ML18033B022
| ML18033B022 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/30/1989 |
| From: | Michael Ray TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8911090002 | |
| Download: ML18033B022 (30) | |
Text
ACCELERATED DISHUBUTION DEMONSTF&TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8911090002 DOC.DATE: 89/10/30 NOTARIZED: NO DOCKET g FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH.NAME AUTHOR AFFILIATION RAY,M.J.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to NRC 890922 ltr re power ascension hold points.
DISTRIBUTION CODE:
D030D COPIES RECEIVED:LTR L ENCL $
SIZE:
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TITLE: TVA Facilities Routine Correspondence R
D NOTES:1 Copy each to: S.Black,D.M.Crutchfield,B.D.Liaw, R.Pierson,B.Wilson 05000260 S
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TENNESSEE VALLEYAUTHORITY CHATTANOOGA. TENNESSEE 37401 5N 157B Lookout Place U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
Docket No. 50-260 In the Matter of
)
Tennessee Valley Authority
)
BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 POWER ASCENSION HOLD POINTS In response to NRC's letter dated September 22,
- 1989, the following information is being provided to augment TVA's letters dated February 14,
- 1989, and September 8,
1989.
It should be noted that Test Plateaus are defined in the BFN Final Safety Analysis Report (FSAR Section 13.10) which requires a review and approval of test results by the Plant Operations Review Committee (PORC) before proceeding to the next plateau.
Management Assessment Points (MAPs) have been added for the Unit 2 Cycle 6 restart to augment these Test Plateaus and provide for a more detailed evaluation of plant and personnel performance because of the prolonged shutdown of Unit 2.
Enclosure 1 contains a discussion of the sequence and testing to be performed during both the Test Plateaus and the Management Assessment Periods.
The testing to be performed between MAPs is per the current proposed schedule for the Unit 2 restart, and represents the best information available at this time.
This testing is intended to provide manag'ement with sufficient information on both, personnel and material performance to ensure both are ready to support higher power operation.
Although the sequence given in Enclosures 1
and 3 represents our best estimate, situations may arise which require testing to be shifted from one Management Assessment Period to another.
Testing must be completed or deficiencies generated and evaluated by PORC before moving to the next Test Plateau.
Enclosure 2 provides a brief description of each test.
These descriptions come from BFN FSAR Section 13 '0 and TVA's February 14, 1989, letter to NRC.
Routine testing, such as surveillance procedures, chemistry and maintenance procedures, performed on a periodic basis (i.e., weekly, monthly, or quarterly) have not been listed.
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PDC An Equal Opportunity Employer
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U.S. Nuclear Regulatory Commission OCT 80 SBB Enclosure 3 provides a table of the interrelationship of MAPs and Test Plateaus and summarizes testing scheduled to be performed within these periods.
Please refer any questions to Patrick Carier, BFN, at (205) 729-3570.
Very truly yours, TENNESSEE VALLEY AUTHORITY Enclosures cc (Enclosures):
Ms. S.
C. Black, Assistant Director for Projects TVA Projects Division UPS.
Nuclear Regulatory Commission One Hhite Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. B. A. Nilson, Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NH, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637
- Athens, Alabama 35609-2000 Manage Nu le Licensing and Regulatory Affairs
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ENCLOSURE 1
MANAGEMENT ASSESSMENT DURING THE UNIT 2 CYCLE 6 RESTART
~Pur ose This enclosure contains a discussion of the sequence and testing to be performed during both the.Test Plateaus and the Management Assessment Periods; The testing to be performed between Management Assessment Points (MAPs) is per the current proposed schedule for the unit 2 restart, and represents the best ~information available at this time.
This testing is intended to provide panagement with sufficient information on both personnel and material performance to ensure both are ready to support higher power operation.
Although'he sequence given in this enclosure represents our best estimate,,
situations may arise which require testing to be shifted from one Management Assessment Period to another.
Testing must be completed or deficiencies generated and evaluated by the Plant Operations Review Committee (PORC) before moving to the next Test Plateau (as defined in the Browns Ferry Nuclear Plant (BFN) Final Safety Analysis Report (FSAR)).
