ML17341A786
| ML17341A786 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 01/07/1982 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM L-82-5, NUDOCS 8201110389 | |
| Download: ML17341A786 (60) | |
Text
REGULATOR&INFORMATION DISTRIBUTION 8 Y o7EM (RIDS)
" AGGES SION NBR: 8201110389 DOC ~ DATE: 82/01/07
-NOTARI."ZED:
NO DOCKET ¹ FACIL:50 250 Turkey Point IPlantE 'Unit 3r Florida IPower and Light C
05000250 50-251 Turkey Point Planti 'Unit PP Florida tPower and Li'ght C
05000251
'AUTH BYNAME AUTHOR AFFILIATION UHRIGP R ~ E ~
Flor ida ~Power L Li'ght Co.
"REC IP ~ NAME
'RECIPIENT AFFILIATION EISENHUTPDBG, Division of Licensing
SUBJECT:
Forwards status of NUREG-0737 post TMI requirerrents requiring -action up to 8 including 820101,
.DISTRIBUTION CODE:
AORTAS iCOPIES iRECEIVED:LTR,-[ ENCL J.
SIZE:,
2 =
TIlTLE: Response to NUREG 0737/NUREG-0660 TMI Action !Plan Rgmts (OL's)
NOTES:
RECIPIENT ID 'CODE/NA'h',E ACTION ORB ¹1 BC 01 INTERNAL: ELD IE/DEP OIR 33 IE/DEP/EPLB NRR/DE/ADCSE 22 NRR/DE/ADSA 17 NRR/DHFS/DEiPY29 NRR/OL/ADL 16 NRR/OL/ORAB 18 NRR/DSI DIR 24 NRR/DSI/ADPS 25 NRR/DSI/RAB NRR/DST/ADT 32
,EXTERNAL: ACRS 3g INPOiJ,STARNES NRC
',PDR
'02 NT I'
>COPIES L'T~TR ENCL 7
>7 1
'0 1
1 3
3 1
1 1
1 1
1 1
1 3
3 1
1 1
1 1
1 1
1 10 10 1
1 1
1 1
1 RECIPIENT ID CODE/NA'ME IE 1'2 IE/OEP/EPDB NRR/DE OIR 21 NRR/DE/ADMQE 23 NRR/DHFS DIR 28 NRR/DL DIR NRR/DL/ADOR 15 NRR/OSI ADRS 27 NRR/DS I/ADGP 31 NRR/DS I/AORP 26 N
T DIER
'30 G FI "04 FEMA REP DIV LPDR 03 NSIC 05
>COPIES LTiTR ENCL 2
2 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
TOTAL NUMBER OF 'COPIES REQUIRED:
LTTR 52 ENCL
'51
Ci 1
4'
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//Ave FLORIDA'POWER & LIGHT COMPANY January 7,
1982 L-82-5 Office of Nuc lear Reactor Regu'I ati on Attention:
Mr. Darrell G. Eisenhut, Director Division of Licensing U.S.
Nuclear Regulatory Commission Washington, D.
C.
20555
Dear Mr. Eisenhut:
Re:
Turkey Point Units 3
8 4
DOCKET NOS. 50-250 5 50-251 Post-TMI Requirements This letter transmits to you the status of those NUREG-0737 items requiring action up to and including January 1,
1982.
We are working towards meeting all of the remainder of the requirements and will advise you should problems arise in meeti ng any of the long-term dates.
Yery truly yours, Ro ert E. Uhrig Yice President Advanced Systems 8 Technology REU/PKG/cab cc:
Mr. J.
P. O'Reilly, Region II Harold F.
Rei s, Esquire v
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820iii0389 820i07 PDR ADOCK 05000250 P
PDR PEOPLE... SERVING PEOPLE
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1.
PLANT SAFETY PARAMETER DISPLAY CONSOLE ~I,.D.2)
The Satety Assessment System of which the Plant Safety Pardmeter Display Console is a part is expected to be tully operational tollowing the t> rst refueling outage of each unit atter January 1,
1983 based on current scheduled equipment delivery dates.
2.
REACTOR COOLANT SYSTEH YENTS +I~I.B.I Our letter L-81-298, dated July 16, 1981 submitted to you the design description of our RCS Vent System.
At that time we stated that we would submit operating procedures fol lowing the NRC approval ot the design.
In an effort to expedite your review, it is now our intent to submit a set of operating procedures to you by I<arch 1, 1982.
It is intended that the RCS vent system for Unit 3 will be installed prior to startup from the unit's steam generator repair outage.
The Unit 4 system has been only partially installed due to the late delivery of equipment.
Since a unit shutdown is required to"install the system, it is our intent to install it during the Unit 4 steam generator repa> r outage.
3.
PLANT SHIELDING (11~8.2 A.
In letter L-80-16 dated January ll, 1980 we identified potential problem areas in the plant that could require plant modifications to lower postulated post-LOCA radiation exposures to plant personnel.
During the detailed design and engineering ei'fort conducted during the two years since our initial submittal, we have gained additional insight into the specific shielding problems.
As a result, certain solutions differing somewhat from those identified in our earlier letters have been implemented or are scheduled to be implemented.
However, ail of the problem areas identified in our previous submittal have been addressed and resolved.
