ML18033A630
| ML18033A630 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/14/1989 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8902170281 | |
| Download: ML18033A630 (13) | |
Text
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REGULATORY INFORMATXON DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR: 8902170281 DOC. DATE: 89/02/14 NOTARIZED: NO FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee AUTH.NAME AUTHOR AFFILIATION GRIDLEY,R.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFXLIATION Document Control Branch (Document Control Desk)
SUBJECT:
Describes power ascension program to be performed during restart of facility.
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TITLE: TVA Facilities Routine Correspondence NOTES:1 Copy each to: S.Black, J.G.Partlow, B.D.Liaw, F.McCoy DOCKET g 05000260 05000260~
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TENNESSEE VALLEYAUTHORITY CHATTANOOGA. TENNESSEE 37401 5N 157B Lookout Place U.S. Nucleax Regulatory Commission ATTH:
Document Control Desk Mashington, D.C.
20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket No. 50-260 BROtTHS FERRY NUCLEAR PLANT (BFH) UNIT 2 POWER ASCENSION PROGRAM This letter describes the Power Ascension Progxam which will be performed during the restart of BFH unit 2.
The Powex Ascension Program includes three categories of tests:
(1) those which are normally performed after a refueling
- outage, (2) those which axe being pexformed (because of the extended unit 2 outage) to reacquaint operations personnel with integrated plant response to specific evolutions, and (3) certain system tests which could not be performed during the Restart Test Program because of plant cond3.tions and are required by the Baseline Test Requirements documents.
Enclosure 1 prov3.des a
description of the tests in categories two and three.
The Master Refueling Test Instruction (MRTI) will be used to coordinate the unit 2 operational and test activit3.es following fuel load.
The MRTI will sequence the testing and serve as a scheduling guide.
The Power Ascension Program is d3.vided into three plateaus.
Plateau 1
Open Vessel Testing Plateau 2
Initial Heatup to 55 Percent Rated Power Plateau 3
55 Percent to 100 Percent Rated Power After completion of all testing within a given plateau, the Plant Operations Review Committee w3.11 review the results and give permission to proceed to the next. testing plateau.
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8902i7028i 890214 PDR ADDCK 05000260 P
PDC An Equal Opportunity Employer
U.S. Nuclear Regulatory Commission Enclosure 2 provides the Power Ascension Test Summary Matrix which lists the test procedure
- numbers, tests
- names, and the plateau(s) during which they will be performed.
Enclosure 3 provides a list of abbreviations used in the test, names section of the Power Ascension Test Summary Matrix.
Please refer any questions to Patrick Carier, BFN, at, (205) 729-3570.
Very truly yours, TENNESSEE VALLEY AUTHORITY g
R.
G idley, iftanager Nuclear Licensing, and Regulatory Affairs Enclosures cc (Enclosures):
Hs.
S.
C. Black, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission One Mhite Flint,, North 11555 Rockville Pike Rockville, Haryland 20852 Hr. F. R. HcCoy, Assistant, Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Harietta Street, NM, Suite 2900 Atlanta, Georgia 30323 Browns Ferry Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637
- Athens, Alabama 35609-2000
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Fnclosure 1
DESCRIPTION OF POblER ASCENSION TESTS ADDED AS A RESULT OF THE FXTENDED UNIT 2 OUTAGE AND TRANSFERRFD FROM THE RESTART TEST PROGRAM 1.0 Hi h Peessuee Coolant In'ection S stem (HPCI)
The HPCI System will be operated during power ascension to ensure proper perfonnance.
This will be done by running the system via U>o different, flow paths and insuring the system can provide required flows.
The HPCI System will be opeeated at both 150 psig and rated reactor pressure.
Conteoller 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, via the test return path to the CST.
Because oE the signiEicant amount, of work that has been done on the HPCI System, the system will also be operated with the dischaege path to tho eeactoe vessel.
2.0 Reactoe Core T. olation Coolin S stem (RCIC)
The RCIC System will be operated during power ascension to ensure peopoe performance over its expected operating pressure and flow ranges.
This villbe done by running the system with suet,ion Eeom the CST and discharge via 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 peessure.
3.0 Thermal E
ansion This te.,t consists of conducting visual walkdown inspections of selected systems which have been identified by Nucleax Engineering and will bo performed at Eoue.peciEic times in the staxtup pxogram.
