Text

Forwards Info Re Scope of Testing Planned for Restart Test Program & Power Ascension Testing Programs,To Support NRC Review of Programs
	ML20092J308
Person / Time
Site:	Browns Ferry
Issue date:	02/18/1992
From:	Zeringue O; TENNESSEE VALLEY AUTHORITY
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NUDOCS 9202240158
	Download: ML20092J308 (43)
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i U.SI Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-

-Gentlemen ui JInithe hatter'of > Docket Nos. 50-?59 Tennessee Valley Authority ) 50-296

-BROWNS FERRY-NUCLEAR: PLANT-(BFN) - REQUEST FOR ADDITIONAL INFORMATION REGARDING'THE RESTART TEST PROGRAM FOR UNITS 1 AND 3

Reference:

TVA letter dated September 27, 1991, " Browns Ferry Nuclear Plant (BFN)'- Restart Test Program (RTP) Description for ,

Unita 1 and 3" This letter provides-the scope of testing planned for BFN Units 1 and 3

-for both the Restart Tect Program and Power Ascension Testing Program to support NRC Staff review of these programs.- In the: referenced letter, TVA provided an overview of the Unit 2 Restart Test Program and a discussion of. lessons': learned,-and defined-the Restart Test Program ,

planned for= Units _1 Land 3. On. November-5,-1991, the NRC requested TVA provide a comparison of the-similarities'and differences.between planned testing f or Units 1 and 3 and actual Unit 2 testing. - The' Staff also requested that TVA relate how Unit 2_-met Regulatory Guide (RG) 1.68 criteria and how.TVA plans to meet that criteria for~Unita 1 and 3. i Enclosures 1 and 2 provide this information for the Restart Test Program and:the Power-Ascension Testing (PAT) Program-respectively. The criteria review provided as'part of each enclosure apply to both Units 1 and.3.

zEach enclosure has a-table that correlates actual Unit.2 tests to planned LUnit 3 tests _. For the Restart Test Program (Enclosure 1, Table 1), there

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2 U.S. Nuclear Regulatory Commission FEB 181992 are 53 system functions (approximately 20 percent of the total) in which TVA can not conclusively determine whether a Unit 3 RTP test will be required until the Baseline Test Requirements Documents (BTRDs) are completed. The BTRDs contain the evaluation necessary to determine the full extent of testing required. TVA vill provide the NRC an update of the table in Enclosure 1 to reflect the outcome of BTRD evaluations by December 31, 1992. However, this information is not needed to evaluate the Units 1 and 3 RTP and its conformance to RG 1.68 guidelines.

Therefore, TVA requests that the NRC review and approve the RTP and the PAT Program for BFN Units 1 and 3, as described in Enclosures 1 and 2 and the referenced letter by June 1992.

A summary list of commitments contained in this letter is provided in Enclosure 3. If you have any questions, please contact Raul R. Ba. on, Site Licensing Manager, at (205) 729-7570.

Sincerely, n

// ,

, b 0, J. Zeringue Enclosures cc (Enclosures):

NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35611 Mr. Thierry M. Ross, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 l

4 i ENCLOSURE 1 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS L

ENCLOSURE 1 Page 1 of 31 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS

1. BACKr,dOUND TVA's R6 start Test Program Plan for DFN Unit 2 was initially submitted to the NRC by TVA letter dated October 7, 1986 and supplem7ated by letter dated July 13, 1987. The proposed program was also described in Volume III, Section 8.0 of the Nuclear. Performance Plan. TVA presented-the BFN Restart Test Program (RTP) to the NRC Staff during meetings held on April 26, 1988, and June 21, 1988. The NRC conducted several inspections of the implementation of the RTP as documented in Inspection Reports (irs) 50-260/87-12, 87-27, 87-30, 87-33, .

87-37, 87-42, 87-46, 88-02, 88-04, 88-05, 88-10, 88-16, 88-18, and 88-21.

TVA's RTP was accepted by the NRC Staff as documented in Enclosure 2 to NRC

' letter dated-August 12, 1988 and NUREG-1232, Volume 3, Supplement 1 dated October 24, 1989.

During the. April 26, 1988 meeting between TVA and the NRC, the NRC requested additional inforn.ation regarding the dif ferences between typical preoperational test requirements, as described in Regulatory Guide 1.68,

" Initial Test Programs For Water-Cooled huclear Power Planta, Revision 2, August 1978," and the BFN RTP, TVA provided the NRC with information that explained and justified the differences between-the criteria used by BFN to identify systnm functions that require testing and the criteria in Section C.1 of the Regulatory Guido during the June 21, 1988 meeting. TVA also compared the types-of. testing identified in the Regulatory Guide and the types of testing that would be required at BFN Unit 2 prior to restart.

Subsequent to our September 27, 1991 submittal that described the RTp planned for Units 1 and 3, the NRC requested that TVA provide similar information for the Units 1 and 3 programs.' This information is provided in Parts II and III of this Enclosure. Part-II also provides a brief restatement of the Unit I and 3 RTP differences that were identified in the September 27, 1991

-submittal.

II . . UNITS 1 AND 3 RTP DIFFERENCES

.As stated-in TVA's September 27, 1991 RTP submittal, Units 1 and 3 will use the experience gained in the Unit 2 RTP to effect program improvements, to eliminate previous problem areas and to realize program efficiencies. For

those systems that. support safe shutdown, an assessment of the Unit 2 System Test Specifications (STS), test procedures,'and test results will be performed. The results of that assessment, in conjunction with the Unit 3 baseline test requirements, will be the initiating basis for the Unit 3 STS.

The differences between the STS for Units 1 & 3 and the STS for Unit 2 will be the result of differences identified in baseline test requirements, other engineering specified test requirements, system modifications, and system maintenance. The restart tests will be performed as an integrated part of the Startup Test Programs for Units 1 and 3, which will include post modification, post maintenance, restart, and surveillance tests.

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ENCLOSURE 1 Page 2 of 31 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS While there are differences between DFN testing criteria and RG 1.68 testing criteria, there are no differences between the comparison of BFN testing criteria to RG 1.68 testing criteria provided in Part III, below and the similar comparison performed for DFN Unit 2 that was provided to NRC Staff during the June 21, 1988 meeting discussed in Part I, above. There are differences in the comparison of teste performed. These are categorized as follows:

e Testing performed during the Unit 2 RTP that fully satisfy requirements for Units 2 and 3 do not require reperformance.

e Additional testing may be required due to analysis of shared system modes (i.e., functions) between Unita 2 and 3, and between Unita 1, 2, and 3, in comparison to Unit 2 system modes alone.

Additional testing may be required due to the addition of new system modes.

-The extent of additional testing remains to be determined. The evaluation of these system modo differences will be made during the generation of the Bapoline Test Requirements Documents (DTRD's). -A detailed comparison of the BFN Unita 1 and 3 RTP Criteria to RG 1.68 Critoria is provided in Part III below. Correlation of actual Unit 2 RTP tests to planned Unit 3 RTP tests is provided in Table 1 of this Enclosure. The RTP tests for Unit 3 are correlated to system modes (functions tested), as was done for the Unit 2 RTP.

Since the Unit 3 RTP takes into consideration. Unit 2/3 shared system interactions, there are system modes listed in Tab'.e 1 that were not identified during the Unit 2 RTP. The listing addresses only RTP tests.

'Therefore, surveillance tests that are performed to desinstrate operability, as required by Technical Specifications, are not identified. The RTP will take credit for some surveillance tests as satisfying RTP test requirements, however, these will be given an RTP test number.

III. Coeparison of BFN Criteria to RG 1.68 Criteria Regulatory Guide 1.68,' Revision 2, describes the general scope-and depth of initial test programs acceptable to the NRC staff for light-water-cooled nuclear power plants. -Section A, below, explains and justifies'the-differences between the criteria used by BFN to identify system functions that require testing and the criteria in Section c.1 of the Regulatory Guide.

Section B compares the types of testing identified in the Regulatory cuide and the types of testing that will be required at BFN Units 1 and 3 prior to restart. This information demonstrates that the BFN generation of testing requiremente for Units 1 and 3 will cover all required aspects of system performance necessary to confirm functional configuration and, in combination with calculation / analyses will ensure adequate performance.to mitigate FSAR Chapter 14 events as discussed in the DFN Nuclear Performance Plan, Volume 3.

A. ' selection of Plant Features to be tested l- Section C.1 of RG 1.68, Revision 2, provides criteria for selection of

plant features to be tested. The criteria in subsections b, d, and e of

FNCLOSURE1 Page 3 of 31 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS this section are consistent with BFN's approach for identifying the equipment necessary to support safe shutdown of BFN in the Design Baseline and Verification Program (DBVP). The DBVP evaluations for Units 1 and 3 will identify the equipment necessary to support safe shutdown for FSAR Chapter 14 design basis events. The Unit 2/3 Safe Shutdown Analysis (SSA), which is being documented as a BFN calculation, was performed to ensure that systems and portions of systems used to mitigato design basis events were identified. This included a thorough review of BFN event analyses and licensing commitments. (The NRC reviewed a similar calculation during an NRC audit of the DBVP for Unit 2.ead found it to be acceptable). The Unit 3 RTP takes into ce nsideration Unit 2/3 shared system interactions. Multi-unit funed onality will be determined by test or analysis.

The following bases are used for not including features supporting criteria in C.1.a, c, and f of the Regulatory Guide.

e Criteria C.1.a - Those plant structures, systems, and components that will be used for shutdown and cooldown of the reactor under normal plant conditions and for maintaining the reactor in a safe condition for an extended shutdown period.

Basie - Operation of equipment associated with normal plant cooldown will be excluded from consideration by identifying their mechanical / electrical interfaces with the systems required for shutdown from transients, accidents, and special events and ensure the " normal plant cooldown" systems' failure would not prevent the capability of achieving safe shutdown. In addition, the capability of these systems to perform their normal functions were tested during initial plant startup, have been demonstrated during plant operations since then, anJ may be demonstrated in support of other plant testing. This provides assurance that safe shutdown can be accomplished during normal plant conditions.

. Criteria C.l.c - Those plant structures, systems, and components that will be used for establishing conformance with safety limits or limiting conditions for operation that will be included in the facility technical specifications.

Basis - BFN already has Technical Specifications / Surveillance Instructions. The requirements established by Technical Specifications are maintained independent of the test program.

Therefore, this criterion is not-applicable to the RTP.

. Criteria C.1.f - Those plant structures, systems, and components that will_be used to process, store, control, or limit the release of radioactive material.

Basie - This criteria will be applied to those systems required to support safe shutdown from transients, accidents, and special events. For other systems that this criteria may apply to, refer to the basis given for Criteria C.l.a, above.

ENCLOSURE 1 Page 4 of 31 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS B. Comparison of RG 1.68 Appendix A Recommended Testing to BFN Development of Test RG 1,68, Appendix A provideo details concerning the type of testing performed during preoperational testing programs. The following compares Appendix A recommended testing to the BFN development of tests. The headings correspond to Appendix A of RG 1.68, some testing racommended by Appendix A is testing that is only performed prior to initial plant operation and does not apply to the BFN RTP.

Appendix A - Spidp d This section states that testing should include, as appropriate, manual operation, automatic operation, operation in alternate or secondr.ry modes of control, and operation and verification tests to demonstrate expected operation following loss of power sources and degraded modes for which the-systems are designed to remain operational. For the scope of testing defined for BFN in Part III.A above, these operational conditions are being considered and testing is being identified to demonstrate the capability to operate under these conditions, as appropriate, with exception of testing in the degraded mode. This exception is taken based on calculations that demonstrate the systems or portions of systems necessary to mitigate / provide for safe shutdown can perform their required functions during worst case conditions for the events / scenarios th6t they are required for.

This section also states that testing should include, as appropriate, proper function of instrumentation and controls, permissive and prohibit interlocks, equipment protectivo devices

.whose malfunction or premature actuation may shut down or defeat the operation of systems or equipment, and system vibration, expansion, and restraint testing. For.the scope of testing defined

! by BFN~above, testing will be considered to address these functions

with the exception of system vibration, expansion, and restraint testing.

Vibration testing has been evaluated based on the testing performed during the BFN preoperational test program, the corrections made at that time, and the vibration problems identified during plant operation and the corrective action taken. Based on past operational performance and normal surveillance of critical pumps, BFN has identified and initiated appropriate action to address system vibration problems. In addition, the Power Ascension Testing (PAT) Program will monitor a list of specific locations (see Enclosure 2).

System expansion and restraint testing is '.arformed primarily during system heatup/startup testing and .herefore, falls outside the scope of the BFN RTP, which is prior to nuclear heatup.

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I ENCLOSURE 1 Page 5 of 31 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS System / component alarm functions will not be verified as part of the RTP since no credit was taken for them for safe shutdown of the plant (some exceptions to this rule may apply on individual I systems). Primary procese variable indications necessary to perform manual actions are considered for testing where necessary.

RG 1.68 Appendix A, Sections 1.a through 1.o address system specific functional test requirements. The following provides TVA's basis for not performing certain types of tests recommended. It addresses only those !

types of testa not considered for inclusion in the BFN RTP for Units 1 ]

and 3. i

Appendix A. Section 1.a - h9 actor Coolant Svetem (RCS) I Integrated System Test - This test is focused on expansion and restraint test. During initial startup at BFN, testing was )

performed in this area, as described in FSAR Section 13.5.2. Since only minor modifications were made to this system, no additional '

testing is required.

Reactor Vessel Internals - During initial testing at BFN the reactor vessel internals were tested as described in FSAR Section 13.5.2. Due to their passive role in performing their function and the minor modifications performed, no additional testing is.

specified for this portion of RCS.

Both of tnese types of tests were. performed during the power ascension part of the initial BFN preoperational test program.

e Annendix A, Section 1.b - ReactivhyJ.QAtISUyptems Testing identified in this rection was considered for inclusion in tne RTP, as described in 7ppendix A, Section 1 above.

e Appendix A. Section 1.t - Reactor Protection Svetem (RPS) and Encineered Safety Feiture fESF) Systems Response Time Testing - The integrated response time of the reactor protection n/ stem (RPS), in combination with its input sensors, will be verified by analyses / calculations. This is based upon reviews of the actual configuration and the maximum allowable time to perform the integrated function.

Calibration - Credit is taken for the plant instrument calibration program for ensuring the accuracy of the process sensing instrumentation inputs. Calibration frequency and methodology is considered in instrumentation calculations.

e Annendix A. Section 1.d - Residual or Decay Heat Removal Systems Testing identified in this section was considered for inclusion in the RTP, as described in Appendix A, Section 1 above.

ENCLOSURE 1 Pag 6 6 of 31 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS

Accendix A. Section 1.e - Power Conversion Systems This section is outside the scope of testing per the criteria review in Part III.A above.

e Accendix A. Section 1.f - Waste Heat Reiection Systems This section is outside the scope of testing per the criteria review in Part III.A above.

e Accendix A. Section 1.0 - Electrical Svetems Protective Devices / Trip Devices - Current and differential actuated tripping devices were not verified by test. They were verified based upon calculations / analyses being performed to show adequate fault protection and a coordinated protection scheme for the standby AC and DC power distribution system as appropriate.

Load Carrying Capability of Equipment - The capability of the transformers, motor control centers, switchgear, etc. will be verified by calculation / analyses and procurement data. Therefore, no testing was required in this area except for the diesel generators.

, Degraded (minimum / maximum) Voltage Conditions - Only the protective transfers to onsite sources is verified by testing. Capability of loads to perform during degraded voltage conditions prior to transfers will be verified by analyses / calculations.

Design Loading of Battery - A system integrated test to verify the battery loading is not required as part of the RTP. Calculations are provided to ensure the battery capacity is sufficient to perform its design function.

Emergency Lighting - No baseline testing will be required on this system.

