of a General Electric BWR 4 nuclear reactor '
rated at 3292 Mut, coupled to an English Electric Turbine / Generator rated at 1100 MWe, constructed in a Mark I !
containment with a toroidal suppression pool.
This plant is owned and operated by the Detroit Edison Company and the Wolverine Power Cooperative, Incorporated.
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Supplcc:nt 3 Page 1-2 1.4 Startup Test Program Description The Startup Test Phase began with preparation for fuel loading and will extend to the completion of the warranty demonstration.~This phase is subdivided into four parts: ,
- 1. Fuel Loading and Open Vessel Tests
- 2. Initial heatup 3 Power tests
- 4. Warranty demonstration i
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-Cooled Power Reactors."
The overall objectives of the Startup Test Phase are as *
. follows:
- 1. To achieve an orderly and safe intial core loading
- 2. To perform all testing and measurements necessary to 1 determine that the approach to initial criticality and the subsequent power ascension are accomplished safely and orderly 3 To conduct low-power physics tests sufficient to ensure that physics design parameters have been met
- 4. To conduct initial heatup and hot functional testing so that hot integrated operation of specified systems are shown to meet design specifications
- 5. To conduct an orderly and safe Power Ascension Program, with requisite physics and system testing, to ensure
, that when operating at power, the plant meets design intent
- 6. To conduct a successful warranty demonstration program
- Tests conducted during the Startup Test Phase consist of
- Major Plant Transients and Stability Tests. The remainder of tests are directed toward demonstrating correct performance of the nuclear boiler and numerous auxiliary
, plant systems while at power. Certain tests may be ;
identified with more than one part of the Startup Test Phase. Figure 1-1 shows a general view of the Startup Test Phase Program and should be considered in conjunction with 1
1 4
Supplement 3 Paga 1-3 Figure 1-2 which shows, graphically, the various test areas as a function of core thermal power and flow. Note that Figure 1-1 has been modified to reflect certain tests which
- we presently intend to delete from the Startup Test Program, as discussed further in Reference 1.5 3 .
For a more comprehensive description of the testing program refer to Reference 1.5.2.
1.5 References the following is a list of documents that provide supplementary information of the Fermi 2 Startup Test Phase Program:
- 1. Fermi 2 Technical Specifications, Section 6.
. 2. Final Safety Analysis Report, Fermi 2 Nuclear Power
- Plant, Section 14.
f 3 Memorandus VP-86-0141, "Startup Test Program Changes",
dated October 17, 1986, from Frank E. Agosti to James G.
Keppler.
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Supplerent 3 Page 1-4 FIGURE 1-1 STANNP TEST PROGRAM e
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SupplemInt 3 Page 1-5 I
FIGURE 1-2 APPICIIllATE F0lfER PIAlf IIAP SWifIllG STARTUP TEST EX)llDITICIES 1
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C. Analyteent lessor lesnit of snesier peuver mee.eentree TC 6 g _ D. Anelyties' upper famit of snester poseer ,
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Typical Stettup Path -
Region 111 l 0 ' ' ' ' ' ' ' '
3 i s I a a i i i i l
0 10 20 30 40 50 to 70 80 to 100 110 Pereentage of Core Flow 3 Notes:
- 1. See Figure 1-1 for startup test titles.
- 2. Power in percentage of rated thermal Power 3292 miT.
3.coreflowgnpercentageorratedcorerecirculationflow.
100.0 x 10 lb/hr.
- 4. TC = test condition.
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Suppiscsnt 3 '.
Page 2-1 2.0 General Test Program Information 2.1 Otronology of Major Events Date Received (55) Facility Operating 03/20/85 License No. NPF-33 Started Fuel Loading 03/20/85 Completed Fuel Loading 04/04/85 Completed Open Vessel Testing 06/01/85 Initial Criticality 06/21/85 s Received (Full Power) Facility 07/15/85 ~
Operating License NPF-43 Completed Initial Turbine Roll 09/26/85 Bypass Line Replacement / 10/10/85 Environmental Qualification Equipment Upgrade Outage Begins Neutron Source Changeout Complete 05/12/86 Outage Ends 07/24/86 Reactor Restarted 08/04/86 Completed Test Condition Heatup 09/03/86 Entered Test Condition One 09/16/86 Initial Synchronization to Grid 09/21/86 l .
Condenser Repair Outage Begins 11/08/86 l Reactor Restarted 12/18/86 Completed Test. Condition One 01/07/87 Main Steam Line Instrument Tap 01/09/87 Repair Outage Begins Reactor Restarted 01/24/87 Entered Test Condition Two 02/24/87
--- -._,,_,,-,.-.,.m., . , , ., , _ - _ .
i 2.2 metria of Test Completten Dates Test Pre-Fuel Open TC TC TC TC TC TC No Test Title Loed Wesset Heatup One Tee Three Four Five Sla 1 Chemicot and Redlochemical 01/14/85 >>> 07/16/85 10/27/96 >>> >>>
2 Radiation measurements 01/24/85 04/19/85 07/14/85 10/17/86 03/08/87 >>> >>>
l 3 Fuel Loading >>> 04/04/85 >>> >>> >>> >>> >>> >>> >>>
4 Full Core Shutdown margin >>> >>> 04/10/85 >>> >>> >>> >>> >>> >>>
5 CEO >>> 04/05/05 09/16/85 10/23/86 >>> >>> >>> >>>
6 SRM Performance and Control >>> >>> 08/25/85 10/05/86 >>> >>> >>> >>> >>>
Rod Sequence 7 water Levet measurements >>> >>> 08/30/85 >>> >>> >>> >>> >>> >>>
8 IRM Performance >>> >>> 08/04/86 01/07/87 >>> >>> >>> >>> >>>
j 9 LPRM Calibretton >>> >>> 09/24/85 01/03/87 >>> >>> >>>
40 APRM Calibration >>> >>> 08/01/85 01/04/87 $$$ >>>
11 Process Computer 05/30/85 >>> >>> 01/02/87 02/09/87 >>>
12 RCIC >>> >>> 09/01/86 10/15/86 11/03/86 >>> >>> >>> >>>
13 HPCI >>> >>> 09/01/86 12/29/86 >>> >>> >>> >>>
14 Selected ProcessTemperatures >>> >>> 09/01/86 >>> >>> >>>
15 System Espension >>> 06/12/85 09/05/86 11/04/86 >>> >>> >>> >>>
i 16 Core Power Otstributton >>> >>> >>> >>> >>> >>> >>> >>> >>>
l (Deleted) l 17 Core Performance >>> >>> >>> 01/04/87 $$$
19 Core Power Vold Mode Response >>> >>> >>> >>> >>> >>> >>> >>> >>>
4 (Deleted) 20 Pressure Regulator >>> >>> >>> 10/21/86 S$$ >>>
21 Feedwater System >>> >>> 07/09/85 10/21/86 555 >>>
I 4 22 Turbine Valve Survettlance >>> >>> >>> >>> >>> >>> >>> >>>
23 MSIV >>> >>> 07/12/85 10/08/86 >>> >>> >>>
{ 24 Rollef Velves >>> >>> 07/03/05 >>> S$$ >>> >>> >>> >>>
i 25 Turbine Stop Velve and Control >>> >>> >>> >>> $$$ >>> >>> >>>
valve Fast Closure 26 Shutdown from Outside Controt >>> >>> >>> 10/23/86 >>> >>> >>> >>>
1 Room
)
! $$$ = Testing to be performed in this test condition, but not yet completed.
, >>> = No testing necessary for this test condition.
Supplement 3 Page 2-2 9
i j
j 2.2 notets of Test Cegletten Dates (Continued)
Trst Pre-Fuel Open TC TC TC TC TC TC No Test Title Lead Vessel Hestup One Tee Three Four Ftwo $1s 27 Flow Control >>> >>> >>> >>> 03/08/87 >>> >>>
28 Recleculation System >>> >>> >>> >>> 03/05/87 >>>
29 Loss of Offstte Power >>> >>> >>> >>> $$5 >>> >>> >>> >>> <
30 Vtbration Measurements >>> >>> 07/03/85 10/04/86 $55 >>> >>>
31 Rectrc. System Flow Callbretton >>> >>> >>> >>> >>> >>> >>>
32 Reactor water Cloenup System >>> >>> 07/14/85 >>> >>> >>> >>> >>>
33 Residual Heat Removal System >>> >>> 08/31/85 >>> >>> >>> >>> >>>
34 Piping System Dynamic Response >>> >>> 09/24/85 >>> $$$ >>> >>>
i 1
555 = Testing to be performed in this test condition, but not yet completed.
>>> = No testing necessary for this test condition.
Supplement 3
$ Page 2-3 f
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Supplement 3 ,
Paga 3 1-1 30 Test assults sm eary 31 chmaical and nadiochemical 3 1.1 Purpose The principal purposes of this test are to collect information on the chemistry and radiochemistry of the Reactor Coolant and Support Systems, and to determine that the sampling equipment, procedures and analytic techniques are adequate to ensure specifications and process requirements are met.
Specific purposes of this test include evaluation of fuel performance, evaluations of filter desineralizer operation by direct an'd indirect methods, confirmation of condenser integrity, demonstration of proper steam separator-dryer ,
operation, measurement and calibration of the off-gas system and calibration of certain process instrumentation, if required. Data for these purposes are secured from a variety of sources:
plant operating records, regular routine coolant analysis, radiochemical measurements of specific '
nuclides and special chemical tests.
3 1.2 criteria Level 1 Chemical factors defined in the Technical Specifications and Fuel Warranty must be maintained within the limits specified. Water quality must be known at all times and remain within the guidelines of the Water Quality Specifications.
The activity of gaseous and liquid effluents must conform to license limitations.
Level 2 None 1
313 assults Prior to loading fuel, appropriate chemistry data was taken. All data remained within criteria levels except for feedwater conductivity and feedwater copper concentration. These values could have been
, elevated due to low condenser vacuum, minimum filter
- desineralizer flow and low sample flow rates.
Supplement 3 Peg 2 3 1-2 During heatup test condition, these values were within acceptable limits. See Figure 3 1 for specific information on pre-fuel load chemistry data.
During the heatup test cbndition, all chemistry data taken fell within applicable limits except for Control Rod Drive (CRD) dissolved oxygen levels.
These levels are expected to decrease during further test conditions with greater steam flow and the steam jet air ejectors in service which will more effectively purge gases from the condenser. Refer to Figure 3 1 for heatup chemistry data.
The Test Condition One data in general remained within acceptance criteria limits. Reactor water chemistry and radiochemistry measurements were made at a time when plant confitions were fairly stable.
Reactor power was at 17), the turbine was rolling but with no electrical output load. Analysis of the results showed the coolant to be well within the Technical Specification limits on all parameters.
Radiochemistry analyses of the coolant showed activity levels and isotopes present to be normal for this power level and core exposure. The Dose Equivalent I-131 result was far below the Technical Specification limit of 0.2 uC1/ga. In Test Condition One, the staan jet air ejectors were in service resulting in low condensate, Condensate desineraliser effluent, and CRD dissolved oxygen levels. The high CRD dissolved oxygen level which was of concern during the heatup test condition is no longer considered to be a problem.
It should be noted that Reactor Conductivity varied considerably during the Test Condition One period. l Conductivity has, on several occasions, even exceeded the Technical Specification values of 1.0 unho/cm for several hours. It was determined that the increase in conductivity was directly related to placing the Generator on line and increasing Generator load. One possible explanatalon is that operation of the Generator is causing the paint that i
was previously used to coat the internals of the Moisture Separator Reheater (MSR) and the Main l Turbine to be carried into the condenser hotwell, thus causing the increase in Reactor conductivity.
Another contributing factor is felt to be the Krylon coating that was previously used as a preservative coating for the turbine blades, which is now being worn off the blades and into the condenser. Further investigation is continuing to determine the exact i
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Supplement 3 Pagi 3 1-3 cause of the conductivity increase. This situation
' seems to be improving as the plant continues to operate for longer periods at increasing power levels. Efforts were made during the condenser outage to remove paini from accessible areas in the !