Section 13.10 of the BFN.FSAR describes the testing to be performed as part of the Refueling Test Program (RTP).
This program is performed after each refueling outage and is administered by Plant Manager Instruction 26.1.
Because of the prolonged nature of the unit 2 shutdown and amount of work performed during the outage, additional testing requirements have also been added by TVA.
These tests are described in enclosure 2.
Other tests and procedures which are. performed to meet surveillance requirements, routine operations, and maintenance procedures, or are part of the chemistry program, are not included in this listing.
'Discussion The Master Refueling Test Instruction provides a plan by which testing required during restart is sequenced as well as a mechanism for recording completion of tests and documenting reviews.
Three distinctive Test Plateaus are established to ensure that the requirements have been met.
These plateaus are:
Phase I
Phase " II Phase "III Open Vessel Heat-up to -55 percent power 55 percent
)ower tq, 100 percent power Before maneuvering the plant to the next higher plateau, the testing performed in the previous plateau is reviewed by PORC.
Open test deficiencies must be evaluated by PORC along with the test results.
Once PORC has approved the test results, the Plant Manager gives approval to proceed with the next phase of testing.
As an augmentation to the already existing program, additional Management Assessment Periods (hold'points) have been proposed to provide a more detailed assessment of power ascension.
The additional periods are:
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Enclosure 1
Page 2 of 4 a.
After initial criticality and prior to increasing reactor vessel pressure above atmospheric.
b.
Prior to placing the Reactor Mode Switch in "RUN."
cd Prior to exceeding 25 percent power.
d.
Prior to exceeding 80 percent power.
These periods were selected to provide suitable transient and steady-state operating experience for the assessment of plant personnel and equipment performance.
Specific tests may be rescheduled between assessment periods as long as sufficient information exists to accurately evaluate plant performance.
Testing must be completed or test deficiencies written and evaluated by PORC before moving to the next Test Plateau.
Inte rated Schedule When the Test Plateaus from the BFN FSAR are combined with the proposed
- MAPs, seven distinctive evaluation points are established.
These evaluation points are summarized in enclosure 3.
The first evaluation is performed before reactor startup.
This evaluation includes a
PORC review of the Open Vessel Test Results (Phase I testing),
as well as an assessment by management that the plant is ready for restart.
Tests which will be evaluated include the following Technical Instructions (TI):
TI-147 TI-020 TI-135 TI-131 TI-132 TI-190 TI-149 Fuel Loading Control Rod Drive System Testing (Partial)
Process Computer (Partial)
Feedwater Control (Partial)
Recirculation Flow Control (Partial)
System Expansion Monitoring (Partial)
Reactor Water Level (Partial)
The second evaluation is conducted after initial criticality has been achieved.
A one-week period has been scheduled to perform training criticals.
This allows operators to complete their practical factors and management to observe and evaluate their performance.
Additionally, TI-115, Full Core Shutdown Margin (SDM), is performed and can be reviewed to ensure proper core characteristics.
After permission has been re'ceived to proceed from the second MAP, a plant heatup to rated temperature and pressure commences.
The third management assessment is performed before the Reactor Mode Switch is placed in "RUN."
At this point, management is able to review Operation's performance in starting up such key systems as Feedwater, Steam Jet Air Ejectors, and Offgas.
- Also, several key tests are performed to the extent necessary for evaluation of equipment performance at low power.
These tests include:
I l
Enclosure 1
Page 3 of 4 TI-136 TI-190 TI-149 TI-188 TI-189 TI-183 TI-20 Average Power Range Monitor Calibration System Expansion Monitoring (Partial)
Reactor Water Level (Partial)
Reactor Core Isolation Cooling System (Partial)
High Pressure Coolant Injection System (Partial)
Reactor Water Cleanup System Control Rod Drive System Testing (Partial)
After the Reactor Mode Switch has been place in "Run," reactor power is raised to approximately 20 percent, and the containment will be inerted.
The pressure regulator is checked using the Main Turbine Bypass Valves and an initial Turbine Startup is performed.