Accendix A, Section 1.h - Encineered Safety Features Expansion / Restraint Tests - See the discussion for Appendix A, Section 1, above.

e Annendix A. Section 1.1 - Primary and Secondary C2Dtainments Containment Design Overpressure Structural Test - This testing was performed during the initial plant construction part for code compliance as described in FSAR Section 5.2. This is a test that is ordinarily performed once to demonstrate code compliance.

Secondary Containmen* Individual Valve Leakage Test - An integrated test was performed on Socondary Containment during the Unit 2 RTP to verify adequate subaumospheric pressures could be maintained l

i ENCLOSURE 1 Page 7 of 31 DROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 RESTART TEST PROGRAMS with the required equipment since secondary containment is common to all three units. No further baseline testing is considered necessary.

Anoendix A. Section 1.1 - Instrumentation and Control Systgag Testing identified in this section was considered for inclusion in the RTP, as described in Appendix A, Section 1 above.

Annendix A. Section 1.k - RadiatL2n Prgtection Systemg Personnel'Honitors and Radiation Equipment Survey Testing - This sectic.1 is outside the scope of testing per the criteria review in Part III.A above.

Laboratory Equipment - This section is outside the scope of testing per the criteria review in Part III.A above, e Appendix A. Section 1.1 - Radioactive Waste Handlina and Storace Systems This section is outside the scope of testing per the criteria review in Part III.A above.

Appendix A. Section 1.m - Fuel Storaao and Handlina Svetems operability and Leak Testing for gaskets and bellows - This function is not required based on an adequate quantity of water being available were the pool to drain to the level of the gaskets and bellows.

Dynamic and Static Load Testing - Normal surveillance testing is performed on the fuel handling equipment prior to normal usage.

Worst case failures of this equipment is bounded by the fuel handling accident. Equipment required to mitigate tnis event is included and-appropriate testing performed.

Aeoendix A. Section 1.n - Auxiliary and Miscellaneous Systeta control Room Habitability - Pressurization, isolation, and flow rates were verified during the Unit 2 RTP. Leak tightness is verified as necessary for pressurization during the pressurization test. Concerns with leak tightness on control room personnai exposure are addressed by calculations. Testing required as a result of any future modifications to this system would be addressed by the plant modification test program.

e Annendix A. Sectior 1.o - Reactor Components Handlina System See response to Section 1.m above.

' ENCLOSURE 1 Page 8 of 31 TABLE 1 -

CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMMENTS MODE - FOR V3 001 01 PROVIDE MAIN TURBINE STOP VALVES < 90% OPEN TRIP 2 BFN-RTP-OO1 YES SIGNAL TO REACTOR PROTECTION SYSTEM (99) 001 02 PROVIDE MAIN STEAMLINE ISOLATION VALVE < B0% OPEN 2-DFN RTP-OO1 YES TRIP SIGN AL TO REAC10R PROTECTION SYSTEM (99) 001 03 CLOSE MAIN STEAMLINE ISOLATION VALVES ON PRIMARY 2-BFN RTP OO1 YES CONTAINMENT SYSTEM (64) GROUP 1 ISOLATION SIGN AL.

001-04 CLOSE MAIN STEAM DRAIN LINE VALVES ON PRIMARY 2-BFN RTP-OO1 YES CONTAINMENT SYSTEM (64) GROUP 1 ISOLATION Sl(tNAL 001 06- OPEN MAIN TURBINE STEAM BYPASS VALVES ON TURBINE 2 BFN RTP-047 YES CONTROL SYSTEM (47) TURBlNE TRIP SIGNAL 001 06 CONTROLLED MANUAL DEPRESSURIZATION OF RPV BY 2 BFN-RTP OO1 YES OPENING ADS SAFETY RELIEF VALVES (SRVSL 001-07 OPEN S AFETY RELIEF VALVES (SRVS) ON HIGH REACTOR - 2 BFN RTP OO1 YES PRESSURE TO PROVIDE RPV PRtsCURE REllEF.

001-00 AUTO OPENING OF ADS SRVS UPON COINCIDENT SIGNALS OF 2 BFN-RTP OO1 YES 2 CS PUMPS (76) OR 1 RHR PUMP (74) AND EITHER LWL (Li&L3 FROM SYS 03) HIGH DW PRESSURE (SYS 64) AND TIME DELAY OR LWL (L1 FROM SYS 03) AND HIGH DW PRESSURE BYPASS TIME DELAY.

001 09 CLOSE MAIN TURBINE STOP VALVES UPON TURBINE CONTROL 2 BFN RTP 047 YES SYSTEM (47) DIVERSION OF HYDR AULIC PRESSURE DUE TO LOW CONDENSER VACUUM SIGNAL (SYSTEM 47).

001 10 MAIN STEAMLINE FLOW RESTRICTORS PASSIVELY LIMIT THE . NONE NO JUSTlFICATION BY MAS 3 FLOW RATE OF COOLANT BEING EJECTED FOLLOWING ' ENGINEERING THE LINE BRE AK UNTIL MSIV CLOSURE OCCURS. ANALYSIS 001 11- MANUALLY DEACTIVATE NON-ADS SRVS AND MSIV TEST 2-BFN-RTP OO) 'YES CIRCUITS TO PREVENT INADVERTENT RPV DEPRESSURIZATION AND LOSS OF COOLANT.

001 12 PROVIDE LOW PRESSURE SIGNAL (IN MAIN STEAM LINE AT 2 BFN RTP OO1 YES

~

TURBINE) TO PRIMARY CONTAINMENT SYSTEM (64) GROUP 1 ISOLATION LOGIC (RUN MODE).

001 13 PROVIDE REACTOR COOLANT PRESSURE BOUNDARY (RCPB). 2-BFN RTP 068 YES 001 14 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2 BFN-RTP-064 A YES 001 15 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN-RTP-065 NO SECONDARY CONTAIN-

, MENT WAS TESTED AS

! A WHOLE DURING UNIT 2 TESTING

i. 001 CLOSE FEEDWATER PUMP TURBINE STOP VALVES (TO TRIP 2 BFN RTP-OO3B YES FEEDWATER TURBINE) ON LOSS OF HYDRAULIC PRESSURE DUE j TO ENERGIZATION OF FEEDWATER SYSTEM SOLENOID.

TBD TO BE DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT L , _. , ._ _ _ _ _

. ENCLOSURE 1 - .Pye 9 of 11 TABLE 1-CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT l UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMMENT S MODE FOR U3 00147 PROVIDE MAIN STEAM LINE HIGH F60W AND HIGH STEAM 2-BFN-RTP-OO1 YES TUNNEL TEMPER ATURS SIGN ALS TO PRIMARY CONT AINMENT SYSTEM (64) GROUP 1 ISOLATION 001 18 PROVIDE STEAM FOR HPCI (73) TURBINE. NONE NO PASSIVE COMPONENT I FUNCTION I 001 19 PROVIDE STEAM FOR RCIC (71) TURBINE. NONE NO PASSIVE COMPONENT l FUNCTION j 001 21 PROVIDE MAIN TURD lNE STOP VALVE CLOSURE POSITION 2 BFN RTP-047 YES SIGNALS TO TURDINE CONTROL SYSTEM (47) WHICH INITIATES OPiN1NG OF MAIN TURBINE DYPASS VALVES.

001 22 CLOSE MAIN TURBINE BYPASS VALVES UPON TURBINE 2 BFN-RTP-OOI YES CONTROL SYSTEM (47) DIVERSION OF HYDR AULIC PRESSURE DUE TO LOW CONDENSER VACUUM SIGN AL (SYSTEM 47).

001 23 PROVIDE >30% TURBINE FIRST STAGE PRESSURE INTERLOCK 2-BFN RTP-OO1 YES SIGN AL TO REACTOR PROTECTION SYSTEM (99) Fall SAFE -)

LOGIC. .j I

001 24 PROVIDE MAIN STEAM LINE PRESSURE SIGNALTO TURBINE 2 BFN-RTP-OO1 YES CONTROL SYSTEM (47) FOR OPERATION OF MAIN TURBINE BYPASS VALVES.

001 25 MANUALLY CLOSE MAIN STEAMLINE ISOLATION VALVES 2 BFN RTP OO1 YES-(MSIVS) AND MAIN STEAM DRAIN LINE VALVES.

002 02 PROVIOE NORMALLY OPEN WATER SUPPLY TO RCIC SYSTEM NONE NO PASSIVE COMPONENT (71) PUMP.- FUNCTION.

.002-05 PROVIDE NORMALLY OPEN WATER SUPPLY TO RHR SYSTEM NONE NO - PASSIVE COMPONENT (74) PIPING FLOW PATH WHICH CONTINUES TO HPCI SYSTEM . FUNCTION .

PIPING UPSTREAM OF HPCI SYSTEM PUMP.

002 06 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN-RTP-065 NO- SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 002 08 PROVIDE PRIMARY CONTAINMENT BOUNDARY, 2 BFN RTP 064A YES 002-09 CONFIRM AS-BUILT DESIGN MEETS 100 DEG. F ASSUMPTION NONE NO JUSTIFICATION BY FOR MAXIMUM FEEDWATER TEMPERATURE DROP THAT CAN ENGINEERING OCCUR FOR ANY SINGLE ACTION OR FAILURE OF FW ANALYSIS HEATER (S), ALSO SEE EXTR ACTION STM (05) AND FW (03)

SYSTEMS.

003-01 PROVIDE HIGH REACTOR VESSEL PRESSURE TRIP SIGNAL TO 2-BFN RTP.OO3 A YES RE ACTOR PROTECTION SYSTEM (99) FAIL S AFE LOGIC.

003-02 PROVIDE RPV LOW WATER LEVEL (L3) TRIP SIGNAL TO 2-BFN RTP OO3 A YES i REACTOR PROTECTION SYSTEM (99).

.TBD TO BE DETERM:NED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

l u

.'. . ENCLOSURE 1 Pag 10 of 31 TABLE 1 CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMMENTS MODE FOR U3 003 03 PROVIDE RPV LOW WATER LEVEL (L2) SIGNAL TO HPCI 2 BFN-RTP-003 A YES SYSTEM (73).

003 04 PROVIDE RPV LOW WATER LEVEL (L3) PERMISSIVE SIGNAL TO 2-BFN RTP-003 A YES MAIN STEAM SYSTEM (01) FOR ADS.

003 05 PROVIDE RPV HIGH WATER LEVEL (LB) SIGNAL TO RCIC 2-BFN-RTP-003 A YES SYSTEM (71) AND/OR HPCI SYSTEM (73) FOR TURBINE TRIP.

003 06- PROVIDE HIGH REACTOR VESSEL PRESSURE SIGNAL TO 2SO 2-BFN RTP-003 A YES ATWS WAS TESTED BY VDC SYSTEM (S73) TO OPEN RECIRCULATION PUMP M/G SET 2 BFN RTP-068 - RTP 068 DRIVE MOTOR BREAKERS FOR TRIP OF RECIRCULATION PUMPS AND TO CRO SYSTEM (85) TO INITIATE ALTERNATE ROD INSERT.

003-07 PROVIDE LOW CONDENSER VACUUM SIGNAL TO ENERGt2E 2-BFN RTP 003B YES REACTOR FEEDWATER SYSTEM (3) SOLENDID TO CLOSE MAIN STEAM SYSTEM (1) FEEDWATER TURBINE STEAM SUPPLY STOP VALVES. ,

003 08 PROVIDE RPV WATER LEVEL INDICATION AT BACKUP CONTROL 2-BFN-RTP-003 A YES CENTER.

003 09 PROVIDE LOW REACTOR PRESSURE PERMISSIVE $1GNALS TO 2-BFN-RTP-003 A YES.

CORE SPR AY SYSTEM (75) FOR OPENING OF LOW PRESSURE ECCS INJECTION VALVES AND TO RHR SYSTEM (74) FOR CLOSING OF RECIRCULATION PUMP DISCHARGE VALVES.

003 10 PROVIDE RPV LOW WATER LEVEL (L1) SIGNAL TO PRIMARY . 2 BFN RTP 003A YES CONTAINMENT SYSTEM (64) GROUP 1 ISOLATION LOGIC.

003 PROVIDE RPV HIGH WATER LEVEL (L8) SIGN AL TO FEEDWATER 2-BFN RTP-003 A YES CONTROL SYSTEM (46) FOR MAIN TURBINE AND REACTOR i- FEEDWATER PUMP TURBINE TRIPS.

003 12 PROVIDE REACTOR CONTAINMENT PRESSURE BOUNDARY 2 BFN-RTP-068 YES (RCPB).

003 13 PROVIDE PRIMARY CCNTAINMENT BOUNDARY. 2 BFN RTP-064A YES-003 14 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN RTP 065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING

003-15 . PROVIDE RPV LOW WATER LEVEL (L2) SIGNAL TO 250 SYSTEM 2 BFN RTP-068 YES (573) TO OPEN RECIRCULATION PUMP M/G SET DRIVE MOTOR BREAKERS FOR TRIP OF RECIRCULATION PUMPS AND TO CRD SYSTEM (BS) TO INITIATE ALTERNATE ROD INSERT.

'003-16' PROVIDE RPV PRESSURE INDICATION IN MAIN CONTROL ROOM. 2-BFN RTP 003A YES 003-17. PROVIDE PATH FOR RCIC SYSTEM (71) AND/OR HPCI SYSTEM NONE NO PASSIVE FUNCTION (73) FLOW TO THE RPV THROUGH THE FEEDWATER SPARGERS. VERIFIED BY WA LKDOWN TBD TO BE DETERMINED BTPD BASELINE TEST REQUIREMENTS DOCUMENT

a e ENCLOSURE 1 Pegs 11 of 31 TABLE 1 CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYS1EM - MODE DESCRIPTION UNIT 2 TEST TEST COMMENTS MODE FOR U3 003-18 CLOSE MAIN STEAM SY6 TEM (1) FEEDWATER TURBINE STEAM 2 BFN RTP OO3A YES SUPPLY STOP VALVES ON LOW CONDENSER VACUUM OR RPV '

HIGH WATER LEVEL (L8) SIGN ALS.

003 19 INDICATE RPV WATER LEVEL IN THE CONTROL ROOM. 2 BFN-RTP OO3 A YES .

003 20 PROVIDE RPV LOW WATER LEVEL (L2) SIGNAL VIA RHR 2-BFN RTP-OO3 A YES SYSTEM (74) FOR AUTOMATIC RCIC SYSTEM (71) INITIATION.

003 21 PROVIDE RPV LOW WATER LEVEL (L1) SIGNAL TO MAIN STE AM 2 BFN RTP OO3A YES SYSTEM (1) FOR ADS.

003 22 PROVIDE RPV LOW WATER LEVEL (L1) SIGNAL TO CORE SPRAY 2 BFN RTP OO3A YES SYSTEM (75) FOR CORE SPRAY RHR LPCI(74) AND DIESEL GENERATOR (82) START.

003-23 PROVIDE RPV PRESSURE INDICATION AT BACKUP CONTROL 2 BFN RTF 003 A YES CENTER.

003-24 CONFIRM AS BJ1LT DESIGN MEETS 100 DEG F ASSUMPTION NONE NO JUSTIFICATION BY '

FOR MAXIMUM FEEDWATER TEMPERATURE DROP THAT CAN ENGINEERING OCCUR FOR ANY SINGLE ACTION OR FAILURE OF FW ANALYSIS HEATER (S). ALSO SEE CONDENSATE (02) AND EXTRACTION STM (05) SYSTEMS.