MSRs and LP turbine exhausts. Mechanical cleaning l by wire brushing and vacuuming was performed on the ;
4 MSR's interior shell surface and hydro-lasing of the three LP turbine exhaust extensions to the condenser l was performed.
l i
Both Condensate Desineralizer Effluent and Feedwater dissolved oxygen levels at Test Condition One were '
less than 10 ppb, which are outside of the limits of 20 102 1200 ppb. The problem of low condensate /feedwater dissolved oxygen has occurred ;
during the startup of other operating plants. The resolution at this time is to simply continue to monitor these parameters at higher power levels to see if the levels will increase with power. If dissolved oxygen levels do not increase to greater than 20 ppb by 100$ power, it may become necessary to inject oxygen into the feedwater system.
All gaseous and liquid effluent samples obtained during performance of this procedure were within the license limitations. Various radioactive gaseous effluents were analyzed during Test Condition One.
Grab samples were taken in an attempt to correlate t analysis results with actual monitor readings.
However, the activity levels being seen at the j off-gas and ventilation sample points are still too Iow to provide meaningful data. Only one noble gas t
was detected, at a level which was just above the miniaua detection limit. The off-gas monitor readings were also still quite low and variable. '
Low off-gas activity values are normal and expected l at this power level and core exposure.
A sensurement of radiolytic gas in steam was also made at Test Condition One. Analysis results were below the 0.06 cfm/MWt limit. Radiolytic gas is the hydrogen and oxygen formed in the reactor by radiation induced breakdown of water molecules.
Values higher than 0.06 cfm/MWt could exceed the capacity of the off-gas system recombiners.
See Figure 3 1 for more detail regarding the chemistry data taken to date.
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Supplc= nt 3 Page 3 2-1 32 andiation nessurements 3 2.1 Purpose -
The purpose of this test is to determine the background radiation levels in the plant environs for baseline data and activity build-up during power ascension testing to ensure the protection of plant personnel during plant operation.
3 2.2 criteria Level 1 The radiation doses of plant origin and the occupancy times of personnel in radiation zones shall be controlled consistent with the guidelines
.' of the standards for protection against radiation outlined in 10CFR20, ' Standards for Protection
,l Against Radiation", and NRC General Design Criteria.
I (
a' t.
.)
, Level 2
' None J .<
1 '
,/- j. 323 mesults 4 ,
e Radiation measurements were taken in the form of
/' process and area radiation monitor data and site surveys. To date, all data taken has been
' ' ' ' d '
acceptable and personnel radiation protection has
~
/ been provided in full compliance with the criteria.
( ,' .
' 4 i See Figures 3 2-1 through 3 2-3 for applicable i
~
, monitor and survey readings. These Figures reflect l .
l' the results of this test for all the test conditions for which this data has been completed.
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. a. . . -
~. . . - - _ - - . .
l l
Supplement 3 '
Page 3 2-2 FIGURE 3 2-1 (Page 1 of 5)
. Area Radiation Monitor Sensor Isaations thannel No. Imcation (Col.) Floor-Bldg.
1 (F-10) 2nd Fir. Reac. Bldg. (RB) Pers. Air Lock
, 2 (B-9) 1st Fir. RB Equip. Air Lock '
3 (J-13) 2nd Fir. Auz. Bldg. (AB) Access Control 4 i (G-10) 2nd Fir. AB Change Area Control 5 (B-13) 3rd Fir. RB CRD Storage and Maintenance Area 6 (G-13) 3rd Fir. AB Main Control Room (CR) s 7 (F-9) Sub Base. RB 5.E. Corner 8 '. (B-10) Sub Base. RB 5.W. Corner ,
9 (B-15) Sub Base. RB N.W. Corner
. 10 (G-17) Sub Base. RB N.E. Corner 11 (G-11) Sub Base. RB HPCI Rs. )
12 (F-11) 1st Fir. RB Neut. Mon. Eq. Rs.
13 (F-10) 1st Fir. RB Neut. Mon. Control Panel.
14 (A-11) Sub Base. RB Supp. Pool 15 (F-15) 5th Fir. RB Fuel Stor. Pool 16 ,
(F-15) 4th Fir. RB New Fuel Vault 4
17 (F-12) 5th Fir. RB Refuel Area Near Reactor 18 (F-13) 5th Fir. RB Refuel Area Near. Reactor' (High Range) 19 (L-12) 3rd Fir. Turbine Bldg. (TB) Turbine Inlet End 20 (R-10) Base. TB Sump 21 (F-7) 2nd Fir. TB Main Cond. Area 22 (J-4) 1st Fir. TB Decon. Area (M-17) ist Fir. Rad. Naste Bldg. (RWB) Control Rs.
23 24 (N-17) Base. RWB Equip. Drain 3. Pump 25 (P-16) Base. RWB Floor Drain S. Pump
! 26 (R-17) ist Fir. RWB Drum conveyor Aisle Operating Area b Spare 27 28 (G-11) 4th Fir. AB Vent. Equip. Rs. .
29 (B-15) 4th Fir. RB Change Rs. '
30 (H-12) RB Basement Air lack .
31 (B-12) ist Fir. RB Drywell Air Lock Labyrinth 32 (G-13) 1st Fir. AB Near Blowout Pnl.
33 (C-9) 1st Fir. RB South Air Lock 34 (N-2) 2nd Fir. TB Near Off Gas Equip.
35 (R-2) 1st Fir. TB Near 5.J. A.E. Area 36 (K-1) 1st Fir. TB s.W. Corner
~"
37 (M-2) 3rd Fir. TB South End 38 (R-14) Base. RWB Scrap Cement Recovery 39 (L-13) 1st Fir. RWB H.P. Lab 40 (P-16) 1st Fir. RWB Receiving Area 41 (S-17) 1st Fir. RWB Bailing Room 42 (N-16) 1st Fir. RWB Filter Desin. Area 1 l 43 (5-17) Mezz. RWB Washdown Area 44 (S-12) 1st Fir. Service Bldg. (SB) Mach. Shop.
^
Supplement 3 Page 3 2-3 FIGURE 3 2-1 (Page 2 of 5)
Area Radiation Monitor Sensor Locations Giannel No. Imcation (Col.) Floor-Bldg.
845 1st Fir. Inside Drywell
'46 1st Fir. On Site Stg. Bldg. Control Roos 847 1st Fir. On Site Stg. Bldg. Compactor Room
'848 1st Fir. On Site Stg. Bldg. Truck Unloading Station
'The remote indicator is located on Process Radiation Monitor Panel H11-P884 (Relay Room). .
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.b83. I k *: uTe m: 1.y? a p :. * . + l Supplement 3 Page 3 3-1 3 3 ruelImading 331 rurpose The purpose of this test was to load fuel safely and efficiently to the full core size (764 assemblies).
- 332 criteria Level 1 4
The partially loaded core must be subcritical by at least 0 38 percent delta k/k with the analytically s , determined strongest rod fully withdrawn. l There must be a neutron signal count-to-noise count ratio of at least 2:1 on the required operable SRMs , or fuel loading chambers (FLC). The minlaus count. rate, as defined by the Technical Specifications, must be met on the required operable SRMs or fuel loading chambers. Level 2 l hm 333 assults i Prior to fuel loading, all fuel assemblies were inspected and then stored in the fuel pool in such a i way that no rotation of fuel assemblies would be required during their transfer to the reactor vessel and also that no assembly would pass over any other i assembly in the fuel pool during fuel loading. The j only exception to this was bundle LJE 954 which was , ; 4 oriented SW insteed of Sg in the fuel pool, but was verified to be properly oriented in the core. Before the start of fuel load, all control rods were ; fully inserted, all blade guides were positioned as l shown on Figure 3 3-1. seven sb-se neutron sources were installed at locations shown on Figure 3 3-1. i All applicable initial conditions were verified prior to the start of fuel loading. Four times during the fuel loading process, fuel loading was suspended for greater than eight hours, and all applicable initial conditions were reverified before fuel loading was resumed. i i Supples:nt 3 Fag 3 3 3-2 The Bottoa head drain temperature indication was used to obtain the Reactor Coolant Temperature at least once every eight hours (1 15 minutes) during the fuel loading process. Detailed fuel loading sheets, approved by the Reactor Engineer, provided the instructions on each individual fuel assembly to be moved from a specific location in the fuel pool to a pre-assigned location in the core. It also provided the instructions on what control rods were to be exercised for functional and sub-criticality checks for pre-defined core configurations. FLC moves to be made during the fuel loading were also included. Most of the changes required to the fuel loading sheets during fuel loading were to move the FLCs earlier due to high count rates experienced when , fuel assemblies and/or the neutron sources were too close to the FLCs. The only other change involved using Control Ecd 10-27 (instead of 06-27) for a sub-criticality check due to an accumulator problem with Rod 06-27 Four FLCs (one per quadrant) were used to monitor the count rate from the start of fuel loading up to . the point when 532 bundles were loaded in the core. In order to keep the FLC count rate within a desirable range and to accommodate an increasing core size, it was necessary to move the FLCs outward by approximately 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 3-4) ~ The upscale alara setpoint was set at 1 x 105 e and the upscale trip setpoint was set at 2 x 10p ops for each FLC. The downscale rod block setpoint was 3 cps. The FLCs were checked for flux responsa either by control rod pulls during scheduled sub-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 prior to the resumption of fuel loading when fuel loading was delayed for eight hours or more. In addition, the Signal-to-Noise ratio was calculated for each FLC prior to start of fuel load, during any required reverification of plant initial conditions and every time the FLCs were moved to a new location. (See Figure 3 3-2) Supplement 3 Paga 3 3-3 Four S Ms (one per quadrant) were used to monitor 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 SM detectors connected to the SM instrument channels, therodblockandgheupscalegripsetpointswere set down to 1 x 10 and 2 x 10 respectively, i since no previous saturation test was performed on the SM detectors. The down scale rod block setpoint was 3 eps. The SM flux response check was l performed at least once every eight hours during the ' fuel loading process by partially withdrawing each ! l 4 SM. Fuel loading commenced on March 20, 1985 with the l loading of four fuel assemblies around the central neutron source. The loading continued in control cell 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 symmetry about the center source. The thirteen control cells (52 bundles) needed to form a 14 x 14 1 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 symmetric after every four control l cells had been loaded. (See Figure 3 3-3) Control rod functional and sub-criticality checks were performed either after every cell (first 4 cells in the core), or after 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 j cell (s). For each bundle a visual verification was performed to ensure that the bundle was properly grappled before the bundle was lifted from the fuel pool i 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 ungrappled before the hoist was l ' raised. Similar verifications were made for the blade guides lifted out of the core and the FLC moves made during the fuel loading process. i 4 I - - + , . , , - ..---_e-.,.---.m- - - , - - -, . - - .- - -.._. . . .-,-.-.-....._m--w,-. . - - - . - - --,.e-.-. .....--m.. .