Turbine trips are performed while power is still within the capacity, of the Main Turbine Bypass Valves in order to familiarize operators'ith this transient.,
Power will then be increased to just below'5 percent and the process computer,and core 'thermal limit verifications performed.
Before exceeding 25 percent core thermal
- power, another management assessment is performed which should include the following test procedures:
TI-135 TI-130 TI-192 Process Computer (Partial)
Pressure Regulator (Partial)
Turbine Trip Within Bypass Capacity Once permission has been obtained to proceed above 25 percent, the plant is maneuvered to clear the restraints of the Rod Worth Minimizer and Rod Sequence Control System so the remaining control rod drives can be-timed during individual scrams (TI-20).
Power is then raised to just below 55 percent, and the remainder of the tests required to be performed for Phase II testing completed.
The first planned scram is performed with a simulated loss of offsite powers This is used as the initiating event.
Once the plant is
- shutdown, a demons,tration of the Remote Shutdown System is planned.
Once the plant has returned to a Cold Shutdown Condition, the last thermal expansion walkdown can be performed.
The final assessment performed during Phase II is the PORC review and approval of all the test results.
In addition to the testing listed in the MAPs, the following additional tests will also be reviewed.
TI-20 TI-131
- TI-182
- TI-180 TI-188 TI-189 TI-190
- TI-186 Control Rod Drive System Testing Feedwater Control system (Partial)
Loss of Offsite power Cooldown from Outside the Control Room Reactor Core Isolation Cooling System High Pressure Coolant Injection System (Partial)
Thermal Expansion Control Rod Drive System
- These tests may be deferred to the end of the program.
The test results review for the 0 to 55 percent power plateau must be completed before the plant exceeds 55 percent power.
0 Enclosure 1
Page 4 of 4 After PORC has approved the test results for Phase II testing and the Plant Manager has given his permission to proceed above 55 percent power, reactor power is increased to approximately 80 percent power.
At this power level, the fuel is preconditioned in accordance with the Preconditioning Interim Operating Management Recommendations requirements.
Preconditioning of the fuel is required to support the tuning on the key control systems (Feedwater, Electro-Hydraulic Controls Pressure Regulator, and Recirculation Flow Control).
Once these systems have been satisfactorily adjusted, another assessment is conducted before exceeding 80 percent power.
The tuning procedures provide ample opportunity to ensure that both equipment and personnel are ready to support full power operation.
This sixth assessment period is the last hold point in the test program before attaining 100 percent power.
This assessment includes:
TI-130 TI-131 TI-132 TI-137 TI-149 Pressure Regulator (Partial)
Feedwater Control System (Partial)
Recirculation Flow Control (Partial)
Core power Distribution (Partial)
Reactor Water Level Measurements (Partial)
The final assessment point is the PORC review and approval of the completed test program.
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- 1. 0 TI-189 2.0 TI-188 3.0 TI-190
- 4. 0 TI-192
- 5. 0 TI-191
- 6. 0 TI-193 7.0 TI-182 8.0 TI-181 9.0 TI-180
- 10. 0 TI-183
- 11. 0 TI-184 12.0 TI-186 13.0 TI-147 14.0 TI-115 15.0 TI-20
- 16. 0 TI-136
- 17. 0 TI-135
- 18. 0 TI-137
- 19. 0 TI-130
- 20. 0 TI-131
- 21. 0 TI-132
- 22. 0 TI-82
- 23. 0 TI-144 24.0 TI-174 ENCLOSURE 2
DESCRIPTION OF POWER ASCENSION TESTS High Pressure Coolant Injection System Reactor Core Isolation Cooling System Thermal Expansion Turbine Trip Within Bypass Capacity Feedwater Pump trip Testing Turbine Trip Loss of Offsite Power Recirculation Pump Tri p Cooldown from Outside the Control Room Reactor Hater Cleanup System Reactor Building Closed Cooling Hater System Control Rod Drive System Fuel Loading After a Complete Core Unload Full Core Shutdown Margin Control Rod Drive System Testing Average Power Range Monitor Calibration Process Computer and Core Performance Core Power Distribution Pressure Regulator Feedwater Control System Recirculation Flow Control Drywell Atmospheric Cooling System Reactor Hater Level Measurements Recirculation System Flow Calibration NOTE:
This enclosure provides a brief description of each test.
