ML20154Q730
| ML20154Q730 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/22/1988 |
| From: | Richardson S NRC OFFICE OF SPECIAL PROJECTS |
| To: | Thadani A Office of Nuclear Reactor Regulation |
| References | |
| TAC-00450, TAC-450, NUDOCS 8810040050 | |
| Download: ML20154Q730 (1) | |
Text
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f f
I Docket No. 50 260 Distribution f
DesEE"WTi~
Projects Reading QSims September 22, 1988 NRC PDR OSP Reading AMarinos Local PDR DNoran/GGears SRichardsonl JWatt MEMORANDUM FOR: Ashok C. Thadani, Assistant Director t
for Systems j
Division of Engineering and System Technology Office of Nuclear Reactor Regulation
[
FROP:
Steven D. Richardson, Director l
TVA Projects Division j
Office of Special Projects t
SUBJECT:
REVIEW OF BROWNS FERRY NUCLEAR (BFN) PLANT, UNIT 2. TVA BFN TECHNICAL SPECIFICATION NO. 254 (TAC 00450)
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The Office of Special Projects (OSP) is requestino NRR review of (Tennessee ValleyAuthority)BFNTechnicalSpecification(TS)No.254whichupdatesthe
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BFN, Unit 2 TS to reflect the reactor core operating limits for the present l
core configuration.
The TVA submittal dated August 26, 1988, requesting arrendment to Operating License DPR.52 is included as Enclosure 1.
We have i
also included a work request as Enclosure 2.
i l
Proposed TS changes for the initial Cycle 6 reloa were submitted to NRC by i
TVA letter dated August 23, 1984 and were approveti by the staff Amendment i
No. 125 by letter dated August 19, 1986 (Enclosure C.
1 i
As you know, the Cycle 6 reload configuration originally approved by the staff i
has been reconstituted as a result of TVA's inspection and reconstitution I
i program which was completed in July 1988.
Because of this effort to improve l
j the quality of the core for Cycle 6. the resulting core configuration represents a change from that approved on August 19, 1986. We, therefore.
(
i request that you review the attached submittal and provide a safety evaluation I
i by January 1989, l
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}
This review schedule is believed to be reasonable based on the enclosed 1
docurentation received fron TVA in support of the TS change for the reload.
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}
Sg If you have any questions, please contact Dave Moran at 2 0766.
I j
ME l ll 5 MLtJ s -
g Steven D. Richardson, Director of TVA Projects Division i
Office of Special Projects 3
Enclosures:
I k 1.
Letter fron H. Ray dated l
August 26, 1988 I
[i 2.
Work Recuest 3.
Letter from M. Grotenhuis datedAugust]/,1986 g
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- sri con DATE :09/tA/88
- 09/
88
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- 09/2,1/A8
- 093.7488
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s l OFFICIAL RECORD COPY I
I Enclosuro 1 8 J' L43 t 3 M,
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TENNESSEE VALLEY AUTH8MITY d % - pI4 t^'u 6 i
CH ATT ANOQQA, TENNESSEE 374ot
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Alj0161988 10 CFR 50.90 TVA 3rW.TS 254 I
t,1 U.S. Nuclear Regulatory Cov ission ATTH:
Document Control Desk e;,
washington, D.C.
20555 Contiemen:
i in the Matter of
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Docket Nos. 50-260 Tennessee Valley Authority
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l 3ROWN3 FERRY NUCLEAR PLANT (BFN) - TV A BFN TECHNICA1. SPECIFICAT10W Wo. 254 L
In accordance with the provisions of 10 CFR 50.4 and 50.90. we are submitting O request for an amendment to operating license DFR-52 to change the SF'N f
l Technte41 specificationc for unit 2 (enclosure 1).
t These proposed changes (enclosure 1) will update unit 2 technical l
specifteations to reftvet the Redetor Core operating 1.imits for cycle 6 operatLons The initial cycle 6 reload proposed technical specification changes were l
submitted to NRC by TVA letter dated August 23, 1984, anJ was approved by the issuance of BFW Tvehnical Specification 199 dated August 19, 1986.
Since our i
initial subeltt41, the current cycle 6 fuel loading has changed as a result of
(
the fuct inspection and reconstitution program which was completed in 1
July 1988.
i The descriptlon, reason for chan6o, and justifisation are provided in enclosure 2.
A proposed deter.ination of no significant hazards consideration l
Le provided in enclosure 3.
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Enclosed is a checs for the $150 amendment fee required by 10 CFR part 170.12.
We request that these specifiestions be made effective 90 days after issuance because of the number of procedure revisions required.
Very ttviy yourc, TENWESSEE vat.t.EY AUTit0RITY
.h M.IJ. Ray, Manas -
Site Licensing staff I4 sworn')dWesylo aip s'ubse ed before me thi,d 1/&
988 2
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A/_ EN) iotary Public
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,N 11y Commission \\ Expires Enclorures cc See page 3 W-G 9
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o U.S. Nuclear Regulatory Consnission g gg g
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i Ms. S. C. Black, Assistant Director for Projects i
TVA Projects Division i
I U.S. Wuclear Regulatory Comission one White Plint, North l
11555 Rockville Pike Rockville, Maryland 20852 Mr. P. R. McCoy, Assistant Director f
for Inspection Programs y
TVA Projects Division j
U.:. Wuclear Regulatory Comission toglon 11 101 Marietta Street, NW, Suite 2900 Atlanta, coorgia 30323 l
?
Browns Ferry Resident Inspector I
Browns Terry ttuclear Plant l
Route 12, Box 637 Athens, Alabar.a 35611 l
t Mr. Charles R. Christopher i
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Chair.an, Limstone County Comission l
P.O. Box 188 Athens, Alaba.a 35611 l
Dr. C. E. Fox 1
State Health officer
(
l State Departeent of Public Health State Office Building Montgom ry Alabara 36104 Mr. J. E. Jones General Electric Company l
j No. 1 Union Square i
808 Krystal Building Chattanooga, Tennessee 37402
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TECHNICAL SPEC 1r! CATION CMANCES t
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l BROWS TERRY UNIT 2, CYCLE 6 l
BASED CW l
380WS FttRY WUCLEAR PLANT RELOAD LICEWSINC REPORT l
t UNIT 2. CYCLE 6 j
TVA-RLR-002 REVIS10M 2 6
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L S /4.5 C0ff AND CONTAINMENT cocLING SY3 TEM 3 LIMITING CCNDITIONS FOR OPERJ.T!CN SURVE!LLANCEREQUIRhENTS 3.5.!
Averate Planar Linear Heat 4.5.1 Maximum Averane Planar GeneratJJn Rate Linear Heat Generation Rate (MAPLHCR)
During steady-state. power operation.
The MAPLHGR for each type of the Maximum Average Planar Linear fuel as a function of average Neat Generation Kate (MAPLHGR) for planar exposure shall be i
sach type of fuel as a function of determined daily during reactor I
average planar exposure shall not operation at 1 25% rated exceed the limiting value shown in thermal power, t
Tables 3,5.!-1, 2, 3, and 4 If at l
any time during operation it is determined by normal surveillance i
that the liatting value for APLHOR is being exceeded, action shall be l
initiated within 15 minutes to l
restore operation to within the l
prescribed 11raits.
If the AFLHOR is I
not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillancy and t
corresponding action shall sentinue until reactor ope' ration is within l
the prescribed linits.
J. Linear Hest Generatien Rate (LMGRS J.
Linear Heat Generatien Pate (LHCR) i During steady-state power operation.
The LHOR shall be checked the linear heat generation rate (LEGR) daily during reactor fuel I
of any rod in any fuel assembly at any operation at 2. 25% rated I
axial lecatien shall not exceed 13.4 ther=al power.
kW/tt.
If at any time during l
operaticr. it is determined by norsal surveillance that the liatting value for 1. HOR is being exceeded, action shalt be initiated within 15 minutes l
l to restore operation to within the t
l Prescribed limits.
If the !JGR is j
l not returned to within the prescribed i
limits within two (2) hours, the l
reactor _shall.be brous.ht.to,the l
Cold Shutdovn condition within 36
= = _
hours.
Surveillan:e and l
correspending action shall continue
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untti reactor cperaticn is within l
the prescribed linits.
l E T'i 3.3/4.!-13 Unit i
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Table 3.5.I-1 MAPLHOE VER3US AVERACE PLANAR. EXPOSURE
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MAPLHCR Average Planar (kW/ft)
Enosure (mwd /T1 11.2 200 11.3 1,000 11.8 5,000 12.0 10.000 12.0 15,000 11.8 20,000 11.2 25.000 30.000 10.8 10.2 35,000 9.5 A0.000 4.8 45.000 Tablo 3.5.I-2 M FLHCR VERIUS AVERACE PLAMAX EXPOSURE Tual Types T 3DR324,5.1 HAPLHCR Average Planar Exisomt.tra (mwd /T)
(kW/ft) 200 11.5 f
11.6 1,000 5,000 11.9 f
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10,000 12.1 15,000 12.1 20.000 11.9 25,000 11.3 30,000 10.7 35,000 10.2 9.6 40,000
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MAPLMOR VER3US AVERACE PLANAA EXPOSURE -
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Fuel type:
P8DRB284
- Avera5e' P1anar ~.....---..w...--------.-------.