-- -- Supplement 3 Paga 3 3-4 A day-by-day account of the fuel load progress is given in Figure 3 3-5 Most of the problaas that caused delays were related to the refueling bridge (limit switch, power loss, grapple indication, air hose break, etc.). Fuel loading was halted on Sundays in order to perform required weekly surveillances on FLC/SRMs, IRMs, APRMs and the refueling bridge. During the fuel loading process, FLC/SRM count rates were monitored 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 1/M plot was used to project the estimated number of bundles for criticality. If criticality was projected during the next loading increment then the increment size was reduced , . between 1/M calculations. Strong geometric effects -- were seen, particularly during the first few bundles loaded in the core and also when the bundles were loaded near and FLC. These geometric effects resulted in erronious (but highly conservative) projections which often resulted in very small , increment sizes (1 - 2 bundles) between 1/N calculations. After eighty bundles were loaded in the core, the maximum increment size between 1/M calculations was reduced to one cell (4 bundles except 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 fastener that had to be replaced. Some ' debris was identified in the core on bundles LKJ 398, LJK 506 and LJK 957. After fuel loading was completed, these bundles were pulled out of the core to correct the respective problems and reinserted back into the core. After the 12 x 12 square array of bundles was completed, 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 0 385 delta ' I/K shutdown margin for the existing core configuration. Because the calculated Keff for the 12 x 12 array with the two rods withdrawn was ; 0 9758, and the calculated Keff for the full core j Supplement 3 Pag 2 3 3-5 with only the strongest rod withdrawn is 0 97, sub-criticality for the partial core demonstrated that the shutdown margin would be met throughout the remaining fuel loading process. The fuel loading was completed after fifteen days on April 4, 1985. All criteria were satisfied. t o l l l ) Supples:nt 3 Paga 3 3-6 FIGURE 3 3-1 BUTION SOURCE I4X:4TICII AMD B.ADE WIDE ORIEllTATICII Falon M FullL IAADING ^ N ~ ~~~~ ' f f i / 4 7d d d 4 d d - 7- 7 7 7 7 274 d 7d 4 70 q ~ T /s $ '$ f b k 7 /? 7 '~ 7h 7?~ 5 /. l - X ,' sV i- N} / Mf ?'f 7(V nV Ff # $ XX '/ 1 u_ / / /s 1 / fs < i V_ f4 ri E M f f f/ V1 /s g ' i /s /s I v i t t v d y_ 3_ s_j y ,y v / ?r /ss /s A / hR /s _ ? 7s 7u 7e 7s_ 7sT 7s In -- ?s t/ fs V /, l/ f V 7 l / / N /e / V 7~ 7~ / l~ /~ f V V Vt h f' f f ft ' is ' /o $7s ft /~ 78 78 78 /: fs O y_ t r s v r v v v r id y_ f~ v v / 7 7e / 7: ~ / / / / /t 71 / ' 4 /: /R 7:~ ~ 7:~ 7~ 7~ 7~ 7: 74 ~ 7 / /s ~ X; F H ' f 7' M n7 ns ! If f' n' fg ff f f f f ' - -l f? ic M 7' 7(- zt 7(A>ll 71f, Fif-M7fM7(-M1 l l l l l l i l I i i I l l ll l
- SOURCE (7) / BLADE GUIDE (185) t I
j SupplemInt 3 Paga 3 3-7 FIGURE 3 3-2 l Signal to Boise Measurement , DATE A B C D # OF BUNDLES (TIM) BETECTOR CPS 3/N ,. CPS 3/N CPS SM CPS S/N IAADED ! 03-20-85 FLC 10 24 10 99 10 32 3 10 24 Prior to (2019) fuel load 03-21-85 FLC' 50 49 60 59 50 49 80 79 4 (0005) 03-22-85 FLC' 50 249 50 99 60 149 70 174 48 (0340) s 03-22-85 FLN 6.8 16 3.8 9.8 6.5 64 6.0 5 96 (2005) - 03-22-85 FLC# - - 70 34 - - - - 96 (2227) ! 03-23-85 FLC# 5 4 12 11 - - - - 144 (2110) 03-25-85 FLC 10 19.0 11 14.7 12 19 0 12 14.0 156 (1420) 03-26-85 FLC' 10 49 0 20 89 9 - - - - 196 (0020) 03-26-85 FLC' 38 189 32 159 40 159 4.8 15 260 (1915) 03-28-85 FLC' 30 99 4 39 35 116 2.5 73 388 (1116) 03-29-85 FLC 300 999 100 999 150 374 90 299 440 (0907) 04-01-85 SRM 16 159 12 119 40 399 15 149 532 (1528) #S/N Ratios obtained during FLC moves -FLC not moved l , - - _ , . , . - . - - - - - - . . . , , , . , - . - - . -._.,n.-,,,-._,.. . _ . . . _ - , _ . . . - - _ , _ _ . _ - - . , , . . - . _ _ _ , , _ _ , _ , . , - _ . _ - - . . ._ . ... , .. - . . . . - ~ ~ - Supplement 3 Pag 3 3 3-8 FIGUR3 3 3-3 cone LOADING SEQUENCE y 2 4 6 6 7 43 h ., ,, u , . }.*4,i . - w.m .. g , h At t tj f): u,4. 197sef ,,,4SL)w 'i y11 g,9 did .,4,, d) 7 ) M @ d 9 ' 3 * 'D et M . . . .'y, , , , , , 27 e a' " 7't s'4u3r')16 ... ,o. 4f; __ h 6IgL, 4 p,i , 58
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=. -=_:Z. _-- . _-,. . =5g_. _ _ _ 5 . - ,3 .;-]__ __ l l NUSER OF BUNDLES LOADED i ___..__.,,--__._,a___-..-.-_ Supplement 3 Pass 3 3-12 FIGURE 3 3-7 (Page 1 of 2) FIEL IACATION TERIFICATICII ggM M dii! -:4 M M M MB i. i ~4 M M M M Msi
- t R M R R M & e
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WA t x , f ~, SupplesGnt 3 i/ , Page 3 4-1 j /, N /, a:, .3.4 Full Core S utdown Ilargin , I + 3 4.1 Furpose The purpose of this test is to assure that the ( < reactor.will be subcritical throughout the first i ' cycle with any single control rod fully withdrawn and all other rods fully inserted with the core in , , its maalaus reactivity state. f* i ',7 :[ 3.4.2 criteria b 4- Level 1 j , The shutdown margin of the fully loaded core with 3 the analytically determined strongest rod withdrawn P aust be at least 0 38 percent delta k/k plus R (an , . additional margin for exposure) where R = 0.5
- ;6 percent delta k/k.
jy , ' I4 vel 2 l Q .. , , , , ff Criticality should occur within + 1.0 percent delta k/k of the predicted critical. 'W t ' 4 i- ;343 assults l V The fully loaded core was made critical by withdrawing control rods following the B sequence, using the Reduced Notch Worth Procedure. This y^ .. sequence contained the analytically strongest Rod f 06-39, which was fully withdrawn before reaching criticality. Prior to performing the shutdown .f, margin demonstration, as required by Technical ! , Specifications, the shorting links were removed to s put the Reactor Protection System in the i non-coincidence scram mode. , , . > j The point of criticality was demonstrated by i - - withdrawing control rods following the order given i ', in the rod pull sheets until an (approximate) 300 l second period was observed with Group 3 Rod 18-51 i withdrawn to notch Position 08. Iloderator ' I ~ temperature was recorded at %'F. Later, with moderator temperature still at 960 F, the reactor j '% ' was then made supercritical by withdrawing Control Rod 10-43 to Position 08. SRM A,B,C and D - l asasurements were taken every 30 seconds for three 1 and one half minutes. Period analysis was performed by fitting the data linearly on a semi-log plot and l ~ % s i l l ! ..--___..m___.,,__.,...____ w - Supplement 3 Paga 3 4-2 '. t measuring time to increase cne decade from which , j period was calculated. The average period was determined to be 76.5 seconds.- . The shutdown margin of the fully loaded core at s 680 F with the analytically strongest rod withdrawn ; i was determined te be 2.T2% delta k/k. Level 1 ,d 1:g9.' criteria were satisfied since the measured shutdown margin was larger than R + 0 385 s 0.88% delta k/k where R is defined here as the analytical difference in shutdown margin (cold) at the most limiting point in the cycle and Beginning .of Life - of the core.J 2 e difference in keft between the theoretical s si critical configuration and the actual measured critical configuration was found to be 0.28% delta l! , ~ k/k. This sat,isfies Level 2 criteria since , criticality occured within 15 delta k/k of the theoretical critical eigenvalue. .- P k i ' p . s 4 l Supplement 3 . Pag) 3 5-1 35 control mod Drive System - 3 5.1 Purpose , Each control rod drive (CRD) was tested to measure insert / withdraw and scram times and friction dP levels in the CRD hydraulic system. This was done to demonstrate that the CRD system operates properly over the full range of primary coolant temperatures and pressures. 3 5.2 criteria Level 1 Each CRD 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. The mean scram time of all the operable CRD's with v functioning accumulators must not exceed the following times (scram time is measured from the time the pilot scras valve solenoids are de-energized). Position Inserted From Fully Withdrawn Scram Time (Seconds) ' 46 0 358 36 1.096 26 1.860 6 3 419 The mean scraa time of the three fastest CRD's in a two-by-two array must not exceed the following times 4 (scram time is measured from the time the pilot , scras valve solenoids are de-energized). ' Position Inserted x, From Fully Withdrawn Scram Time (Seconds) 46 0 379 36 1.161 i 26 1 971 6 3 642 1 ) 4 ~~~ ~ ~ . . _ _ _ _. Z ~ Supplement 3 Page 3 5-2 Level 2 Each CRD aust have a normal withdrawal speed less than or equal to 3 6 inches per second indicated by a full ,12 foot stroke in 40 to 60 seconds. If the differential pressure variation exceeds 15 psid for a continuous drive-in, a settling test must be performed. In this case the differential settling pressure should not be less than 30 psid, nor should it vary by more than 10 paid over a full stroke. 353 Essults Insert / withdraw timing, friction testing, and scraa timing were performed on the CRDs at the conditions , specified in Figure 3 5-1. , All of the individual control rods were scram time l tested, friction tested and insert / withdraw timed during the Open Vessel test condition. Adjustments l to some CRDs had to be done in some cases to bring
- insert / withdraw times into acceptance limits.
During the friction testing, no pressure . differential measurements exceeded the criteria of 15 psid and no settling tests had to be performed. The four slowest rods in each sequence were also scrammed at reduced accumulator pressure. All test criteria were satisfied.
- During Heatup, the four slowest rods in each sequence were scraa timed at 600 psig and at 800 psig. Upon reaching rated temperature and pressure conditions, all CRDs were scram timed. The eight slowest rods determined during Open Vessel and Heatup testing were then insert / withdraw timed, friction tested, and scrammed at reduced accumulator pressure. Figure 3 5-2 shows the average scram time of the eight slowest rods, four in each sequence, at various reactor pressures compared to the maximum permissible.
J supplement 3 Paga 3 5-3 The specific results from our rated pressure testing are as follows: 1 Mean Scram Times l l Rod Position l 46 1 36 l 26 1 06 l l Nean Scram Time for all. l 0 302 1 0.852 l 1 398 l 2.501 l l 88 Sea. B rods (sec) l l l l l , 1 Nean Scras Time for all I 0.288 l 0.802 l 1 340 1 2.436 l l 97 Sea. A rods (sec) l l l l l l Mean Scram Time for ALL 10.295 1 0.826 l 1 368 l 2.467 l l rods, Seq. A and Seq. 8 (sec) l l l l l l (core average) l l l l l l Nean Scram Time of the 1 0 325 1 0 900 l 1.481 1 2.655 I I 3 fastest CRDs in a two-by-twol l l l l , I array for AU. rods, Seq. A andl l l l l l Seq. B (core average) l l I l l In conjunction with the planned scram for the Shutdown from Outside the Control Room test performed in Test Condition One, the scram times for the four (4) slowest Sequence "A" control rods were determined. All the scram times were within the acceptance criteria. l l l l l l l l i i.-__..._.._.._. ., .. _. ___._ _ _._ ___ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ , , , _ . _ _ _ _ . _ ~ Supplement 3 i Paga 3.5-4 FIGURE 3 5-1 (XMrrROL-MMkDRITE SYSTEM TESTS Reactor Pressure with Core Loaded Test Accumulator Preop pois Description Pressure Tests 0 600 500 rated Position All All Indication Normal Stroke Times All All 4(a) Insert / Withdraw Coupling All All , Friction All 4(a) Scran Normal All All 4(a) 4(a) All Scran Ninlaua 4(a) 4(a) Scran Zero 4(a) Scram (scrandischarg{c) Normal volume high level) Scram Normal 4(b)
- a. Refers to four CRDs selected for continuous monitoring based on i slow normal accumulator pressure scraa times, or unusual operating characteristics, at sero reactor presssure. The four selected CRDs must be compatible with rod worth minialmer, RSCS systems, and CRD sequence requirements.
- b. Scram times of the four slowest CRDs will be determined at Test
! Conditions 1 and 6 during planned reactor scrams.
- c. The scran discharge volume fill time will be determined at Test Conditions 1 and 6 during planned reactor scrams.