The fiest inspection will be pexformed before the initial heatup.
Another inspection will bo peefoemed at an intermediate temperature (approximately 150 psig eeactor pressure) and the third inspection will be peeEormed 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 vill be no obstructions which will interfeeo with the thermal expansion of the piping systems.
Selected hangers will be checked to insure they aro not bottomed out ox have their spring fully stretched.
4.0 Turbine Txi Mithin B ass Ca acit The purpose of this test, is to acquaint Operations personnel with U>o response of the reactor and its control systems to protective teips.
A geneeator teip will be performed at low powee level such that nucleae boiler steam geneeation is within the bypass valve capacity (approximately 30 percent) and below the power level at which a turbine trip scram is inhibited (approximately 30 percent power as read by turbine Eirst stage pressueo).
Enclosure 1 (Cont'd)
Page 2
5.0 Reactor Peed um Tri The purpose of Uxis test is to acquaint Operations personnel with the integrated plant response to a tvip of one feedwater pump.
In addit.ion, the recix;culat.ion flow control system will be monitored to demonstrate the capability to reduce x:eactor powev to px.event a low water level scram.
Mith the reactor operating between 90 and 100 pex..cent power, one of the normally operating feedwatex.
pumps will be tvipped and the automatic vecivculation runback circuit will act to lower reactov power Co within the capacity of the x.emaining fecdwater pumps.
6.0 Turbine Tri /Reactor Scram The purpose of this proceduve is to acquaint Operations pex.sonnel with the integrated plant response to a turbine trip with a reactor scram.
In addition, all safety systems will be monitored to insuve that they function properly without manual assi tance.
This test, may he deleted if, as a result of othev testing or unplanned
- events, the Plant Manager feel" that sufficient opevator experience has been gained with the associated systems.
The turbine txip will be performed at. sufficient power to cause a reactor scram and energize the end of cycle recivculation pump trip cix.cuitry.
The subsequent transient px..essuve rise will be limited by the bypass valves initially, and the.,afety relief valves if necessax.y.
7.0 Loss of Turbine Generator and Offsite Power The purpose of this test is to acquaint Operations personnel with the integx.ated plant response to a loss of offsite power (500 kV) to unit 2 coincident with a turbine generator trip.
In addition, all safety systems will be monitoved to insure that, they function pvopevly wiU>out manual assistance.
The loss of turbine generator and off-site power will be pevformed with unit 2 operating between 20 and 30 percent.
Tho 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 veactor pressure vessel (RPV) watex.. level control has been established or 30 minutes, whichever is longer.
During the loss of power transient, temperature and pressure of the reactor and drywell and tempex;ature of the suppression pool will be monitored as well as any required manual opevator actions.
8.0 Recivculation S stem Performance The purpose of this test is to acquaint Operat,ions personnel with the integrated plant x.'esponso to a tx;ip of a reactov recix;culation pump.
addition, the feedwater contvol system will be obsevved for. propev performance and the reactor and vecivculation system tempevatuve monitoved.
Enclo -ure 1 (Cont'd)
At intermediate powoe levels (below 80 povcent rod line), the individual eecirculation pump will be tripped.
Tho reactor vessel and xecirculation system tempeeatuees will be monitored to insuee that tempeeatuee diffoeential limitations Eor restarting the recirculation pumps aro not exceeded.
The feedwater system will be monitored to vevify U>at the feedwatex control system can satisfactorily contvol the water level without a resulting turbine trip ox xoactor sceam on low level.
9.0 Cooldown Outside the Conteol Room This test is divided into two portions.
The first is started with the plant in hot shutdown and the main steam isolation valves (MSIVs) shut.
An operat.ions crew (called the test crew) will assume control of the RCIC System and the steam relief valves (SRVs) at the eemote shutdown panel and opevate them as necessaey to control plant pressure and level.
An independent opeeations cvew who axo not members of the test crew will be in the contxol room and will be in control of all evolutions not dixectly eelated to vessol pressure and level.
There will be a communicat,ions link set up to allow infoxvmtion to flow from the eemote shutdown panel to allow the conteol room cvew to monitov the peogress of the test and regain contvol if a problem arises.
Aftex sat.isfactory pressuee and level contvol feom outside the control xoom has been demonst.vated, thi" portion oE the test will be terminated and conteol of plant peossuee and lovel returned to the control coom ceew.