003 25 THE UPPER LIMIT ON FEEDWATER FLOW MUST RESTRICT FLOW NONt TBD UNKNOWN AT THIS TO ABOUT 130% (RELOAD ANALYSIS ASSUMPTION). TIME. PENDING BTRD ISSUANCE 003 26 PROVIDE UNITS 2/3 CORE COVERAGE PERMISSIVE SIGNAL TO NONE TBD UNKNOWN AT THIS RHR SYSTEM (74) FOR CONTAINMENT COOLING (DRYWELL TIME. PENDING BTRD SPRAY, TORUS SPRAY OR POOL COOLING) MODE. ISSUANCE 005-01 CONFlRM AS-BUILT DESIGN MEETS 100 DEG. F ASSUMPTION NONE NO- JUSTIFICATION BY FOR MAXIMUM FEEDWATER TEMPERATURE DROP THAT C AN ENGINEERING OCCUR FOR ANY SINGLE ACTION OR FAILURE OF FW ANALYSIS HEATER (S). ALSO SEE FW (03) AND CONDENSATE (02)

SYSTEMS.

010-01 PROVIDE PATH FOR MAIN STEAM SYSTEM (1) SRVS STEAM 2 BFN-RTP OO3 A YES TEST VACUUM SLOWDOWN TO PRIMARY CONTAINMENT SYSTEM (64) BREAKER VALVES ONLY.

SUPPRESSION POOL.

010-02 PROVIDE REACTOR COOLANT PRESSURE BOUNDARY (RCPB). 2 BFN-RTP 068 YES 012-01 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2-BFN-RTP-065 NO SECONDAR / CONTAIN-MENT WAS TESTED AS A WHOLE DURNG UNIT 2 TESTING 012-02 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2-BFN RTP-064A YES TBD TO DE DETERM!NED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

-*- G- ENCLOSURE 1 Np 12 of 31 TABLE 1-CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT-UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TFST COMMENTS MODE FOR U3 018 01 PROVIDE DIESEL FUEL Oil TO DIESEL GENERATOR SYSTEM 2-BFN RTP-082 NO UNIT 2 TEST (82). ADEQUATELY SATISFIED REQUIREMENT 018-02 MAINTAIN 7 D AY (LONG TERM) SUPPLY OF FUEL OIL IN 2-DFN RTP 082 NO . UNIT 2 TEST STORAGE TANKS IN SUPPORT OF DIESEL GENERATOR SYSTEM ADEQUATELY (82). S ATISFIED REQUIREMENT -

018-03 MAINTAIN SHORT TERM SUPPLY OF FUEL OIL IN 7 DAY NONE TBD PENDING ISSUANCE OF

STORAGE TANKS BY TRANSFERRING FUEL OtL BETWEEN BTRD AVAILABLE FUEL OIL STOR AGE TANKS IN SUPPORT OF DIESEL GENLRATOR SYSTEM (82).

023 01 PROVIDE COOLING WATER TO RHR SYSTEM (74) HE AT 2-BFN RTP 023 YES EXCHANGERS.

023 03 PROVIDE COOLING WATER TO EECW SYSTEM (071 UPON 2-BFN-RTP 023 YES START OF THE RHRSW PUMPS.

023 04 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2-BFN RTP 06S NO SECONDARV CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING !

023 06i PROVIDE WHEELER LAKE LEVEL INDICATIONIALARM AT 2 BFN RTP 023 NO UNIT 2 TEST S ATISFIES ELEVATION $$8 FEET AND RISING AS WELL AS AT ELEVATION REQUIREMENTS S64 FEET AND RISING.

023 07 PROVIDE PRIMARY CONTAINMENT BOUNDARY. CLOSE RHR 2-BFN RTP 023 NO UNIT 2 TEST SATISFIES SERVICE WATER SYSTEM VALVES 1 FSV 23 SS AND REQUIREMENTS 2-FSV 23 $6 ON RHR SYSTEM (74) UNIT CROSS: TIE VALVES OPEN POSITION SIGNAL.

023 08- MANUAL RHRSW SYSTEM OPERATION TO PROVIDE COOLING 2 BFN RTP 023 YES WATER TO RHR SYSTEM (74) HEAT EXCHANGERS FROM OUTSIDE OF MAIN CONTROL ROOM.

023-09 PROVIDE SUMP PUMP CAPABILITY 2 DFN RTP 023 NO UNIT 2 TEST S ATISFif S REOUIREMENTS 024 01- PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2-DFN-RTP 065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 024 02 PROVIDE PRESSURE BOUNDARY INTEGRITY TO EECW SYSTEM 2 BFN-RTP 024 NO UNIT 2 TEST S ATISFIES (67). REQUIREMENTS 024-OS PROVIDE RCW SUPPLY HEADER LOW PRESSURE PERMISSIVE 2-BFN RTP 024 NO UNIT 2 TEST S ATISFIES SIGNAL TO 4-KV POWER DISTRIBUTION SYSTEM (S75) FOR REOUlREMENTS STARTING OF RHRSW SYSTEM (23) PUMPS.

TBD TO BE DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT i

- - _ , _ . _~

. ~, -_. .,- . ,

i ENCLOSURE 1 Pegs 13 of 31 TABLE 1 !

l CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST 1EST COMMENTS MODE FOR U3

'024 06 PROVIDE MANUAL CONTROL FROM OUTSIDE THE MAIN 2-DFN R1P BUC YES CONTROL ROOM OF RCW PUMPS 10 AND 3D 10 PREVENT .

OVEr4 LOADING OF DIESEL GENERATORS (SYSTEM 82).

025 01- PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFNr 365 NO SECONDARY CONTAIN-MENT WAS TESTED AS .

A WHOLE DURING UNIT >

2 TESTING O25 03 PROVIDE PRESSURE BOUNDARY INTEGRITY TO RCW SYSTEM NONE NO JUSTlFICATION BY (24)IN SUPPORT OF EECW SYSTEM (67) PRESSURE ENGINEERING BOUNDARY. ANALYSIS O2S-04 PREVENT AUTOMATIC START OF HIGH PRESSURE FIRE 2-BFN RTP-025 40 UNIT 2 TEST S ATISFIES PROTECTION SYSTEM (26) PUMPS { LOCK-OUT) TO PREVENT REQUIREMENTS OVERLOADING THE DIESEL GENERATORS (SYSTEM 82).

025 05 PREVENT START OF HIGH PRESSURE FIRE PROTECTION 2 BFN-RTP BUC NO UNIT 2 TEST SATISFIES SYSTEM (26) PUMPS (FROM OUTSIDE THE MAIN CONTROL - REQUIREMENTS ROOM) TO PREVENT OVERLOADING THE DIESEL GENERATORS (SYSTEM 82).

026-01 SUPPORT SECONDARY CONTAINMENT FUNCTION. 2-BFN RTP-065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING O27-01 PROVIDE WARM WATER CHANNEL LEVEL INDICATION IN THE ' - 2 BFN-RTP-027 NO UNIT 2 TEST SATISFIES MAIN CONTROL ROOM. REQU3FMENTS O27 02 PROVIDE FOREBAY LEVEL INDICATION IN THE MAIN CONTROL 2 BFN RTP 027 NO UNIT 2 TEST SATISFIES ROOM FOR MANUAL ACTIONS TO REDUCE POWER OR IF REQUIREMENTS NECESS ARY INITIATE SCRAM.

027 03 PROVIDE COOLING TOWER LIFT PUMP DISCHARGE WATER NONE TBD UNKNOWN, PENDING HIGH TEMPERATURE SIGNAL TO 4 KV POWER DISTRIBUTION BTRD ISSUANCE SYSTEM (575) FOR TRIPPING OF THE CORRESPOtJnNG COOLING TOWER LIFT PUMP.

027 04 PROVIDE MANUAL VACUUM BREAKING CAPABILITY TO ^ 2-BFN RTP-027 YES PREVENT BACKFLOW OF COOLING TOWER WARM WATER DISCHARGE INTO THE FOREBAY UPON TRIP OF THE CCW PUMPS.

027 OS ' PROVIDE FOREBAY / WARM WATER CHANNEL DIFFERENTIAL 2-BFN RTP-027 NO UNIT 2 TEST S ATISFIES -

LEVEL INDICATION IN THE MAIN CONTROL ROOM. REQUIREMENTS O29 PROVIDE SECONDARY CONT AINMENT BOUNDARY. 2.BFN-RTP-065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 030-01 PROVIDE VENTILATION TO UNITS 1 AND 2 DIESEL GENERATOR 2 BFN RTP 030 NO UNIT 2 TEST S ATISFIES BUILDING. REQUIREMENTS TBD TO BE DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

h i

,;. . . ENCLOSURE 1 Proe 14 of 31 q

TABLE 1 CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST . COMMENTS MODE FOR U3 030 02 PROVIDE VENTILATION TO UNIT 3 OlESEL GENERATOR 2 BFN RTP-030 NO UNIT 2 TEST JATISFIES BUILDING. REQUIREMENTS - ,

030-03 PROVIDE VENTILATION TO 250V BATTERY ROOM 3EB IN THE 2-BFN RTP 030 NO UNIT 2 TEST S ATISFIES UNIT 3 DIESEL GENERATOR BUILDING TO PREVENT A BUILDUP REQUIREMENTS OF HYDROGEN GAS DURING BATTERY CHARGING.

031-01 ISOLATE SUPPLY DUCTS AND SUPPLY PRESSURIZED FILTERED 2-BFN RTP-031B NO UNIT 2 TEST SATISFIES OUTDOOR AIR TO MAIN CONTROL ROOM ON PRIMARY REQUIREMENTS CONTAINMENT SYSTEM (64) GROUP 6 ISOLATION SIGNAL OR RADIATION MONITORING SYSTEM (90) INITIATION SIGNAL.

031-02 PROVIDE VENTILATION TO REACTOR BUILDING BOARD ROOMS 2 BFN RTP 031B NO UNIT 2 TEST SATISFIES AND CONTROL BAY MECHe.NICAL EQUIPMENT ROOMS. REQUIREMENTS 031-03 PROVIDE RECIRCULATION AIR COOLING TO REACTOR BUILDING 2 BFN-RTP 031B NO UNIT 2 TEST S ATISFIES BOARD ROOMS, REQUlf.EMENTS 031 04 PROVIDE VENTILATION AND AIR CONDITIONING TO UNIT 3 2-BFN RTP 031B NO UNIT 2 TEST SATISFIES DIESEL GENERATOR EUILDING BOARD ROOMS. REQUIREMENTS 031 05 PROVIDE RECIRCULAllON AIR CONDITIONING TO CONTROL 2-DFN RTP 031B NO UNIT 2 TEST S ATISFIES ROOMS AND AUXILIARY INSTRUMENT ROOMS. REQUIREMENTS 031 06 PROVIDE VENTILATION TO BATTERY ROOMS. 2 BFN RTP-031B NO UNIT 2 TEST SATISFIES REQUIREMENTS 031 08 PROVIDE MANUAL LINEUP OF HVAC EQUIPMENT WITH TOTAL 2-BFN RTP 031 A NO UNIT 2 TEST S ATISFIES LOSS OF CONTROL AIR. REQUIREMENTS 031-09 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2-BFN-RTP-065 NO- SECONDARY CONTAIN.

MENT WAS TESTED AS A WHOLE DURING UNIT .

2 TESTING 032-01 PERFORM ISOLATION ACTION (S) UPON RECEIVING PRIMARY 2 BFN-RTP 032 YES CONTAINMENT SYSTEM (641 GROUP 6 ISOLATION SIGNALS.

032-02' PROVIDE COMPRESSED AIR TO MAtN STEAM SYSTEM (01) AOS 2 BFN-RTP 001 YES SAFETY RELIEF VALVES (SRVS).

l --

032 03 PROVIDE COMPRESSED AIR FOR CLOSURE OF MAIN STEAM 2-BFN RTP-001 .YES l l SOLATION VALVES (SYSTEM 01).

032 04 ' PROVICE COMPRESSED AIR TO EQUIPMENT ACCESS LOCK 2-BFN RTP 032 NO UNIT 2 TEST S ATISFIES SEALS TO PROVIDE SECONDARY CONTAINMENT BOUNDARY. REQUIREMENTS

l. 032-05 PROVit9 091 MARY CONTAINMENT BOUNDARY. 2 BFN RTP 064A YES l.

032-06 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN RTP 06S NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING t TBD' TO BE DETERMINED L ' BTRO ' BASELINE TEST REQUIREMENTS DOCUMENT v w4e-* - + - e-+.rWe sese m. craw t ir- i 3 -- --

I

.- .- ENCLOSURE 1- Pes: 15 of 31 I TABLE 1 -

CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMMENTS MODE FOR U3 032 07 PROVIDE CONTAINMENT ATMOSPHERIC DILUTION SYSTEM (84) NONE TBD BTRDISSUANCE COMPRESSED G AS (NITROGf Ni TO MAIN STE AM SYSTEM (1) PENDING FOR UNIT 3, ADS S AFETY RELIEF VALVES (SRVS) FOR LONGER TERM JUSTIFIED BY OPERABILITY. ANALYSIS ON UNIT 2 032-08 PROVIDE FLOW PtsTH INTEGRITY FOR SUPPLY OF CONTROL NONE NO PASSIVE COMPONENT NITROGEN TO CAD VENT PATH VALVES VERIFIED BY WALKDOWN 033-01 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2-BFN.RTP-064 A YES 033 02 PROVIDE SECOND ARY CONTAINMENT BOUNDARY. 2 BFN RTP-065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 037-01 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2.BFN RTP-06S NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING -

039 01 INPIBIT SPURIOUS CO2 INITIATION SIGNAL WHEN VENTIL ATION 2 BFN RTP 039 NO UNIT 2 TEST S ATISFIES (SYSTEM 30)IS REQUIREO IN DIESEL GENER ATOR BUILDINGS. REQUIREMENTS

.040 01 PROVIDE SECOND ARY CONTAINMENT BOUNDARY. 2 BFN-RTP 065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 040-02 PROVIDE VALVE CLOSURE OR PIPING GEOMETRY TO PREVENT NONE NO NO BTRD ISSUED FOR WATER WHICH FLOODED THE BASE OF THE STACK STANDBY UNIT 2 - JUSTlFICATION GAS TREATMENT AND OFF GAS BUILDING FROM FLOWING BY ENGINEERING INTO THE RADWASTE BUILDING.- ANALYSIS 043 01 PROVIDE REACTOR COOLANT PRESSURE BOUNDARY (RCPB). 2 BFN RTP-068 YES 043 02 CLOSE SWO SYSTEM ISOLATION VALVES ON PRIMARY 2 BFN RTP-069 YES CONTAINMENT SYSTEM (64) GROUP 1 ISOLATION SICNAL (ONLY ON RPV LOW WATER LEVEL (L1) AND MAIN STE AM LINE HIGH RADIATION].

043 03 PROVIDE PRIMARY CONTAINMENT BOUNDARY, 2 BFN-RTP 064 A YES 04344: MAINTAIN RHRSW SYSTEM (23) PRESSURE BOUNDARY 2-BFN-RTP-023 YES INTEG RITY. -

, 043-05 PROVIDE CAPABILITY OF MANUAL BACKUP CONTROL 2 BFN RTP BUC YES ISOLATION (VALVE CLOSURE) TO PREVENT LOSS OF REACTOR i WATER INVENTORY.

043 06 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2-BFN RTP-06S NO SECOND ARY CONTAIN-MENT WAS TESTED AS j

A WHOLE DURING UNIT l 2 TESTING l

l I

TBD TO BE DETERMINED

('

BTRD- BASELINE TEST REQUIREMENTS DOCUMENT

+- "

-w'i-+ T7 7-tw - -

1 1

_3_ .- ENCLOSURE 1 Pro 16 of 31-TABLE 1 CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS

__ -i SYSTEM MODE DESCRieTlVN UNIT 2 TEST 1EST COMMENTS .

MODE FOR U3 043 07 CLOSE POST ACCIDENT SUPPLY SYSTEMISOLATION VALVES ' NONE YES NEW MODE TESTED BY UPON RECEIVING PRIMARY CONTAINMENT SYSTEM (641 PMT ON UNIT 2 GROUP 6 ISOLATION StGNAL.