Note: Single CRD scrans should be performed with the charging valve closed (do not ride the charging pump head). l l e , - , - - - - - . - - - - ~ <-- ,, - - - - - - , - - - - - - ~ ~ , - - - - - - - - - - - - Supplement 3 Pasa 3 5-5 FIGURE 3 5-2 Scram Testing Ensults i: . i. r .J t.' > - a := ilt:.2 2: . .f-.. 44 geod e $.1slas Based Pqes.nsa.s. . 4: - ~ -"' ~ . , . a:at _. . j.. , . . . .. . .. , . . . . . ~ . .t .. e . .t . .f . . . . * 'i . I l 1_. s5 - .. .k . .3 *: z .: ? *. i. 'I ' .... I! 5 :
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=,.i..w.i-[e . ;r :.p ,- - - . . ,-j.. .. ') .- aj- ; 7 :--i r!:-- . J.) :, t ... -. . -- . .3. r - . s . I . - . ,.- e . ,, g ._g . 3 . l - l ,t ._7_ q. . :. . . , . . . .i. .. . . j - _. . . ;e . .. . .. _._ a . i . I + .__t . . . . . . . .i . . 2. 00. ,i . .M. }0. . . ... .,e. 40,0. . ..j, 800 ; .e. - 30.,% I deactor Freasure Tesia) !- I O - ._: . . ;_. l _2_ j .. . . . . - I l....- . . . . . . . . . j : ;. . . . .$ ._ a. . . ..;. i i _2. _: . ; - .: I 1 ; ! ! i , t I l l i Supplement 3 Pag 3 3 6-1 36 Source aanse Monitor Performance and Control Rod Sequence Exchange 3 6.1 rurpose The purpose of this test was to demonstrate that the operational sources, source range monitor (SRM) instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner. 1 The effect of typical rod movements on reactor power l was also determined. , . 3.6.2 Criteria Level 1 There must be a neutron signal count-to-noise count ~7 ratio of at least 2:1 on the required operable SRMs. There must be a minimum count rate as defined by Technical Specification on the required operable SRMs. Level 2 None 363 nasults 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 SRMs (A, B, C and D) were 20, 15, 40 and 15 cps respectively. These were well above the i ainlaus Level 1 criteria required of 0.7 cps. SRM readings were also taken periodically during
- initial criticality in both sequences and IRM readings were obtained during the initial heatup in sequence B. All test criteria were satisfied.
Performance data was gathered during power ascension to 20% in Control Rod Sequence A and Sequence B. At the end of each rod worth minlaizer group, APRM, feed flow, and steam flow values were recorded. i Supplement 3 Page 3 7-1 37 unter invel usasurement 3 7.1 rurpose The purpose of this test is to measure the reference leg temperature and recalibrate the instruments if the measured temperature is different from the value assumed during the initial calibration. 3 7.2 criteria Level 1 None Level 2 s The difference between the actual reference leg
- temperature (s) and the value(s) assumed during initial calibration shall be less than that amount '
that will result in a scale endpoint error of 1 percent of the instrument span for each range. 373 mesults Testing of the level instrumentation accuracy showed that scale end point errors when actual drywell temperatures and assumed calibration temperatures were compared were 0.7085, 0.5545, 1.05075 and 0 320% for wide range (Div. I), wide range (Div. II), narrow range (Div. I) and narrow range (Div. II),respectively. The slight Level 2 criteria violation for Div. I narrow range level instrumentation was found acceptable following an evaluation performed by General glectric. It was previously intended to repeat this test to obtain another set of data with all the drywell , coolers in operation. However, based on an evaluation performed by General Electric, the test results are acceptable and no further testing is required. Supplement 3 Pag 3 3 8-1 38 Im Performance 3 8.1 Furpose l The purpose of this test is to adjust the intermediate range monitor system to obtain an optimus_ overlap with the SM and APM systems. 3 8.2 criteria l Level 1 Each IM channel must be on scale before the SMs l exceed their rod block setpoint. Each APM must be on scale before the IRMs exceed their rod block setpoint. I Level 2 None s 383 Besults During the initial criticality, all IMs except IRM 1 D gwed response prior to the SM's reaching 5 x
- 10 cps. IRM D was repaired and tested
, satisfactorily at a later date. Range 64 overlap i calibration was also coupleted for each IRM, except IRN G which was reading erratically. This IRN was l replaced and retested successfully. ! IRMs G and H underwent repairs during the outage that required retesting of the ra' ige 64 overlap. After some adjustments, overlap was again sucessfully demonstrated for both. All APRMs were shown to be onscale prior to any IRM exceeding its rod block setpoint during a plant shutdown in Test Condition One. It was noted that IRM channels C, E, F and H were not reading one-half decade below their range 9 rod block setpoints. Although Technical Specification verification of overlap was satisfactorily performed in conjunction with Plant Surveillance procedures, the test will be reperformed after APRMs are adjusted in Test Condition 2. - - - ~ , , , ..wo.,.----,--,--ww.%--_ ,,,my_ ,.w c, _w. _ ,m---- Supplement 3 Pcge 3 9-1 39 LPRM Calibration 1 1 391 Purpose I The purpose of this test is to verify LPRM response to flux changes and proper LPRM connection to neutron monitoring electronics and to calibrate the l LPRM's to their calculated valves. i 3 9.2 Criteria , Level 1 ) None Level 2 Each LPRM reading will be within 10 percent of its calculated value. - 393 nasults The initial LPM verification test was performed while the Reactor was at rated pressure in the heatup test condition, in conjunction with scram
- time testing. Specific control rods were selected to be used for flux response checks based on their
- proximity to the LPRM strings. The withdrawal of these rods from Position 00 (FULL IN) to Position 48 (FULL OUT) was observed in teras of the LPRM flux j response as the rod was withdrawn past each of the four LPRMs for the associated LPRM string. All 172 4 LPRMs (43 LPRM strings with 4 LPRMs per string) were l observed, using Brush Recorders and STARTREC System for flux response. Initially, no fluz response was observed on 25 of the 172 LPRMs. Por the LPRMs that i showed flux response, the proper order of the LPRM response (D, C, B, A) was observed.
f During 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 be repaired. During Test Condition One all remaining LPRMs were observed to show proper flux response following repair efforts. An initial LPRM calibration utilising the Traversing In-Core Probe (TIP) System and the Backup Core l Limits Evaluation (BUCLE) program was conducted in
- Test Condition One. Utilizing TIP traces, local l LPRM readings, and heat balance information, a gain l
l l l l Supplement 3 Pags 3 9-2 adjustment factor (GAF) was determined for each LPRM. Rose GAFs were then used to adjust the gains of the LPFJe and a followup test was performed to verify criteria. Due to non-steady state conditions, a total of four full sets of TIP traces were made. Upon completion of the test, a total of 23 LPRMs did not meet the above criteria. The majority of the failures were reasonably close to the criteria, or were in the low power region of the core where criteria can be ignored. h is test will be reperformed at a higher power level during Test Condition Two. supplement 3 raga 3 10-1 3 10 AveraSe Power Range Monitor calibration 3 10.1 rurpose The purpose of this test is to calibrate the APM system. 3 10.2 criteria l Level 1 In the startup mode, all APM channels must produce I 1 a scram at less than or equal to 15 percent of rated
- thermal power.
The APM channels must be calibrated to read equal , to, or greater than the actual core thermal power. . Becalibration of the APM system is not necessary from a safety standpoint if at least two APM .- channels per RPS trip circuit have readings greater than or equal to core power. Technical Specification and fuel warranty limits on APM scram and rod block shall not be exceeded.
- Level 2 If the above criteria are satisfied, then the APM channels will be considered to be reading accurately i
if they agree with the heat balance to within (+7, i -0) percent of rated power. 3 10 3 assults During heatup, each APM channel was calibrated to read greater than or equal to a annual calculation l of Core Thermal Power based upon a constant heatup rate analysis. The APM scram trip setpoints were i also adjusted to produce a scram at less than 15% of rated power. The Level I criteria was satisfied. , 4 An initial APM calibration was performed during l Test Condition One at a Reactor Power of 13 3%. All = AP Ms were adjusted to read within (+3, -0)$ of calculated core thermal power, as determined by a manual heat balance calculation. A second APM calibration was performed later in Test Condition One when core thermal power (CTP) was determined to be 15 565 as determined from a manual heat balance - calculation. APM gain adjustments were then evaluated and the APMs adjusted to read 16.0% which is 4 .445 above CTP and satisfies the above Level 2 criteria. Supplement 3 Paga 3 11-1 3 11 Process computer 3 11.1 Pepose The purpose of this test is to verify the - performance of the process computer under plant operating conditions. 3 11.2 criteria Level 1 None Level 2 Programs OD-1, P1, and 00-6 are considered operational when the MCPR, the maximum LHGR, the . 3 maximus APLHGR, and the LPRM gain adjustment factors calculated by BUCLE and the process oceputer agree with the tolerances specified in the FSAR. Remaining propans will be considered operational on the successful completion of the static and dynamic testing. 3 11 3 mesults The TIP System consists of five identical probes used to sensure and record the axial neutron flux i profile at 43 radial core locations. The recorded information is used by the Process Computer to calibrate the fixed in-core Iscal Power Range Monitors. Each probe is driven into and withdrawn from the core by its associated drive mechanism. In order to operate automatically, the TIP drive control units must be programmed with the probe position at top and bottom of the core. 1hese top and bottom limits are programmed and verified in the ; TIP cold alignment. This portion of the test was performed successfully by hand-oranking the TIPS to the top of the core and setting the core limits based on the resulting position readings. In order to follow and read data from the TIP machines, the Process Computer must receive position information and flux signals from the TIP System. . ! This interface is tested in the Static System Test , Case by running the TIP machines in various l configurations and verifying the proper responses on 4 the Process Computer. I i 1 I Supplement 3 Pag 3 3 11-2 The Static System Test Case had two objectives: verification of the program logic and checkout of the TIP interface. The first objective was 1 successfully achieved, but the TIP interface checkout was unsuccessful due to a probles with the TIP System that resulted in the loss of TIP position indication. This original position indication 1 - probles"was repaired. As part of the Test Condition One testing, the TIP top and bottom core limits were reverified under hot conditions, and the TIP interface with the I-Y plotter was also verified to function properly. Following repairs to TIP 'C" ball valve, a process computer interface probles, and TIP '8" Logic, a successful 0D-1 was obtained from the process computer. It was noted that a three (3) second delay was occurring between X-Y plotter traces and s the machine normalized, full power adjusted TIP array. This probles was corrected prior to the OD-1 portion of the Dynamic System Test Case. The Dynamic System Test Case was performed during steady state conditions with reactor power at approximatley 20%. The testing includedi
- 1. Verification of the Computer Outage Recovery Monitor (CORM) to initialise necessary variables and esposure arrays as part of initial plant computer startup and to allow for controlled set of data in further system testing.
l
- 2. Verification that all required plant sensors for NSS programs are being properly scanned.
3 Verification of the heat balance subroutine used by OD-3 and F1 by coopering it with a manually 1 calculated heat balance. .
- 4. Performing an LPRM calibration to verify the operation of 00-1 prior to the verification of thermal limit calculations.
- 5. Verifloation of thermal limits calculations and core power distribution. ,
l
- 6. Verification of the exposure updating programs .
P4 (10 Minute Core Energy Increment), P1 (Periodic Core Evaluation), P2 (Daily Core Performance Summary) and P3 (Monthly Core j Performance Suasary). - - - - . , - . - --- 2.-__.-.-.-_- _-- _ _ _ _ supplement 3 Page 3 11-3
- 7. Verifying key variable memory locations and performing annual calculations to verify the remaining NSS software at steady state operation and syneetric rod pattern.