These descriptions come from Browns Ferry Final Safety Analysis Report, Section 13.10, and TVA's February 14, 1989 letter to NRC.
Routine
- testing, such as surveillance procedures, chemistry procedures, and maintenance procedures performed on a periodic basis (i.e., weekly, monthly, or quarterly) has not been listed.
Enclosure 2
Page 2 of 7 1.0 Hi h Pressure Coolant In ection S stem" (HPCI)
The HPCI System will be operated during power ascension to ensure proper performance.
This will be done by running the system by the different discharge flow paths and ensuring the system can provide required flows.
The HPCI System will be operated at both 150 psig and rated reactor pressure.
Controller settings for both manual and automatic operation will be determined and the system tested with the suction lined up to the condensate storage tank (CST) and discharging by the test return path to the CST.
Because of the significant amount of work that has been done on the HPCI System, the system will also be operated with the discharge path to the reactor vessel.
2.0 Reactor Core Isolation Coolin S stem (RCIC)
The RCIC System will be operated during power ascension to ensure proper performance over its expected operating pressure and flow ranges.
This will be done by running the system with suction from the CST and discharge by the test return path to the CST.
Control settings for both manual and automatic operation will be determined and the system tested at both 150 psig and rated reactor pressure.
3.0 Thermal Ex ansion This test consists of conducting visual walkdown inspections of selected systems which have been identified by Nuclear Engineering and will be performed at four specific times in the restart program.
The first inspection will be performed before the initial heatup.
Another inspection will be performed at an intermediate temperature (approximately 150 psig reactor pressure) and the third inspection will be performed at rated system pressure and temperature.
The final inspection will be performed after the first thermal cycle when the plant has been shutdown and the temperature of the piping has returned to ambient conditions.
These inspections shall ensure that there wi 11 be no obstructions which will interfere with the thermal expansion of the piping systems.
Selected hangers will be checked to ensure they are not bottomed out or have their spring fully stretched.
4.0 Turbine Tri Nithin 8 ass Ca acit The purpose of this test is to acquaint Operations personnel with the response of the reactor and its control systems to protective trips.
A generator trip will be performed at low power level such that nuclear boiler steam generation is within the bypass valve capacity (approximately 30 percent) and below the power level at which a turbine trip scram is initiated (approximately 30 percent power as read by turbine first stage pressure).
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Enclosure 2
Page 3 of 7 5.0 Reactor Feed um Tri The purpose of this test is to acquaint Operations personnel with the integrated plant response to a trip of one feedwater pump.
In addition, the recirculation flow control system will be monitored to demonstrate the capability to reduce reactor power to prevent a low water level scram.
Hith the reactor operating between 90 and 100 percent
- power, one of the normally operating feedwater pumps will be tripped and the automatic recirculation runback circuit wi 11 act to lower reactor power to within the capacity of the remaini'ng feedwater pumps.
6.0 Turbine Tri /Reactor Scram N
The purpose of this procedure is to acquaint Operations personnel with the integrated plant response to a turbine -trip with a reactor scram.
In addition, safety systems will be monitored to ensure that they function properly.
This test may be deleted if, as a result of other testing or unplanned
- events, the Plant Manager feels that sufficient operator experience has been gained with the associated systems.
The turbine trip will be performed at sufficient power to cause a reactor scram and energize the end of cycle recirculation pump trip circuitry.
The subsequent transient pressure rise will be limited by the bypass valves initially, and the safety relief valves if necessary.
7.0 Loss of Turbine Generator and Offsite Power The purpose of this test is to acquaint Operations personnel with the integrated plant response to a loss of offsite'power (500 kv) to unit 2
coincident with a turbine generator trip.
In addition, safety systems will be monitored to ensure that they function properly.
The loss of turbine generator and off-site power will be performed with unit 2 operating between 20 and 30 percent.