0~
'I il' I
't II
It is our intent to have all remaining modifications complete by March 31, 1982 with the following exception:
The containment isolation valves CV.2819 and CV 2826 (air bleed valves used in the "pump back system")
are to be replaced with valves that are qualitied to operate in a post-LOCA envi ronment.
The manufacturer's shipping date for these valves is February 20, 1982.
The installation ot the valves requires a uni t outage.
It is our intent to install these new valves in Unit 3 during the next refueling outage following receipt and to install them in Unit 4 during its upcoming steam generator replacement outage.
B.
Radiation qualification ot safety related equipment is being addressed through our program to address the NRC concerns expressed in ISE Bul letin 79-01B.
4.
POST ACCIDENT SAt1PLI NG CAPABIL IT YQI I. B~3 The Post Accident Sampling System will not be instal'led and operational on January 1,
1982 as required by NUREG-0737 due to problems with equi pment availability.
The online chemi stry sampl.i ng analyzer is riot scheduled to arrive on site until January 31, 1982.
It is our intent that the system will be installed and operational prior to the Turkey Point Unit 3 startup from the steam generator repair outage.
At that time, the system will be fully operational and environmentally qualified with the exception of the Unit 4 containment isolation valves.
The new valves which are quali:fied to operate in a post-LOCA environment did not arrive on site until after Unit 4 completed its recent refueling and maintenance shutdown.
Since the replacement of these valves requires a
reactor shutdown, it is our intent to replace the valves during the Unit 4 steam generator repai r outage.
Proposed technical specifications will be submitted to you vollowing installation of the system.
4i.
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5.
TRAINING FOR MITIGATING CORE DAMAGE (-II.B.4$
The training program we discussed in L-81-183, dated of April 28, 1981 was completed as required prior to October 1,
1981.
6.
SARETY/RELIEf YALYE TE~STING I I.D.1')
It is Florida Power 8 Light Company's intent subject to the schedular constraints of the EPRI Safety and Relief Valve Test Program, to comply with the revised implementation schedule set forth in Mr. Eisenhut's letter of September 29, 1981 (Generic Letter No. 81-36).
7.
VALVE POSITION I ND ICATION I I. D. 3)
Our vendor has completed the envirOnmental qualification tests of the equipment.
The test reports will be available for inspection at the Turkey Point site as ar;e the test reports that are required to conform to I 5 E Bulletin 79-01B.
The results of the environmental qualification tests determined that the charge converter must have an additional enclosure added.
The new enclosure assemblies are now on site.
It is our intent to add the enclosures on Turkey Point Unit 3 prior to startup from the steam generator repair outage and to add the enclosure assemblies on Unit 4 during its steam generator repair outage.
8.
AUXILIARY FEEDWATER SYSTEM EVALUATION I I.E.1.1)
A.
AFW SYSTEM FLOW,RATE DESIGN BASES AND CRITERIA In letter L-80-419 dated December 26,
- 1980, we stated that the analyses required to document the design bases system flow requirements for the AFW system were underway and would be supplied
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upon completion.
Attacliment 1 coritains Florida Power 5 Light Company's final response on the subject.
The evaluation addresses Enclosure 2 of the NRC letter of October 16, 1979 as well as pos>tron (3) of NUREG-0737 item II.E.l. 1.
The loss of main feedwater transient serves as the design basis t'or the minimum tlow required tor the smallest capacity single auxiliary feedwater pump for Turkey Point Units 3
8 4.
B.
SHORT TERN ITEMS Our short term moditications to the AFH system have oeen completed for both units except.for those specit'ic items listed below.
The redundant Condensate Storage Tank level indicator has riot been installed in Unit 3 but will be operational prior to star tup trom the steam generator repair outage.
The modification ot two out ot three steam supply. valves from A.C. to D.C.
MOV's has been delayed because of equipment availability problems.
The required equipment has a
shipping date of March 8, 1982.
The modification to provide Lube Oil Cooli.ng from the discharge of the AFM pumps has been delayed due to equipment availability problems.
All equipment is on site with tlie exception of a relief valve that has a shipping date ot June 1983.
'It is intended that the lube oil modi fication be made within one month of receipt ot the valve.'he automatic t'low control modifications for Unit 3 will be completed prior to startup t'rom the steam generator repair outage.
C.
LONG TERN ITEMS The steam and feedwater piping modit'ications to insure-redundancy in what are now common sections of piping are planned tor installation and operability sometime in 1983.
A more precise date cannot be given at this time.
The new auxiliary feedwater control valves are expected to be Shipped by July 2, 1982.
It )s our intent to pertorm the majority ot the modifications tor both units during the Unit 4 steam generator repair eftort." It should also be noted that some ot
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1,
the modi tications can only be pertormed while both plants are shutdown.
It is also our intent to phase in the new equipment and not replace it all at one time.
Removal of non-seismic piping from tr>e suction lines t'or Unit 3 is planned to be completed prior to startup trom the steam generator repair outage.
The modification on Uni.t 4 is planned to be done during its steam generator repair outage.
9.
AFW INITIATION AND FLOW (II.E.l.~2.
2~C The moditication for satety
- grade, redundant flow indication rras been completed for Unit 4 and wil I be completed on Unit 3 prior to startup from steam generator repair outage with the t'ol lowing exception.