3 MAP'.H O R rreosure (WWd/?)
(kV/ft) i 200 11.2 1.000 11.2 5.000 11.7 10.000 12.0 15.000 12.0 20.000 11.s 25.000-li.;
30.000 10.4 25.000 g.8 40.000 g,1 45.000 g.5 i
Table 3.3.;.4 1
'* MAPLHOR VERSU3 AVERACE PLANAR EXPOSURI I
Fuel Type
- 8DRE284L fe l'
1 Average Planar KAPLMOR Er2epy're fMWe/7)
IkV/ft) 200
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11.3 3.000 11.s 10.000
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10.0 20.000 ti.g 25.000 11.2
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Table 3.5.I-1 KAPLHCR VERSUS AVERACE PLANAR EXPOSURE Fuel Type:
PSDRB284L/ QUAD +
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Average Planar KAPLHCR E xt e r.u re (mwd /T)
'(kW/f t) 200 11.2 1,000 11.3 5,000 11.8 10,000
.2.0 15,000 12.0 11.8 20,000 25,000 11.2 30,000 10.8 35,000 10.2 40,000 9.5 45,000 8.8 Table 3.5.I-2 i
MAFLHCR VERSUS AVERACE PLANAR EX?cSURE Fuel Type:
P80RB265H Average Planar KAPLHOR Extesure (mwd /T)
IkW/ft) 200 11.5 i,000 1*.6 5,000 1 *. 9 10,000 12.1 15,000 12.1 20,000 11.9 25,000 11.3 e
i 30,000 10.7 I
25,000 10.2 40,000 9.6 I
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TAU MCPR Umits forhure 3.5.K-18 X BR/8 X 8R/ Q F1
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The peck cledding temperature following a postulated Icac-of2 ooltnt p
accident is primarily a function of the average heet asneration rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod-to-rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than 20*T relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix K limit.
Th: limiting value for MAPLHGk is shown in Tables 3.5.I-1, 2, I
3, and 4 The analyses supporting these limiting values are presented in Reference 1.
3.5.J. Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated.
The LHGR shall be checked daily during reactor operation at 125 percent power to determine if fuel burnup, or control rod movement has caused changes in power distribution.
For LHGR to be a limiting
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l value below 25 percent rated thermal power, the R factor vould have to l
be less than 0.241 which is precluded by a considerable margin when employing any per=issible control rod pattern.
l 3.5.K. Minimum Critical Power Ratio (MCPEl At core thermal power invels less than or equal to 25 percent, the reactor vill be operating at minimum recirculation purp speed and the moderator void content vill be very small.
For all designated control rod patterns which.,4y be e= ployed at this point, operating plant experience tnd thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With l
this lov void content, any inadvertent cort flow increase would only l
place operation in a = ore conservative mode relative to MCPR. The l
daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shif ts are very slev vhen there have not been significant power or control rod changes. The require =ent for calculating MCPR vhen a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
3.5.L. APRM Setooints ostration.is constrm4 W to A envirus 111G2. o f 1MkW/.f t for.8x8. fuel _ ___ __.]
This limit is reached when core maximum fraction of li=iting power density (CMFLPD) equals 1.0.
For the case where CMTLPD exceeds the fraction of rated thermal power, operation is per=itted only at less than 100-percent rated power and only with APRM scra= settings as required by Specification 3.5.L.1.
The scram trip setting and rod block trip setting are adjusted to ensure that no combination of CMTLPD and FRP vill increase the LHGR transient peak beycnd that allowed by the 1-percent plastic strain limit. A 6-hour time period to achieve this condition is justified since the additienal margin gained by the setdown adjustment is above and beycnd that ensured by the safety analysis.
- 3. 5 / 4,!- 31 l Unit 2 1
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ENCLOSURE 3 DESCRI PTION, REASON AND JUSTIFICATION FOR CMANCE PROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 Description Of Cha're The BEN Unit 2 Technical Specifications are being updated to reflect the, limits for cycle 6 operations.
The cycle 6 core loading has been changed Locause of the results of inspection and reconstitution of the fuel completed in July 1988.
The actual changes are a slight adjustment in the Minimum Critical Power Ratio (MCPR) and the addition of two Tables of Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) versus average planar exposure.
Reason For Change The Minleum Critical Power Ratio as a function of scram time (figure 3.5.K-1) has changed because of the reanalysis performed to include BFN Unit 1 fuel in the Unit 2 reload.
The MAPLHCR for each type of fuel as a function of average planar exposure is presented in tables 3.5.I-1, 2, 3, and 4 These tables have changed because of the inclusion of a different fuel type from BFN Unit 1, and the pressurized and unpressurized MAPLHCR have been separated into two tables.
Technical specification 3.5.1 and the bases are changed to reflect the addition of the two MAPLHCL Tables (Table 3.5.I-3 and 4).
)
Justification Oor Channe
- The initial cycle 6 reload was submitted to NRO by letter dated August 23, 1984, and was approved by the issurance of BFN Technical Specification 199 dated August 19, 1986.
The cycle 6 core loading has shanged as a result of the fuel inspection and reconstitution program completed in July 1988.
The justification and safety analysis results for the changes are presented in TVA-RLR-002 Revision 2, July 1988, "Reload Licensing Report for Browns Ferry Unit 2 Cycle 6."
A summary is presented below.
Figure 3.5.K-1 MCPR vs TAU is changed because of the reanalysis.
The reanalysis indicated the bounding accidents are rod withdrawal error and generator load reject without bypass.
All of the accidents and the bounding envelope are shown in figure 14 of the Reload Licensing Report.
Four MAPLHGR figures are required to define the limits for all fuel to be loaceo in cycle e.
Tne current technical specification have two of those figures. Table 3.5.I-3 is specific to fuel type P8DRB2842.
This fuel type was not in the initial cycle 6 fuel load but was added as a result of the fuel inspection and reconstitution program.
Table 3.5.1-4 was added t: separate the prescurized (P80RB284L/ QUAD + shown in Table 3.5.I-1) and the non-pressurized fuel (8DRB284L).
The pressurized fuel allows higher exposures.
The changes to specification 3.5.1 and the 3ases are administrative in nature to reference the additional MAPLHCR tables.
ENCLO3URE 3 DETERMINATION OF NO CICHIFICANT RU.ARDS CONSIDERATION BROWHS FERRY KUCLEAR PLANT (BFN)
)
UNIT 2
' Description of Pioposed Amendment *
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The BFN Unit.2 Technical Specifications are being updated to reflect the limits for cycle 6 operations.
The changes con'sist of a slight revision to the Mininum Critical Power Ratio (MCpR) and the' addition of two Kaximum Average planar Linect Heat Generation Rate (MAPLHGR) versus average planar exposure tables.
Basis for Proposed No Sir.nificant Hazards Consideration Determination NBC has provided standards for determining whether a significant hazards consideration exist? as stated in 10 CFR 50.92 (c).
A proposed amendment to an operating license involves no cignificant hazards considerations if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from an accident previoucly evaluated, or (3) involve a significant reduction in a margin of safety.
1.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Operational transients analyzed in the Final Safety Analysis Report have been reevaluated in detail.
The Reload Licensing Report for Browns Ferry Unit 2, Cycle 6. Revision 2, provides a summary of the limiting operating transient, stability, and selected accident analyses for the proposed core a rrangement.
The 8xS fuel assemblies to be installed in the core are not significantly diffurent from the 8x8 fuel assemblies they are replacing.
The NRC ctaff has previously approved the design of the CE p8x8R assemblies as described in the CESTAR document (NEDO-24011-p-A-8).
The NEC staff has previously evaluated and approved the use of four Westinghouse designed QUAD + demonstration assemblies in the low power region of the core.
The NRC staff has also opproved the analysis methods used by TVA.
2.
The p'roposed amendment does not create the possibility of a new or different accident.
This reload changes the initial conditions and/or final condition used in the existing analyses and does not create any new accident mode.
a.
Ane propose:5 amend.mt aoes not invoWc a significan!~T51MEden in a margin of safety because the plant will be operated under the same safety limits with MCpH and RApLHCR operating limits comparable to those currently established.
The Reload Licensing Report provides a summary of the limiting operating transient, stability, and selected accident analyses for the proposed core arrangement.
The HCpR and MApLHCR limits luve been revised to assure the.urgin of safety is maintained as demonstrated in the Reload Licensing Report for Browns Ferry Unit 2 J
Cycle 6 Revision 2.
Based on the above reasoning, TVA has determined that the preposed amendment doe.s not involve a cignificant hazards consideratien.
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4 i,6 1,I
.i ETN LICEFSING TRANSM2 MAL TO NRC-Str.ARY AND CONCURRINCEr6KrrT Q
v DATE s/8/88 DATE DUE NRC R/17/88 ACTION NO.
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JdBMITAL PREPARED BY J. McCarthy TIES REQUIRED TES I NO R*.'$~5 PROJE/,T/ DOCUMENT.I. D. - TS 251.-EFF Unit 2 cycle 6 reload gp i..
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sed PJP PURPOSE /
SUMMARY
te update Pni Unit 2 Tech Seers For eyele 6 ederation O1I(*
7me QHE RESPONSE TO N/A (RIMS'NO.)
COMPLETE RESPONSE YES 2 NO PROB!.".M OR DEFICIENCY DESCRIPTION Plant Tech Seecs incereerste various recuirements for fuel / core verformance for the fuel characteristics of the present core. When a refueline occurs these core eerfor ance recuirements must be uedated in the Tech Seces.
CORPICTIVE ACTION /C MMITMENT Submit and obtain secrevel of evele 6 Tech Seecs.
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CONCURPINCE NAME ORGANIEATION SIGNA"'URE DATI J. G. Walker Plant Manater
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4D bO H. P. Pemrehn Site Directer. BFN
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s Independent Verificatien Per Verification Ce:plete SDSP 15.10 Section 6.8 Required YES L/ NOL/
If Required Date/By SL'EMITTA:., APPRO*.'ID DATE DIi,LFA !U. NAGER Wanged/Telecopied Receipt Acknowledged ty Date
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TVA-RLR-002 Revision 2 July 1988
. 1-.. -
TENNESSEE VALLEY AUTHORITY BROWNS FERRY !TUCLEAR PLANT UNIT 2 CYCLE 6 RELOAD LICENSING REPORT 5503B 680-citr$ cato WI
1 Revision 2 July 1988 I.
Introduction This reload' licensing report presents the results of the core redesign and safety analyses performed for Browns Ferry Nuclear Plant (BFN) unit 2, cycle 6 operation.
The current licensed design is documented in references 1 and 2.
The methodology and technical bases employed in the perforTnance of these analyses are discussed in references 3-8.