The reactor is expected to scram and isolate as a result of automatic tripping of the reactor protection system motor generator sets on low voltage.
The transient will be extended until the relief valve action (if required) shows adequate pressure control and reactor pressure vessel water level control has been established or 30 minutes, whichever is longer.
During the loss of power transient, the temperature and pressure of the reactor and drywell and temperature of the suppression pool will be monitored, as well as any required manual operator actions.
8.0 Recirculation Pum Tri The purpose of this test is to acquaint Operations personnel with the integrated plant response to a trip of a reactor recirculation pumps In
- addition, the feedwater control system will be observed for proper performance and the reactor and recirculation system temperature monitored.
14 Enclosure 2
Page 4 of 7 At intermediate power levels (below 80 percent rod line),
an individual recirculation pump wi 11 be tripped.
The reactor vessel and recirculation system temperatures will be monitored to ensure that temperature differential limitations for restarting the recirculation pumps are not exceeded.
The feedwater system will be monitored to verify that the feedwater control system can satisfactorily control the water level without a resulting turbine trip or reactor scram on low level.
9.0 Cooldown Outside the Control Room This test is being performed to acquaint Operations with the procedures used to cooldown the plant from outside the control room.
The test is divided into two portions.
The first is started with the plant in hot shutdown and the main steam isolation valves shut.
An operations crew (called the test crew) will assume control of the RCIC System and the steam relief valves at the remote shutdown panel and operate them as necessary to control plant pressure and level.
An independent operations crew who are not members of the test crew will be in the control room and will be in control of all evolutions not directly related to vessel pressure and level.
There will be a communications link set up to allow information to flow from the remote shutdown panel to allow the control room crew to monitor the progress of the test and regain control if a problem arises.
After pressure and level control from outside the control room has been demonstrated, this portion of the test will be terminated and the control of plant..pres. sure and level returned to the control room crew.
The second portion of this test will start when reactor pressure has been lowered to the point where the shutdown cooling mode of residual heat removal can be initiated.
Control of reactor vessel pressure and level will be shifted to the remote shutdown panel and all actions required to place the plant in shutdown cooling will be performed from outside the control room.
All other operator actions not related to reactor vessel water level and pressure control will be performed in the main control room.
After shutdown cooling has been initiated, reactor vessel temperature will be lowered sufficiently to demonstrate adequate control of the cooldown rate from outside the control room.
10.0 Reactor Hater Cleanu S stem The purpose of this test is to verify that a flow path to the vessel will be available during the RCIC injection by ensuring that 2-CKV-69-579 can shut.
The test will be performed to meet the requirements of Restart Test Program (RTP)-069 and will check the operation of valve 2-CKV-69-579 using reactor feedwater discharge pressure.
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Enclosure 2
Page 5 of 7 Reactor Bui ldin Closed Coolin Water This test will verify the ability of the drywell atmosphere cooling system to maintain the bulk volumetric average drywell temperature below 147 F during operating conditions.
Control Rod Drive (CRD)
S stem This test will collect data from a scram performed at less than 50 percent control rod density.
This scram can be either from a previously planned evolution, or from an unplanned transient.
This data will then be used to close out an open test item from the RTP.
FUEL LOADING AFTER A COMPLETE CORE UNLOAD This procedure was used to perform a controlled loading of the complete core.
After loading, a core verification was performed to ensure proper loading configuration.
FULL CORE SHUTDOWN MARGIN After core loading is complete, the shutdown margin (SDM) is demonstrated to be at least 0.38 percent
> K/K throughout the upcoming cycle with the analytically determined strongest rod withdrawn.
The SDM is demonstrated by evaluating core reactivity after achieving criticality.
The actual amount of rods withdrawn are compared to the predicted amount to verify that 0.38 percent g K/K or better margin is available.
CRD SYSTEM TESTING This test is separated into the testing done at zero psig and the testing done at normal operating pressure.
Following maintenance on a
control rod, that rod is functionally checked by stroking it to full length.