044 01 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2-BFN RTP 065 NO SICONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 046-01 PROVIDE RPV HIGH WATER LEVEL (LB) SIGN AL TO ENERGlZE 2 BFN R1P OO3B YES RE ACTOR FEEDWATER SYSTEM (3) SOLENOID TO CLOSE MAIN STEAM SYSTEM (1) FEEDWATER TURBINE STEAM SUPPLY STOP VALVES.

047-01 PROVIDE MAIN TURBINE CONTROL VALVE FAST CLOSURE 2 DFN RTP 047 YES SIGNAL TO REACTOR PROTECTION SYSTEM (99).

047 03 PROVfDE HYORAULIC CLOSURE OF MAIN STEAM SYS (01) 2-BFN RTP 047 YES TURBINE SYOP VALVES UPON LOW CONDENSER VACUUM (APPROXIMATELY 20 HG) SIGNAL.

047 04 PROVIDE HYDRAULIC CLOSURE OF MAIN STEAM SYS (01) MAIN 2 BFN RTP-047 YES TURBINE BYPASS VALVES UPON LOW CONDENSER VACUUM (APPROXIMATELY 7 HG) SIGNAL.

047 OS PROVIDE HYDRAULIC CONTROL TO OPEN MAIN STE AM SYS 2 BFN RT* 047 YES (01) MAIN TURBINE BYPASS VALVES ON TURBINE TRIP (MAIN TURBlNE STOP VALVE CLOSURE) SIGNAL.

047 06 MAXIMUM STEAM FLOW THROUGH TURBINE PLUS BYPASS - NONE TBD THIS WAS A POWER VALVES WITH THE PRESSURE REGUL ATOR FAILED OPEN IS ASCENSION TEST ON 125% PER UFSAR ANALYSIS (CONTROL ROOM MANUALLY UNIT 2. UNIT 3 TEST ,

ADJUSTED LIMIT). UNKNOWN PENDING BTRD ISSU ANCE 0S0 01- PROVIDE PRESSURE BOUNDARY INTEGRITY TO EECW SYSTEM NONE NO PASSIVE FUNCTION (67). VERIFIED BY WALKDOWN OS3701_ PROVIDE SECOND ARY CONTAINMENT BOUNDARY. 2-BFN RTP 065 NO SECONDARY CONTAIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING "O63-01 MANUAL INJECTION OF BORON SOLUTION INTO REACTOR 2-BFN RTP-063 YES GIVEN INDICATION OF INCOMPLETE INSERTION OF CONTROL RODS (CRD SYSTEM 85) AND REACTOR NOT BEING IN SUBCRITICAL CONDITION (NMS SYSTEM 92).

063-02 PROVIDE SLCS INITIATION SIGNAL TO RWCU SYSTEM (69) FOR 2-BFN RTP 069 YES ISOLATION OF RWCU SYSTEM FROM THE REACTOR TO PREVENT ENTRY OF BORON SOLUTION INTO RWCU SYSTEM.

063 03 PROVIDE PEACTOR COOLANT PRESSURE BOUNDARY. 2 BFN RTP 068 YES 063 04 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2 BFN-RTP-064 A YES L

l' .TBD- TO B'E DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

q c <

ENCLOSURE 1 Pep 17 of 31 TABLE-1 CORRELATION DETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION - UNIT 2 TEST TEST COMMENTS MODE FOR U3 064-01 PROVIDE SIGNAL TO OLOSE GROUP 1 PRIMARY CONTAINMINT 2 BFN RTP 064A YES ISOLATION VALVES { MAIN STEAM SYSTEM (1) AND SWO SYSTEM (43)l.

064-02' PROVIDE SIGNAL TO CLOSE GROUP 2 PRIMARY CONT AINMENT 2-BFN RTP 064 A YES ISOLATION VALVES IRHR SYSTEM (74) CORE SPRAY SYSTEM (75) AND RADWASTE SYSTEM (7711.

064 03: PROVIDE SIGNAL TO CLOSE OROUP 3 PRIMARY CONTAINMENT 2 BFN RTP 004A YES ISOLATION VALVES IRWCU SYSTEM (69)l, 064-04 PROVIDE SIGNAL TO CLOSE GROUP 6 PRIMARY CONTAINMENT 2-BFN RTP 064 A YES ISOLATION VALVES (SYSTEMS 3243647684 AND 90) ISOLATE AC SYSTEM (31) SUPPLY DUCTS TO MCR INITIATE EMERGENCY PRESSURIZATION SYSTEM (31) TRIP FANS & POSITION DAMPERS (64) & INITIATE SGTS (6S).

064 OS . PROVIDE SIGNAL TO CLOSE GROUP 8 ISOLATION VALVES. 2 BFN RTP 064A YES SYSTEM (94)IS NOT TO PERFORM ACTIVE ISOLATION FUNCTION.-

06446 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2-BFN RTP-064 A YES 064 08' PROVIDE HIGH DRYWELL PRESSURE TRIP SIGNAL TO REACTOR 2-BFN RTP 064A YES PROTECTION SYSTEM (99).

l '064 09 PROVIDE HIGH DRYWELL PRESSURE SIGNAL TO RHR SYSTEM 2-BFN-RTP 064 A YES (74) FOR LPCIINITIATION LOGIC AND TO CORE SPR AY JYSTEM (76) FOR SYSTEM INITIATION LOGIC 480V LOAD SHED LOGIC DIESEL GENERATOR START LOGIC AND HPCl SYSTEM (73)

INITIATION LOGIC.

064 102 PROVIDE VACUUM RELIEF SYSTEM (VACUUM BREAKER 2-BFN RTP 064A YES VALVES) TO PREVENT DRYWELL OR SUPPRESSION CH AMBER (TORUS) NEG ATIVE PRESSURE FROM DAMAGING -

CONTAINMENT STRUCTURE.

064 11. PROVIDE DRYWELL TEMPERATURE INDICA FlON IN MAIN 2 BFN RTP 064A YES CONTROL ROOM IN SUPPORT OF RHR FYSTEM (741 DRYWELL SPRAY (CONTAINMENT COOLING) MODE.

064-12 PROVIDE GUPPRESSION POOL TEMPERATURE INDICATION IN 2 BFN RTP 064A YES l' MAIN CONTROL ROOM IN SUPPORT OF RHR SYSTEM (74)

CONTAINMENT COOLING (10RUS COOLING AND I DRYWELL/ TORUS SPRAY) MAIN STEAM SYSTEM MANUAL RPV DEPRESSURIZATION AND RPS SYSTEM (99) MANUAL SCRAM, 064 13 PROVIDE SUPPRESSION POOL LEVEL INDICATION IN MAIN 2-BFN-RTP-064 A YES CONTROL ROOM IN SUPPORT OF RHR SYSTEM (74)

CONTAINMENT COOLING AND MAIN STE AM SYSTEM (1)

MANUAL RPV DEPRESSURIZATION. PROVIDE PRESSURE BOUNDARY INTEGRITY TO HPCI SYSTEM (73).

TBD TO BE DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

ENCLOSURE 1 Psge 18 of 31 TABLE 1 CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS -

SYSTEM MODE DESCRIPTION UtHT 2 TEST TEST COMMENTS MODE FOR' U3

., g _

064-14 PROVIDE DRYWELL PRESSURE INDICATION IN MAIN CONTROL 2-BFN RTP 064A YES RDOM IN SUPPORT OF RHR SYSTEM (74) CONTAINMENT COOLING (DRYWELLITORUS SPRAY) AND CONTAINMENT ATMOSPHERE DILUTION SYSTEM (641 CONTAINMENT VENTING AFTER A LOCA.

064 15- PROVIDE ORYWELL TEMPERATURE INDICATION OUTSIDE THE 2 BFN-RTP-064 A YES MAIN CONTROL ROOM IN SUPPORT OF MAIN STEAM SYSTEM (1) MANUAL RPV DEPRESSURIZATION AND RHR SYSTEM (74)

OPERATION FROM OUTSIDE THE MAIN CONTROL ROOM, 064 16 PROV:DE SUPPRESSION POOL TEMPER ATURE INDICATION 2-BFN-RTP-064 A YES OUTSIDE THE MAIN CONTROL ROOM (MCR)IN SUPPORT OF MAIN STEAM SYSTEM (1) MANUAL RPV DEPRESSURIZATION RHR SYSTEM (74) OPERATION AND REACTOR PRO 1ECTION SYSTEM (99) MANU AL SCRAM FROM OUTSIDE THE MCR.

064 17 PROVIDE SUPPRESSION POOL ! EVEL INDICATION OUTSIDE THE 2-BrN RTP 064 A .YES MAIN CONTROL ROOM IN SUdORT OF MAIN STE AM SYSTEM (1) MANUAL RPV DEPRESSURf2ATION RCIC SYSTEM (71)

OPERATION AND RHR SYSTEM (74) OPERATION FROM OUTSIDE THE MAIN CONTROL ROOM, 064 18 PROVIDE DRYWELL PRESSURE INDICATION OUTSIDE THE MAIN 2 BFN RTP-064A YES CONTROL ROOM IN SUPPORT OF RHR SYSTEM OPERATION FROM OUTSIDE THE MAIN CONTROL ROOM, 0641P PROVIDE FLOW (VENT) PATH FOR THE CONTAINMENT NONE TBD UNIT 3 TEST PENDING ATMOSPHERE FROM EITHER THE SUPPRESSION CHAMBER OR BTRD ISSUANCE, MODE DRYWELL TO THE CONTAINMENT ATMOSPHERE O!LUTION WAS REMOVED FROM SYSTEM (84). RESTART SCOPE FOR UNIT 2 064-20 PROVIDE PRESSURE SUPPRESSION BY - NONE TBD UNIT 3 TEST PENDING COOLING / CONDENSATION OF S AFETY RELIEF VALVES (SRVS) BTRD ISSUANCE. TEST STEAM (FROM BOILER DRAINS AND VENTS SYSTEM (101) AND - NOT REQUIRED FOR RCIC SYSTEM (71) AND HPCI SYSTEM (73) TURBINE EXHAUST UNIT 2, STEAM. ACCEPT RCIC & HPCI SYSTEM PUMP MINIMUM BYPASS FLOW.

064 21 PROVIDE SECONDARY CONTAINMENT BOUNDARY, 2-BFN RTP 065 NO SECONDARY CONTAIN-l- MENT WAS TESTED AS

!' A WHOLE DURING UNIT 2 TESTING 064-23 PROVIDE FORCED AIR COOLING FOR RHR SYSTEM (74) AND BFN RTP-030 YES CORE SPRAY GYSTEM (75) PUMP MOTORS, 064-24 PROVIDE WATER SUPPLY TO HPCI SYSTEM (73) CORE SPRAY NONE NO JUSTIFICATION BY SYSTEM (75) AND/OR RHR SYSTEM (74) PUMPS. ENGINEERING ANALYSIS l

064 25 PROVIDE HIGH DRYWELL PRESSURE SIGNAL TO MAIN STEAM 2-BFN-RTP-064 A YES SYSTEM (1) FOR AUTOMATIC DEPRESSURIZATION SYSTEM

( ADS) LOGIC.

TBD- TO BE DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

4 . ENCLOSUREI Pegs 19 of 31 TABLE 1 CORRELATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMMENTS MODE FOR V3 064 26 PROVIDE STRUCTURAL SUPPORT FOR THE CONTROL ROD NONE NO PASSIVE COMPONENTS DRIVE SYSTLM (85) HOUSINGS. JUSTIFIED BY ANALYSIS 064 27 CLOSE PRIMARY CONTAINMENT VENTILATION SYSTEM 2 BFN RTP-065 YES ISOLATION VALVES ON PRIMARY CONTAINMENT SY6 TEM (64)

GROUP 6 ISOLATION SIGNAL.

064-28 PERFORM ISOLATION ACTION (S)(TRIP FANS CLOSURE OF 2 BFN'RTP 064A YES DAMPERS OPENING OF DAMPERS TO SGTS (65)) ON PRIMARY CONTAINMENT SYSTEM (64) CROUP 6 ISOLATION SIGNALS.

065-01 MAINTAIN NEG ATIVE PRESSURE IN SECONDARY 2 BFN-RTP-065 1BD PENDING ISSUANCE OF CONTAINMENT ON PRIMARY CONTAINMENT SYSTEM (64) UNIT 3 BTRD GROUP 6 ISOLATION SIGNAL FILTER A13 BORNE PARTICULATE

& GASES IINCLUDING TH AT FROM HPCI SYSTEM (73) & CAD SYSTEM (34)) PRIOR TO DISCHARGE TO OFF G AS SYSTEM (66).

065-03 MAINTAIN NEG ATIVE PRESSURE IN SECONDARY 2-BFN RTP 065 TBD PENDING ISSUANCE OF CONTAINMENT ON PRIMARY CONTAINMENT SYSTEM (64) UNIT 3 BTRD SIGNAL DUE TO RADIATION MONITORING SYSTEM REFUELING ZONE HIGH RADIATION SIGNAL FILTER AIRBORNE PARTICULATE & GASES PRIOR TO DISCHARGE TO OFF GAS SYSTEM (66).

065 PROVIDE SECONDARY CONTAfMMENT INTEGRITY. 2-BFN RTP 065 NO UNIT 2 TEST SATISFIES REQUIREMENTS 066 02 PROVIDE FLOW PATH INTEGRITY FOR THE RELEASE OF THE 2--BFN-RTP-065 TBD PENDING ISSUANCE OF FILTERED STANDBY G AS TREATMENT SYSTEM (65) G ASES TO UNIT 3 BTRD THE STACKS.

066-03 PROVIDE VALVES OR PIPING GEOMETRY TO SUPPORT NONE TDD PENDING ISSUANCE OF RADWASTE SYSTEM (77) TO PREVENT RADWASTE BUILDING UNIT 3 STRD FLOODING.

066-04 OFF-GAS DILUTION FAN ISOLATION DAMPERS SHALL BE 2-BFN RTP-065 NO UNIT 2 TEST SATISFIES

- CLOSED ON INITIATION OF THE SGT SYSTEM (65). REQUIREMENTS 067 01- PROVIDE COOLING WATER TO AC SYSTEM (31) CHILLERS RHR 2-BFN-RTP-067 YES SYSTEM (74) PUMP SEAL COOLERS CIS (76) 02 & H2 G AS ANALYZERS DIESEL ENGINES (82) RHR & CORE SPRAY EQUIPMENT ROOM COOLERS (64) & FUEL POOL (79). MAINTAIN EECW SYSTEM (23) PRESSURE EOUNDARY.

067 PROVIDE SECONDARY CONfAINMENT BOUNDARY. 2-BFN RTP-065 NO UNIT 2 TEST S ATISFIES HEQUIREMENTS 067-03 FOR SHUTDOWN FROM OUTSIDE OF MAIN CONTROL ROOM: (1) 2 BFN RTP-067 YES PROVIDE COOLING WATER TO RHR & CORE SPRAY EQUIPMENT ROOM COOLERS (64) RHR SYSTEM (74) PUMP SEAL COOLERS DIESEL ENGINES (82) & FUEL POOL (79) (2) MAINTAIN EECW SYSTEM (23) PRESSURE BOUNDARY.

TBD TO BE DETERMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

. , INCLOSURE 1 Pm 20 of 31 TABLE 1 CORREt AliON DE1 WEEN DROWNS ((RRY NUCLEAR PLANT UNITS 2 AND 3 RESTART 1EST PROGRAMS SYSit M MODE DE SCRIPilo UNIT 2 TIST TEST COMMINTS MODE FOR U3 067 04 RHR SERVICE WA1ER PUMPS MUST MAINTAIN CAPAtULIT) FOR NONE TDD PtNDINu ISSUANCE OF SHUTDOWN HE AT RtMOVAL UNDIR TORN ADO DLSIGN UNIT 3 BTRD CONDITIONS (P AS$lVt),

067 0$ ATT ACH A fiRt H0$t 70 t(CW TO MAINT AIN WATER LlVilIN 2 (JN RTP 067 TDD PtNDING ISSU ANC[ OF THE FUtl POOL. UNIT 3 HTRD 068 01 CLO$t Rf CIRCULATION PUMP DISCHARGE VALV!S ON RHR 2 DFN RTP 068 YES SYSitM (74l AUTOMATIC LPCl MODf, INITIATION SIGN AL.