Items specifically addressed here include the nuclear fuel assemblics and core loading to be used in cycle 6, the reload core nuclear design characteristics, the transient and accident safety analysis results, and the proposed operating thermal limits.
The cycle 6 reload core will include four Westinghouse QUAD +
demonstration assemblies located in nonlimiting core peripheral locations.
A complete description of the demonstration assemblics is contained in Westinghouse Report WCAP-10507 (reference 9).
The cycle 6 core loading has been changed based on results of inspection and reconstitution of the fuel available for use in cycle 6.
The unit 1 once-burned fuel will replace the unit 2 once-burned f or us.it 2, cycle. 6.
Also, 212 twice-and thrice-burned bundles to be loaded were inspected and reconstituted as needed.
II.
Reload Cycle Information A.
Design Basis Exposures 1.
Actual cycle 5 core average exposure at end of cycle:
20.8 GWd/07 2.
Minimum cycle 5 core average exposure at end of cycle from cold chutdown considerations:
20.8 GWd/ST 3.
Assumed cycle 6 core average exposure at depletion of reactivity (DOR)*:
17_.9 GWd/ST B
Reload Fuel Assemblics Fuel Type Cycle Loaded Number Irradiated
---80RB 2 8 4 L, U 2 R 2 --- -----
- U2CYS -
9 3-
~
P8DRB284L,U2R3 U2CY4 159 P8DRB265H,U1R5 UICY6 160 P8DRB284L,UlRS UICY6 80 P80RB284%,U1R5 UICY6 8
New P8DRB284L,U2R5 U2CY6 300 QUAD + Demo U2CY6 J
TOTAL 164
'UUR - t.nd of f ull power capability 5503B
2 255 Revision 2 July 1988 Descriptions of the nuclear and mechanical design of the General Electric, irradiated and new fuel assemblies to be loaded in cycle 6 are contained in reference 10.
The nuclear,. mechanical, and.
~ "thermal-hydraulic design ' descriptions for the Westinghouse
- ~~
demonstration assemblies are contained in reference 9.
C.
Reference Core Loading Pattern The reference loading pattern is the basis for all reload licensing and operational planning and is comprised of the fuel assemblies designated in item II.B of this report.
It is based on the core condition at the end of the previous cycle, the number and type of fuel assemblies suitable for use, and on the desired core energy capability for the reload cycle. The reference loading pattern is designed with the intent that it will represent, as closely as possible, the actual core loading pattern.
Figure 1 shows the reference core loading pattern for cycle 6.
The reference loading pattern includes four Westinghouse QUAD +
demonstration assemblies loaded in peripheral locations.
These locations satisfy the criteria specified in references 2 and 9.
Evaluations performed by Westinghouse (reference 9) show that the results of licensing analyses for the lead P8x8R fuel
's embly bound those for the QUAD + demonstration assemblies.
Cycle specific analyses performed by TVA confirm this conclusion.
A total of 212 twice-and thrice-burned assemblies were inspected and reconstituted for use in cycle 6.
Prior to the reconstitution project, guidelines were implemented to ensure that performance of the reconstituted assemblies would not differ significantly from the original assemblies.
Consequently, the safety analysis results reported in this document were generated with the reconstituted assemblics modeled as original assemblias.
Following completion of the reconstitution work, this modeling assumption was verified by individually analyzing each reconstituted assembly and by performing core-wide analyses to specifically address the effects of reconstitution.
These analyses confirmed that all design criteria
'are satisfied and that operating limits reported in this document remain valid.
D.
Special Conditions Tim use tyt-trcreatrWrdre floJTIcn is planned for cycle 6 noeration.
Saf ety analyses were perf ormed f or both 100 percent and 105 percent of rated core flow with the most conservative results used for determining the operating limits.
The conclusions regarding LOCA analysis, reactor internals pressure drop, and flow-induced vibration as discucced in reference 11 are applicable to cycle 6.
The flow-biased instrumentation for the rod block monitor will be signal clipped for a setpoint of 106 percent since flow rates higher than rated would otherwise recult in a SCpR higher than reported for the rod withdrawal error.
5503B
3 Revision 2 July 1988 III.
Nuclear Design Characteristics
' ~ "
~
A.
~
~ ~ ~
The reference core is analyzed in detail to ensure that adequate shutdown margin exists.
This section discusses the results of core calculations for shutdown margin (including the standby liquid i
control system).
1.
Core Effective Multiplication and Control Rod Worth Core effective multiplication and control rod worths were calculated using the TVA BWR simulator code (references 4 and 6) in conjunction with the TVA lattice physics data generation code i
(references 5 and 6) to determine the core reactivity with all i
rods withdrawn and with all rods inserted.
A tabulation of the results is provided in table 1.
Th>sse three eigenvalues (effective multiplication of the core: uncontrolled, fully controlled, and with the strongest rod out) were calculated at the beginning-of-cycle 6 core average exposure corresponding to j
the actual end-of-previous-cycle core average exposure.
The core.was assumed to be in a xenon-free condition.
Cold keff was calculated with the strongest control rod out at various exposurcs through the cycle.
The value R is the i
difference between the strongest rod out keff at BOC and the i
maximum calculated strongest rod out keff at any exposure point. The maximum strongest rod out keff at any exposure point is equal to or less than:
1 SRO SRO Maximum keff a keff (BOC) +R 2.
Reactor Shutdown Margin r
i Technical Specifications require that the refueled core must be j
capable of being made suberitical with 0.38-percent ak margin i
in the most reactive condition throughout the subsequent 3
operating cycle with the most reactive control rod in its full out position and all other rods fully inserted. The shutdown margin is determined by using the BWR simulator code to 1
-..._. eal'eulat.e~t.hiiTore mulf.TpirenITorarserscTers~xp6surT points with the strongest rod fully withdrawn.
The shutdown margin for SRO the reloaded core is obtained by subtracting the maximum kerf i
from the critical keff of 1.0.
resulting in a calculated 4
minimum cold shutdown margin of 1.0-percent ok for DFN unit 2, cycic 6.
i j
l 5503B i
i
k~
4()
Revision 2 July 1988 Table 1 CALCULATED CbHE EFFECTIVE MULTIPLICATION - NO VOIDS, NO,XENONu20*C UNC Uncontrolled.
K gg (BOC) 1.120 e
CON Fully Controlled K,gg (BOC) 0.956 SRO Strongest Control Rod out, K,gg (BOC) 0.985 R. Maximum Increase in Cold Core Reactivity 0.005 With Exposure Into Cycle Ak 3.
Standby Liquid Control System The standby liquid control system (SLCS) is designed to provide the capability of bringing the reactor, at any time in a cycle, from full power and a minimum control rod inventory (which is defined to be at the peak of the xenon transient) to a suberitical condition with the reactor in the most reactive xenon-free state.
The SLCS shutdown margin is determined by using the BWR simulator code to calculate the core multiplication for the cold, xenon-free, all-rods-out condition at the exposure point of maximum cold reactivity with the soluble boron concentration given in the Technical Specifications.
The resulting k-effective is subtracted from the critical k-effective of 1.0 to obtain the SLCS shutdown margin.
The results of the SLCS evaluation are given in table 2.
Table 2 STANDBY LIQUID CONTROL SYSTEli CAPABILITY Shutdown Margin (ak)
"^
660 0.029 B.
Reactivity Coefficients The reactivity coefficients associated with the nuclear design of DFN unit 2, cycle 6 are ieplicit in the 1-D cross sections used for the cafety analyccc.
As cuch, reactivity coefficients are not separately calculated for input to the transient analyses.
lioweve r,
a void coefficient is generated in the 3-D to 1-D cross section collapsing process and is used as a verification check.
For DFN unit 2, cycle 6 the following results were obtained:
100% core flow, DOR
-0.0734
%Sk/% void 105% core flow, EDOR1
-0.0745
%ak/% void
& EDOR-extendeddepletionofreactivityresultingfromincreasedcoref{$0'B 3
es 5(O Revision 2 July 1988 c.
Fuel Performance The BFN unit 2. cycle 6 fuel performance is predicted by projecting the fuel burnup to the end of cycle with the 3-D simulator code.
The calculated peak pellet exposures for the various fuel types are less than the limits specified in references 9 and 10.
Furthermore, peak linear heat rates satisfy the assumptions made in the fuel vendors' thermal-mechanical integrity analyses (references 9 and 10).
All fuel types loaded in cycle 6 are predicted to operate within these bounding assumptions.
Additionally, the QUAD +
demonstration assemblies are predicted to have substantial margin to the lead P8x8R assembly in steady-state bundle power and thermal limits throughout cycle $ (figures 20-22).
The minimum margin for bundle power is 27 percent which satisfies the requirement for at least a 20-percent margin specified by NRC (reference 2).
For MCPR the minimum margin is 43 percent and for LHGR Lt is 32 percent.
IV.
Transient Analyses A.
Pressurization Events The RETRAN computer code (reference 12) is used to analyze both the reactor system and hot channel responses during core wide pressurl:ation transients.
The analytic models used in these analyses are described in reference 7.
A description of the CPR correlation and its application to Browns Ferry is contained in reference 13.
Analyses are performed for the potentially limiting events at the most adverse initial conditions expected during the cycle.
Reload unique initial conditions and transient analyses resulta are su==arized in the following tables.
NSSS Initial Conditions Steam Flow Core Flow Gap Conductance Expocu re
(% Rated)
(% Rated)
(BTU /ft2 hr..p)
EDOR 105 105 674 liot Channel Initial Conditions (Limitinr. Event)
Fuel Bundle Bundle Gap Conductance Type TCPR Power (HW)
, Flow (Kib/hr)
P-Factor (BTU /f t2-hr *F)
P8X8R 1.295 6.416 123.7 1.051 1287 5503B
es 6O Revision 2 July 1988 Pressurization Event Analysis Results Peak Power Peak Heat Peak Vessel ACPR1 System
'"~~
Translent
~~(% Rated)
Flux (% Rated)
Press'l (psia)
P8x8R
Response
Load 403.4 121.6 1235.3 0.225 Figures kej ection w/o Bypass 2-5 Feedwater 234.8 115.5 1215.'1 0.149 Figures Controller Failure 6-9 B.