Thermal limit and LPRM calit, ration factor calculations were verified in conjunction with the D57C. The verification was performed by taking the same data that is input to the F1 program, for its calculation, and inputting it into an approved offline oosputer program (Backup Core Limits Evaluation (BUCLE), which also performs the F1 calculations. The resulting thermal limits and LPRN onlibration factors were verified against the criteria. In all instances the results were in the same fuel assembly and the results are as follows: Parameter Location F1 Results Sucle Results 5 trror ~ Max LNGR 33-52-13 3 78 3 78 05 Max MAPLHGR 27-10-13 3 30 3 30 05 Min CPR 27-10 3 877 3 876 .02% P1 Result - Bucle Result $ Error a # 100$ F1 Result The Local Power Range Monitor (LPRM) gain adjustaent factors calculated by BUCLE and the process computer were verified to agree within 25 Programs 00-1, P1, 00-6 and the remaining NS5 programs were considered operational upon the satisfactory performance of this procedure. l \
- I
l 3 12 aCIC system - 3 12.1 Furpose The purpose of this test is to verify the proper operation of the EIC systes over its espected operating pressure range. 3 12.2 estar 15 Level 1 The a Wrage pump discharge flow must be equal to or , greater than the 100-percent-rated value after 50 seconds have elapsed from initiation on all auto starts at any reactor pressure between 150 psig and rated. With pump discharge at any pressure between 250 peig and 100 poi above rated pressure, the , required flow is 600 spa. (The 100 psi is a conservatively high value for line losses. The measured value may be used if available). - The E IC turbine shall not trip or isolate during auto or manual starts. Level 2 To provide a margin on the overspeed tr'ip and isolation, the first and subsequent speed peaks on the transient start shall not esoeed the rated speed of the RCIC turbine by more than 5 percent. For small speed or flow changes in either manual or automatic mode, the decay ratio of each recorded RCIC systes variable must be less than 0.25. The turbine gland seal condenser systes shall be capable of preventing stema leakage to the atmosphere. The delta P switch for the RCIC steam supply line high-flow isolation trip shall be adjusted to actuate at 300 percent of the maximum required , steady state flow, with the Reactor assumed to be near the pressure for main relief valve actuation. i Supplement 3 Pag) 3 12-2 3 12 3 mesults During the Heatup Test Condition, the RCIC pump suction and discharge was lined-up in a closed loop with the condensate storage tank. The system was subjected to negative and positive 10% step changes in flow.at system flows of 600 spa and 270 spa using 1 both a step generator and the RCIC flow controller. ' Minimum flow data was also taken at a speed of 2000 ; rps and a RCIC quickstart was performed. The RCIC system was able to supply 600 spa at a discharge pressure of 1140 psig in 35 seconds when automatically started using 940 psig steam from the vessel. The g72 time delay relay was set down from 10 see to 5 see to prevsnt the RCIC turbine from coasting down excessively before the opening of the Steam Admission Valve, thus reducing the experienced . transient. The RCIC turbine did not isolate or trip during the auto and maunal starts. In addition, there were no RCIC turbine speed peaks or oscillations in RCIC system variables in the transient testing. The RCIC system was also subjected to an extended run at rated flow conditions. RCIC performed satisfactorily with all system temperatures stabilized below alara levels and a negative pressure maintained on the gland seal condenser system. All Level 1 and Level 2 criteria were satisfied except the RCIC steam supply high flow isolation trip setting. During the Outage for the replacement of the Main Steam Bypass Lines, engineering modifications to the instrument lines were completed that were espected to solve the problems found with the instrument sensing lines. Upon recommencing Heatup in August of 1986, the RCIC 3GH module was found malfunctioning and was replaced. Because of this and the instrument line modifications discussed above, the RCIC system was subjected to further testing including 105 positive and negative step changes in both speed and flow, and a quickstart. With the reactor pressure at 955 psig, the RCIC system was able to supply 600 sps at a discharge pressure of 1143 pais in 33 seconds. All Level 1 and Level 2 criteria were satisfied except the i turbine gland seal system verification and the RCIC steam supply high flow isolation trip setting. i . _ . . _ . _ , . . , . - _ _ _ . - , - . - , . _ . ~ . _ _ . _ _ _ _ . _ _ . _ _ . - - . _ . _ . . . _ _ . _ . _ . . _ _ _ _ . _ , . . _ _ _ _ _ . . _ _ _ , , _ _ _ _ _ _ . l f Supplement 3 l Pag 3 3.12-3 i Due to a failure of the RCIC Barometric Condenser Vacuum Pump, data did not show the existance of a vacuum on the vacuum tank as required by the test criteria. Subsequent work on the Barometric Condenser Pump corrected the problems and it was retested successfully. Data was also taken during this test to determine
- the actual 3005 value for the RCIC steam supply line high flow isolation trip setpoint. However, the trip setpoints were not adjusted to these settings, )
- but are being left at the current trip setpoints .
given in the Technical Specifications. The current i settings as specified by the Technical Specification are set conservatively compared to the value calculated by the performance of this testing, yet provide ample margin to prevent spurious RCIC isolations on system automatic initiations. . l During Test Condition One, RCIC system testing consisted of a hot manual vessel injection, two (2) cold quick start vessel injections, a 150 psig CST to CST run, a 150 psig vessel injection, and a CST i to CST run at rated pressure for baseline data. The only probles of any significance during any of these runs was a turbine speed peak 29 rps above the Level ; 2 limit of 4725 rpm, which occurred during the
- initial hot manual vessel injection. Minor adjustments were made to the RCIC control circuitry and the probles did not reoccur in subsequent tests.
i' For the hot manual vessel injection, with the reactor supplying steam at a pressure of 915 psis, the RCIC pump delivered a flowrate of 3 600 spa at a discharge pressure of 965 psig in 28.4 seconds. As l discussed above, the turbine reached a maximum speed peak of 4764 rps, which exceeded the Level 2 criteria. Based on data taken in conjunction with this test, it was determined that the actual line ; loss value for the RCIC system was 50 paid. i For the first cold 'essel injection, with the i reactor supplying steam at a pressure of 918 psig, l the RCIC pump delivered a flowrate of 3 600 sps at a 4 discharge pressure of 970 pois in 26.5 seconds. The . annimum speed peak was 4686 rps for the RCIC turbine. , 4 i i i - .. . . . ~. - _ .-. . - . __ Supplement 3 Pag 3 3 12-4 For the second cold vessel injection, with the reactor supplying steam at a pressure of 910 psig, the RCIC pump delivered a flowrate of 3600 spa at a discharge pressure of 970 pais in 29 2 seconds, with a maximum speed peak of 4488 rps. During the 150 psig CST to CST run, with the reactor supplying steam at a pressure at 165 psig, the RCIC pump delivered a flowrate of > 600 gym at a discharge pressure of 271 psig in 22.0 seconds, with a maxisus speed peak of 2818. During the rated reactor pressure CST to CST run, with the reactor supplying steam at a pressure of ! 920 psig, the RCIC pump delivered a flowrate of 3 600 spa at a discharge pressure of 1095 psig in 29 seconds, with no discernable speed peak as the , turbine ramped up smoothly to a final speed of 4500 rpm. The 150 psig vessel injection was conducted with the ! reactor supplying steam at 160 psig. The systes reached 3 600 spa in an elapsed time of 21.5 seconds at a discharge pressure of 215 psig, with a marinua speed peak of 2641 rpm. RCIC testing was successfully completed with a 150 psig cold CST to CST baseline data test. With the reactor supplying steam at a pressure of.165 psig,
- the RCIC pump delivered a flowrate of 3600 spa at a discharge pressure of 360 psig in 19.5 seconds, with an initial speed peak of 1418 rps followed by a i smoo6h ramp to a final maximum speed of 2766 rpm.
4 i i i 4 l Supplement 3 Pag) 3 13-1 3 13 e cI system , 3 13 1 Purpose The purpose of this test is to verify proper l operation of the High Pressure Coolant Injection (HPCI) system over its expected operating pressure range. 3 13 2 criteria i Level 1 The average pump discharge flow aust be equal to or greater than the 100-percent-rated value after 25 seconds have elapsed from initiation on all auto starts at any reactor pressure between 150 psig and s ! rated. With pump discharge at any pressure between . 250 psis and 100 psi above rated pressure, the flow should be at laast 5000 sp. (The 100 psi is a - conservatively high value for line losses. The asasured value may be used if available). ThaiHPCI turbine shall not trip or isolate during auto or manual starts. Level 2 i The turbine gland seal condenser system shall be capable of preventing stosa leakage to the atmosphere. , The delta P switch for the HPCI steam supply line high flow isolation trip shall be adjusted to actuate at 300 percent of the maximum required steady-state flow with the reactor assumed to be near main relief valve actuation pressure. 1 For small speed or flow changes in either manual or ,
- automatic mode, the decay ratio of each recorded l HPCI system variable must be less than 0.25.
To provide a margin on the overspeed trip and j isolation, the transient start first speed peak 1 shall not come closer to the overspeed trip than 15 percent of rated speed, and subsequent speed peaks shall not be greater than 5 percent above the rated ! turbine speed. - i l 1 l ~_ - - _ - - _ _ - - - _ - . ___ Supplement 3 Pago 3 13-2 3 13 3 assults Following setup of the control system, initial coupled turbine performance runs were performed on the HPCI systen during initial heatup. Dynamic stability checks were conducted with the HPCI pump suction and discharge lined-up in a closed loop with the CST in which 500 spa flow step changes were manually and automatically introduced by the flow [ controller with HPCI system flows at 5000 sps and 2700 spa. . During the automatic initiation testing of HPCI, a i discharge flow of 5000 sps was reached in 23 4 seconds. Twenty-five seconds after the automatic initiation HPCI flow had reached 5310 spa at a discharge pressure of 1140 peig, 190 psig greater than reactor pressure. HPCI did not trip or isolate , during any manual or automatic starts. There was
- also adequate margin on turbine speed peaks and oscillations of system variables. An extended run was also performed in which systes temperatures '
stabilized at acceptable levels and the gland seal systen performed satisfactorily. All Level 1 and Level 2 criteria are satisfied except the steam supply isolation trip setpoint. During the extended Outage which started in the Fall of 1985, engineering modifications were completed i that were espected to correct the problems experienced with the instrument sensing lines. Because of this modification, the gGR bypass line Installation, and other modifications that were made j to the HPCI Systes during the Outage, the Startup i - Tests were repeated for this system when the plant ! restarted in August of 1986. ' l Dynamic Stability checks were again completed using 500 spa step changes introduced in both manual and l automatic flow control modes with the NPCI System operating in a closed loop to the CST. Level 2 criteria was exceeded when HPCI System flow had a ' measured decay ratio of 0.28 when a mid-flow speed decrease step change was inserted in the manual mode. This is currently considered to be acceptable but will be esamined closely in HPCI testing at higher test conditions. I i l ! l l Supplement 3 Paga 3 13-3 During a NPCI automatic initiation in the CST closed loop lineup, a HPCI System flow of 5000 sps was achieved in 21.2 seconds. Twenty-five seconds after the automatic initiation occurred, RPCI flow was 5003 sps at 1185 psig pump discharge pressure, 225 psig greater than the 960 pois reactor pressure. Data was also taken during this test to determine the actual 3005 value for the NPCI steam supply line high flow isolation trip setpoint. However, the trip setpoints were not adjusted to these settings, but are being left at the current trip setpoints given in Technical Specifications. The current isolation settings as specified in Technical Specifications are considered acceptable as they are conservative yet provide ample margin to prevent spurious HPCI isolations on system automatic initiations. , All other 14 vel 1 and 2 criteria were set. During the 9/86 retesting of NPCI, sluggish response was noted in the HPCI control valve. As a result, it was decided to replace the gGR component in the hydraulic portion of the NPCI control system. As a result, the 1000 pois hot CST injection was repeated to verify proper control system operation. On the quick start HPCI discharge flow reached the 100-percent-rated value (5000 spa) in 21.0 seconds. Following the automatic initiation, MFCI flow leveled out at 5100 sps with a discharge pressure of '1190 pois. The initial speed peak was 2134 rps and the maximum peak was 4114 rpm. All other Level 1 and Level 2 criteria were met. l supplement 3 Fage 3 14-1 3 14 selected process Temperatures 3 14.1 Furpose The purposes of this procedure are to establish the proper setting of the low speed limiter for the recirculation pumps to avoid coolant temperature stratification in the reactor pressure vessel bottoe head region, to provide assurance that the esasured bottom head drain temperature corresponds to bottom head coolant temperature during normal operations, and to identify any reactor operating modes that . cause temperature stratification. 