068 02 OPIN Rf CinCUuTION PUMP MOTOR BRtAktR$ ON Rt ACTOR 2 DFN RTP 068 Yi$

PROTECTION SYSitM (99) SIGtJAL DUE TO >30% TURDINE flRST STAGE PRESSURE AND TITHER MAIN TURD NE CONTROL VALVE FAST CLOSURt OR MAIN TUnt.INE STOP VALVtB < 90%

OPIN MANUALLY (NABLtD END Of CYCLE RPT).

068'03 CLOSE RECIRCULATION PUMP DISCHARGE VALVES MANUALLY 2 Of N-RT P-068 Yt$

IN SUPPORT OF MANUALLY INITIATtO RHR SYSitM (74)

SHVf DOWN COOLING MODI AND LPCI M90t (FROM MAIN CONTROL ROOM AND OUTSIDE MAIN CONTROL ROOMI.

068 04 PROVIDE RE ACTOR COOL ANT PRES $URt COUNDARY (RCPBl. 2 DFN RTP 068 YtB 06805 PROVIDE LOW Rt APTOR PRESSURI PERMISSIVE SIGNALS TO 2 Of N RTP 068 YES CORE SPRAY SYSTE' A 175) AND RHR SYST[M (74L 068-06 TRIP RECIRCULATION PUMP MOTOR MOTOR GitJf RATOR StT 2 BFN RTP 068 YES ON OPINING OF 4KV POWER D1%TRIDUTION SYSitM (S75) M G StT DRIVE MOTOR DRt AKERS DUE TO HIGH Rt ACTOR PRES $URE OR LOW WAf tR LEVEL (L2h OSS07 PROVIDE RECIRCULATION FLOW SIGNAL TO THE NEUTRON NONC TBD FitMOVID FROM Utili 2 MONITORING SYST[M (92)IN SUPPORT OF ROD BLOCK REST ART SCOPE MONITOR TRIP SIGtJAL (VARYING WITH RECIRCULATION FLOW)

TO PREVtNT CONTROL ROD WITHDRAWAL, PtNDING ISSUANCE OF UNIT 3 BTRD 068 08 PROVIDE PRIMARY CONT AINMENT BOUNDARY, 2-DF N RTP,064 A YES 068'09 ASSURE THAT MOTOR GENtRATOR IM Gl SET SPt[D CHANG [S NONE TBD REMOVtD TROM UNIT 2 ST AY WITHtN ANALY7tD LIMITS (ACCEL [R ATION AT REST ART SCOPt MAXIMUM RATE OF 25% OF FULL SPtf D PER SECOND MAXtMUM TLOW OF 10S% OF RATED FLOW AT 100% R ATED PENDING ISSUANCE OF POWERL ,

UNIT 3 DTRD 068 10 PLANT TECHNICAL SI(CS AtJD PROCEDURES REQUIRC f40NE NO JUSTIFICATION BY WARMUP OF THE LOOP DEFORE PUMP ST ART, t NGINE tRING ANALYSl$

039 01 PROVIDC PRIMARY CONTAINMENT BOUNDARY. 2 DFN RTP 064 A Yt$

069 02 PROVIDE StCONDARY CONTAINMINT COUNDARY. 2 DIN RTP 06b NO St COND ARY CONT AIN MENT WAS TISTfD AS A WHOlt DURING UNIT 2 TESTING TBD TO DE DETERMINED BTRD BASELINE TEST RtOUIREMENTS DOCUMENT

i

. . ENCLOSURE 1 Peo* 21 *f 31 l i

TADLE 1 ,

t CORRELATION BETWEEN BROWNS FERRY NUCCEAR PLANT UNITS 2 AND 3 HESTART TEST PROGRAMS t

i SYST[M MODF DtSCRIPTION UNIT 2 f tST TEST COMMf NTS MODI FOR US _

069 03 CLOSE RWCU SYSTLM ISOLATION VALVL$ ON PRIMARY 2 Bf N RTP 069 Y!$

CONTAINMENT S YSTEM (641 GROUP 3 ISOL AllON S10N AL.

069 04 CLOSC RWCU SYSTtM SVCTION LINE ISOLATION VALVf B ON 2-Df N RTP 069 Vf8

ST ANDDY LIOUID CONTROL SYSTEM (63) INITIATION SIGNAL TO PREVENT ENTRY OF BORON SOLUTION INTO RWCU f SYSTEM.

06905 PMOVIDE HIGH 6dCU SYSTEM EQUIPMENT ARLA ATMOSPHERC 2 BfN R1P-069 Yts AND DRAIN T[MPERATURE SIGNALS TO PRIMARY CONTAINMENT SYSTEM (64) GROUP 3 ISOL ATION LOGIC 4 069 06 PROVIDE REACTOR COOLANT PRESSURE BOUNDARY (RCPDL 2 BFN RTP 06R YES 069-07 PROVIDE SYSTEM PRtSSURE BOUNDARY SUPPORT tCHECK 2 Bf 4 RTP 069 Yfs VALVf)10 HPCI SYSTEM (73) TO PREVENT DIVERSION OF HPCI SYSitM CO8tt COOLING WATf R FROM Rt ACTOR VESSIL, 063-08 PROVIDE FLOW PATH FOR RCIC SYSTEM (71) CORT COOLINO 2 DFN-RTP 060 YtB WATER TO RCACTOR Ff tDWATER SYSTEM 13) SPARGERS.

069 09 PROVIDE CAPABILITY OF MANUAL BACKUP CONTROL 2 BFN RTP DUC Yt3 ISOLATION (VALVE CLOSURE) TO PREVENT LOSS OF Rf ACTOR i WAT[R INVENTORY,

[

070-01 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2 DfN RTP 064 A YES .

070 02' PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN RTP 068 NO StCONDARY CONT Altd. :

MENT WAS TESTfD AS l A WHOLE DURING UNIT !

2 TESTING 070-03 PROVIDI DRYWELL COOLING WHEN POWER AtJD COOLINO 2 BrN RTP-030 TBD lNKNOWN, P[NDING WATER ARE AVAILABLE. ISSUANCE OF UNIT 3 '

BTRD

'071 01 AUTOMATIC RCIC SYSTEM INITIATION ON REACTOR 2 BrN RTP-071 YES 1 FttDWATER SYSTEM (3) RPV LOW WATER LEVIL (L2) SIGN AL [

TRANSMITTED VIA RHR SYSTEM 174). AUTOMATIC RCIC SYSTEM SHUTOFF (IF OPERATING) ON Rf ACTOR FttDWATf R SYS1tM (3) RPV HIGH WAlf R LEVEL (L8) SIGNAL, i 071 02- MANUAL RCIC INITIATION AND TRIP TO CONTROL LEVEL. 2 BrN RTP-071 YES (NON LOCA UNIT) 071-03 CLOSE RCIC SYSTEM STE AM SUPPLY LINE ISOLATION VALVES 2-BfN RTP 071 YiS

ON RCIC SYSTEM GROUP 5 ISOLATION SIGN ALS (HIGH STE AM '

LINE DirFERENTIAL PRESSURE HiGH STEM LINE SPACE TEMPERATURE OR LOW STf AM LINF PPLSSURE).

071+04 MANUALLY CLOSE RCic SYSTEM STE AM SUPPLY LINE 2 BfN RTP-071 YfS '

ISOLATION VALVCS ON Rf ACTOR FEEDW ATER SYSTEM (3)

INDICATION OF LOW RPV hitsSURE, 071 OS PROVIDE RE ACTOR COOL ANT PRES $URE BOUNDARY (RCPRL YES 2 BrN RTPf68

. 071 07 PROVIDE PRIMARY CONTAINMfNT BOUNDARY. 2 BrN FTP 064A YFS TBD TO BE DETERM6NED -

BTRD BASILINE TEST REQUIREMENTS DOCUMENT

,a _____.___..__.,_-.u-...__....____._.._.--. , _ _ _ _ _ _ - - . _ . - _ _ _ -

ENCLOfiURE 1 Pao 22 W 31 TABLE 1 CORREL ATION BETWEEN DROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 HESTART TEST PROGRAMS i l

l I

SYSTEM MODE DESCRIPTl1N UNIT 2 TEST TEST COMME NTS i MODE FOR l 03 ;

071 08 PROVIDE SYSTEM PRES $URE BOUNDARY IN SUPPORT OF RHR 2 DFN RTP 068 YES ,

SYSTEM (74) CONT AINMENT (TORUSl COOLING FUNCTION. ,

t 071-09 MANUAL RCIC SYSTEM 0PERATION FROM OUTSIDE THE MAIN 2 DFN RTP 071 YLS CONTROL ROOM TO MAINT AIN NORMAL RPV WATER INVENTORY WHILE RPV PRESSURE 19 ADOVE 100 PSlo.

071 10 PROVIDE POWER TO ECCS DIVISION I AND 11 AN ALOG TRIP 2 BFN RTP-071 YES UNITS (RE ACTOR FEEDW ATER SYSTEM (3) PRIMARY r CONTAINMENT SYSTEM (64) RE ACTOR WATER RECIRCULAfl0N SYSTEM (68) RCIC SYSTEM (71) AND HPCI SYST E M (73}l.

071 11 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN RTP-06S No SECONDARY CONTAIN-NENT WAS TESTED AS A WHOLE DURING UNIT #

2 TESTING 073-01 AUTOMATIC HPC( SYflTEM INITIATION ON RE ACTOR 2 BFN RTP-073 YES '

FEEDWATER SYSTEM (3) RPV LOW WATLR LEVEL (L2) SIGN AL, AUTOMATIC HPCI SYSTEM SHUT 0FF ilF OPER ATING) ON RE ACTOR FEEDWATER SYSTEM (3) RPV HIGH WAllR LEVEL (L8) SIGN AL, 073 03 CLOSE HPCI SYSTEM STE AM SUPPLY LINE ISOLATION VALVES 2 BFN RTP 073 YES ON HPCI SYSTEM GROUP 4 ISOL ATION SIGNALS (HIGH STEAM LINE DIFf ERENTIAL PRESSURE HIGH STE AM LINE SPACE TEMPERATURE OR LOW STE AM LINE PRESSURE). i 073-04 MANtjALLY CLOSE HPCI SYSTEM STE AM SUPPLY LINE 2 BFN RTP 073 YES ISOLATION VALVES ON REACTOR f EEDWATER SYSTEM (3)

INDICATION OF LOW RPV PRES $URE. , ;

073 OS ; PROVIDE RE AC10R COOLANT PRESSURE BOUNDARY (RCPD) 2 DFN RTP 068 YES DURING HPCI SYSTEM STANDDY.

073 06 PROVIDE RE ACTOR COOLANT PRESSURE BOUNDARY (RCPD) 2 BFN R1P 068 YES DURING HPCI SYSTEM OPERATION, l I

073 07 PROVIDE PRIMARY CONTAINMENT BOVNDARY DURING HPCI 2 BFN RTP 064A YES SYSTEM STANDDY.

073 08 PROVIDE PRIMARY CONTAINMENT 00VNDARY DURING HPCI 2 DFN RTP-064A YES SYSTEM OPERATION.

073 09 MANUALLY TRIP HPCI SYSTEM FROM OUTSIDE THE MAIN 2 DFN RTP 073 YES-CONTROL ROOM TO PREVENT RPV OVERF!LL.

073 10' LIMIT THE LOSS OF COOLANT THROUGH HPCI SYSTEM STE AM NONE NO PASSIVE COMPONENT '

SUPPLY LINE BRE AK (PASSIVE FLOW RESTRICTOR DUILT INTO JUSTiflED RY STEAM LINE). ENGINEERING ANALYSIS TBD To BE DETERMINED BTRD . BASELINE TEST RE0ViREMENTS DOCUMENT

_ _ _ _ _ . , _ _ . _ _ _ . . _ , . _ -._. . ~ _ _ . , _ . . . , . _- . . _ . - -

ENCLOSURE 1 Pos ** *t 31 l

TADLE1 ;

1 CORRELATION DETWEEN BROWNS f ERRY NUCLEAR PLANT l UNITS 2 AND 3 RESTART TEST PROGRAMS i l

SYSTEM MODE DESCRIPil0N UNif 2 TEST TEST COMMENTS MODE FOR U3 2 073 11 PROVIDE SECONDARY CONT AINMENT 00VNDARY. 2 BFN RTP OS$ NO SECONDARY CONTAIN-MENT WAS TESTED AS [

A WHOLE DURING UNIT i 2 TESTING 073 12 PROVIDE PrWER TO PRIMARY CONTAINMENT SYSTEM (64) 2 BFN RTP 064 A YES DRG'!*i8, PRESSURE INDICATOR $ IN OUPPORT OF RHR SYSTEM DRYWELL/ TORUS SPRAY MDDE AND CONT AINMENT ATMOSPHERE DILUTION SYSTEM POST 40CA CONTAINMENT ,

VENTING MODE.

074 01 AUTOMATIC LPCI MODE INITIAfl0N ON RPV LOW WATER 2 BFN RTP-074 YES t LEVEL (Lil $10NAL OR HIGH DRYWELL PRES $URE SIGNAL WITH CONCURRENT LOW RPV PRES $URE PERMIS$1VE SIGNAL. j MANUAL LPCI MODE INITIATION FROM THE MA!N CONTROL ROOM.

074-02 PROVIDE SUPPRESSION POOL WATER COOLING TO MAINTAIN 2 EdH RTr 974 YES SUPPRES$10N POOL WATER TEMPER ATURE DELOW LIMITS TO ASSURE THAT PUMP NPSH REQUIREMENTS ARE NET AND THAT COMPL'it CONDENSATION OF DLOWOOWN STEAM {

FROM A DEsta BAtlS LOCA CAN BE EXPECTEDc _

074-03 PROVIDE SPRA) (0 DRYWELL AND TORUS FOR CONF AINMENT 2 BFN RTP-074 YES COOLING AND LOWERING OF CONTAINMENT PRESSURE UNDER '

POST ACCtDENT CONDITIONS.

074-04 PRov1DE SHUTDDWN COOLING MODE (MANUAL) TO RESTORE 2 OFN RTP 074 YES REACTOR TEMPERATURf TO NORMAL.

074 09 PROVIDE SECONDARY CONT AINMENT BOUNDARY. 2 BFN RTP-065 NO SECONDARY CONT AIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 014-10 PROVIDE REACTOR COOLANT PRESSURE BOUNDARY 1RCPB). 2 0FN RTP-068 YES 074 11 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2-BFN RTP-064 A YES 074-12 PROVIDE SIGN AL (TH AT A PHR PUMP l$ RUNNING) TO MAIN 2 BFN RTP 074 YES STEAM SYSTEM (1) AUTOMATIC DEPRESSUh.tATION SYSTEM.

(ADS) INITIATION LOGIC.

074 14 PROVIDE RHR SYSTEM PIPING FLOW PATH FOR TRANSMISSION NONE NO PASSIVE COMPONENTS OF CONDENSATE AND DEMINERAllZED WATER SYSTEM (2). VERIFIED BY WATER CUPPLY TO HPCI SYSTEM (73) PIPING UPSTRE AM OF WALKDOWN HPCI SYSTEM PUMP.

