ML18033B263
| ML18033B263 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/12/1990 |
| From: | Wallace E TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9004230071 | |
| Download: ML18033B263 (23) | |
Text
ACCELERATED D
BUTION DEMONS aI ON SYSTEM REGULATORY INFOK41ATZON DISTRIBUTION SYSTEM (RZDS)
ACCESSION NBR:9004230071 DOC.DATE: 90/04/12 NOTARIZED: NO DOCKET ¹ FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH.NAME AUTHOR AFFILIATION WALLACE,E.G.
Tennessee Valley Authority RECZP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to NRC 900212 request for info re power ascension program at Unit 2.
DISTRIBUTION CODE:
IE26D COPIES RECEIVED:LTR I
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TITLE: Startup Report/Refueling Report (per Tech Specs)
NOTES:1 Copy each to: S.Black,D.M.Crutchfield,B.D.Liaw, R.Pierson,B.Wilson D
05000260 S
RECIPIENT ZD CODE/NAME LA ROSSiT.
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TE>INESS:-E VALLEYAUTHORI I v CHATTANOOGA. TENNESSEE 37401 5N 157B Lookout Place P,PR i P 1980 U.S. 'Nuc 1 ear Regu 1 atory Commi s s i on ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket No. 50-260 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 -
POWER ASCENSION AND RESTART TEST PROGRAMS
References:
(a) Letter from S.
Black to O.
D. Kingsley, Jr.
dated February 12,
- 1990, "Request for Information Regarding the Power Ascension Program of Browns Ferry Nuclear Plant, Unit 2" (b) Letter from R. Gridley to NRC dated February 14,
- 1989, "Power Ascension Program" (c) Letter from M. J.
Ray to NRC dated October 30,
- 1989, "Power Ascension Hold Points" As requested in your letter dated February 12,
- 1990, Enclosure 1 provides a
response to each of the four questions regarding BFN's Power Ascension and Restart Test Programs.
Enclosure 2 provides a matrix for the scope of planned testing during the Unit 2 Cycle 6 postoutage startup compared to those tests performed during a Near Term Operating License Plant startup with a discussion of significant differences.
Enclosure 3 provides a list of commitments.
The last formal presentation of the power ascension testing was made in the summer of 1988.
Since that presentation, TVA has refined the Power As=ension Program to be more integrated and comprehensive (e.g.,
thermal expansion testing and piping vibration testing have been expanded to include the scope of work involved with the Inspection and Enforcement Bulletin 79-14 Program).
TVA looks forward to presenting the program to you.
TVA considers that as a result of the testing perfor.med through the Restart Test
- Program, the BFN Operability Testing
- Program, and the Power Ascension
- Program, there is reasonable assurance that the safety systems as modified will perform as designed.
9004230071 900412 PDR ADOCK 05000260 P
An Equal Opportunity En11)loyor
~.g6
U.S. Nuclear Regulatory Commission If there are any questions please telephone Patrick P. Cari er, BFN, at (205) 729-3566.
Very truly yours, TENNESSFE VALLFY AUTHORITY E.
G. Wallace, M nager Nuclear Licensing and Regulatory Affairs cc (Enclosures):
Ms.
S.
C. Black, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Comaission One White Flint, North 11555 Rockvi lie Pike Rockvi lie, Maryland 20852 Mr. B. A. W/lson, Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resi~'ent Inspector Browns Ferry Nuclear Plant Route 12, Box 637
- Athens, Alabama 35609-2000
Responses to NRC Questions on Power Ascension Testing (PAT) and Restart Test Program (RTP)
Enclosure 1
Statement of NRC Question ¹1:
Since the Browns Ferry Final Safety Analysis Report (FSAR) is currently being updated and may not completely represent the reconstituted design basis for Unit 2, the PAT program should be reevaluated by comparing it against the forthcoming changes to the updated FSAR (e.g.,
Section 13.10).
Answer to NRC Question ¹1:
TVA has completed the verification review for FSAR Section 13.10 and compared the updated FSA.". against the PAT program.
No changes to the PA'rogram resulted from this evaluation.
The descriptions found in Section 13.10 correspond to testing which will be performed during each refueling startup.