3 14.2 criteria Level 1 The reactor recirculation pumps shall not be ,,, restarted nor flow increased unless the coolant temperatures between the steam does and bottom head drain are within 145'F. The recirculation pump in an idle loop must not be started, active loop flow must not be raised, and power must not be increased unless the idle loop suction temperature is within 50 7 of the active loop suction temperature. If two pumps are idig, the loop suction temperature must be within 504 of the steam done temperature before pump startup. Level 2 During operation of two recirculation pumps at rated oore flow, the bottom head temperature as sensured bytheboltosdrainlinethermocoupleshouldbe within 30'T of the recirculation loop temperatures. 3 14 3 annults For the initial testing conducted in 1985, the coolant temperatures sensured at 305 Recirculation pump speed satisfied the Level 1 criteria. The instability of the rectro speed controller that occurred during this test precluded an effective investigation of the stratification phenomenon at low flows. The test also allowed setting of the low speed limiter based on flow controller variations oft 25 of rated speed. Flow controller variations of 2 55 were esperienced prior to stratifloation so the test was terminated. Supplement 3 Page 3 14-2 The minimum recirculation pump speed data collection was resumed in August, 1986 following completion of the preceding Outage. In subsequent heatup testing, the Rectro NG 5ets were hand cranked down to speeds of about 20%. The Level 1 criteria was satisfied at all times during this test. The low speed limiter setting was chosen to be 285 speed based on the previously observed controller instability below that level. The remaining testing in this section will be completed at hitner test conditions, including those tests intended to verify the Level 2 criteria at rated core flow. O e l l l r Supplement 3 Pag) 3 15-1 3 15 system espansion 3 15.1 rurpose The purpose of this test is to verify that selected - plant piping systems are free and unrestrained with regard to thermal espansion, and to verify that the thermal movement of the piping and associated support system components is consistent with the analytical prediction of the piping system stress analysis. 3 15 2 criteria Level 1 The measured displacements at the instrumented locations shall be within the greater of the , specified allowable tolerance of the calculated - values, or 1 0.25 inches for the specific points. There shall be no obstruction which will interfere ! with the espected thermal expansion of the piping system. glectrical cables shall be able to accommodate espected thermal espansion of the piping systes. Instrumentation and branch piping can accommodate espected thermal expansion of the piping system. The constant hanger shall not be bottooed or topped out. The spring hanger shall not be bottomed or topped out. The snubber shall not be bottooed or topped out. Level 2 The measured displacements at the instrumented locations should be within the greater of the specified espected tolerance of the calculated values, or 1 0.25 inches for the specific points. The installed cold position of the constant hanger must be within 1 5% of the design cold load. The installed cold position of the spring hanger must be within 1 55 of the design cold load. SuppSemen3 3 Pag 3 3 15-2 The snubber may deviate from its design cold l position setting + 1/2", providing the position is l not less than 1/23 from bottoains out. 3 15 3 assults Piping Inspection nasults Selected piping systems were walked down at various plant conditions to identify possible restraints to projected thermal expansion. These walkdowns occured at ambient temperature, 2500F and rated temperature. Manger and snubber settings were i recorded and thermal expansion (PVDET) sensors were I verified to be intact. No restraints to projected thermal expansion were identified. One-hundred and forty-three (143) ' supports were identified as being out of tolerance or topped or bottooed out. Following re-veriff?etion and engineering evaluation, sixteen (16) supports were adjusted or modified and the remainder accepted as is. The East and West Main Steam Bypass Lines were replaced during the Outage which started in the Fall of 1985, because of cracks which were discovered in these lines. During subsequent testing following reactor restart in August, 1986 these lines were visually inspected to verify that they were unrestrained with regards to projected thermal expansion. These walkdowns occured at ambient tes rature; and at recirc loop temperatures of 350 and rated. No restraints to bypass line thermal expansion were identified. Five supports were found out of tolerance, and upon engineering evaluation were accepted as-is. Third thermal cycle visual inspections and hanger readings were made on all system piping including the replaced Main Steam Bypass Lines. There were no restraints to thermal expansion identified. Two-hundred-ninety-five (295) supports were identified as not being within their proper working range. Following engineering evaluation and reverification, eight (8) supports were reset and the remaining supports accepted "as-is". - Supplemeft Q , Pas] 3 15 ') '< ~ 1 System Emmansion Result 3-Selected points on tae piping systens were w!?ed with remote sensors to beM toe the'thereally l.nduced , piping museents during stystem operation. fte ' oint d ^ monitoret movementa'op'r er.a were espected perience to undergo large thermal stressas.large I After establisi:ing initist readings ter the sensors - at ambient conditions, the sensors were monitored > during the ini ial heatup of the plant. Onta uns recorded at 50 ' intervals until the remeter reached operating temperature. The evalintions 1, f ound several criteria esoeedanoes, but upon engineering evaluation of the earmdances, r.11 were four.d acceptable. - In addition, initial ambient senaar readisce taken . j before Heatup were compared to ambient memor - readings after a Neatup and cooldown oy de w u ocapleted, llo appreciable difference in tho beforo and after' readings were noted, indicating piping i movement was not restrained. Thermal gapansion data was again taken at 500F j intervals at moderator temp tatures b? sinning at 100*F during the subsequent! estup cycle Iellowing ~ l fnitial heatup. The'deta na evaluated at each j temperattree plateau before groceeding to the* nest level. Upon reaching rated temperature, four Level 2 criteria vioistions existed, but these were very , l minor and accepted as-is.
- the gast and West Main Steam Pypass Lines that were i -
replaced in the fall of 1985.were also monitored.fpr espected therani espansion during the subsequent heatup after the Outage. The westep and cooldown ! sensor readings satisfied all 14 vel 1 and Level 2 criteria except at Lt.e 350"F recirc loop i temperature plateau. At that imint there was one imysl 2 failure which resulted from inadequate tw: sting of the bypass piping dce te* tm bypass
- valves being closed at the tiet the test was performed. At higher temperatures data was taken
- with the bypass va2ves open, and all criteria were satisfied. 4 e
1 I ,- a.- -- , j Supplement 3 o Pago 3 16-1 s , 3 16 care Power Distribution ? NDTE: As discussed in sonorandus VP-86-0141, "Startup Test Program Changes", dated October 17, 1986, from Frank c E. Agosti to James G. Esppler, it is our intention to delete this test. ,N$ j .c g I I f g ) \ ,o, \ d i o + e 4r $ k ? J l , I. l r-- -- -, e ,,,,,-m -~,,~.--r----- Supplement 3 Pas) 3 17-1 3 17 Core Performance 3 17 1 Purpose
- a. To evaluate the core thermal power.
- b. To evaluate the following core performance parameters:
- 1. Maximum linear heat generation rate (MLHGR)
- 2. Minimum critical power ratio (MCER) 3 Maximum everase planar linear heat generation rate (MAPLHGR).
3 17.2 criteria Level 1 The maximum linear heat generation rate (MLHGR) during steady-state conditions shall not exceed the t . allowable heat flux as specified in the Technical Specifications. The steady-state minimum critical power ratio (MCPR) shall be maintained greater than, or equal to, the value specified in the Technical Specifications. ;' The maximum average planar linear heat generation s rate (MAPLHGR) shall not exceed the limits given in the plant Technical Specifications. , Steady-state reactor power shall be limited to full l rated maximus values on or below the design flow control line. l l Core flow should not exceed its rated value. l Level 2 None 3 17 3 Results BUCLE computer analysis of whole core TIP-traces i obtained at 15.6% reactor power showed that all criteria were met. 1 Supplement 3 Paga 3,13 9 3 18 steam % gg , This test was previously deleted from the FSAR (Section l 14.1.4.8.18). l l s W., o Supplement 3 , Paga 3 19-1 ) 3 19 Core Power-void seode nasponse EDTE: As discussed in memorandus VP-86-01361, "Startup Test Program Changes", dated October 17, 1986, from Frank E. Agosti to James G. Keppler, it is our intention to delete this test. . 1 e i I l l o Supples:nt 3 Paga 3 20-1 3 20 pressure mesulater 3 20.1 purpose The purpose of this test is to:
- a. Doherminetheoptinuasettingsforthepressure control loop by analysis of the transients induced in the reactor pressure control system by means of the pressure regulators.
. b. To demonstrate the takeover capability of the backup pressure regulator on failure of the controlling pressure regulator and to set spacing between the setpoints at an appropriate value, s
- c. To demonstrate smooth pressure control . .
transition between the control valves and bypass values when the reactor generates more steam than is used by the turbine. 3 20.2 criteria . Level 1
- The decay ratio aust be less than 1.0 for each' process variable that exhibits oscillatory response to pressure regulator changes.
Level 2 In all tests the decay ratio aust be less than or equal to 0.25 for each process variable that exhibits oscillatory response to pressure regulator changes when the plant is operating above the lower limit setting of the master flow controller. Pressure control deadband, delay, etc., shall be small enough for steady-state limit cycles, if any, to produce turbine steam flow variations no larger than 0 5 percent of rated flow. During the simulated failure of the controlling
- pressure regulator along the 100 percent rod line, the backup regulator shall control the transient so that the peak neutron flux or peak vessel pressure remainsbglowthescraasettingsby75percentand 10 lb/in. , respectively.
l C ~- ~ Supplement 3 Paga 3 20-2 After a pressure setpoint adjustment, the time between the setpoint change and the occurrence of the pressure peak shall be 10 seconds or less. (This applies to pressure setpoint changes made with the recirculation system in the master or local manual control mode.) 3 20 3 nasults' Proper pressure regulator operation was demonstrated in Test Condition One by analysis of system response to step increases and decreases in pressure demand , with the bypass valves open and generator not on the ' line. Additional steady-state measurements were taken with the generator loaded and bypass valves closed. All Level 1 and Level 2 criteria were met. The pressure setpoint changes on each regulator, , while significant in magnitude (11-13 pais), were
- stable and well damped. As such no systen tuning was performed in this test condition; initial tuning of the Pressure Regulators will be done in Test Condition 2.
The Regulator failure tests yielded significantly different responses (14 psig change for failure of #1; 6 pais change for failure of #2). This discrepancy in response is likely attributable to differences in the time delay circuitry for each channel in the High Value Gate and difference of 1.7 psig in the sensed pressure being fed to each regulator channel. The discrepancy in regulator failure response will be further investigated during Test Condition 2 testing. The time delay component - in the regulator high value gates has since been removed; the effect of this will be evaluated during Test Condition Two tuneup testing which is in progress but not yet completed as of this report date. .-- ,..-,-..-..,_,..,c,.,. .Q_..._..., - SuppleaInt 3 Paga 3 21-1 3 21 Feedwater system 3 21.1 Purpose
- a. To adjust the feedwater control system for acceptable reactor water level control.
~
- b. To demonstrate stable reactor response to subcooling changes.
- c. To demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedwater Pump,
- d. To demonstrate adequate response to feedwater heating loss.
- e. To determine the maximum feedwater runout ,
capability. 3 21.2 criteria Level 1 The response of any level-related variable to any test input change, or disturbance, must not diverge during the setpoint changes. For the feedwater temperature loss test, the maximum feedwater temperature decrease due to a single 0 failure case must be less than or equal to 100 F. The resultant MCPR aust be greater than the fuel thermal safety limit. For the feedwater temperature loss test, the increase in simulated heat flux cannot exceed the predicted Level 2 value by more than 2 percent. The predicted value vill be based on the actual test values of feedwater temperature change and power level. The feedwater flow runout capability must not exceed the assumed value in the FSAR. Level 2 Level control systen-related vaiables 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 0.25, as a result of the setpoint change testing. Supplement 3 Pagi 3 21-2 A scram must not occur from low water level following a trip of one of the operating feedwater pumps. There should be a greater than 3-in. water-level margin to scram for the feedwater pump trip. For the'feedwater temperature loss test, the increase in simulated heat flux cannot exc9ed the predicted value referenced to the actual feedwater temperature change and power level, which will be taken from the Transient Safety Analysis Design Report. The average rate of response of the feedwater actuator to large (>20 percent of pump flow) step disturbances shall be between 10 to 25 percent of pump rated feedwater flow /sec. This average response rate will be assessed by determining the , time required to pass linearly through the 10 _ percent and 90 percent response points of the flow transient. The dynamic flow response of each feedwater actuator (turbine or valve) to small (<10 percent) step disturbances shall be the following:
- a. Maximum time to 10 percent of a step disturbance
$1.1 sec.