Uue to an oversight, the power supplies for the tlow indication and t'low control are not environmental ly and seismical ly qualitied.
gualifi,ed power supplies will be installed as soon as possible.
10.
CON'I'AINMENT ISOLA I'ION DEPENDABILITYQI I.E.4. 2)
A.
In our letter L-80-419, dated December 26, 1980 we stated that some of our shorter term modifications original,ly planned for completion by January 1,
1981 have been rescheduled due to longer then expected lead times tor the delivery of safety related valves.
These moditications (described in L-80-88, dated triarch 19, 1980) have been completed for Turkey Point Unit 4.
The modifications wil I be completed on Uni.t 3 prior to startup trom the steam generator rema> r outage.
B.
A Satety Evaluation Report enclosed in a letter tram S.
A. Varga to R.
E. Uhrig dated August 31, 1981 concluded that the requirements of Item II.E.4.2(5) of NUREG-0737, with th<< additional guidelines
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developed by the statt, have been met for the Turkey Point Units.
We therefore consider Item II.E.4.2. (5) complete with no modifications necessary.
11.
ADDITIONAL ACCIDENT-MONITORING INSTRUMENTAT~ION I I~F.I NOBLE GAS IIONITORS/IODINE, PARTICULATE SAMPL~ING II.F.1~1
.AND II.F.~22 )
A.
All of the plant effluent monitors have been installed as of January 1,
1982.
The plant vent effluent monitor may requi re additional modi fications to provide i soki netic sampling.
Our engineering department is currently evaluating whether a modification is necessary.
If the modification is determined to be necessary, we will inform you when it is complete and implemented.
Although the effluent.monitors have been installed, the system must yet be tested and calibrated, procedures written and our technicians trained by the manufacturer in its proper use.
We intend to have the system operational (except for i sokinetic sampling in the plant vent) by March 1, 1982.
We will submi t our proposed technical specifications to you prior to that date.
B.
CONTAINMENT HIGH RANGE RADIATION MONITOR@II.F.lb33)
The containment high range radiation monitors have been instal led in Turkey Point Unit 4 and will be insta'lied in Unit 3 prior to the startup from the current steam generator repair outage.
Our proposed technical specifications wi'll be submittedI to you prior to March 1, 1982.
C.
CONTAINIIENT PRESSURE IIONITOR I I ~F.I 4
)
The wide range portion (0-180 psig) of the containment pressure monitors has been installed in Turkey Point Unit 4 and will be i nstal'led in Unit 3 prior to startup from the steam generator repair outage.
Delivery of the transmitters for the vacuum portion
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(0 to -5 psia) of the system is currently scheduled for May 22, 1982.
The installation of the transmitters is not outage related and will be done promptly following receipt of the equipment.
Our proposed technical specifications for the monitors will be submitted to you once the system is installed.
D.
CONTAINMENT WATER LEVEL MONITOR I I.F.1'
)
The containment water level monitors have been instal led and are operational in Turkey Point Unit 4 and will be instal led in Unit 3 prior to the startup from tne steam generator repair outage.
Our proposed'echnical specifications f'r the monitors will be submitted prior to March 1, 1982.
E.
CONTAINMEHT HYDROGEN MONITORS II.F. 1 6
The containment hydrogen monitors have been partial ly instal led in both Turkey Point Units 3 and 4.
The delay in installation has b~en caused by equipment delivery problems.
It is our intent to have the system installed and operational by March 1, 1982 with the exception ot heat tracing required on the inlet sample lines.
The heat tracing is not expected to be shipped by the manufacturer until April 24, 1982.
The instal,lation of the heat tracing is not outage related and will be installed promptly foll.owing receipt.
Technical specifications for the monitor will. be submitted to you once the system is c'ompletely installed.
12.
INSTIIUI<ENTATION FOR DETECTION OF
~INADE UATE CORE COOLING {II.F.2)
INSTALLATION OF LEVEL INSTRUMENTATION Purchase Orders for tne C-'E designed heated junction. thermocouple system have been-issued to the appropriate vendors.
It is our i ntent to have the system installed and operational following the first refueling
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outage for each unit arter January 1,
1983 dependent upon equipment avail abi 1 ity.
13.
THERMAL MECHANICAL REPORT--EFFECT UF HIGH I RESSURE INJECTION OH VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF-COOLANT ACCIDENT WITH HO AUXILIARY FEEDNATEK (II.K~2.13 This item requires a detailed analysis of the thermal-mechanical conditions in the.reactor vessel during recovery from smal I breaks with an extended loss of al I feedwater.
Westinghouse (in support ot the Westinghouse Owners Group) is performing an analysis for generic Westinghouse.plant groupings to address this issue which is scheduled to be submitted to the HRC by the end of 1981.
This generic study will be applicable-to Turkey Point Units 3 8
4 and wi I 1
be referenced ds necessary to completely address HRC concerns.
14.
POTENTIAL FOR VOIDING 'IH THE RCS DURING TRANSIENTS+I I.K.2~17 Westinghouse
{in support of the Westinghouse Owers Group) has performed a
study.,which addresses the potential for void. formation in Westinghouse designed nuclear steam supply systems during natural circulation cooldown/depressurization transients.
Th'is study has been submitted to the NRC by tiie Westinghouse Owners Group (Letter OG-57, dated April Zu, 1981 from R.