Nonpressurization Events The nonpressurization events analyzed for reload licensing are either steady-state events or relatively slow transients that can be analy:ed in a quasi-static manner using a 3-D BWR siculator (reference 4).
The methods used to analyze these events are described in reference 3.
Results are summarized below.
Nonpressurization Event Analysis Results ACPR*
Peak LHCR(kW/ft)'
Event P8x8R/8x8R/00AD+
P8x8R/8x8R/0UAD+
Loss of 0.18 17.5 Feedwater Heating (100*F)
Rod Witadrawal Error 0.208 20.8 Rotated Bundle Error 0.198 15.3 Hislocated Bundle Errar 0.13 14.4 1 Results presented were calculated for P8x8R fuel and will be conservatively applied to 8x8R.
8 For increased core flow based on a signal clipped rod block setpoint of 106 percuit.
8 includes 0.07 penalty required when using the variable water gap method (reference 10).
- Results presented were calculated for the P8x8R fuel typc and conservatively bound the results calculated for the 8x8R fuel type.
The results are also bounding for the QUAD + demonstration assemblies which will be loaded into nonlimiting, peripheral core locations.
55039
e 7
Revision 2 July 1988 C.
Overpressure Protection The mair$ steamline isolation valve closure. with failure of direct
~
scram is analyzed tb demonstrate sufficient overpressure protection (peak vessel pressure must be less than 110 percent of design pressure - 1390 psia).
The event is analyzed using the models and methods described in reference 7.
Results are summarized below.
MSTV Closure (Flux Ocram) Results Peak Vessel Peak Steamline System Pressure (psia)
Pre.esure (psia)
Respong 1281.0 1242.5 Figures 10-13 V.
MCPR Operatinr. Limit Sumary The methods used to detennine the required OLMCPR values for each event analyzed are described in references 3 and 7.
The application of Options A and B limits in determining the cycle OLHCPR is described in the unit Technical Specifications.
Results are summarized below and in figure 14.
0LHCPR for Pressurization Events (BOC6-EOC6)
Option A1 OptionJ 1 P8x8R/8x8R/0VAD+
P8x8R/8x8R/OUAD+
Load Rejection Without Bypass 1.35 1.26 (GLRWOB)
Feedwater Controller Failure (FWCF) 1.27 1.23 OLMCPR for Nonpressurization Events (BOC6-EOC6)
P8x8R/8xBR/CUAD+1 Lots of Feedwater Heaters (LFa'H) 1.25 Rod Withdrawal Error (RWE) 1.27 Rotated Bundle Error (RBE) 1.26
.---- w.,-.
Mislocated Bundle Error (MBE) 1.20 Results presented were calculated for the P8x88 fuel type and 1
conservatively bound the results calculated for the Bx8N fuel type.
The QUAD + demonstration asse:mblies will be loaded inte nonlin! ting core locations and monitored to the ca e OLMCPR.
15 C 2 '-:
i I
8 Revision 2 I
July 1988 VI.
Accident Analyses A.
-- * * * ~'
Loss of Coolant Accident (LOCA)
MAPLHGR limits for the unit 1 P8DRB2842 fuel type (from i
reference 14) still apply for fuel being transferred to unit 2 since i
i the LOCA responses for the two units are identical (reference 15).
The limits for remaining fuel types are taken from reference 16.
1 Reference 9 indicates that the MAPLHCR limits for fuel type I
P8DRB284L can be conservatively applied to QUAD + demonstration assemblies.
Tables of MAPLHCR liraits for all fuel types in unit 2 cycle 6 are presented below.
t LOCA Limits for CUAD+ Demonstration Assemblies i
Average Planar MAPLHCR
[
Exposure (mwd /t)
(kW/ft) 200 11.2 1,000 11.3
[
5,000 11.8 10,000 12.0 l
15.000 12.0 20,000 11.8 25,000 11.2 30,000 10.8
[
35,000 10.2 40,000 9.5 45,000 8.8 LOCA L bits for CE Fuel Type P80RB284Z Average Planar MAPLHOR Exposure (mwd /t)
(kb/ft) 200 11.2
[
1,000 11.2 5,000 11.1 10,000 12.0 15,000 12.0 20,000 11.8 25,000 11.1
- -3 W IC a- -
35,000 9.8 l
40,000 9.1 I
45,000 8.5 I
[
{
t h
t t
t
?
I ssea t
J Revision 2 July 1988 LOCA Limits for GE Fuel Type P80RB265H
~
j Average Planar
-MAPLHCR g,,,yp, gyggfg)
(kW'/ f t )
200 11.5 1.000 11.6 l
5,000 11.9 l
10.000 12.1 l
15,000 12.1 20.000 11.9 25.000, 11.3 30.000 10.1 l
35.000 10.2
(
40.000 9.6 1
LOCA Limits for CE Fuel Type 80RB284L
[
Average Planar MAPLHCR 1
Exposure (mwd /t)
(kW/ft)
L i
f 200 11.2 j
1.000 11.3 i
{
5.000 11.8 10.000 12.0 l
15.000 12.0 j
20.000 11.8 r
25.000 11.2 i
j 30.000 10.8 j
35.000 10.2 j
40.000 9.5 I
]
LOCA_ Limits for CE Fuel Type P80RB284L t
j Average Planar MAPLHGR 4
Exposure (mwd /t)
DOL /f t) 200 11.2 1.000 11.3 L
5,000 11.8
'l 10,000 12.0 i
15.000 12.0 l
M,000 1176 l
~
i 25.000 11.2 I
i 30.000 10.8 l
j 35.000 10.2 40.000 9.5 45.000 8.8 1
l a
)
I
(
3
}
t
}
$503B 4
l
i 1
Revision 2 July 1988 B.
Rod Drop Accident (RDA)
'-~
The methodol'ogy used to analyze the rod. drop, accident is described in appendix A of referen'ce 8.
Results for unit 2, cycle 6 are surrarized below.
Results for the f.imiting RDA Condition:
3 75'F, MOC Exposure Rod Worth:
1.05 percent ok Rod Position:
38-15 Peak Fuel Enthalpy:
194.5 cal /gm Core Response:
Figures 15-18 VII.
Stability Analyses The methodology used to analyze core and channel stability is described in appendix B of reference 8.
The mininom stability margin occurs at the intersection of the natural circulation line and the 105-percent rod line (the flow biased scram line also passes through this point).
Results for BFN unit 2, cycle 6 are summarized below and in figure 19.
Stability Analysis Results at Limitint Initial Conditions Maximum Analysis Decay Ratio Core stability 0.841 Channel Stability 0.592 P8x8R/8x8R/ QUAD +
& Includes 0.14 uncertainty adder as described in appendix B of reference 8.
8 Results presented are for the P8x8R fuel type and concervatively bound the 8x8R fuel type and the QUAD + demonstration assemblicc.
5503B
cIN Revision 2 July 1988 References
\\
~
~
'1.
TVA-RLR-002, Rev. 1 dated April 1985, Reloa2' Licensing Report for Browns
~
Ferry Unit 7, Cycle 6," TVA.
t 2.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting i
Amendment No. 125 to Facility Operating License No. DPR-52. Tennessee
[
Valley Authority, Browns Ferry Nuclear Power Plant. Unit 2, Docket No. 50-260.
3.
TVA-EG-047 dated January 1982.,"TVA Reload Core Design and Analysis Methodology for the Browns Ferry Nucicar Plant," TVA.
L l
4 TVA-TR78-03A dated January 1979, "Three-Dimensional LWR Core Simulation I
Methods," TVA.
5.
TVA-TR78-02A dated April 1978, "Methods for the Lattice Phytics Analysis of LWRs," TVA.
6.
TVA-TR79-01A dated January 1979, "Verification of TVA Steady-State BWR Physics Methods," TVA.
7.
TVA-TR81-01A dated De*, ember 1981, "BWR Transient Analysis Model Utilizing the RETRAN Program," TVA.
8.
TVA-RLR-001 dated January 1984, "Reload Licensing Report for Browns Ferry Unit 3. Cycle 6 " TVA.
E; 5
9.
WCAP-10507 dated March 1984, "QUAD + Demonstration Assembly Report,"
l Westinghouse Electric Corporation.
10.
NEDE-24011-P-A-8 dated May 1986, "General Elcetric Standard Application j
for Reactor Fuel," General Electric, i
11.
NEDO-22245 dated October 1982, "Safety Review of Browns Ferry Nuclear Plant Unit No. 2 at Core Flow Conditions Above Rated Core Flow During Cycle,5." General Electric.
t l
12.
EPR1 NP-1850-CCM dated May 1981, "RETRAN02 - A Program for Transient i
Thormal-Hydraulic Analysis of Complex Fluid Flow Systems," Electric Power f
i Research Institutc.
l i
j 13.
NEDE-24273, "GEXL Correlation Application to TVA Browns Ferry Nucicar
{
Power Station," General Electric.
i j
14 NEDO-24056. Rev. 1 dated May 1983, "Loss-of-Coolant Accident Analysis for Browns Ferry Nucic.ar Plant Unit 1 " General Electric.
1 j
15.
DGC:88-146 Letter from D. C. Churlik to J. D. Robertson dated July 13 f
l 1988, "Telecon of 1/13/88," General Elcetric.
l 1
i 1
16.