- b. Maximum time from 10 to 90 percent of a step disturbance $1 9 sec.
- c. Peak overshoot (percentage of step disturbance) 115 percent.
3 21 3 mesults During the initial heatup, the feedwater system performed satisfactorily in both the manual and automatic modes. All level-related variables did not diverge during testing and all system related variables did not exceed a 0.25 decay ratio for their oscillatory responses in the level setpoint changes. All applicable test criteria were satisfied. During Test Condition One, as previously done during the heatup testing, the Startup Level Controller setpoint was adjusted to simulate step changes of three inches for Reactor water level. During the setpoint increase water level increased in a smooth ; Supplement 3 1 Pagi 3 21-3 l l manner with little overshoot and stabilized within , 75 seconds. During the setpoint decrease water j' level decreased and overshot the three inch down step by 2 to 3 additional inches. This overshoot , dampened rapidly and water level stabilized within ' 110 seconds. The Tesf, Condition One test was completed satisfactorily. The criteria that the decay ratio of level control system-related variables being less than .25 was met for all portions of this test. The remaining tests involving the Feedwater system will be completed in future test conditions. Test Condition Two testing is in progress but not yet completed as of this report date. l l Supplement 3 Paga 3 22-1 3 22 Turbine valve surveillance 3 22.1 Furpose To demonstrate acceptable procedures and maximum power levels for surveillance testing of the main turbine. control and stop valves without producing a reactor scras. 3 22.2 criteria Level 1 None Level 2 Peak neutron flux aust be at least 7 5 percent below the scram trip setting. remain at least 10 lb/in. y below vesselthe pressure must high-pressure scraa setting. Peak heat flux must remain at least 5.0 percent below its scram trip point. Peak steam flow in the high-flow lines must remain 10 percent below the high-flow isolation trip settings. 3 22 3 Results The Turbine Valve surveillance test has not been completed to date. ~ Supplement 3 Paga 3 23-1 3 23 main stama Isolation valves 3 23 1 rurpose
- a. To check functionally the main stema line isolation valves (MSIVs) for pr v r operation at selected power levels.
- b. To determine reactor transient behavior during and after simultaneous full closure of all MSIVs.
- c. To determine isolation valve closure time. j 3 23 2 criteria Level 1 The MSIV stroke time (ts) shall be no faster than
- 3 0 seconds (average of the fastest valves in each steaaline) and for any individual valve 2 5 seconds
<ts <5 seconds. Total effective closure time for ~ any Individual MSIV shall be tsol plus the maximum instrumentation delay time and shall be <5 5 seconds. The positive change in vessel done pressure
- occurring within 30 seconds after the simultaneous full closure of all MSIVs must not exceed the Level 2 criteria by more than 25 psi. The positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2 percent of rated value.
Flooding of the main steam lines shall not occur following the full MSIV closure test. The reactor must scran during the full simultaneous MSIV closure test to limit the severity of the neutron flux and simulated fuel surface heat flux ' transient. Level 2 During full closure of individual valves, peak vessel pressure must be at least 10 psi below scran, peak neutron flux must be at least 7 5 percent below scras, and steam flow in individual lines must be at least 10 percent below isolation trip setting. The peak heat flux laust be at least 5 percent less than its trip point. The reactor shall not scram or isolate as a result of individual valve testing. I --e,_ _ _ _ _ _ ._ supplement 3 Pega 3 23-2 The relief valves must reclose properly (without leakage) following the pressure transient resulting from the simultaneous MSIV full closure. The positive change in vessel done pressure and simulated heat flux occurring within the first 30 seconds after the closure of all MSIV valves must not exceed the predicted values in the Transient Safety Analysis Design Report. Predicted values will be referenced to actual test conditions of initial power level and done pressure and will use - beginning of life nuclear data. The predicted values will be corrected for the app ~;priate sensured parameters. These values will be generated before fuel loading. After the full MSIV closure, the initial action of the RCIC and HPCI shall be automatic if L2 is reached, with RCIC capable of establishing an - average pump discharge flow equal to or greater than 600 sps within the first 50 seconds after automatic initiation and HPCI capable of establishing an average pump discharge flow equal to or greater than 5000 sps within the first 25 seconds after automatic initiation. If the low-low set pressure relief logic functions after the simultaneous full MSIV closure test, the open/close actions of the SRVs shall occur within +20 psi of the low-low set design setpoints. The Eotal number of opening cycles, for the safety / relief valves opening on low-low setpoint, after initial blowdown is not to exceed four times during the initial 5 minutes following isolation. - If any safety relief valves open as a result of this test, only one valve may reopen after the first blowdown. Recirculation pump trip shall be initiated if L2 is reached after the MSIV full closure test. 3 23 3 assults i l During the Heatup Test Condition, with the RPV at rated temperature and pressure conditions, each of the inboard and outboard isolation valves were successfully closed slowly to the approximately 90% open position and then fully reopened, without any noticeable change in reactor pressure, APRM readings or reactor water level. - ' ' ~ ~ ~ ~ Pass 3 23-3 In To t Condition One, eith the Reactor et 75 power, a fast full closure of each individual MSIV was performed. All applicable Level 1 and Level 2 criteria were met. The closure times are shown in the table below, using a calculated earlaus instrument delay time of 0.299 seconds. f ~ ~ Test Condition One' I I I I I l MSIV l ts I tsol l Total l I I I I l l F022A l 4.298 l 4.611 1 4.910 l l F022B l 3.505 l 3.703 1 4.002 l l F022C l 4.798 l 4.904 1 5.203 l l F022D l 3.205 l 3.301 1 3.600 l l F028A l 4.294 1 4.387 I 4.686 l .- l F028B l 3.809 1 3.839 l 4.138 l l F028C I 3.617 1 3.899 l 4.198 l l F028D l 4.057 l 4.226 1 4.525 i e All recorded times are measured in seconds. The remaining Level 1 and Level 2 criteria are associated with the MSIV simultaneous full closure and will not be verified until that test is performed during a higher test condition. l l l l Supplement 3 Paga 3 24-1 3 24 nelief Valves 3 24.1 Purpose The purposes of this test are to verify that the Safety Relief Valves (SRV) function properly (can be opened and closed manually), reset properly after operation, and that there are no major blockages in the relief valve discharge piping. 3 24.2 criteria Level 1 There should be a positive indication of steam discharge during the manual actuation of each valve. Level 2 Variables related to the pressure control systen 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 0.25 The temperature measured by thermocouples on the discharge side of the valves shall return to within 100 F of the temperature recorded before the valve was opened. If pressure sensors are available, they shall return to their initial state upon valve closure. During the 250 psig functional test, the steam flow through each relief valve as measured by the initial and final bypass valve (BPV) position shall not l differ by more than 10 percent from the average relief valve steam flow as measured by bypass valve position. During the rated pressure test, the steam flow through each relief valve as measured by change in M (e) is not to differ by more than 0.5 percent of rated M (e) from the average of all the valve responses. , 3 24 3 assults During the heatup testing, all 15 SRVs were manually actuated. There was positive indication of steam discharge upon actuation of each SRV. As each SRV was operated there was a sudden temperature rise on the SRV discharge tailpipe, the appropriate pressure l I --_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . - _ _ _ _ . . . _ _ _ , _ _ _ _ Supplement 3 Paga 3 24-2 switch responded, and BPV position decreased to control reactor pressure. The Level I criteria was satisfied. All pertinent variables related to pressure control did not exhibit any oscillatory responses with decay ratios greater than 0.25. The 3RV discharge line temperatures for five 3RVs did not return to within 10 0F of the temperature recorded prior to actuation as quickly as the other discharge lines; however, they did cool down sufficiently to indicate that the 3RVs were not leaking. Shortly after the performance of this test a reactor scram occurred and on the subsequent startup, the SRV tailpipe temperatures remained low, further verifying that the SRVs did properly reclose. Three 3RVs had steam flow values, as measured by BPV position change, that differed from the average relief valve steam flow by greater than 10$. The bypass valve position was inadequate to get a proper value of steam flow from BPV position c.hange. Upon the actuation of each SRV the BPV closed completely. Bad there been more bypass steam flow, the BPV would not have closed completely and there would be a more accurate value of SRV steam flow. This steam flow variance will be :sevaluated during the Test Condition 2 3RV tasting. The remaining Level 2 criteria involving W(e) variation resulting from individual relief valve opening shall also be verified when the relief valve j testing is performed at Test Condition 2. l i l' I l . - . _ . - . _ - - - . . - . . _ . - . - - _ . . - - _ _ .- =- _ _ _ ._ _ .. Supplement 3 Pag 2 3 25-1 3 25 Turbine stop valve and control valve Fast closure Trips l l 3 25.1 Purpose ! The purpose of this test is to demonstrate the , response of the reactor and its control systems to l protective trips in the turbine and generator. 3 25.2 criteria Level 1 For turbine / generator trips, there should be a delay of no more than 0.1 seconds following the beginning of control or stop valve closure before the beginning of bypass valve opening. The bypass valves should be opened to a point corresponding to greater than or equal to 80 percent of their capacity within 0.3 seconds from the beginning of control or stop valve closure action. ~ , Flooding of the main steam lines shall not occur 4 following the turbine / generator trips. , The positive change in vessel done pressure occurring within 30 seconds after either generator or turbine trip must not exceed the Level 2 criteria by more than 25 psi. The positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2 percent of rated value. Level 2 There shall be no MSIV closure in the first 3 minutes of the transient, and operator action shall l not be required in that period to avoid the MSIV I trip. The positive change in vessel done pressure and in simulated heat flux that occur within the first 30 seconds after the initiation of either generator or l turbine trip must not exceed the predicted values in the Transient Safety Analysis Design Report. For the turbine / generator trip within the bypass valves capacity, the reactor shall not scram for initial thermal power values less than or equal to. 25 percent of rated. I e L " - - - ~ ~ Supplement 3 Paga 3 25-2 If the low-low set pressure relief logic functions, the open/close actions of the SRVs shall occur within + 20 psi of their design setpoints. If any safety relief valves open, only one valve may reopen after the first blowdown. 3 25 3 assults Neither of the tests involved with this section have been completed to date. i e -------,,---.,-----_-w r Supples:nt 3 . Paga 3 26-1 3 26 Shutdown from Outside the Control Room 3 26.1 Purpose To demonstrate that the reactor can'be brought from a normal, initial, steady-state power level to the hot shutdown condition and to verify that the plant has the potential for being safely cooled from hot ~ shutdown to cold shutdown conditions from outside the control room. 3 26.2 criteria Level 1 None Level 2 During the cold shutdown demonstration, the reactor aust be brought to the point where cooldown is initiated and under control. During the slaulated control room evacuation and hot shutdown demonstration, the reactor vessel pressure and water level are controlled using equipment and j controls outside the control room. 3 26 3 Results l During the slaulated control room evacuation and hot-shutdown test performed during Test Condition One, the designated Shutdown Crew, consisting of the minimum shift complement, performed Q activities . associated with the reactor shutdown and control of the reactor vessel water level and pressure from outside the Control Room. The reactor vessel pressure and water level were controlled for a period of over thirty minutes following successful reactor shutdown and isolation from outside the Control Roca by the minimum shift complement, which successfully seeta all tast criteria and performance objectives of the applicable governing documents. 1 l -. .. .= . ._. - . . _ . . . _ . .~ s _ Supplement 3 Page 3 26-2 The test sequence of events was as follows: Time Event l l 1223 Test Start Time (Hi Come Announcement) j l 1224 -" Shutdown Crew" Evacuation of Control Roon l l 1224 APRMs A&B to Standby (to initiate Reactor Scram) 1224 Relay TTR-2 manually tripped (to initiate Main Turbine Trip) 1225 Main Steam Line Radiation Monitors to Standby (to initiate MSIV Isolation) 1226 Restoration of APRMs A&B and the Main Steam Line Radiation Monitors to the Operate positions
- 1226 Exit Relay Room 1228 Transfer Switches operated at Remote Shutdown Panel (RSP) (RSP Control) 1230 RHRSW started at Remote Shutdown Panel (RHR Service Water Pumps A&C)
I 1233 RHR Pump A started at Remote Shutdown Panel 1233 Div II Transfer Switch operated (Div II D.C. ESF Power) 1234 RcIc initiated from Remote Shutdown' Panel 1235 RcIc at rated flow (600 spa) 2 1237 "A" SRV oycled from Remote Shutdown Panel (0 pen for approximately seven seconds) ! 1238 'B" SRV cycled from Remote Shutdown Panel
- (0 pen for approximately nine seconds) 1239 Start of Stable Control Period in Hot i Shutdown 1313 Completion of Stable control Period in Hot Shutdown i
5upplC_utt 3 Paga 3 26-3 Time Event 1313 Transfer switches operated (RSP Transfer to Control Room Control) 1313 Test Termination . The remaining testing within this section, involving a demonstration of the plant's capability to reach cold shutdown conditions from outside the control room, is scheduled to be performed in Test Condition Six. Supplement 3 Paga 3 27-1 3 27 Flow control 't 3 27 1 Purpose
- a. To determine the correct gain settings for the individual recirculation controllers,
~
- b. To demonstrate plant response to changes in recirculation flow in both local annual and master annual mode.