While the rod is being stroked, a visual check of the Rod Position Indication System position indication is made;
- and, upon reaching the end of travel, the coupling is checked for over travel.
The functional/position indication check and the coupling check are performed for each rod.
Additionally, friction testing is performed as a diagnostic tool on suspected problem rods.
After reaching rated pressure but before 40 percent
'local power range monitor (LPRM) hookup to the process computer is verified by observing the LPRM response to control rod motion.
1 AVERAGE POHER RANGE MONITOR (APRM) CALIBRATION, Enclosure 2
Page 6 of 7 The APRM's are initially adjusted to maximum amplifier gains.
During the heatup
- phase, the APRM's are calibrated to a core thermal power determined either by a constant heatup rate heat balance or by a bypass valve comparison.
APRM's will be calibrated during a heat balance after feedwater flow becomes reliable.
PROCESS COMPUTER AND CORE PERFORMANCE Phase I
Following completion of the refueling monitor update program (OD-20),
the nuclear steam supply system data will be verified for accuracy and proper location in the computer memory.
Phase II After reaching 10 percent power but before reaching 25 percent
- power, the core thermal power calculated by the process computer is compared to a detailed manual heat balance.
The thermal limits for minimum critical power ratio (MCPR),
maximum average planar linear heat generation rate (MAPLHGR), and maximum linear heat generation rate (LHGR) are compared to manual methods or an off-line computer system.
Phase III The testing done in Phase II is repeated after calibrating the LPRM's.
CORE POHER DISTRIBUTION At least two full Traversing Incore Probe (TIP) sets will be performed in order to measure the TIP uncertainty.
In addition, several partial TIP sets (or OD-2's) will be needed to determine TIP random noise.
The data from these TIP sets will be compared statistically to determine total TIP uncertainty.
In addition, TIP data and Pl/OD10 data taken after TIP sets will be analyzed to determine TIP assymmetry and core power symmetry.
One of the TIP sets must be taken at greater than 75 percent power level, and it is recommended that neither TIP set be taken below 50 percent powers The core must be in an octant symmetric rod pattern to perform these tests.
PRESSURE REGULATOR The pressure setpoint will be varied (both increased and decreased) to produce step changes in pressure and the response of the system is measured.
These tests will be performed at a high enough load limit setting so that control valves alone will control the transient.
The load limit setpoint will then be reduced to test bypass valve capability
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Enclosure 2
Page 7 of 7 to handle the excess transients.
In addition, the ability of the backup pressure regulator will be tested and regulator settings will be optimized.
Simulated failure of each primary and secondary regulator will be performed to verify backup capability of the remaining regulator.
FEEDWATER SYSTEM To provide an observable feedwater system and reactor
- response, two methods are used to initiate test disturbances.
The first is to change the input level setpoint about 4 to 6 inches.
The second is to change one pump flow approximately 10 percent in the manual mode.
The remaining pumps are in automatic. mode to control reactor vessel water level.
Both one eleme'nt and three element modes of level control will be tested.
Tests'are to be performed between 35 percent and 70 percent of rated power.
RECIRCULATION MOTOR-GENERATOR (M-G) SET CONTROL Pre-heatup tests are performed to test individual components and make other preparations for the tests at power.
Once the unit has reached the 100 percent rated core flow point, several small steps changes to M-G set speed will be made and applicable data recorded.
DRYWELL ATMOSPHERIC COOLING SYSTEM The drywell atmosphere cooling system will be placed in operation and its ability to maintain normal operating temperatures inside the drywell verified.
For this test, 8 of 10 fans (and associated coils) are on,
- thereby, demonstrating greater or equal to 25 percent standby heat removal capability.
REACTOR WATER LEVEL MEASUREMENTS This instruction will be performed before and during restart of unit 2.
It addresses two actions:
(1) the evaluation of the reference leg condition which is performed before closing the drywell and pulling control rods, and (2) the recording of pertinent vessel level data during the startup from atmospheric pressure in the reactor to rated pressure and flow.
Recirculation S stem Flow Calibration During the testing program with the recirculation system at rated flow and rated
- power, the jet pump flow instrumentation will be adjusted to provide the correct flow indications.