NEDo-24088-2 (as amended) dated May 1985 "Loss-of-Coolant Accident Analysis f or browns Ferry Nuc1 car Plant Unit 2." General Elt etric.
e i
e 12 E'3 Ravision 2 FIGURE 1
% tus REFERENCE LOADING PATTERN BROWNS FERRY UNIT 2 - CYCLE 6 60 Bl B B'B BI B B B BB Bl B Bl B 58 l8 AI F AF AI F F: F Fl A FI A Fl B Bl 56 BL BI B FI A Fl C F! C FI F CF CI F AI F BIB B
54 B Bl F CI F CI F CI F Cl C F
C FI C Fi C Fj B B
52 Bl B F D Fl C Fl C F B FF BF CF CF DF BB 60 Bl B Aj o CF CF CI F CF E
E F. C Fi C F. C F
C O A BIB 48 B! 8 FI C FI C Fj C FC FD F F DF CF CF CF CFBB
~
46-lB Bl F DI F CI F Di F DF DF BB FD F: D FL D FC F
D F<B Bl 44 - Bl 8 FC Fl C FI D BI D BIC B D Fl F Dj B Cl B Di B D! F C
F CI F BI B 42 BI F AF CI F CI F Dl F Dj F C. F CI C Fj C Fj D FID FI C F
C FI A Fl B 40 - BJ A FC FI C FI D A! D F'D Bl C F'F CB D
F DI A D' F ' C F CI F AI B 38 BI F C, F CI F Cl F CI F DF Dl F B A F D F
D Fl C F
C F
C FI C FI B 36 - BI A F C FC FI D Bl C Bl D FI A F: F A'F DI B CI B DF CI F CI F A! B 34 - Bl F C. F B, F Di F DI F CI F AI F C C FA Fi C FI D F D FI B FI C FI B 32 - Bl F F
C F
E F
B Fl C F B FC AA CI F B( F CI F BI F E
F CI F Fl B 30 - AI F F: C F
E: F' B Fl C FB F: C AI A CI F BI F CI F BI F EF CI F FI A 28 - BI F CI F Bl F Di F Dj F CI F AI F CI C FI A FI C FI D FD Fl B FI C FI B 26-Bl A. Fi C FI C, FI D Bl C Bl 0 FI A FIF AI F Di BI CI B D, F CI F CI F Al B 24 - BI F CI F CI F CI F CI F Di F DI F Bl B FI D FI D Fi C Fl C FI C F! C Fl B 22 - BI A FI C FI C. FID AfD FI D BI C FI F Cl B DI F DjA DIF Cl F CI F Al B 20 - Bl F AI F CI F Cj Fl DI F Dl F Cl F CI C FI C FI DI FI D Fl C FI C FIAjFIB 18 Bl B Fi C Fi C FIDIBlD AtC BjD FI F DjB ClAjDjB Di F CI F CI F BI B 10 -
B B! F ' Ol F CI F Dl F Dj F Di F BI B FI D FI D FI D Fl C FI D Fl B B) 14 B! 8 Fl C Fi C FI C FI C FI D FI F O! F Ci F CI F Cl F CI F Bl B l
12 81 B i Al O CI F CI F CI FL C! F El E FI C FI C Fl C FI C 01 Aj Bi B 10 BI B F! D FI C FI Cl FI B FI F BI F; CI F CI F DI F Bl Bl 8
B Bl F Cl F C
, j 0
8 Bl B FIA F C FC FF CI F CF AJ F A B _B, 4-j lB AI F A F A F F
F FI A FI A FB B
2 BI B B,
B BB A9 BI B Bl B BB 3
5 ~
9 ~
13-
~2 F 25 20~ ~33~
7-4 45 49 53 57T 3
7 11 15 19 23 27 31 35 39 43 47 51 55 59 A= 8DRB284L,U2R2 B= P8DRB284L,U2R3 C= P80RB265H,U1R5 D= P80RB284L,U1R5 Ea P80RB2842,U1R5 F= P80RB284L,U2R5 Q= OUAD+ DEMO U2RS
e 13 d
Rcvision 2 FIGURE 2
- BF2CY6, GLRWOB - ICF 500 1.egend 400--
To7AL powrm (z) s wsc.w.w.m..
c.m.ww.n.m s............
300-Cear 'NtET syecoctive (x),
200~
~
1o0-
),' "" '.
~
0 0
1 2
3 4
5 6
7 BME (SEC)
FIGURE 3
- BF2CY6, GLRWOB - ICF 200
'50 --
100-I e
t'
_50:
leged yrsstt 04Ess R'st (PM f
/
IQM.Sd.VMV.E.%f (.%)..
/
M?e.i.S. K4.L.Y.E. G 0.?,. ( %. ),....
0 0
1 2
3 4
5 6
7 TlME (SEC)
14
.d Rcvision 2 FIGURE 4
- BF2CY6, GLRWOB - ICF 15 0 10 0 -
i i
l
,'i,,.,,.
i l,,
50- ::
e,,
s, %
. ~ _-
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-10 0 0
1 2
3 4
5 6
7 TIME (SEC)
FIGURE 5
- BF2CY6, GLRWOB - ICF 2
Legend TOTAL mtActivtTV ($)
Ka.A.W.a.f.497JW. CO..
0-
'~~....#.g'6 0g 6
2.-
~4-0 0.'5 i
1.5 2
2.'5 3
BME (SEC)
y-pd' u
Revision 2 FIGURE 6 BF2CY6, FWCF - ICF 250 200-
\\
15 0 -
b s
,nn _
- 6. = = n =. - ~
.k_'.I'.,
. ~.
Legend tout powen ts) 3g, fRW.5r ewfwt (z.)..
.co g,i,N,g g,n,9,w,1x),,,,,,,,,,,,,
coat INtrT gysecouwe (x1, O
i 0
5 10 15 20 25 TIME (SEC) 4 FIGURE 7 BF.2CY6, FWCF - ICF 15 0 Legend vesset auss est (no W h W.ELKCLP?.Ol.
100--
RTP.t.57..V.SUJ\\A '.Gl......
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- \\
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10 15 20 25 TIME (SEC) i i
t f
1
es 16 d
Revision 2 FIGURE 8 BF2CY6, FWCF - ICF 15 0
........................'.........,d'~
li,,,
s e
!l i5l
{~-
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n,i 50-
- . e '
I*
- '8 e Q~
- .M Legend LtVR frNCM-aff-S f p Su tR9 l
KUI( PJ't.Nef.L51....
l yg,
,,y,,,,,g, g79 w,,ge,y,,g,,,,,,,,
rtrowatro n,ew (c) q
-10 0 -
0 l'O 15 20 25 BME (SEC)
FlGURE 9 BF2CY6, FWCF - ICF 2
0-
.~~....*.,'
4 es d<
g
's legend =
TC'Al REACT 1vlTY ($)
@.'.W.TI.A.C11yQ! J D,,,
~6, 15 16 17 18 19 BME (SEC)
I
17 7J Rsvision 2 July 1988 RGURE 10
- BF2CY6, MSIVC - ICF 600 Legend TOTAL POWER (X) eK EW.A.Cf.H.E AI,W,W,,
cp. R,E,,t,N(G,,G,Q,W,,( Q,,,,,,,,,,,,
400-CORE INtET,5.UDC00UNO (3),,
200-
~
r
~ ~
\\
0, 0
2 4
6 8
TIME (SEC) r I
l RGURE 11
- BF2CY6, MSIVC - CF 250 4
f 200--
l l
15 0 -
100-
[
Legend
.e' 50-vrsstt oREss eSt (as5 IoT.A.LN.R.V.^.tM.%C.* !O..
(
/
0-BYf6sy,,ssg,gew (N.,,,,
j
?
O 2
4 6
8 T1ME (SEC) l l
l
18 cO Rsvision 2 July 1988 FIGURE 12
- BF2CY6, MSIVC - ICF 15 0 Legend uvet fmew-arr-tra trian YC5J MUM GM.151.....
10 0 -
~;.
IN M.?3. M " Y M.L N.......
s
- i r
l 'lsl' '.....t tow arte rtow it)..-....T....
's,
e s
,e i.
50-l
^'
v
$ I i
e, i
i, t
-50 5
0 2
4 6
8 l
TIME (SEC)
\\
.l FIGURE 13
- BF2CY6, MSIVC - ICF 4
2 Legend
)
.tpfat starvey ($1 K8.'?H.* F.T!?. W..
0--
c-......
\\
l
..g 2-
'. g r
4
'g a
i I
r i
~6'.
0 1
2 3
4 5
~1ME (SEC) i l
L l
19 c
'J R vision 2 Figure 14 July 1988 OLMCPR for P8X8R 8X8R./ QUAD +
1.35 1.34-1.32 -
GLRWOB 1.30-4 1.28-
[
'-----------*'-----------------------------.*1.27 o.-
2 1.25-,- - - - - - - - - - - - - - - R B E--------------------- --- -.
~~~.
.................L.R..YH...............'....-
l 1.24 -
RVCF 1.23 1.22-i 1.2 0 - -----------------MBE t
gg 0
0.'2 0.'4 0.'6 0.'8 1
TAU =
= SCRAM Speed Interpolet icn Poramter es Defined in the Tecnnicci Specificctions 4
I y
Figure 15 BF2CY6 Rod Drop Accident 10000 6000-t
~
t 6000 o
?:E Eie 4000-Legend O Power 2000-
)
b tu a
0 0.5 1
1.5 2
2_5 3
3.5 4
4.5 5
Time (sec)
P ll* g
'u 5o go m
4
Figure 16 BF2CY6 Rod Drop Accident i
12 0
( )
10 0-
~
e0-C CP o3 o
13, 60-c)
O.
[3
'o 40-Legend O Core Avercqe Temperature Rise 20-I CD l
o t.
i i
0 0.5 1
1.5 2
2.5 3
3.5 4
4.5 5
Time (sec) l
&T#
M O
g=
FJ ll
Figure.17 BF2CY6 Rod Drop Accident 2-1-
i O(
4 v5 v
.7 b
's i}o
.3 m
(Y.
Legend O
Core Recctivity '( )
{3.