- c. To set the limits of range of operation for the recirculation pumps.
3 27.2 criteria Level 1 The transient response of any variable related to -- the recirculation system to any test input must not diverge. Level 2 The decay ratio of the speed loop response shall be <0.25 at any speed. Flow control systen limit cycles (if any) must produce a turbine steam flow variation no larger than _4 5 percent of the rated steam flow value. , The APRM neutron flux trip avoidance margin shall be >7.5 percent, and the test flux trip avoidance margin shall be _>5.0 percent as a result of the
- recirculation flow control maneuvers.
3 27 3 assults In Test Condition Two, t 54 step change testing was performed on both recirculation systen speed control i i loops in the local manual mode at 38.85 Reactor I power and 47 55 core flow. A review of the data recorded indicates no variables related to the recirculation system were divergent. A qualitative review of the speed response of the A Reactor Recirculation MG Set verified that the decay ratio was < 0.25 for the 1 4% speed steps performed. I l Supplement 3 Pcg3 3 27-2 1 The B Reactor Recirculation NG Set exhibited a limit cycle of approximately 2 1/25 speed peak-to-peak when operating at 385 speed. Due to this limit cycle, the "B" speed loop response Decay Ratio could not be verified and will be retested when controller - optimization is performed in Test Condition Three. Flow control systes limit cycles were verified and the peak-to-peak change in gross generator output during steady-state conditions was less than 1 0.5% of rated generator output or 11.5 We peak-to-peak. ', This criteria was satisfied with the largest observed generator output limit cycle of 10.55 We peak-to-peak (g .46% of rated output). The peak APM neutron flux was 57 715. This APM reading includes an APM gain adjustment factor of 1 1.25 which was required due to a high core peaking factor. The calculated APM neutron flux trip avoidance margin was 60.29, satisfying the 3 7 55 criteria. The minimum heat flux trip avoidance margin was 22 395 for the increasing speed steps, satisfying the criteria of 2 5.05 e i 4 i -t ..-,,-.v__ ._ , -..- _ _ , , _ . ,__.___-,,, - _,_._,...r-.,_ _.__ _ ,.__,-.__ ,,_,._.,,m.. - . _ _ . _ , , . , _ _ - _ _ _ , - . Supplement 3 Paga 3 28-1 l l \ 3 28 anotroulation system 3 28.1 Purpose
- a. To verify that the feedwater control system can satisfactorily control the water level without a ,
resulting turbine trip / scram and obtain actual pump speed / flow.
- b. To verify recirculation pump startup under pressurised reactor conditions.
, c. To obtain recirculation systes performance data.
- d. To verify that no recirculation system cavitation occurs in the operable region of the power-flow map. g 3 28.2 criteria ,
Level 1 The response of any level-related variables during pump trips must not diverge. Level 2 The simulated heat flux margin to avoid a scraa shall be greater than or equal to 5.0 percent during the one pump trip recovery. The APRM sargin to avoid a scraa shall be greater than or equal to 7 5 percent during the one pump trip recovery. During the noncavitation verification, runback logic shall have settings adequate to prevent operation in areas of potential cavitation. During the one pump trip, the reactor unter level margin to avoid a high-level trip (L8) shall be greater than or equal to 3 0 inches. During the simulated loss of a feedwater pump test, the recirculation pump NG sets shall run back upon a trip of the runback circuit. Supplement 3 Pag 3 3 28-2 3 28 3 assults During Test Condition Two, recirculation system baseline performance data was recorded at 38.8% reactor power and 47 55 core flow and at 48% reactor power and 55.7% core flow. Further" testing will be performed in Test Conditions three, Four and Six which relate to the above criteria. E l e l 1 Supplement 3 Pag 3 3 29-1 3 29 Imss of t rbine e m tor and offsite Power 3 29.1 Purpose
- a. To determine the reactor transient performance l during the loss of the main generator and all j offsite power.
- b. To demonstrate acceptable performance of the station electrical supply system.
3 29.2 Criteria Level 1 The reactor protection system, the diesel-generator, RCIC and HPCI must function properly without manual assistance. HPCI and/or RCIC system action, if necessary, shall keep the reactor water level above , the initiation level of low-pressure core spray, LPCI, and automatic depressurisation systems. Level 2 If the low-low set pressure relief logic functions, the open/close actions of the SRVs shall occur within 120 psi of their design setpoints. If any safety relief valves open, only one any reopen after the first blowdown. 3 29 3 assults None of the testing associated with this section has been completed to date. O I i ~ ~~~ ~ ~ Supp3esint 3 ~ Paga 3 30-1 3 30 steady-state vibration 3 30.1 Purpose To determine the vibration characteristics of the primary pressure boundary piping (NSSS) and ESF (ECCS) piping systems for vibrations induced by recirculation flows, hot two-phase forces, and hot hydrodynamic transients; and to demonstrate that flow-induced vibrations, similar in' nature to those expected during normal and abnormal operation, will , not cause damage and excessive pipe movement and vibration. 3 30.2 criteria Level 1 The measured vibration levels of the piping shall not exceed the acceptable specified values. . Level 2 i None 3 30 3 assults During Test' Condition One, the RCIC Steam Supply Line inside the drywell and the RCIC Pump Discharge Line near its connection to the Feedwater Line were monitored for vibration using installed sensors during a vessel injection at rated conditions. Evaluation of the data showed that all vibration levels were within acceptable values. During Test Condition Two, steady state vibration is l measured for selected piping systems at 25% (1 5%) of rated steam flow and 50% (15 5) of rated core flow. As of this report date, data has been ' gathered for seven piping systems consisting of Feedwater, Main Steam, Reactor Recirculation, RHR, I SRVs D&J, HPCI and RCIC but results have not yet i been analyzed and will be reported in the next l supplement of this report. . t i Supplement 3 Paga 3 31-1 3 31 nociraulation system riov calibration 3 31.1 rurpose To perform a complete calibration of the installed recirculation system flow instrumentation. 3 31.2 criteria < Level 1 None Level 2 Jet pump flow instrumentation is adjusted so that the jet pump total flow recorder provides a correct oore flow indication at rated conditions. ~ The APRM/RBN flow-bias instrumentation is adjusted to function properly at rated conditions. The flow control systes shall be adjusted to limit maximum core flow to 102 5 percent of rated flow by limiting MG set scoop tube position. 3 31 3 masults . None of the testing associated with this section has been completed to date. l l l l - y ,, .-..._,_. , , , , , _ _.__. . _ _ _ _ _ _ _ _ _ - . _ _ _ . _ - . _ , _ _ . , , _ , , , _ _ , _ , . Supplement 3 P433 3 32-1 3 32 meneter unter cleanup system 3 32:1 rurpose 'The purpose of this test is to demonstrete specific aspects of the mechanical operabiltty of the reactor ~ water cleanup system. 3 32.2 criteria Level 1 , None' Level 2 1 , The temperature at the tube sids outlet of the non-regenerative heat exchtngers .(NRHX)'shall not ^ exceed 1300 F in the blowdenn mode and shall not
- exceed 120 F in the normal mode.
The coolin6 water supplied to the non-regenerative heat exclamgers shall be lees than 6 percent above the flow corresponding to the heat exchangers capacity (as determined from the process diagram) and the existing temperature-differential across the - heat exchangers. The outlet temperature shall not exceed 180 F. . The bottom head flow indicator will be recalibrated against the RWCU flow indicator if the deviation is greater than 25 spa. The pump available NPSH 1a 13 feet or greater during the hot shutdown with loss of RPt recirculation pumps mode defined in the processi cIagrus. 3 32 3 mesults Daring the Beatup test condition, the EUCU system was placed in 4 configuration oc that flow was taken from the bottan-drain and directly fed back to the vessel, bypassing the desineralizers. In this - configurat,1co'G33-610, bottom drain flew, should read the sano as G33-609, system inlet flow. Our data showed n anzimum deviation of 62 gym. Botton drain flow was recalibrated such tilat the Level 2 criteria could be satisfied. Altio during Heatup, the RWCU syster. was operated in both the normal and blowdown modes with the reactor at rated temperature and pressure. Process ~ --- - _--__m __. _ _ _ __ _
- y. . . _ . . _ . . _ _
I I, 0' Paga 3 32-2 ,Y . ,,,. ? , , 6' , / i variables were recorded in order to demonstrate the Pc;;,/ t .
- t. - proper performance of the RWCU system in each of F - -
these modes. The non-regenerative heat exchange c tube side outlet temperatures for the normal and blowdown mode were 112% and 122% s
- .', respectively. These values were within the Level 2
! \ 'l criteria limits of 120% and 130 % for each [j j , , mode. Using temperature measurements from the RBCCW side of the non-regenerative heat exchangers (NRHX) "\ ' l the cooling water flow was calculated to be less l :" than 65 above the NRHX capacity. The non-regenerative heat exchanger cooling water outlet l' ,; ' , temperatures were well within our Level 2 criteria O' , of 1804. All applicable Level 2 criteria were (, < satisfied. ~ ' The remaining testing for the Reactor Vater Cleanup i System (Hot Standby Operation) will be completed in s i' l f Test Condition Four, I n '% $ r l t, ' t b 8 5. t i, l' [- ,, ; g 5 I 's i r f I
- e l
l,? u s' t \ r / [l ~ ~~ supplement 3 , Pago 3 33-1 ,; 3 33 Posidual Heat Removal System i - ,, - l's 3331 Purpose h e purpose of'this test is to demonstrate the ability.of the Residual Beat Removal (RHR) System to - rtaove residual and decay heat free the nuclear .. system so that refueling and nuclear servicing can c be performed. 3332 criteria Level 1 ; Y ~ Level 2 The RHR System is capable of operating in the-suppression pool cooling and shutdown cooling modes - at the flow rates and temperature differentials indicated on the process diagrams. 3 33 3 assults , During the Heatup test phase, each division of the RHR system 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 indicated an excess capacity of 67.5%. This was expected since in early-hest exchanger life the heat transfer coefficient is larger and capacity was determined to accommodate some deterioration.
4 s g ) "' Supplement 3 Page 3 34-1 3 34 Piping System Dynamic Bosponse Tasting ~ 3 34.1 Purpose Verify that piping system structural behavior under probable transient loadings is acceptable and within the limit predicted by analytical investigations. 3 34.2 criteria Level 1 The measured vibration levels of the piping shall not exceed the acceptable specified values. Level 2 The seasured vibration levels of the piping must not exceed the expected specified values. 3 34.3 mesults Piping dynamic transient vibrations were monitored during Heatup, in conjunction with Relief Valve testing, for two SRV lines and selected Main Steam Lines. All vibration data recorded was within the acceptable and expected limits as defined by the 14 vel 1 and Level 2 criteria. . l l i [ l t L . .}}