074 15- PROVIDE RHR SYSTEM P! PING FLOW PATH FROM HPCI SYSTEM NONE NO PASSIVE COMPONENTS (73) PUMP MINIMUM FLOW BYPASS LINE TO PRIMARY VERIFIED BY CONTAINMENT SYSTEM (64) SUPPRESSION POOL. WALKDOWN 074 16 PROVIDE RHR SYSTEM PIPANG FLOW PATH FROM RCIC SYSTEM NONE NO PASSIVE COMPONENTS (711 PUMP MINIMUM FLOW BYPASS LINE TO PRIMARY VERIFIED BY Cf?NTAINMENT SYSTEM (64) SUPPRESSION POOL. WALKDOWN TBD TO BE DETERMINED -

BTRD BASELINE 1[ST REQUIREMENTS DOCUMENT '

-. ,_ ._ . . . . _ . . . . . . . _ _ _ _ . . . . ..___._._au,-.-___, __ ,.. .--

i

ENCLOSURE 1 Pep 24 et 31 i TABLE 1 j CORRELATION DETWEEN BROWNS TERRY NUCLEAR PLANT !

UNilS 2 AND 3 RESTART TEST PROGRAMS I

t 1

SYSTEM MODE DESCRIPTION UNIT 2 TEST 1tST COMMENTS !

MODE FOR l U3 j 074 17 PROVIDE AUTOMATIC LPCI MODC INITIATION SIGidAL FOR 2 BFN RTP 074 YES CLOSURE OF Rf ACTOR WATER RECIRCUL Afl0N SYS TEM I68)

E PUMP OlSCHARGE VALVi$. _

074-19 MANUAL RHR SYSTLM OPERATION (I PCI TORUS COOLING AtJD 2 BFN RTP BUC ;Yt3 ,

SHUTDOWN C00LifdG MODis) FROM OUTSIDt THE MAIN CONTROL ROOM. ,

074 2D PROVIDt FLOW PATH AND PRESBURE BOUNDARY INTEGRITY 2 BFN RTP 074 YES FOR RHR SERVICE WAttR SYSTEM 123) COOLANT TO THE MAIN RHR SYSitM HE AT EXCHANGERS.

074-21 PROVIDE REACTOR FIEDWATER SYSTEM (3) RPV COW WATER 2 BF N R TP-07 ' YtS i LEVEL (L2) SIGNAL FOR AUTOMATIC RCic SYSTIM 171)

INITIATION.

074 22 PROVIDE RHR SYSitM UNIT CROSS. Tit VALVES OPEN ~ 2 BFN RTP-023 TBD PtNDING ISSUANCt OF POSITION SIGN AL TO RHR SERVICE WATER SYSTEM (23) FOR UNIT 3 DTRD CLOSURE OF RHR SERVICE WAT[R VALVES TO MAINTAIN ,

PRIMARY CONTAINMENT BOUNDARY. !

074423 RHR ISOLATION SIGNAL TRIPPED ON SloN AL FROM PRIMARY 2-BFN RTP 064 A YtB i CONTAINMENT SYSTEM (64L _

074 24 INHlBIT AUTOMATIC INITIATION OF FIHR IN UN!! 2 GIV(4 A NONE TED PENDING ISSUANCE OF l LOCA SIGNAL FROM THf RHR SYSTEM OF UNIT 1. (UNIT 1 UNIT 3 BTRD ,

ACCIDENT SIGNAL ASSUMED TO DE DIS ACLED). .

l 074 25 PROVIDE A LOCA SIGNAL FROM UNIT 2 TO INHIBIT NONE TBD PENDING ISSUANCE OF AUTOMATIC INITIATION OF RHR (74) OF UNIT 1. UNIT 3 DTRD 076 01- SUPPLY COOLING VlATER TO RtACTOR AUTO INITIATION. 2 BFN RTP 075 YES t

075 03 PROVIDE CS PUMP POWER DISLONNEct'f ROM OUTSIDE MCR. 2 BFN RTP 075 YES 075 04 PROVIDE REACTOR COOLANT PRESSURE DOUNDARY. 2 BFN RTP-068 YES 075 05 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2 BrN RTP,064A YES 075-06 PROVIDE ACCIDENT SIGNAL INPUT TO 480V LOAD SHED 2-BFN RTP-075 YES LOGIC.

,. 075 07 PROVIDE START SIGN AL TO DIESEL GENERATOR ON LOW 2 BFN RTP-076 YES L. LEVEL (L1) OR HIGH DRYWELL PR[SSURE, l

075 08 PROVIDE PRIMARY CONTAINMENT SYSTEM (841 H!GH 2-BFN RTP 073 YES DRYWELL PRESSURE SIGNAL FOR AUTOMATIC HPCI SYSTEM l (73) OPER ATION.

i TBD- TO BE DETEPMINED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

L

ENCLOSURE 1 Peps 28 of at :

i TADLE 1 i CORRELATION DETWEEN BROWNS FERRY NUCLEAR PLANT :

UNITS 2 AND 3 RESTART TEST PROGRAMS l

w. .._

SYSTEM MODE DESCRIPfl0N UNIT 2 TEST TEST COMMENTS MODE FOR U3 ,

078-09 PROVIDE CORE SPRAY SYSTEM PihNG FLOW PATH FROM tJONE NO PASSIVE COMPONENTS PRIMARY CONTAINMENT SYS TEM (64) SUPPRESSION POOL TO VERIFIED BY .

RCIC SYSTEM (71) PIPING UPSTREAM OF RCIC SYSTEM PUMP WALKDOWN !

FOR MANUAL RCIC SYSTEM OPERATION FOR OUTSIDE THE r M AIN CONTROL ROOM.

075 10 PROVIDE SIGNALS ITHAT CORE SPRAY PUMPS ARE RUNNING) 2 BFN RTP 078 YES TO MAIN STE AM SYSTEM (il AUTOMATIC OEPRESSURl2Atl0N SYSTEM (ADS 11NITIAT10N LOGif 07S-11 PROVIDE REACTOR FEEDWAT[6 '.YSiN g '

M .*TFR 2 BFN PTP474 YES I LEVEL (L1) SIGNAL TO RHR SYS1 18 0 4: L'il +.: ,Y "ON LOGIC.

.wy .-e= w 07S-12 PROVIDE REACTOR TEEDWA1ER SYSTEM tw .No RE AC10R 2 BFN RTP 006 Yts t WATER RECIRCULATION SYSTEM (BB) LOW R BCTOR PRES $URE SIGNALS TO RnN SYSTEM (741 LPC ' f A0DE INITIATION LOGIC.

07S 13 PROVIDE SECONDARY CONTAINMENT DOUNDARY. 2 BFN RTP 06S NO SECONDARY CONTAIN- i MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 07S-14 PERFORM ISOLAtl0N ACTIONS UPON RECEIVING ISOLATION 2 BFN RTP 064 A - YES SIGNAL (LOW LEVEL L3 OR HIGH DRYWELL PRESSURE) FROM THE PRIMARY CONTAINMENT SYSTEM (64).

07 S-I S PROVIDE 4. LOC A SIGNAL FROM UNIT 210 INHIBIT NONE 100 PENDING ISSUANCE OF AUTOMATIC INITIATION OF CORE SPH AY (75) 0F UNIT 1. (UNIT UNIT 3 BTRD 1 (CCIDENT SIGNAL ASSUMED TO DE DIS ABLED),

07S-16 INHiblT AUTOMATIC INITIATION OF CORE SPRAY IN UNIT 2 NotJE TBD PENDING ISSUANCE OF GIVEN A LOCA SIGNAL FROM THE CORE SPRAY SYSTEM (75) UNIT 3 BTRD OF UNIT 1.

076 01 CLOSE CONTAINMENT INERTING SYSTEM ISOLATION VALVES 2 DFN RTP 084 YES-ON PRIMARY CONTAINMENT SYSTEM (64) GROUP 6 ISOLATION SIGNAL.

076-02 PROVIDE OXYGEN AND HYDROGEN G AS ANALYZERS AND 2 BFN RTP 084 YES INDICATORS TO MONITOR QAS CONCENTRATIONS INSIDE THE t PRIMARY CONTAINMENT IN SUPPORT OF CONTAINMENT ATMOSPHERE DILUTION SYSTEM ($4) OPERATION.

076 03: PROVIDE PRIMARY CONTAINMENT DOUNDARY. 2 BFN RTP 064A YES 076 04. PROVIDE SECONDARY CONTAINMENT BOUNDARY, 2 BFN RTP 06S NO SECONDARY CONTAIN-

! MENT WAS TESTED AS A WHOLE DURING UNIT ,

2 TESTING

t. .

l 077 01 CLOSE RADWASTE SYSTEM ISOLATION VALVES ON PRIMARY 2-BFN RTP 024 YES

! CONT AINMENT SYSTEM (64) GROUP 2 ISOLATION SIGN ALS.

I' TBD To B5 DETERMANED l: BTRD BASELINE TEST REQUIREMENTS DOCUMEt4T

I

..

ENCLOSURE 1 Pepi to of 31 !

TABLE 1 i

CORRELATION DETWEEN BROWNS TERRY NUCLEAR PL ANT ;

UNITS 2 AND 3 HESTART TEST PROGRAMS f

am. -

SYSTEM MODE DESCRIPil0N UNIT 2 TEST TEST COMMENTS MODE FOR ,

U3 i 077 02 PROVIDE PRIMARY CONTAINMENT 00VNDARY. 2-Bf N RTP 064A YES i 077 03 PROVIDE SECONDARY CONTAINMENT BOUNDARY. 2 BFN RTP 066 ND SECONDARY CONT AIN-MENT WAS TESTED AS A WHOLE DURING UNIT 2 TESTING 077-06 PREVENT DACKFLOODING OF RADWASTE BLOG THRU SGT NONE TDD PENDING ISSUANCE OF DLDO OFF-GAS DLOG & OFF-GAS STACK DRAINS. UNIT 3 BTRD l 078 01 PROVIDE SECONDARY CONTAINMENT 00VNDARY. 2 DFN RTP 066 NO SECONDARY CONTAIN.

MENT WAS TESTED AS !

A WHOLE DURING UNIT 2 TESTING 5

078-02 PROVIDE PRESSURE BOUNDARY INTEGRITY AT RHR/FPC 2 BFN RTP-074 TBD PENDING ISSUANCE OF INTE RF ACE. UNIT 3 BTRO ;

i 078-03 PREVENT INADVERTENT SIPHONtNG OF THE SPENT FUEL POOL. 2 DFN RTP-060 YES 078 04 PROVIDE FUEL POOL COOLING WHEN POWER AND COOLING NONE 700 PENDING ISSUANCE OF WATER ARE AVAILABLE. UNIT 3 DTRO 079-01. PROVIDE SAFE FUEL HANDL!NG USING REFUEL BRIDGE & 2 DFN RTP-079 No UNIT 2 TEST S ATISFIES EQUIPMENT. REQUIRtMEN19 .

079 02 PROVIDE INTERLOCKS TO CRD SYSTEM DURING FUEL 2 0. A RTP-086 YES .:

i MOVEMENT. l

!'079 03 PROVIDE SAFE STORAGE FOR NEW AND SPENT FUEL. NONE NO PASSIVE COMPONENTS JUSTIFIED DY i i

LNGINEERING ANALYSIS 079-06 MAINTAIN SPENT FUEL POOL WATER LEVEL. NONE NO JUSTIFICATION DY ENGINEE RING ANALYSi$

079-07 REFUELING PLATFORMS ARE TO DE TIED DOWN UNDER NONE TUD PENDING ISSUANCE OF TORNADO CONDITIONS, WITH THIS PROVISION THE REFUELING UNIT 3 DTRD ;

PLATFORMS MUST WITHSTAND TORN ADO DEslGN LOADS, l 082 01 START STANDDY AC POWER SOURCE FOR 4KV SYSTEMtS75) F BFN RTP-082 TBD PENDING ISSUANCE OF UNIT 3 DTRD .

i 082 02 PROVIDE POWER TO 4KV SYSTEM'(S7El UPON 0/C 2 DFN RTP-082 YES AVAILABILITY AND LOSS OF OFF SITE POWER.

082-03 PROVIDE D/G POWER TO DIESEL FUEL TRANSFER 2 BFN RTP 082 NO UNIT 2 TEST SATISFIES ,

PUMPSISYSTEM 18), ALL REQUIREMENTS

-TBD TO DE DETERMINED .

BTRD' BASELINE TEST REQUIREMENTS DOCUMENT {

-- ~~. . , . . _ , , , . . _ , , _ _ ,. _ , ..,._,_.__.__.m, _- --.

ENCLOSURE 1 Pepe 27 of at TABLE 1 CORRELATION DETWEEN OROWNS TERRY NUCLEAR PLANT

4 UNITS 2 AND 3 RESTART TEST PROGRAMS j l

SYSTEM MODE DESCRIPfl0N UNIT 2 TtST 1tST COMMt NTS '

MODE FOR V3 i 084 01 PROVIDE DILUTION OF THE PRIMARY Col TAINMENT 2 DIN RTP 084 TBD PtNDING ISSUANCE OF ATMOSPHERt WITH NITROGtN AFTIR A LOCA TO MAINTAIN UNIT 3 DTRD ',

COMBUSTIBLE CAS (0XYGEN AND HYDROGEN) i CDNCENTRATIONS SELOW LtV!LS (0XY0tN 5% BY VOLUMll WHICH COULD) PRODUCE A COMBUSTIBLE C AS MIXTUnt.

084-02 VENT PRIMARY CONTAINMENT ATMOSPHIRt FROM PRIMARY NONt 100 RtMOVED FROM UNIT 2 CONTAINMENT SYSTEM (64) f LOW PATH TO STANDB( O AS Pf ST ART SCort. UNIT '

TREATMINT SYSTEM (65) AFTER A LOCA. SUPPLY CONTROL 31151 PINDING BTRD NITROotN TO OPEN PRIMARY CONTAINMENT SYSTEM (64) ISSUANCE.

ISOLATION VALVf B ON FLOW PA1H TO CAD SYSTtM 1841.

084 03 PROVIDE PRIMARY CONTAINMtNT BOUNDARY. 2 0FN RTP 064A YES 084 PROVIDE SICONDARY CONTAINMENT BOUNDARY. 2 tiFN RTP 066 NO SECONDARY CONTAIN- ;

MENT WAS TESTtD AS A WHolt DURING UNIT 2 TtSTING 084 06 PROVIDE NITROCEN TO THE CONTROL AIR SYSitM (32)(N NONE TBD UNIT 2 TtSTING DY SUPPORT OF LONG TIRN OPERADILITY OF MAIN STtAM POST MODIFIC ATION SYSTEM ADS S AFETY RILIEF VALVES (SRVS). APPENDIX R. TEST. UNIT 3 PINDING ISSUANet OF BTRD. i 084 06 CLOSE CAD SYSTEM VENT VALVts ON PRIMARY 2 DFN RTP 084 YES [

~

CONTAINMENT SYSTEM (64) OROUP 6 ISOL ATION SIGNAL.