O O.5 1
1.5 2
2.5 3
3.5 4
4.5 5
Time (sec) p y::
GS o
O
?"~
- ill l
d
~
C O
{ $8,
[s w*
- o 5
!i i ii y
>5 p
4
!c h
tn E
n iP
> 4 m
d u
n m
i e
xa gM e
5 L
3 t
O n
e d
icc 3
A 8
1 p
eo
)c rr e
uD
.5. (s 2
g e
id m
Fo i
T R
6
>2 YC 2FB
>5 1
i1 if!
l
>5 0
t 0
0 0
0 0
o 0
5 2
0 5
1 1
T 4 uA N )tC 5 J
1 i
llll l
l i
k F' igure 19 i
Decay Ratio vs. Power 1
e With Conservative Adder o.8 -
Natural Circuiation 105% Rod Line o.c -
g
.,o (Y
ti C) o.4 -
o.2 -
i o-o h,o e
40 so 80 70 0 12 0 Power (percent of roled)
~
E Y 't G.&
O 3,",
B Figure 20 i
Bundle Power Comparison: QUAD + vs Lead Bundle 3
I i.e
- 1. 5 -
'O ud
\\,,,. /
- N *
- "x.-.
- s, i
1.3 -
E i.2 -
3 O
Q.
c) 1.1 -
.?
To C) 1-m 0.0 '
0.8 -
O DEMO BUNDLE C) g,_
9 LEAD BUNDLE l
o.o.
h ::' u 6
o i
2 3
4 e
7 8
9 es.
Cyclo Exposuro (GWD/MT) d 2 :'~
i Figure 21 i
MCPR Comparison: QUAD + vs Lead Bundle 3
l
- 2. 8 -
O DEMO BUNDLE.
2.0 O LEAD BUNDLE
'M 2.4 - l
( l 2.2 -
tr O
2-2 i
- 1. 8 -
- 1. 0 -
-S 1.4 g
/g-
%' Neir515iGOOL*
g_
_g 9-O~3'O O Ng, 3== 9 OPERATING LIMIT 1.2 -
CS i
1-i 3
6 6
6 6
4 e
$f5 0
1 2
3 4
5 6
7 8
9 xg Cyclo Exposuro (GWD/MT) i s[
~
i Figure 22 i
MLHGR Comparison: QUAD + vs Lead Bundle 14 -
1 OPERATING LI!11T
+
1 DESIGN GORL c[
'N s.
. N
~~
vI o,
s.
.s,
.~..g 22 l
i
=
10-j s$
M l
xox
- d 0-
' f O-i I
O DEMO BUNDLE (9
. LEAD BUNDLE _
4-f u=~
f j
f f
y 3
0 1
2 3
4 Cyclo Exposuro (GWD/MT) g{
r FJ
ENCLOSURE 2 Date Prepared N/Md3 Requested Target Date < wm;yg fgg V
WORK RE0 VEST To: Waune //m'au
, Branch Chief, Reach SvsTEm.r Branch N22/sRXe THRU: jrhele 5ha d.3ns'
, WOW, # Fer Sur?$ r. DMrim dann,rre4e /serw m/,.
~
FROM: oav// mo,.m
. Project Manager, 40-osp/rva fw/Bih Mail'Stop 7s 2r TAC /T1YLE:
2oo.4s~o //rr.,,em ren w Adw P/.mf delt2 fYL 2<*L 5),e//n $)PdeJ 7*f, kby, Ma v$A.w 7?AlAuf 2&J988r2yu & {
.?mmem/ men?'
Description of Review Requested 7o' opentia filri,sa opp.s*L ts cAn es rie a/wrodSys fs unit *L 7h e ve/a C ev/eadpeeredd kl.bu. cAu lar warisujmiWad Aa a.giss 4
sf or i82rprevad A s flun//Aw Mdmen t/1c' fe WM. rWrs M9)Aue. /9 /fM. 7% einvA mm a
v u v ic,veA G /ue/ /JL wa Ameal n a r e.ruftd fanlikerset<h be.nenttphtdn mm Ja/.
J Priority:*
_f Roses for Priority: Af$n fw3/ h.2//ev mf5f" e
v v
a Please indicate your acceptect o' the Work Request and Target Date by Signature arid Assignment of Reviewerl*).
Work Package should be retained by reviewer (s).
Priority Deterinination Acc,eptable: Yes No Alternative Priority Target Date Acceptable:* Yes,
No Alternative Target Date:
Assigned Reviewer (s)
Phone Phone Section Chief Signature Date Branch Chief Signature Date Return to plant PM within 5 working days of receipt If a revision to previously approved schedule:
New completion date or New priority Weason for revision
't Section Chief Approval:
Branch Chief Aporoval:
i
- Review schedules for the 4 priority cateocries are norrally:
Pricrity 1 InTrediate assigreerit of resources Priority 2 Near term action Priority 3 Long term action Priority 4 Resource dependent action
_tr.msure 3 4
4
'?'
August 19, 1986 l',,,,#
Docket No.:
50-260
,p Mr. S. A. White Manager of Nuclear Power Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, Tennessee 37401
Dear Mr. White:
l The Commission has issued the enclosed Amendment No.125, to Facility /
Operating License No. DPR-52 for the Browns Ferry Nuclear Plant, Unit F.
This amendment is in response to your application dated August 23, 1984 (TVA BFNP TS-199), as supplemented September 4 and November 133 1984. April 3. May 8, June 27, November 20 and December 30, 1965 and April 29, 1986.
The amendment revises the Technical Specifications (TS) of the operating license to: (1) modify the core physics, thermal and hydraulic limits to be consistent with the reanalyses associated with replacing about one-third of the core during the Cycle f core reload outage and (2) reflect changes in various specifications as a result of plant modifications performed during the outage.
In addition. TVA has updated the TS pages involved and made r
administrative corrections.
A copy of the Safety Evaluation is also enclosed.
Notice of Issuance will be included in the Comission's Bi-Weekly Federal Register Notice.
i Sincerely, t-Marshall Grotenhuis, Project Manager BWR Project Directorate *2 Division of BWR Licensing I
J a
Enclosures-1.
Amend. Tent No.125 to i
License No. DPR-52 E
2.
Safety Evaluation FCantrell,RI! RWessman DMuller I
cc w/ enclosures:
CStable TKenyon WLong TAlexion See next page SRConnelly, GZech,RI!
BJYoungblood JHolonich i
BHayes,01 NGrace,RII LSpessard.DI SWeise.RI!
DISTRIBUTION:
HDenton HThompson SRichardson,!E JTaylor.lE Doc 6et File RBernero EJordan DVassallo HThompson NRC PDR SNorris BGrimes OPA LFMB Local PDR MGrotenhuis RClark TBarnhart(4) t JPartlow OGC WJones Plant File I
LHarron ACRS (10) 0FFICIAL RECORD COPY
'bl
' p' : D DBL:
J,0.G M D
DBL:PD*2 DBL:PD*2 DBL:PD*
fr l
j SN&Y MGrotenhuis RClark 08//4/86 OSAv/86 CS/W86 08] 3 /8 0 //f/86
-6 6 4 8 a m er w e scv
Browns Ferry Nuclear Plant Pr. S. A. White Units 1, 2, and 3 Tennessee Valley Authority cc:
s W. C. Bibb H. S. Sanger, Jr., Escuire Site Director BFNP General Counsel Tennessee Valley Authjrity Tennessee Valley Authority Post Office Box 2000 400 Conreece Avenue Decatur, Alabana 35602 E 11B 330 Knoxville, Tennessee '37902 Resident Insnector U. S. Nuclear Reculatory Connission Mr. Ron Rocers Route 2. Box 311 Tennessee Valley Authority Athens. Alabara 35611 5N 130B Lookout Place Chattanooga, Tennessee 37402 2801 Mr. Donald L. Uillians, Jr.
Tennessee Valley Authority Chairran, Linestone County Connission 400 Pest Surnit Hill Drive. U10EP5 Knoxville, Tennessee 37902 Post Office Box 188 Athens. Alabana 35611 Pobert L. Lewis, tianacer, PFNP Tennessee Valley Authority Ira L. Meyers M.D.
Post Office Box 2000 State Health Officer Decatur, Alabana 35602 State Department of Public itealth State Office Buildine Montoonery, Alabana 36130 Mr. K. P. Whitt E3A8 400 West Sunnit 11111 Orive Tennessee Valley Authority Knoxville, Tennessee 37902 Pepional Adninistrator, Region !!
U. S. Nuclear Regulatory Connission 101 Marietta Street, Suite 3100 Atlanta, Georcia 30303 Mr. Steven Roessler U. S. Nuclear Regulatory Connission Reactor Training Center Osborne Office Center Suite 200 Chattanoona, Tennessee 37411
[stD **%qjo, UNITED $TATES NUCLEAR REGULATORY COMMISSION
[
e W ASHINGTON, D. C. 20965 o
's,*....
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMEN 0 PENT TO FACILITY OPERATING LICENSE Amendment No.125 License No. OPR-52 1.
The Nuclear Regulatory Comission (the Comission) has found that:
The application for amendment by(Tennessee Valley Authority (the A.
licensee) dated August 23, 1984 TVA BFNP TS-199), as supplemented September 4 and November 13, 1984. April 3. May 8. June 27 November 20 and December 30, 1985 and April 29, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the h Ith and safety of the publict and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. OPR-52 is hereby amended to read as follows:
fQ=& W Y bk
2 (2) Technical Specification The Technical Specifications contained in Appendices A an B as revised through Arendment No.125, are hereby incorporate in tne license.
The licensee shall operate the facility in accc dance with the Technical Specifications.
3.
This license arendment is effective as the date of its issuance and is to be implemented within 90 days.
ECR THE NUCLEAR REGULATORY COMMISSION
-w ~
<?
Caniel R. Muller, Director BWR Project Directorate *2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: August 'i',1926
ATTACHMENT TO LICENSE AMENCMENT NO.125 FACILITY OPERATING LICENSE NO. OPR-52 00CKET NO. 50-260 Wevise Appendix A as follows:
1.
Remove the folicwirg pages and replace with identically nu".!;ered pages.
2.
The rr.arginal lines en these pages denote the area being chanced.
iv 62 250 vi 63 256 vii 73 257 i
viii 78 262 3
79 263 4
80 330 l
l 9
85 356 19 96 l
23 102 l
25 105 28 105a 32 110 33 110a" 34 159 f
35 160 37 168a 38 169 39 171 1
40 172 41 172a*
42 220 44 231 55 232
'Fage 17Ia is reroved but there is ro replacement.