This section does not include testing which is performed to meet a specific requirement that is unique to a particular unit or startup.
Becaus) of the prolong.d nature of the Unit 2 shutdown and the amount of work performed during the outage, testing in addition to that listed in Section 13.10 is being performed.
These additional tests were presented in References (b) and (c).
Furthermore, although BFN's FSAR is presently being updated, current procedures ensure that the PAT program will support the reconstituted design basis through the Design Baseline Verification Program (OBVP) and the Joint Test Group (JTG).
The objective of'he DBVP, as described in the Browns Ferry Nuclear Performance Plan (BFNPP), is to ensure that the functional plant configuration is reflected on appropriate plant documents and conforms to the safe shutdown design basis requirements.
In order to support t iis objective, design basis documents for the systems or portions of systems required to support the Safe Shutdown Analysis for the plant were developed.
These design documents were then used for in>ut into BFN's updated FSAR and Baseline Test Requirements Documents (BTRDs).
Testing described in the BTRDs has been or will be performed through the RTP or through the PAT program.
Testing which could only be performed during power operation has been included in the PAT program.
Nhen a restart test is completed (DBVP assumptions verified), the test results are reviewed and approved by a JTG, which includes membership by Nuclear Enginee. in; (NE). If a specific test requirement (assumption) is not satisfied, NE evaluates the impact and determines what corrective action is required.
This could result in additional testing or ch nges to the design criteria, changes to the BTROs and/or necessary changes to the plant.
This process is evaluated for FSAR impact.
This ensures that the PAT program supports the design basis as found in the updat.d FSAR.
As an additional check to ensure that the above process results in an adequate PAT program, each PAT identified as a critical test is reviewed and approved by the JTG before its performance.
This review ensures that PAT meets the requirements of the current design basis.
Statement of NRC Question 42:
Does the PAT program rely on any other source documents affected by changes in the design basis?
Have these been reevaluated for their impact upon the PAT program?
Answer to NRC Question 42:
The PAT program relies on two programmatic source documents which are affected by changes in the design basis:
the FSAR, Section 13.10, "Refueling Test Program" and the DB'/P.
As explained in response to question one, controls are in place to ensure that chang s are incorporated into the PAT program if the design basis is modified.
Statement of NRC Question ¹3:
Much of the Unit 2 Restart Test Program (RTP) was completed last
- summer, in 1989.
The current TVA schedule projects a restart date for the early part of summer 1990.
During the year in between, a considerable number of significant modifications have been or will be accomplished.
TVA should explain their basis for not repeating applicable RTP tests for the affected
- systems, given the recent regulatory history of post-modification testing problems at BFN Unit 2.
Furthermore, how is TVA assuring that the safety systems of BFN, Unit 2, are being maintained in a condition consistent with the as-tested condition considering the extended period of time since the RTP was completed.
Response
to NRC Question ¹3:
The primary purpose of the
- RTP, as described in Volume III of the NPP, is to verify systems are capable of meeting their safe shutdown requirements.
The DBVP generated the testing requirements necessary to verify the system design functions utilized in satisfying the safe shutdown analysis.
The RTP established test objectives to meet these test requirements.
The RTP has been executed by taking a "snapshot in time" approach for each system with test requirements.
Due to the requirement for the RTP program to collect data to support the Safe Shutdown Analysis, it was determined to perform the testing at the earliest opportunity in order to identify problems early in the program.
Testing has been performed as system conditions
- allowed, and system status during testing was documented within the RTP The test release process for the restart test ensured that the system was ready to verify the design functions required by the DBVP.
Hhen new design changes or calculations are issued after the RTP has been completed, test scoping documents are generated to delineate any additional testing requirements necessary to validate these changes.
These test requirements are satisfied by development of Post Modification Testing (PMT) instructions for design changes and by special tests for calculations.
For this reason RTPs do not have to be repeated.
The system pre-operability checklist (SPOC) process is the method used to recommend and document system operability.
Areas evaluated by the SPOC process include modification closures, temporary alterations,
- PMTs, maintenance, licensing commitments, condition adverse to quality re>.orts, surveillance instructions,
- drawings, system status control and restart testing.