W. Jurgensen to P.S.
Check).
In addition, the Westinghouse Owners Group has developed a natural circulation cooldown guideline that takes the results of the study into account so as to preclude void formation in the upper head region during natural circulation cooldown/depressurization transients, and specifies those conditions under wh>ch upper head voiding may occur.
These Westinghouse Owners Group generic guidelines have been submitted to the NRC (Letter OG-43 dated November 30, 1981 from R
W. Jurgensen tu D.
G.
Ei senhut).
The generic guidance developed by the Westinghouse Owners Group (augmented as appropriate with plant specific cons)derat]on) has
41 A
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been utilized in the preparation of the Turkey Point plant specific operati ng procedures as described in L-81-513, dated December 4,
1981.
15.
SEQUENTIAL AUXILIARY FEEDWATER;FLOW ANALYSIS (II. K.2.1+9 Subsequent to the issuance of NUREG-0737 and as documented in a letter dated July 1, 1981 from S.
A. Varga to R.
E. Uhrig, the NRC has completed a generic review on the this subject and concluded that the concerns expressed in Item II.K.2.19 are not applicable to NSSSs with inverted U-tube steam generators such as those designed uy Westinghouse.
Therefore, this item is not applicable to Turkey Point and no tuther action is necessary.
16.
AUTOl'1ATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT Westinghouse (in support of the Westinghouse Owners Group) has performed an analysis of delayed reactor coolant pump trip during sma'l l-break LOCAs.
This analysis is documented in 'Reference 1.
In addition, Westinghouse (again in support of the Westinghouse Owners. Croup) has performed test predictions ot LOFT Experiments L3-1 and 'L3-6.
The results of these predictions are documented in References 2,3, and 4.
Based on:
- 1) the Westinghouse
- analysis,
- 2) the excellent predict,ion of the LOFT Experiment L3-6 results using the Westinghouse analytical
- model, and
- 3) Westinghouse simulator data related to operator response time,, tne Westinghouse and Florida Power 5 Light Company position is that automatic reactor coolant pump trip is nut necessary since surtic>ent time is available for manual tripping of the pumps.
Our understanding of tiie schedule for rinal resolution of this issue is:
A.
Once the NRC formally approves the Westinghouse
- model, a 3-month study period will ensue during which, the Westinghouse Owners Group
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wil k attempt to demonstrate compliance witty some NRC acceptance criteria for manual RCP trip.
The NRC acceptance criteria we'll accompany tkieir formal approval ot the West>ngk>ouse models.
B.
If,. at the end of the 3-month period, the Westinghouse Owners Group cannot show compliance wi th tne acceptance
- crsteria, tne NRC wi 1k formally notify uti 1-ities that they must submit an automatic RCP trip design.
This letter expands our position transmitted to you on August 6, 1981 (Letter L-81-343).
References:
( 1)
"Analysis of Delayed Reactor Coolant Pump Trip During Small Loss of Coolant Accidents for Westinghouse Nuclear Steam Supply Systems,"
WCAP-9584 (proprietary) and WCAP-9585 (non-proprietary), 'August 1979.
(2)
Letter OG-49, dated f1arch 3, 1981, R.W. Jurgensen (Cha,irman, Westinghouse Owners Group) to 0. F. Ross, Jr.
(NRC).
(3)
Letter OG-SV, dated flarch 23, 1981,.
R.
W Jurgensen (Chairman, Westinghouse Owners Group) to D. F.
Ross, Jr.
(NRC).
(4)
Letter OG-60, dated June 15,
- 1981, R.
W. Jurgensen'Chairman Westinghouse Owners Group) to P.S.
Check (NRC).
17.
EFFECT OF LOSS OF AC POWER ON PUt4P SEALS III.K.3.2+5 This item requires tkIat the consequences of a loss of RCP seal cooling due to a loss of AC power (defined as loss of offsite power) for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be demonstrated.
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During normal operation, seal injection flow from the chemical and volume control system is provided to cool the RCP seals and the component cooling water system provides flow to the thermal oarrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals.
In the event of loss of oftsite power the RCP motor is deenergi zed and both of these cooling supplies are terminated; however.,
the diesel generators are automatically started and'ither seal injection flow or component cooling water to the thermal barrier heat exchanger is automaticai ly restored within seconds.
Either of these cooli ny supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during a loss of offsite power tor at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
18.
REVISED St<ALL-BREAK LOCA METHODS TO SHOW COMPLIANCE WITH iOCFRb0, APPENDIX K (II.K.3.30)
This item requires that the analysis methods used by NSSS vendors and/or fuel suppliers for small-break LOCA analysis for compliance with Appendix K to 10 CFR Part 50 be revised, documented, and submitted ror NRC approval.
Westinghouse feels very strongly and Florida Power 5 Light Company agrees that the small-break LOCA analysis model currently approved by the NRC for use on Turkey Point is conservative and in conformance with Appendix K to 10 CFR Part 50.
- However, (as documented in Reference 1)
'Westinghouse believes that improvement in the realism of smali-break calculations is a worthwhile effort and has committed to revise its small-break LOCA analysis model to address NRC concerns (e.g.,
HUREG-
- 0611, NUREG-0623, etc.).
This revised Westinghouse model is currently scheduled ror submittal to the HRC by April 1, 1982 as documented in Reference 2.