"Page Added
i
'i Pane No.
seettee I
30$
3.
Core Monitoring C.
Spent Fuel Pool Water 11$ l ~
D.
Peactor BJ11 ding Crane......... 3C7 307 E.
Spent Fue) Cask............
F.
Spent Fuel Cask Handling Refueling 308 Floor.................
4 315 3 11/a.11 Fire Protection Systems A.
High Pressure Fire Protection System.
31$
B.
C0; Fire Protection System......
319 C.
Fire Detectors............. 32 0 321 D.
Roving Fire Watch 322 E.
Fire Protection Systems Inspection.
4 322 F.
Fire Protection Organisation.
I 323 C.
Air Masks and Cylinders l
32 3 I
H.
Coettmueum Fire Watch i
3.
Open Flamas Welding and Duanime. in Il 32 3
[
the Cable Spreading Room.
r 330 50 Major Design Testures 51 Site Features.............
330
)
I
$.2 Deactor................ 330 j
1 330 l
3 53 Poactor vessel 5.4 c on t a t np e n t.............. 330 33:
5.s Fuel stora.e 331
{
S.6 Selsric Design 6.0 A$ministrative Controls 332 332 6.1
- -tsation 6.2 Re..a= and Au11%
333 i
i tv i
t Amendment No. if.125 i
i
L!$T OF TA4LES Title Page No.
Table 3.1. A Reactor Protection System (SCRAM) Instrumentation Requirements.................
33 4.1.A Reactor Protection System (SCRAM) Instrumentation Functional Tests Minimum Functional Test Frequencies for Safety Instrumentation and Cont rol Circu i t s.................
37 4.1.8 Reactor Protection System (SCRAM) Instrument Calibration Minimum Calibration Frequencies for Reactor Protection Instruwent Channels.......
40 3.2.A Prtnary Containment and Reactor Building Isolation 55 Ins t rumen ta tion.................
3.2.8 Instrumentation that Initiates or Controls the Core 62 and Containment Cooling Systems 3.2.C Instrumentation that Initiates Rod Blocks 13 3.2.0 Of f-Gas Post Treatsent isolation Instrunentation..
76 3.2 E Instrumentation that Monitors Leakage into Drywell.
77 3.2.F Surveillance Instrumentation............
.8 3.2.G Control Room 1 solation Instrumentation.......
81 3.2.H Flood Protection Instrumentation..........
82 3.2.1 Meteorological Monito'ing Instrwnentation......
83 r
3.2.J 5eismic Monitoring Instrumentation.........
84 l
4.2.A Surveillance Requirements for Primary Containment and Reactor Building Isolation Instrumentation..
85 t
4.2.8 Surveillance Requirements for Instrumentation that Initiate or Control the CSC5...........
46 I
4.2.C Surveillance Requirements for Instrumentation that 102 Initiate Rod Blocks 4.2.0 Survetilance Requirements for Off Gas Post Treatment 103 Isolation Instrumentation 4.2.E Minimum Test and Calibration Frequency for Drywell Leak Detection Instrumentation...... 1 04 vi Amendrent No. 125
l 1.tST OF TAILES (Cont'd)
I L
Table Iltie pane Ne, l
l 4.2.F Minimum Test and Calibration Frequency for
[
Surveillance Instrumentation.......
. 105 j
4.2.G Surveillance Requirements for Control Roon
! solation Instrumentation.........
106 4.2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation.....
107 r
4.2.J 5eismic Monitoring instrument Surveillance.. )C9 3.5 1 Minimum RHR5W and EECW Puup Assignrent.... l4ta l
- posure...
171 3.5.!
i l
t 3.7 A Primary Containment Isolation Valves....
150 l
3.7.3 Testable Penetrations with Double 0. Ring Seals 756
)
3.7.C Testable Penetrations with Testable tellows 257 3.7,0 Air Tested Isolation Valves.........
25R
[
3.7.E Prirary Containeent Isolation Valves which
(
Terminate telow the Suppression Pool Water e
Level...................
262 l
3.7.F Prieary Contain-4nt Isolation Valves 1.ocated I
in Water Sealed Seismic Class 1 Lines...
261 I
3.7.G Deleted...,...............
264 j
3.7.H Testable Electrical Penetrations......
265 i
4.8.A Radioactive Liquid Waste Satpling and Analysis PR7 i
4.8.8 Radioactive Gaseous Waste Sampling and Analysis?!B
{
t 4.9.A.4.c Voltage Relay Setpoints/ Diesel cenerater start 29's j
3.ll.A Fire Protect',on 5.nter Wydraulic Reavirements. 3?4
[
1 l
j 6.8,A Mini um Shift Crew Requirener.ts...,,..
i
)'
f Arendment No.125 vii F
4 4
g_ti er tttutttAttent t
Pese le
$1g rg AHr fie et eceete Stree see Asap nog steg, Il f
t.l.1 lettlagt....................
14 Aeea flee lieg $ tree 16, Beetter (see-fle.
!.l.)
(eesmit &ie to the $elettien of am A4eswete 49 a.1 1 laterval $ttween telts.............
119 4,7 1 liste= Waevettsoitity.....
a reetsweete teletisa velisme teaseasest6ea 1)t
- 3. e.1 Seeiv eeswicearann..
sesi,. reateweeie netwtisa te=,ceatw e 139 e
i a.
..............172 l
seevieew ats 3.S.K.1 McPh Lietts 17 )
3.g ;
t, rette...
a lesserstv e 't Ase,e Change in Teemiteet e
), f.. I Ristav 194 l esent e e t w e e...................
3.6 1 Cheage in Cae*Pr e 1eentittoa teseerstvee vs.
195 hewtrea (esolvee................
TVA 0'fite of Powee 01anteettoa fee Cwerettoa M1 4.1.)
of metlese Pomoe Pleats.............
MI
- f.. t. ?
Tw*tttoaal Oegaattettoa...
36) f..P.1 6eview eat A.ett twm tion M4 G.).)
In.ptent Ftre Frogras Segoottattoa stii Ment. ent No. If.125 M
I A
1.3 Off1N1710SS (temt'.i g,
cretaste - caer tb titty = A syntes. whsvstes. train, c' epenent.
er sevtse ar.a.a so w'retable er have operabiltty when t r, 3 t s capable et perf orming its specified twnstten(s).
itst in this det Laitten shall be the asswrstten trat all nesseeery attandant instrumentatten, controls, nerial and energency electrical prwet seurtes, toeling er seal water, lubrication ei ethat awa1114ty equiseent that are requtted for the syster.
ewheyotes, tratn. testesent er device to perf ors its f watt 174. <;
are also capable of,perforsang their relates swipett (wattats(s).
T.
Cperattet - Cperattag seams that a syntaa er eersenent is perf erstet its antenced funattena la its regwared manner.
C.
Imr44 tate. !asediate means that the required attien will be 1
Lastsates as seca as practicable tensidering the saf e operatism af the unit ana the trportante of the required attian.
H.
Reacter tewer cieratten - Reactor power operaties 13 any epetatten vtta sne acte switan an the "Startsp" or "Run" positaan with the teatter arttstal and above 1* rated power.
1.
Not Stamitv Carditter.. Hot standby tendition seans opetatten with too. ant te=7eratwre greater tr.an 212'T. systen tressure less than 10 $ $ p e g g. t h e E413 stess isolatica valves alosed and the sete switch in the Startup/Not Star.dby positica.
t J.
Cold s'er.dttles Reaster coolant tes;erature etwal to et less than N
112't.
f i
K.
Het !* t d ess a The reactet != la the ehwttrwn scoe and the teatter l
j tecient tesyetature greater tr.an 210'T.
C414 t h u t f ess - The r e a t t e r is sa the thwtJewm tode and the res.ter l
tealant tes;eratwre etwat to or less tr.an 21**T.
i L
.M.
Feev et Sterat tes - A reactor asse evitch selects the prepet laterlacks tot the operattar.41 statwa af tr.e watt.
Th e f e l ;s.t.*.3 are the sedes aLJ 1rterletga prev 14eal 1.
S t a r t'.s tWa t Stantiv Fees = In this asse the reacter protettien
[
r syntes la energized vtth thK newtren ser.itatte; systes ttty.
the A?KX 13* high f1wa trip. Ena ecstrel rea withstawal l
&aterlacLa la service. This is ettet teisfr** tv as 'wst
{
5tartwa Xaae. This as intendea to Leply tr.e startwpisct Stanaty pestilan et the sede switth.
j L
r l
[
3 I
s.
A ene ent No. 82,125 l
l
(
1.6,n2r,1 N1,T,t nNt ( Ce nJ'd),
2.
Rum t*ada - in this mode the reactor syntes pressure la at er above a25 reta and the featter Protettien ayatem to eneretted with AFe.M protectlen (cuetuding the 13% hir.h flus tetr) and the RM teterlocks in serview.
3.
5%utdewn Nede - Placing the mode switch to the shutdown potatten initiates a teatter etram and power to the cofstrol red delves is removed. After a short time perted (about 10 seJ). the scram stanal to feaoved allowin;t a scram reset and restertas the moreal valve itneup in the control red drive hvdraulic system.
t I
4 Muel t'ade - Litth the mode switch in the refuel Poettien teterlocne are estabitehed so that ene control rod only saw te withdrawn when the Source Ranre Mentter indicate at least 3 cps and the refuelina crane is not over the reactor eRCept as specified by TS 3.10.B.1.b 2.
If the refueling crane it over the rea: tor, all rods must be fully inserteti and none can be withdrawn.
l s ted Pever - Rated power re(ere to operatteft at 4 reactor rower of H.
a -
T.2vT Sti thts is as.. ier.ed les percent rever and to the east.,
power level authertsed by the eserstint ltesese. Rated steam flow.