For those systems that have undergone substantial modifications (i.e., high pressure coolant injection system and control bay ventilation system),
integrated system testing is planned.
Additionally, major control systems which could be affected by the long outage are being tested thoroughly as part of the PAT program.
TVA has evaluated the concerns related to PMT and determined that most of the problems identified have been primarily related to the modification activities, not the PMT.
That is, the PMT discovered errors in the performance of modification work as they are intended.
TVA acknowledges that there have been isolated cases where PliTs did not discover existing hardware deficiencies related to modification activities.
In order to correct this, programmatic changes have been implemented to reinforce the PMT program.
For
- instance, responsibilities for developing test requirements have been clarified; test guides and matrices have been incorporated into the program to assist in making PMT determinations; and documentation of testing, including
- signoffs, have been added to design change packages.
These
- changes, in conjunction with the operability testing performed during the system "return to service" process, will provide reasonable assurance that BFN Unit 2 safety systems are adequately maintained beyond the completion of the RTP.
In summary, TVA considers that the objective of the RTP (to ensure that plant systems are capable of meeting their safe shutdown requirements) has not been invalidated by the modification activities performed since the completion of the applicable RTP tests.
TVA also considers that as a result of the testing performed through the Restart Test Program, the BFN Operability Testing
- Program, and the Power Ascension
- Program, there is reasonable assurance that the safety systems as modified will perform as designed.
Statement of NRC Question ¹4:
Except for Reference (b), which infers that RG 1.68 was the basis for establishment of "Management Hold Points",
there is no apparent correlation between the guidance of RG 1.68 and the PAT program proposed for BFN, unit 2.
TVA should identify any significant differences between RG 1.68 (including its supplements) and the P,T program, and provide appropriate explanations for these differences.
Answer to NRC Question ¹4:
References (b) and (c) provided an overview of the testing which is currently planned to be performed during the Unit 2 Cycle 6 startup.
TVA considers that this testing compares closely to the testing described in Regulatory Guide (RG) 1.68, revision 2 for an initial startup test program.
Enclosure 2
provides a detailed comparison of RG 1.68 and BFN's PAT program (in the format of Inspection and Enforcement tianual Inspection Procedure
- 72300, Appendix 2),
and a detailed discussion of the significant differences.
In general, significant differences fall into the following categories:
Performing tests at the plateaus specified by BFN FSAR in section 13.10 versus those plateaus listed in RG 1.68.
Not performing baseline determination testing for those parameters unaffected by the long outage.
Not performing selected transients (e.g.,
natural circulation, loss of feedwater
- heating, main steam isolation valve closure) to verify specific dynamic core response which could only be affected by a new core design and for which sufficient data is already available.
Not performing testing on equipment which is not installed on BFN Unit 2 (Inclined Fuel Transfer, Suppression Pool Makeup, Partial Scram, etc).
During the development of the test program for Bro'ms Ferry Unit 2, TVA conducted a review of the restart testing planned.or Peac'ottom and Pilg:im
(>>hich were also performing startups following exten ied ou ;ges).
Now that these programs have been completed, BFN is conducting anoti~.~r review in order to take full advantage of the experiences gained from these startups.
Lessons learned from this review will be incorporated into the PAT Program.