(1)
Letter HS-TI1A-2318, dated September 26,
- 1980, T.
H. Anderson (Westinghouse) to D.
G. Eisenhut (NRC).
(2)
Letter HS-EPR-2524, dated November 25,
- 1981, E.
P.
Rane (Westinghouse) to D.
G. Eisenhut (HRC).
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ENCLOSURE 1
AHF SYSTEM FLOH RATE OESIGN BASES ANO CRITERIA In the question and answer format that follows,. the questions are taken rrom enclosure 2 of the NRC letter of October 16, 1979.
12
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Question 1
p transient and accident conditions con
'd d
lashing AFMS flow requirements, including th fol 1)
Loss of iiain Feed
(~NF.I) 2)
L~iFW w/loss of offsite AC power 3)
Li~FM w/loss of onsite and offsite
~C power 4)
Plant coo ldown 5)
Turbine trip with and without bypass 6)
<lain ste m isolation valve closure 7) iMain feed line break 8)
Main steam line break 9
Smal 1 break LOCA
- 10) Other transient or accident conditions not lis ed above.
b.
Oescribe the plant protection acceptanc crtteria and corresponding technical bases used for each initiating event identified above."
The acceptanc criteria should address plani limits such as:
1)
Maximum RCS pressure (POR~/ or safety valve actuation) 2)
Fuel tempera ure or damage limits (ON8, PCT, maximum fuel central temperature) 3)
RCS cooling rate limit to avoid exc ssive coolant shrinkage 4) i~linimum steam generator level to assure sufficient steam genera-tor heat transfer surface to remove decay heat and/or cool down ihe pf imary system.
Response
to ~.a Tne Auxiliary Feedwater System serves as a backup system for supplying feedwater to 'the seconoary side of the steam generators at times when the feedwaier system is not available, thereby maintaining the heat sink capabilities of the steam generator.
As an Fngine red Safeguards Sys-tem, the Auxiliary Feedwater System is directly relied upon to prevent core damage and system overpressurization in the event of transients such as a loss of normal feedwater or a secondary system pipe rupture, and to provide a means for plant cooldown following any plant transient.
Following a reactor trip, decay heat is dissipated by evaporating water in the steam generators and venting the generated steam either to the condensers through the steam dump or to the atmosphere through the steam generator safety valves or the power-operat d relief valves.
Steam generator water inventory must be maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat removal process.
The water level is maintained under these circumstances by the Auxiliary Feedwater System which delivers an emergency wat r supply to the steam generators.
The Auxiliary Feedwater System must be capable of functioning for extended
- periods, allowing time either to restore normal feedwater flow or to proceed with an orderly cooldown of the plant to the reactor coolant temperature where the Residual Heat Removal System can assume the burden of decay heat removal.
.The Auxiliary Feedwater
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I System flow and e'i~rgency water supply capaci z
t be sufficient to remove core deca~eat, reactor coolant jump heat, and sensible heat during the plant cooldown.
The Auxiliary Feedwat r System can also be used to maintain the steam generator water levels above the tubes fol-lowing a LOCA.
In the latter function, the water nead in the steam generator s serves as a barrier to prevent leakage of fission products from the Reactor Coolant System into the secordary plant.
OESIGH COiHDITIOi'IS The reactor plant conditions which impose safety-related performance requirements on the design of the Auxiliary Feedwater System are as follows for Tur!<ey Point Uni s
3.and 4
Loss of Main Fe dwater Transient Loss of main feedwater with offsite oower available Station blackout (i.e., loss of main feedwat r'ithout offsite power available)
Ruptur of a Main Steam Line
'oss of all AC Power Loss of Coolant Accident (LOCA)
Cooldown Loss of Main Feedwater Transients The design loss of main feedwater transients are those caused by:
Inte. ruptions of the Hain Feedwater System flow due to a malfunction in the feedwater or condensate system Loss or offsite power or blackout with the consequential shutdown of the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a rapid reduction in steam generator water levels which results in a reactor trip, a tur-bine trip, and auxiliary feedwater actuation by the protection system logic.
Following reactor trip from high power, the power quic!<ly fal'Is to decay heat levels.
The water levels continue to decrease, progres-sively uncovering the steam generator tubes as decay heat is tr ansferred and discharged in the form of steam either through the steam dump valves to the condenser or through the steam generator safety or power-operated relief valves to the atmosphere.
The reactor coolant temperature increases as the residual heat in excess of that dissipated through the steam generators is absorbed.
With increased temperature, the volume of reactor coolant expands and begins filling the pressurizer.
Without the addition of sufficient auxiliary feedwater, further expansion will result in water being discharged through the pressurizer safety arid relief valves.
If the temperature rise and the resulting volumetric.
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expansion of the priapuscoolant are permitted to cont%pe, then (1)
> pressurizer safety valve capacities may be exceeded causing overpres-surization of the Reactor Coolant System and/or (2) the continuing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually in core uncovering, loss of natural circulation, and core damage.
If such a situation were ever to occur, the Emergency Core Cooling System would be ineffectual because the primary coolant system pressure exceeds the shutoff head of the safety injection pumps, the nitrogen over-pressure in the accumulator
- tanks, and the design pressure of the Residual Heat Removal Loop.