The second portion of this test will start when eeactoe pressuxe has been loweved to the point wheeo the
.hutdown cooling mode of residual heat eemoval (RllR) can be init,iated.
Control of reactox vessel peessuve and level will be shifted to the eemote shutdown panel and all actions eequired to place the plant in shutdown cooling will be performed from outside the conteol xoom.
All other operator actions not eolated to reactor vessel water level and pvossuee conteol will be performed in the main contvol room.
After shutdown cooling has been initiated, eeactoe vessel temperatuee will be lowered sufficiently to demonsteate adequate control of the cooldown cate from outside the control coom.
10.0 Reactoe Mater Cleanu S stem This test, was teansfereed to power ascension fvom the restart test peogeam and will check the opeeation oE valve 2-CKV-69-579 using roactov feedwater discharge peessuve.
The purpose of the test is to vorify that a Elow path to the vessel will be available during, tho RCIC injection by insuring that 2-CKV-69-579 can shut.
11.0 Containment Ineet.in S..tern This test, was teansfoexod to powee ascension feom peogram.
During reactor startup, the containment be monitored to in"uve it can provide niteogen to rate oE 90 to 100 scfm with a minimum containment tank pressure of 100 psxg.
tho ve"tact tost ineeting system will the containmont at the atmospheric dilution
Enclosure 1 (Cont'd)
Page 12.0 Reactor Buildin Closed Coolin Mater RBCCM The portion of the RBCCW test which checks the performance of the drywall cooling system was transEerred to power ascension Erom the restart test, program.
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 condit,ions.
13.0 Control Rod Drive S stem This test is being transferred to power ascension from the Restart Test Program.
It. will verify that the scram discharge system will function to allow a reactor scram by providing sufEicient volume for the control rod drive over piston area and seal leakage water during a reactor scram.
Additionally, each control rod will be timed to insure compliance with plant technical specifications.
ENCLOSURE 2 POMER ASCENSION TEST SUHHARY HATRIX TEST TI 147 TI 115 TI 20 SI 4.2.C-3 SI 4.1.B-3 TI 136 SI 4.1.B-3 TI 135 RTP/PA 071 RTP/PA 073 TI-149 RTP/PA EXP TI 137 SI 2.1 TI 130 TI 131 RTP/PA-FPT SI 4.7.D SI 4.6.D RTP/PA-TTB RTP/PA-TGT RTP/PA-RSD TI 132 RTP/PA-RPT RTP/PA-LOP TI 82 RTP/PA-069 RTP/PA-070 RTP/PA-084 RTP/PA 085 TEST NAHE FULL CORE LOAD FULL CORE SDH CRD IRH LPRH CAL APRH (CONSTANT HEATUP)
APRH CAL PROCESS COHPUTER RCIC HPCI MATER LEVEL HEASUREHENTS SYSTEH EXPANSION CORE POMER DIST CORE THERHAL LIMITS PRESSURE REGULATOR FEEDMATER TUNING FEEDPUHP TRIP HSIV SRV (Functional)
TURBINE TRIP MITHIN BYPASS CAPACITY TURBINE TRIP COOLDOMN OUTSIDE CONTROL ROOH RECIRCULATION SYSTEH TUNING (RUNBACK)
RECIRCULATION PUHP TRIP LOSS OF OFFSITE
- POMER, TURBINE TRIP DRYMELL TEHPERATURES REACTOR MATER CLEANUP REACTOR BUILDING CLOSED COOLING MATER CONTAINHENT INERTING CONTROL ROD DRIVE OPEN VESSEL X
X X
X X
X X
X X
X X
X X
X X
X X
X X
55-100'le X
X X
X X
X X
X
- Te..ts li"ted in enclo.",ur.'o 1.
EHCLOSURE 3
LIST OF ABBREVIATIOHS USED IH THE TEST HAMES SECTION OF THE PO>lER ASCENSIOH TEST
SUMMARY
MATRIX CAL CRD MPCI LPRM MSIV RCIC SDM SRV Average Power Range Monitor Calibration Control Rod Drive High Pressure Coolant Injection Intermediate Range Monitor Local Power Range Monitor Main Steam Isolation Valve Reactor Core Isolation Cooling Shutdown Margin Steam Relief Valve