<<)
ENCLOSURE 2
CORRELATION BETNEEN BROHNS FERRY NUCLEAR PLANT'S PAT PROGRAM AND RG 1.68 BFN Test TEST NAliE OPEN VESSFL 1-55'/
55-100%
- TI 136 SI 4.1.8-3
- TI 135 TI 188 TI 189 TI 149/TI 181 TI 190
'TI 137 SI 2.1 Not Applicable ATI 130
- TI 132 NA NA TI 201 NA TI 174 TI 183 NA
- TI 82 TI 184 SI 4.8.8.l.a.l TI 139
<NA)
Chemical/Radiochemical X/N Radiation Measurements Full Core Shutdown Margin Control Rod Drive System Source Range Monitor Hater Level Measurements Intermediate Range Monitor Local Power Range Monitor Calibration Average Power Range Monitor (Constant Heatup)
Average Power Range Monitor Calibration Process Computer Reactor Core Isolation Cooling System High Pressure Coolant Injection System Selected Process Temperature System Expansion Core Power Distribution Core Performance Core Power Void Mode Response Pressure Regulator Feedwater System Turbine Surveillance Main Steam Isolation Valve Safety Relief Valve Turbine Trip Cooldown Outside Control Room X
Recirculation System Tuning X
(Runback)
Recirculation System Loss of Offsite Power Turbine Trip Drywell Piping Vibration Reactor Pressure Vessel In.ernals Vibration Recirculation Flow Calibration Reactor Hater Cleanup System Residual Heat Removal System Drywell Temperatures Reactor Building Closed Cooling Mater System Offgas System Containment Inn.rting X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N N
X/N X/N X/N X/N X/N X/N X/>>
X/N N
X/N X/N N
X/N X/N X
X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N X/N N
X/N X/N X/N N
X/N X/N X/N X
8rouns Ferry Test ii Required 8y Rear Term Operating License (RG l.68)
'Required in FSAR Section 13.10
SIGNIFICANT DIFFERENCES BETWEEN BROWNS FERRY NUCLEAR PLANT'S PAT PROGRAM AND RG 1.68 Control Rod Drive (CRO)
S stem During an initial plant startup, additional control rod data is collected for selected rods at various temperatures to verify that thermal expansion of the vessel internals will not affect control rod performance.
These rods are also monitored during planned reactor scrams to verify proper performance.
As this was verified in the initial startup program and no significant work has been done to the reactor internals, these tests will not be repeated.
The following table summarizes the planned operations for the CRO system.
Pressure 0
Rated Rated Descri tion Control Rod Cou lin Check Insert and Withdrawal Timin Functional Check of Position Indication Runnin and Stall Flow Friction Testin Core Scram with less than 50 ercent densit~
Hain Steam Isolation Valve (HSIV) Testing During an initial plant startup, a closure of all HSIVs would be performed at high power (90-100 percent) to verify plant performance.
This test was satisfactorily demonstrated during the initial startup of Unit 2.
No modifications have be n made that would significantly affect plant performance; therefore, this test does not need to be reseated.
Feet'.>ter System During an initial plant startup, trips of reactor feedpumps would be performed to verify plant performance.
This testing was satisfactorily demonstrated during the initial startup of Unit 2.
No modifications have been m de that would significantly affect plant performance; therefore, this test does not need to be repeated.
Turbine Generator Teston During an initial plant startup, a turbine generator load reject (TGLR) would be performed at Hig'.~ Power (90-100 percent) to verify plant performance.
Additionally, turbine trips within the capacity of the turbine bypass valves would be performed to verify that the reactor does not scram.
This testing was satisfactorily demonstrated during the initial startup of Unit 2, and as no modifications h.ve been made that would significantly affect this performance, they do not need to be repeated.
Jg
Loss of Turbine/Generator and Offsi te Power During a NTOL star."up, a loss of offsite power coincident with a turbine generator trip would be performed to verify electrical and reactor system transient performance during a loss of auxiliary power.
A significant amount of integrated electrical system testing has been performed as part of the RTP which included four Loss of Offsite Power demonstrations.
In order to minimize electrical transients to plant switchgear as well as transients to Balance of Plant Systems (i.e., feedwater, condensate, turbine support
- systems, etc.),
TVA does not plan to perform this test at power.
Shutdown from Outside the Control Room A scram from outside the control room was successfully performed during the Unit 3 test program on October 1,
1976.
The scram was initiated by shutting the HSIVs from the remote shutdown panel.
As it is fairly easy to scram the plant from outside the control room (e.g., trip reactor protection system
[RPS] channels, RPS motor generator
[HGl sets, etc.)
and there is no manual trip feature provided at the remote s,utdown panel that would require operational
- testing, TVA considers that no additional training would be gained by performing a reactor scram as part of this evolution.
This coupled with the desire to minimize the transients inflicted on the plant during the Unit 2 startup lead to the decision to only perform the hot standby and controlled cooldown portions of this test during the power ascension program.
Dr well Vibration During the initial startup of a Reactor Plant, all safety related piping systems in the drywell are moni'.ored during scheduled transients to develop baseline vibration levels.