Hence, the timely iJitroduction of sufficient auxi li'ary feedwater is necessary to arrest the decrease in the steam generato'r water levels, to reverse the rise in reactor coolant temperature, to prevent the pres-surizer from fillin'g to a water sol'id condition, and eventually to establish stable hot standby conditions.
Subsequently, a decision may be made to proceed with plant cooldown if the problem cannot be satis-factorily corrected.
The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equip-ment.
The loss of power to the electric driven condenser circulating water pumps results in a'oss of condenser vacuum and.condenser dump valves.
- Hence, steam formed by decay heat is relieved through the steam generator safety valves or the power-operated relief valves.
The calcu-lated transient is similar for both the loss of main feedwater and the
- blackout, except that reactor coolant pump heat input is not a consider-ation in the blackout transient following loss of power to the reactor coolant pump bus.
.l The Loss of tdain Feedwater transient serves as the basis for the minimum flow required fear the smailest capacity s-;ngle auxiliary feedwater pump for Turkey Point Uni ts 3 and 4.
The pump is sized so that any single pump will provide sufficient flow against the steam generator safety vtilve set pru'"ure (with 3~ accumulation) to prevent water relief free the pressurizer.
The same criterion is met for the Station Blackout transient, where A/C power is assumed to be unavailable.
Rupture of a t1ain Steam Line Because the rupture of a main steam 1'inc may result in the complete blowdown of one steam generator, a partial loss of the plant heat sink is a concern.
The main steamline rupture accident conditions are char-acterized initially by plant cooldown, and hence, auxil'iary feedwater f1ow is not needed during the early stage of the trans'ient to remove decay heat from the Reactor Coolant System.
Provisions must be made ii the design of the auxiliary feedwater sys em to al1ow termination of flow to the faulted loop and to provide flow to the intact steam genera-tors during the controlled cooldown following the steamline break acci-dent.
4~
~ ~
Loss of All AC Power The loss of all AC power is postulated
<.b Table 18-1 summariz s the criteria which are the general design bases for each
- event, discussed in the response t" question l.a, above.
Specific assumptions used in th~ analyses to verify that tho design bases are met are discussed in response to guestion 2.
The primary function of the Auxiliary Feedwater System is to provide sufficient heat removal capability for heat' accidents following reactor trip to remove the decay heat gener ted by the core and prevent system overpressurization.
Other plant orotection systems are designed to meet short term or pre-trip fuel failure criteria.
The effec.s of excessive coolant shrinkage are bounded by the analysis of the rupture of a main steam oipe transient.
The maximum flow reouirements deter-mined by other bases are incorporated into.his analysis, resulting in no additional flow requirements.
~
~
uestion 2
Oescribe the analyses and assumotions and corresponding technical justification used with plant condition considered in 1.a above including:
a.
Maximum reactor power {inc!uding instrument error allo.varce) at the time of the initiating tr nsient or acc..d nt.
b.
Time delay from initiating event to reactor trip.
c.
Plant parameter(s) which initiates AFMS flow and time delay between initiating event and introduction o
ASS flow into st am genera-tor(s).
d.
Minimum steam gene. ator water level wnen initiating event occurs.
e.
Initial steam generator water inventory and depletion rate before and after AF'.tS flo:v cc,
...ences identi.y reactor decay heat rat used.
f.
Maximum pressure at which steam is released frcm steam generator(s) and agains-which th AF'rl pump must develop sufficient head.
g.
Minimum number of steam generators that must receive AFM flow; e.g.,
1 out of 2?
2 out of 47 h.
RC flow condition continued ooeration of RC pumps or natural circul at ion.
i.
Maximum ATM inlet temperature.
j.
Following a pos ulated st am or feed line break, time delay assum d
. to isolate break and direct AFM flow to intact steam generator( s).
AFM pump flow capacity allowance to accommodate the time delay and maintain minimum s team generator water level.
Also identify credit taken for primary system heat removel due to bl'owdown.
k.
Yolume and maximum temperature of water in main feed lines between steam generator(s) and AFMS connection to main feed line.
l.
Operating condition of steam generator normal blowdown following initiating event.
m.
Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
n.
Time at hot standby and time to cooldown RCS to 'RHR system cut in temperature to size AFlA water source inventory.
ii I
'7 II, II
Analyses have been performed for the Loss of Main Feedwater and the loss of offsite AC power to the Station, the transients which define the AFWS performance requirements.
These analyses have been provided for review and have been approved in the Applicant's FSAR.
In addi tion to the above analyses, calculations have been performed specifically for Turkey Point Units 3 and 4 to determine the plant cocl-down flow (storage capacity) requirements.
The LOCA analysis, as dis-cussed in response 1.b, incorporates the system flow requirements as defined by other transients, and therefore is not performed f'r the purpose of specifying AFWS flow requirements.
Fach of the analyses listed above are explained in further Detail in the following sections of this response.
Loss of Hain Feedwater (81ackout)
A loss of main feedwater analysis was performed in FSAR Section 14.l.ll.
for the purpose of showing that a single auxiliary feedwater pump delivering flow to two'team generators does not result in filling the pressurizer.
Furthermore, the peak RCS pressure remains below the criterion for Condition II transients and no fuel failures occur (refer to Table 18-1).
Table 2-1 summarizes the assumptions used in this analysis.