After the relationship between drive flow and core flow is established, the flow biased APRM/Rod Block Monitor system will be adjusted to match this relationship.
In
- addition, the total core flow and recirculation drive flow signals to the process computer will be calibrated.
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4
0 Enclosure.
3 I
Summar of Test Plateaus and Mana ement Assessment Points This enclosure provides a table of the interrelationship of Management Assessment Points and Test Plateaus and summarizes testing scheduled to be performed within these periods.
Although the sequence given in this enclosure represents our best estimate, situations may arise which require testing to be shifted from one Management Assessment Period to another.
Testing must be completed or deficiencies generated and evaluated by the Plant Operations Review Committee (PORC) before moving to the next Test Plateau (as defined in the Browns Ferry Final Safety Analysis Report).
I.
Open Vessel Test Plateau (Phase I Testing)
TI-147 TI-20 TI-135 TI-131 TI-132 TI-190 TI-149 Fuel Loading Control Rod Drive System Testing (Partial)
Process Computer (Partial)
Feedwater Control (Partial)
Recirculation Flow Control (Partial)
System Expansion Monitoring (RTP/PA-EXP) (Partial)
Reactor Water Level (Partial)
PORC review test results and management assessment to ensure plant ready for restart.
II.
0 55 Percent Test plateau (Phase II Testing)
A.
Second Mana ement Assessment Period
( Initial'; Restart)
TI-115 Full Core Shutdown Margin Operator Training Criticals Management assessment to ensure plant ready to support Heatup to Rated Temperature B.
Third Mana ement Assessment Period (Heatu to Rated)
TI-020 TI-136 TI-149 TI-183 TI-188 TI-189 TI-190 Control Rod Drive System Testing (Partial)
Average Power Range Monitor Calibration Reactor Water Level (Partial)
Reactor Water Cleanup System Reactor Core Isolation Cooling System (Partial)
High Pressure Coolant Injection System (Partial)
System Expansion Monitoring (Partial)
Management assessment to ensure plant ready to support change of Operational Mode and Initial Turbine Roll.
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Enclosure 3
Page 2 of 3 C.
Fourth Mana ement Assessment Period (10-25 Percent Power)
TI-135 TI-130 TI-192 Process Computer (Partial)
Pressure Regulator (Partial)
Turbine Trip Within Bypass Capacity Management assessment to ensure plant ready to increase power to 55 percent.
D.
Fifth Mana ement Assessment Period (25-55 Percent Power)
TI-20 TI-131
- TI-182
- TI-180 TI-188 TI-189 TI-190
- TI-186 Control Rod Drive System Testing Feedwater Control system (Partial)
Loss of Offsite Power Cooldown from Outside the Control Room Reactor Core Isolation Cooling System High Pressure Coolant Injection System (Partial)
Thermal Expansion Control Rod Drive System PORC Approval of Phase II Test Results Management assessment to ensure plant ready to increase power to 80 percent.
- These tests may be deferred to the end of the program.
III.
50 100 Percent Test Plateau (Phase III Testing)
A.
Sixth Mana ement Assessment Period (55-80 Percent Power)
TI-130 TI-131 TI-132 TI-137 TI-149 Pressure Regulator (Partial)
Feedwater Control System (Partial)
Recirculation Flow Control (Partial)
Core Power Distribution (Partial)
Reactor Water Level Measurements (Partial)
Management assessment to ensure plant ready to increase to full power.
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Enclosure 3
Page 3 of 3 B.
Seventh Mana ement Assessment Period (80-100 Percent Power)
TI-130 Pressure Regulator TI-131 Feedwater Control System TI-132 Recirculation Flow Control TI-135 Process Computer TI-137 Core power Distribution TI-149 Reactor Water Level Measurements TI-189 High Pressure Coolant Injection System TI-191 Feedwater Pump Trip TI-181 Recirculation Pump Trip TI-193 Turbine Trip TI-82/184 Drywell Cooling System TI-174 Recirculation System Flow Calibration PORC approval of completed test program
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