086-01 PROVIDE SCRAM (98) AND CLOSE SDV VENT DRAIN VALVIS. 2 DFN RTP-086 YtB ;

OCS O2 PROVIDE PRIMARY CONTAINMENT BOUNDARY. 2 BFN RTP O64 A YES t

09503 PROVIDE SECONDARY CONTAINMENT 00bNDARY. - ' 2 DFN RTP-065 NO SECOND ARY CONTAIN-MENT WAS TESTED AS ,

A WHOLE DUR!NO UNIT ;

2 TESTINO 08604 PROVIDE REACTOR COOLANT PRES $URE BOUNDARY (RCPBl. 2 BFN RTP-068 YES i

085 05 ~ PREVENT ROD WITHDR AWAL. 2-DFN RTP 085 YES 085 06 PROVIDE HOUSING SUPPORT TO KitP RODS IN PLACE. NONE NO JUSTIFICATION BY ENGINitRING ANALYSl3 085 07 LIMIT ROD DROP RATE TO LESS THAN 3.11 FT/ SIC, NONE NO JUSTIFICATION DY ENGINEtRING ANALYSIS i

085 08 PROVIDE MCR ROD POS(TION INDICATION. 2 BFN RTP 085 YiS 08S-09 PROVIDE SCRAM DISCHARGE HIGH WATER LEVEL SIGNAL. 2 BFN RTP 085 YES TBD: TO BE OtTERMINED BTRD BASELINC TEST REQUIREMENTS DOCUMtf4T t

,..,.,_%,,,____.,_._,-._.,,-._.-.-.~ . _ . . . , . . . - _ _ _ _ - . _ _ _ _ . . _ _ . _ . . . . _ . , . , _ - -

ENCLOSURE 1 Ps9e 28 c( 31 TABLE 1 ;

CORRELATION LIETWEEN BROWNS TERRY NUCEEAR PLANT i UNITS 2 AND 3 RESTART TEST PROGRAMS l SYSitM MODt DtSCRIPfl0N UNIT 21[ST ltST COMMINTS !

MODE - FOR U3 085 10 PROVIDE SCRAM DISCHARot LOW AIR HEADER PRES $VRE 2-liFN RTP 085 YtS ,

SIGNAL.

l 085 11 PROVIDE REMOTE DACKUP CONTROL FRDM OUTSIDE THE MAIN 2 DIN RTP DUC YES I CONTROL ROOM. l 088 12 PROVIDE SYSTEM PRCSSURE DOUNDARY SUPPORT TO MAIN NONE No PASSIVE FUNCTION STE AM 0YSTEM (1) > 30% TURDINt FIRST STAot PR[SSURL JUSTIFitD BY ,

INST RUMINt ail 04. Al4ALYSis ?

085 18 PROVIDE SYSTf M PRESSURE COUNDARY IN SUPPORT OF RCIC NONE NO. PASSIVE FUNCTION 4 SYsitM (71) AUTOMATIC INITIATION MODE AND MANUAL JUSTIFifD BY OPERAtl0N FROM OUTSIDE THE MAIN CONTROL ROOME ,A"ALYSit 085 14- PROVIDE ALTERNATC ROD INSERTION BY OrtNING BACKVP 2 DFN R1P 085 YES !

SCRAM VALVES ON FitDWATER SYSitM (3) RPV LOW WATtR ,

LEVEL (L2) SIGNAL OR HIGH Rt ACTOR VisstL PRESSURE SIGNAL.

086-01. PROVIDE DitStL STARTING AIR To Dif StL Gt NtRATOR 2 DFN RTP 082 NO 11 NIT 2 TEST SATisFits -

SYSTEM (82). RE QU1HtME NT S 090 01 PROVIDE MAIN STEAM LIVE H10H RADIAtl0N $1GNAL TO 2 BFN R1P 090 Ytt Rt ACTOR PROTECTION SYSTEM 1991. 1 090-02 PROVIDE PRIMARY CONTAINMtf4T BOUNDARY (UP TO 2.DFN RTP-064 A Yts 2 FCV-90 254 AB 2bS 257AD).

090-03 PROVIDE REACTOR BUILDING VtNTIL ATION EXHAUST LINC AND 2 DFN RTP-090 YES REFutLING ZONE AREA (ADJACINT TO THE FUCL POOLS) HIGH RADIATION SIGN ALS TO PRIMARY CONT AINMENT SYSTEM (64)

CROUP 6 ISOLATION LOGIC.

080 04 ' PROVIDE CONTROL ROOM INTAKE AIR DUCTS EXCESSIVE No UNIT 2 TEST SATISFits !

2 0FN RTP-0318 R ADIATION SIGNAL TO AIR CONDITIONING SYSTtM (311 f OR Rt QUIREMtN16 INITIATION OF CONTROL ROOM iMERGENCY VtNilLATION (ISOLATION OF INTAkt DUCTS AND SUPPLY OF PRISSURiftD FILTERED OUTDOOR AIR).

090 0S CLOSE VALVES ON SUCTION AND RETURN LINES TO THE 2 DFN RTP 090 YES DRYWLLL RADIDACTIVE PARTICULATE 10 DINE AND OASIOUS MONITOR ON PRIMARY CDNTAINMENT SYS1tM (64) OROUP 6 ISOLAtl0N SIGNAL. ,

090 06 PROVIDE SYSTEM PRtSSURE BOUNDARY INT (CRITY (WITH ALL NONE No PASSIVE FUNCTION I MICHANICAL JOINTS AND COMPONENTS ASSOCIATfD WITH JUSTIFif 0 DY j THE OFF-LINE LIQUID MONITORS)10 RAW COOLING WATER [NGINEERING SYST[M (24)IN SUPPORT OF LECW SYSTLM 167) PRtSSURL AN ALY516 -

BOUND ARY INTEGRITY.

090-07 PROVIDE SYS1[M PRESSURE SOUND ARY INTEGRITY (WITH AlL NONE NO PASSIVE FUNCTION MECHANICAL JOINTS AND COMPONENTS ASSOCIATED WITH JUSTIFitD DY THE OFF LINE L10VID MONITORS) TO RHR StRVICE WAT[R [NGINE[ RING SYSTEM (23) COOLING WATER FOR RHR SYSTt M (74) HE AT ANALYSIS i EXCHANGERS.

TBD. TO DE DETERM:NED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

, _ - . _ - . . . . _ . . . .. . _ , - . _ _ . _ _ , , . . . .__.._~.....-......,,_u.~. _._; . - . - _ ,

l l

"* ENCLOSURE 1 Psos 28 of 31 i l

TADLE 1 l CORRELATION DE1 WEEN BROWNS TERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS SY$1tM MODE DESCRIPil0N UNIT 2 TEST TEST COMMENTS MODE FOR U*

090 09 PROVIDE SLCONDARY CONTAINMENT BOUNDARY 2 BFN RTP 005 NO SECONDARY CONTAIN-MENT WAS TLSitD AS i A WHolt DURING UNIT 2 TESTING l 052 01 PROVIDE IRM HIGH NtVTRON FLUX 1 RIP SIGNAL T O Rt AC70ft 2 DFN RTP 092 YtS 1 PROTECTIO _N SYSTEM 092 02 PROVIDE APRM HIGH NtVTRON FLUX TRIP SIGNAL TO 2 BFN RTP 092 YtB Rt ACTOR PROTtCil0N SYST[M 1991.

092 04 PROVIDE MOD BLOCK MONITOR TRIP SIGNAL IVARIADLE WITH NONt 100 REMOVID FROM UNIT 2 Rf AC10R WATER RfCIRCULATION SYST[M(68) FLOWI TO THE RISTART SCOPt. UNIT REACTOR MANUAL CONTROL SUBSYSitM OF THE CONTROL 3 DECisl0N PENDING ROD DRIVE SYSTEM (88) TO INHIBif CONTROL ROD ~lSSUANCE OF BTRD.

WITHDRAWAL.

092-05 PROVIDE INDICATION IN MAIN CONTROL ROOM OF NONE TBD REMOVtD IROM UNIT 2 POWERINIUTRON FLUX LEVtl AS MONITORED ON THE SRMS RESTART SCOPt. UNIT 1RMS OR APRMS (AS APPLICABLt) AS THE (VINT IS IDENTIFitD 3 DECISION PtNDING AND THE STANDBY LIQUID CONTROL SYf,TtM (63) INJECTS ISSUANCE OF BTRD.

THE DDRON SOLUTION INTO THE Rt ACTOR.

09247 PROVIDI RE ACTOR Co0L ANT PRtSSURT BOUNDARY (RCPBL 2 DFN RTP-068 Yt$ ._

094-01 PROVIDE PRIMARY CONT".%ENT INTEGRITY. 2 BFN-RTP 064A YES 094 03 PROVIDE RtACTOR COOLANT PRESSURt DOUNDARY (RCPD) - 2-DFN RTP 064 A 'YES (PASSIVE FUNCTION ONLYi 099 01 PROVIDE AUTO SCRAM 4. SDV VENT / DRAIN VALVE ISOL ATION 2 BFN RTP-099 YtS SIGNAL TO CRD SYST!M(86L 099 02 PROVIDE MANUAL SCRAM SIGNAL AND SDV VENT / DRAIN 3 BrN RTP-099 YES VALVE ISOLATION SIGN ALS TO CRD SYSTIMl8Sg_

099 03 PROVIDE 'RUN' MODE SGNL TO PCIS 164) FOR LOW STE AMLtNE 2 BrN RTP-099 YES PRES $URE ISOLATION PERMISSIVE.

099 04. PROVIDE REFUtlINTERLOCK TO RtACTOR MANUAL CONTROL 2 DFN RTP-OBS YES SYSTEM.(BS) 099 05 PROVIDE TRIP SIGNAL TO Ric!RC PUMP MOTOR BREAKtRS 2 DFN RTP 068 YES (SYSTEM 68L 099 06 PROVIDE SIGNALS TO PRIMARY CONTAINMENT ISOL ATION 2 DFN RTP-099 YtS-SYSTEM (64) LOGIC.

(; 111 01 Rt ACTOR DUILDING CRANE IS TO DE Tit 0 DOWN UNDER 'NONE' TDD PASS!VE MODE ,

TORNADO CONDITIONS WITH THIS PROVISION THE CRANE PtNDING ISSUANCE OF NiUST WITHSTAND FULL OCSIGN LOAos IPASSIVil. BTRD l .- 244 01 PROVIDI COMMUNICATION FROM LOCAL PANtLS FOR 2 DFN RTP 244 NO UNIT 2 TEST SATISFIES SHUTOOWN FROM OUTSIDE THE MCR. At0VIREMENTS l'

TBD TO BE DETERMINtD BTRD _ BA$tLINE TEST REQUIREMENTS DOCUMENT

'* ENCLOSURE 1 Peps 30 of 31 i TABLE 1 CORRELATION BETWEEN BROWNS TERRY NUCLEAR PLANT UNMS 2 AND 3 RESTART TEST PROGRAMS i

SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMME NTS j MODE FOR ;

3 303 01- MAINTAIN CONFIGURATION INTIORITY OF STRUCTURIS NONE NO JUSTIFICATION BY DURiNG t ARTHOUAKl. ENGINttRING <

ANALYSIS 303 02 PROVIDE PR01tCT10N AGAINST THE (FFLCTS OF FLOODING. Notit NO JUSTIFICATION BY

[NOINttRING ANALYSIS 303 03 MAINTAIN CONFIGUR ATION INTIORITY OF STRUCTURIG NONE No JUSTIFICATION BV .

DURING TORN ADO E NGINttRING ANALYSIS 303 04 SECONDARY CONT AINMENT LE AKAGE RATE CRITERIA MUST 2 LIFN RTP-065 No UNIT 2 TEST $ ATISFits BE MAINT AINED BY THE REACTOR BUILDING. At0VIREMENTS

$7101: PROV10E 126V DC CONTROL POWER TO D0 CIRCUlTRY 2 DFN RTP S71 TBD PENDING ISSUANCE OF (SYSTEM 82h UNIT 3 BTRO 672-01 PROVIDE 208/120V l & C BUS POWER DISTRIBUTION. NONE TBD PINDING ISSUANCE OF UNIT 3 BTRD

$72 02 PROVIDE UNIT PREFERRED POWER DISTRIBUTION. 2-BFN RTP 67 2 TBD PENDING ISSUANCE OF UNIT 3 BTRO ,

$72 03 PROVIDE 120V AC POWER FOR RPS SYSTEM. 2 PFN-RTP-57 2 TBD PENDING ISSUANCE OF UNIT 3 BTRD 573 01 PROVIDE CONTROL & LOOlC POWER TO 4KV & 400V 2-BFN RTP-57+3 100 PENDING ISSUANCE OF SWITCHGEAR. UNIT 3 BTRD 573 02 PROVIDE SWITCHY'ARD (LOOKV 161KV) REL AYING & TRIPPING NONE TBD REMOVED FROM UNIT 2 POWER. SCOPE, UNIT 3 PENDING ISSU ANCE OF 7 UNIT 3 BTRD, 573 03 PROVIDE MOTIVE POWER L LOGIC POWER To t0VIPMtNT, - 2 BFN RTP 67 3 TDD UNIT 2 TESTED BATTERIES ONLY, UNIT 3 PENDING ISSUANCE OF UNIT 3 BTRD.

573 04- PROVIDE DISTRIBUTION POINT FOR NUMEROUS ELECTRIC NONE TBD REMOVED FROM UNIT 2 SYSTEMS. SCOPt. UNIT 3 PENDINO ISSUANCE OF UNIT 3 BTRD.

573 06 . PROVIDE LO".1 POWER TO 480V LOAD SHED LOGIC. 2 DFN-RTP 57-3 TBD UNIT 2 TESTED BA1TERit$ ONLY, UNIT-3 PENDING ISSUANCE OF UNIT 3 BTRD.

574 01 PROVIDE 480V SWITCH 0 TAR DISTRIDUTION. 2 DFN RTP-57 4 TDD UNIT 3 SCOPE PENDING ISSUANCE OF DTRD ;

$74-02 PROVIDE 480V MCC DISTRIBUTION. 2 DFN RTP-67 4 TBD UNIT 3 SCOPE PENDING ISSUANCE OF BTRD TisD TO BE DETERM!NED BTRD BASELINE TEST REQUIREMENTS DOCUMENT

, _ . _ _ .,__._ ,; ._..a_-_

. c_,.. - , , _ . _ _ . . . ..., . _._ _ . _; _ . ,, , . . -

+

ENCLOSUREI Peo. 31 of 31 TABLE 1 i

CORRR.ATION BETWEEN BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 RESTART TEST PROGRAMS i

~

SYSTEM MODE DESCRIPTION UNIT 2 TEST TEST COMMENTS MODE .FOR U3 i 674 03 PROVIDE 480V LOAD SHED LOGIC SYST(M. 2-BrN RTP 67 4 TBD UNIT 3 SCOPE PENDING ;

ISSUANCE OF BTRD 574-04 PROVIDE 480V AC DISTRIBUTION BACKUP CONTROL. 2<BFN RTP BUC TDD UNIT 3 SCOPE PtNDING . !

ISSUANCE OF DTRD 578-01 PROVIDE 4 KV POWER DISTRIBUTION FOR DG LOADING 2 BFN R1P 67 6 TBD UNIT 3 SCOPE PENDING (SYSitM 82h ISSUANCE OF BTRD 576 02- PROVIDE RECIRCULATION PUMP 1 RIP UPON SIGN AL. 2 BFN-RTP 068 YtB i

$7603 PROVIDE INSTRUMENTATION FOR DG PARAltiLING (SYSitM 2 BFN RTP 67 6 TBD UNIT 3 SCOPE PENDING - -!

82) ISSUAfdCt OF BTRD

$75 04 PROVIDE INITIATION SIGNAL TO Dit$tLS {SYSitM 82). 2 DFN RTP 67 5 TBD UNIT 3 SCOPE PENDING ISSUANCE OF BTRD

$76-05 PROVIDE COOLING TOWER LIFT PUMP TRIP ON CONDENSER NONE TBD UNIT 3 SCOPE PENDING ;

CIRCULAT10f4 WATER SYSTEM (27) COOLING TOWER LIFT ISSUANCE OF B1RD l PUMP DISCHARGE WATER HIGH YtMPERATURE SIGNAL.