The transient analysis begins at the time of reactor trip.
This can be done because the trip occurs on a steam generator level
- signal, hence the core power, temperatures and steam generator level at time of reactor trip do not depend on the event sequence prior to trip.
Although the time from the loss of feedwater until the reactor trip occurs cannot be determined from this analysis, this delay is expected to be 50-60 symonds.
<he analysis assumes that the plant is initially operating at 102K (calorimetric error) of 2300 MWt.
A very conservative assumption is made in defining decay heat and stored energy in the RCS.
Thu reactor is assumed to be tripped on low-low steam generator
- level, allowing for level uncertainty.
The FSAR shows that there is a
considerable margin with respect to filling the pressurizer.
A Station Blackout transient with the assumption that the smallest auxi liary feedwater pump operates results in even more margin.
This analysis establishes the capacity of the smallest single pump and also establishes train association of equipment so that this
~: alysis remains valid assuming the most limiting single failure.
Plant Cooldown Minimum flow requirements from the previously discussed transients meet the
, low requirements of plant cooldown.
This operation,
- however, defines the basis for tankage
- size, based on the required cooldown
- duration, maximum decay heat input and maximum stored heat in the
4
'!lt I
0 system.
As. previously discussed in response
'.A, the auxiliary feedwater system partially cools the system to the point where the RHRS may com-plete the cooldo.~n,, i.e.,
350oF in the RCS.
Table 2-1 shows the assunpticns used to determi'ne the cooldown heat capacity of the auxi-liary feedwater system.
The cooldown is assured to ccmnence at 2345
!<Mt power, and maximum. trip delays and decay heat source terms are assuned when tlute reactor is tr ipped.
Primary metal, primary water, secondary system metal and sec-ondary system water are all included in the stored heat to be removed by the AF'AS.
See Table 2-2 for the items const',tuting the sensible heat stored in the NSSS.
This operation is analyzed to establish minimum tank siz r quirements for auxiliary feedwater fluid source which are normally aligned.
0 4,
UESTION 83 Verify that the APW pumps in your plant will supply the necessary flow to the steam generator(s) as determined by items 1 and 2 above considering a single failure.
Identify the margin in sizing the pump flow to all.ow for pump recirculation fl'ow, seal. leakage and pump wear.
RESPONSE
TO f13 Figure 3-1 schematical'ly shows the major: features and components of the Auxiliary Feedwater System for Turkey Point Units 3 and 4.
Plow rates for the design transients described'n
Response
2 are tabulated in Table 3-1 considering the following single fai'lures.
A.
B.
A/C Train Failure Pump Failure C.
APTS Flow Control Valves Failure (failure to assume proper preset position)
Hodifica tions being made to the system, which include automatic flow control and redundant H.ow,paths, vill automatically provide.a minimum of 200 gpm to each steam generator.
NOTE: Figure 3-1 does not reflect the proposecl modifications.
The Turkey 1%int Units 3
& 4 auxiliary feedwater pumps were procured to supply a net flow of 600 Fpm at 1191 psia.
The minimum recirculation flow angn Criteria Loss of Hain Feedwater Condition II Peak RCS pressure not to exceed design Pressure.
Ho consequential fuel failures Station Hlackout Loss of all A/C Po~~er Loss of Cool ant Cool down Condition II H/A Condition III Condition IV (same as LHFlt) t<ote 1
10-CFR 100 dose 1 admits 10 CFR 50 PCT lim)ts 10 CFR 100 dose limits 10 CFR 30 PCT limits Pressurizer does not fillwith 1 single aux.
feed pump feeding 2 SGs.
100oF/hr 547oF to 350oF
-Ref:
At)SI H18.2 (This information provided for those transients performed in the FSAR).
f
'ote 1
Although this transient establishes the basis for Aft/ pump powered by a diverse power source, this is not evaluated relative to typical criteria since multiple failures must be assumed to postulate this tra>>-
sient.
4t
~ C
TABLE 2-1 SQ?glary of Assumpti ons Used in AF hS Design Yerification Analyses t
Transient a.
Max reac.or power b.
T ime de 1 ay from trip signal to rod mot.'on c.
AF'AS actuation sig-nal/time delay for AFMS flow d.
SG water level at time of reactor trip Loss of Fe dwat r station bl ac<out
- 102~~ of 2300 Mwt 2 sec lo-lo SG level/
'3 minutes lo-lo SG level
'(lo SG lev 1)
Coolooro 2345 t&t 2 sec e.
Initial SG inventory Rate of change before ai ier AF'AS actuat ion 37 314 1 Lm/SG a i. trip 37 200 1 bli/SG (SZ,COO Ibm/SG at trip) 8 516oF Se FSAR Section 14.1.11 and 1!r1.12 decay heat f.
AFW pump d sign g.
Minimun -,'f SGs which must receive AFM flow AHS + 20" 1133 ps-ia 2of3 1'33 psia N/A h.
RC pump status Tripped 9 reactor trip Tripped i.
Maximum AF'ii temperature j.
Operator action 120oF N/A 100oF l.
informal bl owdown m.
Sensible heat n.
Time at s andby/time to cooldown to RHR none assuned see cooldown 2 hr/4 hr k.
MFM purge volume/temp.