Browns Ferry Nuclear Plant (BFN) is devel:ping a
list of specific locations to be monitored consistent with the modifications which have been performed during this outage.
Reactor
".essel Internals Vitration This testing requires special equipm nt to be installed inside the Reactor Vessel.
It was performed during the initial test
- program, and no work has been performed inside the vessel which would require repeating this test.
Residual Heat Rer oval (RHR) System During an initial test p~..>gram, RHR heat exchanger performa,>ce
~ auld be verified with the system operating in the shutdown cooling mode and the suppression pool cooling mode.
This data was taken during the initial startup test pro.ran and need not be rep ated.
During a NTOL startup, individual and dual trips of recirculation pumps would be performed to verify dynami". core response.
This testing is not planned as it was satisfactorily demonstrated during the initial startup test program.
Additionally, current plant technical specifications (3.5.N and 3.6.F.4) and procedures (2-AOI-68-1, Recirculation Pump Trip) req iire a scram to be immediately initiated should both recirculation pumps trip.
~
~
J
Selected Process Tem erature During a
NTOL startup, recirculation flow is lowered to ensure that temperature stratification does not occur in the reactor at the lowest possible recirculation flow.
As the setpoint for the Recirculation MG Set low speed limiter has not been
- changed, there is no requirement to repeat this test.
Core Power Void Mode Res onse Test This test was performed on early boiling water reactor plants to prove that the transient response of the reactor to a reactivity perturbation was sufficiently stable.
This test was performed during the initial startup of BFN Unit 2, and no changes to the basic core design have been made during this outage which would af'feet the dynamic stability of the core.
Additionally, the test was normally performed in test condition 4 (natural circulation),
and BFN is not allowed to operate in this region by its current version of technical specifications.
Reactor Core Isolation Conti,ng
<RCIC> System During an initial plant startup, baseline readings are taken on RCIC steam supply line high-flow isolatiot, circuitry to provide an accurate value for the setpoint.
As this value was obtained during the initial startup test program, and no work was accomplished which would affect this data, these setpoir ts will not be adjusted.
During an initial plant startup, additional "cold start" demonstrations are performed to improve the confidence level in system performance.
As the plant was in operation for several
- years, and the RCIC system performed reliably during this time period, TVA considers that the present program adequately demor strates system reliability.
The following table summarizes the planned operations of the system.
Pressure 150 PSIG Rated Rated 150 PSIG Rated In~ec tion Path Condensate Stor<>ge Tank Condensate Stor~a
~ Tank Conde,rsate Storage Tank Condensate Storage Tank Reactor Vessel Description Rated Flor
~ii th Auxilia'y Boiler Hot (buick-start (2-TI-188)
Cold Quick-start (SI)
Hot Quick-start (2-TI-188)
Hot Quick-start (If R~euired)
High Pressure Coolant Injec..ion (HPCI)
~S si.em During an initial plant startup, baseline readings are taken on H.,I steam supply line high-flow isolation circuitry to provide an acct.rate v<< u'or the setpoint.
As this value was obtained du: ing the initial startup test program, and no work was accomplished which would affect this data, these setpoints will not be adjusted.
During an initial plant startup, additional "cold start" demonstrations are performed to improve the confidence level in system performance.
As the plant was in operation for several
- years, TVA considers that the present program adequately demonstrates system reliability.
The following table summarizes the planned operations of the system.
Pressure 150 PSIG Rated Rated 150 PSIG Rated In ection Path Condensate Stora e Tank Condensate Stora e Tank Condensate Stora e Tank Condensate Stora e Tank Reactor Vessel Oescri tion Rated Flow with Aux Boiler Hot ui ck-start (2-TI-189)
Cold ui ck-start (SI)
Hot uick-start (STEAR 8611)
Responses to NRC (Iuestions on Power Ascension Testing (PAT) and Restart Test Program (RTP)
Enclosure 3
Commitment BFN is conducting a review of Peach Bottom and Pilgrim startup programs in order to take full advantage of the experiences gained from their startups.
Lessons learned from this review will be incorporated into the Power Ascension Testing program.