676 00 BACKUP CONTROL FOR 4KV Ft[ DER BREAKERS OUTSIDE THE 2-BFN RTP BUC NO UNIT 2 TEST SATISFitS CONTROL BAY. REQUIREMENTS 675-07 LOAD SHEDDING TO PREVENT -0VERLOADING OF 4KV SYSTEM 2 BFN RTP 67 5 TBD UNIT 3 SCOPE PENDING (575L ISSUANCE 0F BTRO 676 01 PROVIDE OFF SITE POWER TO 4KV DISTRIBUTION (576) NONE TBD UNIT 3 SCOPE PENDING ISSUANCE OF BTRD ,

t 676 02 PROVIDE 24V DC POWER, NONE TBD UNIT 3 SCOPE PEfdDING ,

ISSUANCE OF BTRD L

l TBD TO BE DETERMINED

BTRD . BASELINE Tt3T REQUIREMENTS DOCUMENT

O e ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 POWER ASCENSION TESTING PROGRAMS

\

ENCLOSURE 2 Page 1 of 7 DROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 POWER ASCENSION TESTING PROGRAMS I RACK 0ROUND TVA's Power Ascension Testing (PAT) Program for DFN Unit 2 was initially described in a TVA letter to the NRC dated February 14, 1989, and supplemented by letters dated September 8, 1989, October 30, 1989, April 12, 1990, August i 10, 1990, Novenbar 16, 1990, January 17, 1990. TVA correlated the DFN PAT l Program to R0 1.68 guidance in the April 12, 1990 letter per NRC request (NRC letter dated February 12, 1990). The NRC's review and acceptance of the Unit 2 PAT Program is documented by NRC letters dated September 22, 1989, February 12, 1990, July 6,-1990 and April 3, 1991. TVA plans to administer similar PAT ,

Programs for Units 1 and 3. The following addresres the differences and j correlates the Units 1 and 3 PAT Programs to R0 1.68 as was done for the Unit 2 PAT Program in the April.12, 1990 letter. Power Ascension testing hold points are not addressed.

II UNITS 1 AND 3 PAT PROGRAN DIFFERENCES ,

The PAT Programs for Unita 1 and 3 will deviate slightly from the specific testing identified for the Unit 2 PAT Program. Table 1 of this enclosure provides a comparison of the actual Unit 2 tests to the planned Unit 3 tests.

In general, for Unit 3,'TVA does not does not plan to perform the Turbine Trip test, the feedwater pump trip, and the backup control system test (Shutdown from Outside the control Room test) during power ascension as was done for Unit 2. The BFN training cimulator has been upgraded to more accurately model the plant's response to transiento. ?This, coupled with the desire to minimize the transients inflicted on the plant, led to the decision not to perform these tests. The reedwater System test will include tuning as was done on Unit 2. A backup control system test (Shutdown from outside the control test) will be performed during open vessel testing to demonstrate the system in functional. These departures from the Unit 2 PAT program are discussed in more detail.in Part !!I, below (III.F, 0, and I).

III Cooparison of BFN Criteria to R0 1.68 Criteria Table 1 also provides a detailed comparison of Regulatory Guide-(RO) 1.68 and BFN's PAT Program. A detailed discussion of the significant differences is provided below. The correlation to R01.68 applies to both Unita 2 and 3 except as noted. This information is provided in similar format to that provided for the U8...

PAT program (provided by TVA's April 12 1990 letter).

In general, significant oifferences between the BFN PAT program and R0 1.68 fall into the following categories:

Performing test at the plateaus specified by DFN FSAR in Section '

13.10 versus those-plateaus listed in R0 1.68.

e Not performing baseline determination testing for those parameters unaffected by the long outage.

, , - , , , - - , , - , , -.,- ,~., ,,, ,, - -..,,n,- ,.-,-,-,~e..,,, , . - - n s +.cv -

,,,--e.ws- ,.- -g_ an

ENCLOSURE 2 Page 2 of 7 DROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 POWER ASCENSION TESTING PROGRAMS !

e Hot 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, o Hot performing testing on equiteent which is not installed on DTH !

Unit 1 and 3 (Inclined fuel Transfer, Suppression pool Hakeup, '

partial Scram, etc.).

A. Control Rod Drive (CRD) Systen l 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 will be done to the reactor internals, these tests will not be repeated. The following table summarizes the planned operations for the CRD system.

PRESSURE DESCRIPTIONS 0 Control Rod coupling Check 0 Insert and withdrawal Timing 0 Functional Chock of Position Indication 0 Running and Stall Flow 0 Friction Testing Rated Individual Control Rod-Scram Timing [:

Rated Core scram with Less than 50% density-

8. Reactor Core Isolation Cooling (RCIC) System During an initial plant startup, baseline readings are taken on RCIC steam supply line high-flow isolation circuitry to provide an accurate value for the setpoint. As this value was obtained during the initial startup test program, and no work will be performed which would affect this data, these setpoints will not be adjusted.

During an initial plant startup, additional " cold starta demonstrations are performed to improve the confidence level in system performance. As the plant t was in operation for several years, and the RCIC system performed reliably during this time period, TVA considers that the present program adequately demonstrates system reliability. The following table summarizes the planned operations of the system.

1

l l

ENCLOSURE 2 page 3 of 7 BROWNS FERRY NUCLEAR PLANT UNITS 1 AND 3 POWER ASCENSION TESTING PROGHAMS PRESSURE INJECTION PATil DESCRIPTION 150 PSIC Condensate Storage Tank Rated Flow with Auxiliary Boiler Rated Condensate Storage Tank Hot Quick-Start Rated Condensate Storage Tank Cold Quick-Start 150 PSIO Condensato Storage Tank flot Quick-Start .

Rated Reactor Vessel llot Quick-Start '

C. High Pressure Coolant Injection (IIPCI) System During an initial plant startup, basoline readings are taken on itPCI steam ;

supply line high-flow isolation circuitry to provide an accurate value for the ,

setpoint. As this value was obtained during the initial plant startup test program, and no work will be performed which will affect this data, these setpoints will not be adjusted.

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 l reliability. The following_ table summarites the planned operations of the system.

PRESSURE INJECTION PATH DESCRIPTION 150 PSIG Condensate Storage Tank Rated Flow with Auxiliary Boiler Rated Condensate Storage Tank Hot Quick-Start ,

Rated Condensate Storage Tank Cold Quick-Start 150 PSIC Condensate Storage Tank Hot Quick-Start Rated Reactor Vessel Hot Quick-Start D. selected Process Temperature During a Hear Term Operating License (NTOL) startup, recirculation flow is lowered to ensure that temperature stratification does no occur in the reactor at the lowest possible recirculation flow. As the setpoint for the recirculation HG set low speed limiter has not been changed, there is no ,

requirement to repeat this test.

E. Core Power Void Mode Response 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 Unita 1 and 3 and no changes to the basic core design have been made

~ . _ . . . . _ .

i

. a .

ENCLOSURE 2 Page 4 of 7 DROWNS FERRY NUCLEAR PLANT l UNITS 1 AND 3 POWER ASCENSION TESTING PROGRAMS during this outage which would affect the dynamic stability of the core.

Additionally, the test was normally performed in test condition 4 (natural circulation), and BPN Units 1 and 3 will not operate in this region.

F. Feedwater System During an initial plant startup, trips of reactor feed pumps are performed to verify plant performance. This testing was satisfactorily demonstrated during the initial startup of Units 1 and 3. No modifications will be made that ;

would significantly affect plant performance. A feedwater pump trip from high power was performed during the power ascension test program of Unit 2 for cycle 6. This test-was performed specifically to acquaint operations personnel with the integrated plant response to this transient. The BFN training simulator has been upgraded to more accurately model the plant's '

response to transients. This, coupled with the desire to minimite the transients inflicted on the plant, led to the decision not to perform this test. Feedwater system tuning will be performed as was done on Unit 2. The reactor feedwater pump turbine (RFFT) trips listed in Enclosure 1 will be performed prior to unit startup or with the RFPT in a condition that is not ,

supplying coolant to tb1 reactor.

t O. Main Steam Isolation Valve (HS1V) Testing i Durir an initial plant startup, a closure of all HsIVs is performed at high powet (90-100 percent) to verify plant performance. This test was satisfactorily demonstrated during the initial startup of Units 1 and 3. No modifications will be made that would significantly affect plant performance; therefore, this test does not need to be repeated.

H. Turbine Trip During an_ initial plant startup, a turbine generator load reject (TGLR) is performed at High Power (90-100 percent) to verify plant performance. p Additionally, turbine tripo within the capacity of the turbine bypass valves nre performed to-verify that the_ reactor does not scram. This testing was

-satisfactorily demonatrated during the initial plant startup of Unito 1 and 3.

No modifications will be made that would significantly affect the response of the plant.

A turbine trip at high power was performed during the power ascension testing program of Unit 2 for cycle 6. This test was performed specifically to acquaint operations personnel with the integrated plant response to this transient. The BFH_ training simulator has been upgraded to more accurately model the plant's response to transients. This, coupled with the desire to minimize the transients. inflicted on the plant, led to the decision not to perform _this test.

1. Shutdown from Outside the Control Room This test was successfully performed at power during the initial startup test program of Unita 1 and 3.- A Unit 3 backup control system will be demonstrated to be functional during open vessel testing. This testing will utiliae voltage checks, indicating lights, annunciations, and visually inspecting l

I'

- - , _ _ _ ~ _ . . _ - ~ - ~ _ . , . . _ _ _ _ , _ , _ - . . , . - - - , - _ . _ .

o * . -

ENCLOSURE 2 Page 5 of 7 I BROWNS TERRY NUCLEAR PLANT UNITS 1 AND 3 POWER ASCENSION TESTING PROGRAMS components. A backup control system test at power was performed to ;

demonstrate personnel and procedural adequacy during the power ascension test program of Unit 2 for cycle 6. The DFN training simulator has been upgraded to include a fully functional backup control panel which is utilized for procedure validation and personnel training. This, coupled with the desire to minimiae the transients inflicted on the plant, led to the decision not to perform this test.

J. Recirculation Systes Testing During a NT0b startup, individual and dumt trips of recirculation pumps are performed to verify dynamic core response. This testing is not planned as it was satisfactorily demonstrated during the initial startup test program.

t K.

Loss of Offsite Power and Turbine Trip During a NTOL startup, a loss of offsite power coincident with a turbine generator trip are performed to verify electrical and reactor system transient ;

performance during = loss of auxiliary power. 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.

L, Drywell Vibration j During an initial startup, all safety-related piping systems in the drywell are monitored during scheduled transients to develop baseline vibration levelm.- Browns Ferry Nuclear Plant (BFN) is developing a list of specific locations to be monitored, consistent with the modifications which will be performed.

H, Reactor Vessel Internals Vibration This testing requires special equipment to be installed inside the: reactor vessel. It was performed during the initial test program, and no work will be performed inside the vessel which would require repeating this test.

H. Residual Beat Removal (RHR) System During an initial test program, RHH heat exchanger performance is 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 program and need not be repeated.

i

. _ . . , _ _ , _ _ _ _ . _ . _ _ _ . . . _ _ _ _ . _ . _ . _ . . _ , . . . . _ _ _ , - , . ~ . . . . , . . _ _ . , _ _ _ . , , . _ _ , .

= " - ENCLOSURE 2 Pags 6 of 7 J TABLE 1 ;

CORREL /sTION BETWEEN BrN UNITS 2 AND 3 PAT PROGRAMS

AND REGULATORY GUIDE 1,08 Bf N TEST TEST NAME UNIT 2 TEST OPEN 0-b b % bS. j TEST FOR VESSEL 100%

UNIT 3 ;

__-, g- - .

. ~- ,

$1 4.6.0.1 4 C'4tMICAL/RA010CHEMIC AL - Y10 Yts X/N X/N X/N RCl 1 R ADIATION ME ASUREMENTS YES YES X/N X/N i i

614.3.A.1 REACTIVITY MARGIN TEST Yts YtS X/N '

Tl 20 CONTROL ROD DRIVE SYSTEM YtB YES X/N X/N N 0011001 A SOURCE RANGE MONITOH YES YES X/N Tl149 WATER LEVIL MEASUREMENTS YES YES X/N X Sl 4.2 C 3 INTERMEDIATE RANGE MONITOR YtB YES X/N S14.1.0-3 LOCAL POWER RANCE MONITOR C AllBR ATION Yts YES X/N X/N Tl136 AVERAGE POWER RANGE MONITOR ICONSTANT YtS YES X/N HEATUP)

SI 4.1.B 2 AV! RAGE POWER RANGE MONITOR CALIBRATION YtB Ytt X/N X/N Tl136 PROCESS COMPUTER - YES YES X/N X/N 11198 RtACTOR CORE ISOLAtl0N COOLING SYSTEM Yts YES X/N ,

Tl189 HIGH PRESSURE COOL /,NT IN.fECTION SYSTEM YtS Yt8 X/N X/N l Tl149 SELECTED PROCESS Tl:MPER ATURE YCS YES X/N X/N I F

'Ti190 SYSTEM EXPANSION YtS YES X/N j Tl137 CORE POWER DISTRIDufl0N YES YES X/N X/N St 2.1 CORE PERFORMANCE YES YES X/N X/N NA- CORE POWER VOID MODE RESPONSE NA NA N Tl130 PREASURE REGULATOR YES YES X#4 X/N f

e

' Tl 131.. FEEDWATER SYSTEM YES YES X/N X!N ,

SI4.1.A.16 TURBINE SURVEILLANCE YES YES X/N X/N SI 4.7.0 MAIN STE AM ISOLATION VALVE YES YES X/N N ,

SI 4.6.0 SAFETY RELIEF VALVE YES YES X/N

. g NA TURDINE TR'P - YES NO N

$ Tl 73 SHUTDOWN FROM OUTSIDE CONTROL ROOM YES YE r, X N N Tl132 RECIRCULATION SYSTEM TUNING X '

YES YES X/N X/N NA RECIRCULATION SYSTEM NA NA N N X DROWNS FERRY TEST N REQUIRED BY NEAR TERM OPERATING LICENSE 1RG 1.68) ,

DIFFERS FROM BFN UNIT 2 PAT PROGRAM !

~ NA NOT APPLICABLE TBD TO BE DETERMINED

.+, .w.---.-...~..u_.--,-.-- ..-,_.,-..,,:-...--. ,- - , - - - . . . . - - , -

" +

ENCLOSURE 2 Page 7 of 7 TABLE 1 CORRELATION BETWEEN OfN UNITS 2 AND 3 PAT PROGRAMS AND REGULATORY GUIDE 1.00 BFN f t4T TEST NAME UNIT 2 1tST OPIN OM% 65 TEST FOR VisstL 100%

UNIT 3 l e.__. _

,_ i NA LOSS OF Of TSITE POWER TURBINE TRIP NA NA N 1(10 DRYWILL PIPANG VfDRATION Yll Ytt X/N X/N i NA Rf ACTOR PRf SSURE VtSSIL INTIRN ALS VIDRATION NA NA N N T1174 RECIRCULAT10N FLOW CALIBRATIOid .__

YtS Yts N X/N i T1183 REACTOR WATtR Cit ANUP SYSTtM YES YtB X/N 1 I

NA RESIDUAL Ht AT RtMOVAL SYSTEM NA NA N 4 l 1182 DRYWELL TEMPIR ATURIS Ytb Yts X/N X/N

$14.8.B.1.s.1 0FFGAS SYSTEM YtB YES X/N X/N -

I t

[

I i

e X BROWNS FIRRY TEST ,

N REQUIRID BY NE AR TERM OPERATING LICEN$t (RG 1,68)

DIFFERS FROM BFN UNIT 2 PAT PROGRAM NA NOT APPLICABLE TBD- TO BI DITIRMINED y 'r- ~m ew

w w -e p -m ---em.-w-e<-e--

e 3r--+ s'= e - re e

<w=a

m--- w, rw w- H aeY vv-e- = $7 u-t'-

e .. .

ENCLOSURE 3

SUMMARY

OF COMMITMENTS TVA will provide the evRC en update of the table in Enclosure 1 to reflect the outcorno of BTRD evaluations by Deconhor 31, 1992.

I' 4

I I

1 l

l

_ . _ . _ -

ML20092J308

Text

Reference:

SUMMARY

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