182 ft3/440oF 450 f"3/
440oF none assumed Table 2-2 2 hr/4 hr o.
AFA flow rate 600 GPM - constant (min. requi renent) variable
- Yalue shown only if different from Loss of Feedwater
Oi
- (j
TABLE 2-2 Surnnary of Sensible Heat Sources P, imary Mater Sources (initially at 2345HMt power temperature and inventory)
RCS fluid
- Pressurizer fluid (liquid and vapor )
Primary Metal Sources (initially at 2345 Yr/t power temperature')
- React"r coolant oiping, pumps and reactor vessel
- Pressurizer
- Steam generator tube metal and tube sheet Steam generator metal below tube she
- Reactor vessel internals Secondary Mater Sources (initially at 23!5 NMt power temperature and inventory)
- Steam generator fluid (liquid and vapor)
- tlain reedvater purge fluid between steam generator and AF~rlS piping.
Secondary i'fetal Sources (initially at'3!5i'1Mt pouter temperature)
- All steam generator metal above tube she t, excluding tubes.
~k "i
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TABLE 3-1 AUXILIARYFEEDWATER FLOW (1)
TO STEAN GENERATORS FOLLOWING AND ACCIDENT/TRANSIENT WITH SELECTED SINGLE FAILURE GPH Accident/Transient Single Failure Elec. A.C.
Train Failure Pump Failure CV(2)
Failure 1.
Loss of Hain FM 2.
Blac ko<< t 3.
Coo idown 600 (3) 600 (3) 600 600 (3) 600 (3) 600 C
600 600 600 NOTES:
2 ~
3.
Items I thru 3 are minimum expected flows to intact loops.
Including only those CVs in the AFWS.
"Failure" is defined as failure of the valve to assume its proper preset controlled position.
Flow is automatically initiated and controlled to 200 gpm to each steam generator.
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0 0
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1 f
is Attachment foal MESTINGHOUSE NSO CERTIFICATE OF CONFORMANCE Exceptions to Florida Power and Light Company Purchase Order 893000-81417 and OMA 438608 with regards to the NRC questions 1 through 3 on AFMS flow require-ments design basis information.
Mestinghouse-Did not specifically address events (5), (6) and (7) in its response to question 1 Section A.
2.
Did not specifically address events (5), (6), (7) and (8) in its response to question 1 Section B.
3.
Did n'ot, in providing plant protection acceptance criteria, speci'fically address the four "plant limits" cited in guestion 1 Section 8 for each event.
(
Reference:
Table 1B-1) i Did not address events
{2'j, (3), (5), (6), (7), (8) and (9) in its response to question 2.
Did not address margin in sizing the pump flow to allow for recirculation flow, etc.
Justification of exceptions taken to Florida Power and Light Company's order for providing the AFMS design basis flow require.-ents-responses to NRC questions:
The transients and conditions resulting frcm events (5) and (6) which impose safet;-related performance requirements on the design of the AFMS are bounded by the transients and conditions resulting from other events which were specifically addressed by Mestinghouse in its response to question 1 Section A.
Although event (7) typically serves as a design basis for plants which must meet today's licensing requirements, it is not part of the design basis for this plant and is not discussed in the response to question 1
Section A.
2.
Same justification as in (1) above.
See paragraph 3 below for justification for omitting (8) in the response to guestlon 1 Section B.
The "plant limi-ts" cited in the NRC question which should be addressed by the acceptance criteria were cited as examples of plant limits (Note the use of t.he phrase "such as") to be addressed and not a requirement.
In two cases, the so-called "plant limits" cited in Section 8 (Items 3 and
- 4) are not, plant limits under any recognized licensing basis and are somewhat undefined as stated -'.g., in our opinion:there is not such thing as a plant limit on "RCS cooling rate to avoid excessive coolant shr'.nkage" and t,he meaning of "excessive" is, subject to debate in this context.
4~
~ I The Westinghouse response to this question provided the plant protection acceptance criteria which were part of the original design and licensing basis for these events.
Where these criteria can be related to limits on specific equipment or systems, the response included these limits.
4.
a)
Discussion of event (2) was unnecessary in the response to question
. 2 because event (1) is more limiting with respect to defining minimum AFWS flow requirements than event
( 2).
This was noted in the response to question 1 Section,A.
Differences between event (1) and event (2) assumptions are shown in Table 2-1 however.
')
As was noted in the response to question 1 Section A, event (3) does
~
not impact the establishment of AFWS flow requirements.
Its impact is only in determining (or indicating) the necessity for power and control for an AFWS pump which is not dependent on AC power and can maintain the plant at hot shutdown until AC power is restored.
Since this event has no impact on determining AFMS flow requirements, it was not addressed in the response to question 2.
c)
As noted previously in (1) above,'events (5) and (6) are bounded by other events and thus do not determine the limiting (max/min) flow requirements for the AFWS and thus are not addressed in the response to question 2.
d)
As noted previously in (1) above, events (7) and (8) are not a part of the design basis for this plant and therefore are not addressed in question 2.
e)
Event (9),
as noted in the response to question 2, is not used in determining design basis flow requirments for the AFL/S.
5.
Mestinghouse did not address design margin in the response to question 3
because Westinghouse did not design the AFWS.
Westinghouse's role was to specify design criteria which the designer (customer utility or Architect Engineer) had to meet.