('
rated coolant flow, rated neutron flua, and rtted nuclear system Pressure refer to the values of these carameters when the reacter l
to at rated reiwer. rw trn cover, the power to wktch the saf etv analyste arettee, corresponds to 1.440 'Nt.
l 1
n.
ertastv fantalmeent Intce m - Pelaarv contatmacnt inteerttv eeses I
that t > e st r ywe t t and tiremeure suarreeston chamber are intatt and all af tbv followtes condittiins are sattafted e
1.
All non aut enat te cent atn-ent tactation valves on lines cannected to the reactor coolant systene or contalecent which are not re9utred to be open duf tet accident toMittene are closed. These valves eay be orened to perform notessary operational activittee.
l 2.
At least one door in each atr1och is cleerd and sealed.
3.
A11 automatte containment toelation valves are eeerable er deactivated in the teolated pasttien.
4 All httnd flaetes ar.d panwave are closed.
K i
Mend ent No. n.125 t
I
I i
l Lt m!T ! **C AFF.?Y S f it:;tt stMtuc l.
' '
- r
- 8. 4 M t ?
f 4
- i. t
- d t cLA!!!'4e 19tt:Mt?y 2.1 Ft*tL_ftA?a!N: ;n? t O't! ?!
I h.
i Fee es certtaatten of leep, egg,,,,
7 s.
let ten flew f ate and sore toerg,g Pont shall the gaa (!wn 84tes.tttt selttes be ellowed to seeeeg ;g :
el f ated thetaat power.
[
J l
These settinge esswee egetatten l
(wete:
f within the beste thermal hidrawlt d e s a rs,
ettteria. These ettlette ete1.M R 613 4 kwlft l
j end Mcit witman 1 taste el spetttttestan 3.3.h. If to deterstaeJ t>st e ttPet of 6tese It vesign etitet e 16 beleg viola'ed i
e6 4 61 64 I
elweleg operat6em, setted t,itisie,,ii i. n n.,ie ie,et,..e 3
operattee sithia plettetted tietto serveilleese toewstements let Mut etrag settetet are stven in opett!!:stion 6.1.3.
I J
8 73e MpM 304 hiocit t. rip g,
sett&ng small tet f
4 s s ( 0. 6 6 'a + e ll) g I
voetet f
paa tauca oestemt s
e g
an percent of rate!
l tmeteel power i
(J293 MWt) 1 I
s.oop rectr:21staca w
)
flow rate &n percent of rates trated lots i
i recarculat60m i1:*
i rate egests 3*,: a 1:e iv r.ri j
l i
i l
A. end ents hos. H, it, t[, pg, 73,125 j
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l.1 DASES:
LIWITISC 5AFETY SYSTEW SETTINGS RELATED TO FULL CLADD150 INTEGt17Y
~
1he abnormal operetional tranetente epplaceble to operettoa s
of the Browns Ferry Nasleet Plant heve been s n e l y,, e e throtshoot the spostess of pleased opereting eondstlone op to the deetse thereal power sendition of 3440 htt.
The emelysee vere besed upos plant operelion la aesordsi.s e eath the opere'.taa esp gtven la F6gare 3.7-1 of the PSAR la eedities. 32p3 kwt ie the lleensed meetoen pover level of Brovee Ferry Nuclear Plast. and this toprosente the unstans elesdy-seele powor ohiek shall set kaovtagly be eseeeded.
Costerveinas ie lasorporated la the treastent snelyses la eettnetsag the sentrolltag festors, soth as yold r ees t ly s ty see(fassent. sostrol rod eeram oorth, satse delay taae, peaking fseters, and exist pover shafes.
These festors are seteetod conservetively vith reepost to their effeet on the a pyl t se bl e treastest toentts ae detersaged by the surrest enelysie model.
This tremetent model, evolved over easy yeere, hes been subetentieted na operetien aa e senservstave tcol for evsluat6ag reestor dyaeste periormense.
Reauita obtesee4 from a General Electrie boiling eeter reestor have been sonpared vith ptediettone sede by the model.
The comporteens ead tosuite are sessartsed la Referosse 1 The y0id reactivity Cot #ficient and the Scran worth art dtScribed in detail in reference 1.
The astes deley t se e and rate of rod insertson a!!osed by the entlysee are someervettvely set eqeel to the losseat delay sad sloweet insertsen reto aseeptsble by Tesbatsel Spestfisettoos as l
further deestated in R e f e r es s e 1.
The e(fest of stres eorth, scres delay time and rod laserttom rate, all s os s e rv s t av e ly applied. are of greetoet einstfassese la the eerly porties of the avgotive resclivtty taserttoa.
The repad assortion of se s e t iv e resstivity ie eseused by the l
time requirementa for 3% sad 30% taserttoa.
By the ttee the rods sie 60% taserted. epyrostostely four dellare of seastive reeettysty hae been inserted vblah strongly teras the treastent. sad eescopitehee the deetred eflest.
The stees (c'
S0% and 90% tacertion are gaven to aessee proper sceplettom of the eapostod perforsease la the earlier porties of the treasient, and to establish the altiaste futly shutdoes etesdy-etsto scadttion.
For seelysee of
.he Ihermel sensequeswee of the tressteste e a
W L7 W ) lamete etoeafiod to eposaftselion 3.5.h as i
sossereeLivety sieused to estet prior to taittetaen of tha treasteste.
Thia shotse of setsg somservetave vsluee of sostrolltag peresttore sad inattetlag tressiest6 et the 1
deatsa pow e t lewel prodeses more psestatatts easeers thea eowld tosalt by ustag espeeted vsitee of scattel perssetera saJ easlyslag at hagher pover levels.
- A,9e n t"en t s N0 5. 4. II* U
f i
[
t r.s.setit i
j g,,,
t.e n awde, antwelat a ste as,.s t at e eree st lea at the telp.ettle
. e,ee g
the entle, eteleewiattee, flee esage.
Tb, ese gin t o the $414e. Lleg t t,tgeesee j
e, g o, g l e, s e g e e s t e t der the tr*4illed tely settlet goetwo fiw velat enehty; I
the *ee.
ease mtra wheth sw14 ogeog dweing eteady.etate opetellen it thereteet,.
e, inei e eet,.... i, -., i.. e e.. e e, e a. se i, e e4,i s, t.,,,. e t o. a. the ese.at pe e e d i e t e t h 4 6 ao en the tote 66 estabilited hp eyestised on.ittel god see...eeg eng gg ovettesed teatta.eutly tf t he la*, s e e t9 hm s y s t e e.
C. _e..etoe wete, te. iv.et tee'em sad 1%et.iten (raee.t m.ac. s...
.,,)
o l
7%e set pelat f or the les level satan is above the bettee of the separater stle t.
- > S leven has been wies in treatneat anal =ses deeltas with restaat la v e n t o ry aeteesse. The setells tepettet in fla.1 swbsettlea 14.1 spew that setse sad laelattaa el att protest !!aes testept eatn steam) at this lee,4 4Jegeasely pteteegt the f.et saa the pressete tatrier. because NCpt is treat er than 1.07 la all e sses, and erstee reeltwee does met teeth the safety salve settlags. 1he screw settaing is o p p e e s t a.a t e l y )! 4*thee below the aersal opetaling range and is thus adeqwete to 4
e.ete sporte.e stra a.
l
, 6ei.. e t,c,,i s e, s e e e s o
,n. g.
e.e_
t h.* tmbine stop valve Clctwre trip antiticates the pretsure. neutre's flus an.t he et flus inCreates that would result frta Cletuse of the step valeet.
With a trip setting of IC* of valse Cleture f ra n fw11 rasen, the resultant w
increate in heat flut 15 swCh that ade;wate thtrmal N rgin$ are M intainC1 even cfyring the worst Case trarstent that 41spel the turbine bypass valvel
,re ain Closed. (Reference 2) 1 I
J l'. T u r t, I n e re_m t_e s t V a l v e h4t cleeuta or Turbin. Trto Scram l
Twthlre control valve fant cleswee or turbine trip stras antittpates the l
pre sawre, neut ron flus. Sud beat fita tecreate that cc.nlit r e ss! t item timt r al v alve f.o..t c l o q u i t* due to load rejectinn er centrol valve eleswee r
.lue in t urh toe t r ic t ca6h withewt byp.ans valve raphility. Tbc reacter prettstlen system anttistes a scram in less than 33 stiltseconds aftet the start of contfol valve fast closure due to lead rejection er control l
valve closure due to turntne trip. This scess is achtsved by rapidly reiluttet hydraulle centrol oil pressure at the main turtnne Contrel valve attwater 'dist d.-g valves.
Thin Ingt Cf prestwee It lented by Cretture StfitChel whOle Centatt$ fore the one*0wt*0f'* t=0* tsiCe IOMC In:wt to the reatt0F Pr0ICCT,1C'l lytt;'d.
i TMt trip gettino, a ncainally W ttreater C1clure tire #Pd a ellf f erent
- dive Char.tCiti tglit f roe thJt Cf the turbine 5100 valve, C0r4tne tv ProNte trentients very limilar to that fer the stCp valve.
Relevant transient analyses are ditCwined in fleferences I 4F3 2.
Yhts gerse It broenteJ = ben twretne steam flw is telow 3*! cf rated. as egelbre*
by ti.rtiot first state prestwre.
- s i
icend ents Nos. H. N. Nd 125 l
1 I
f
\\
l E'I
- '.t.'
- 3. J. 4 t, se se t os low wat**
level set Kl C, elessN _sein estat fee tastesteen o f 18 ec t am e s t e.s.
one tore e**ar tiet3tioa val e.,
ye stattaat uti
_ y a.
Twaie eestema metateta aset.ete coetent eoellat.ith the enjeet6*e et pecesseLag e.ecessive slee tlavestery sad seeelde s Two seetta of these erstees to edetven ete potters the l t ii.a eeeeestwees.
to te.es en the speetties lev it.. i.
a teWee feat.
est,ses seeeetes sateeet seeen set peine sat taitte.
esists, teenesent tsu steeniteets te t s e e s sa 14 et the e
t he se s eas t i t.a.
...sia. tee 6eth te.e f.en 4 tan erstee reseevees= e4ee.sie estetr ees.tt t.
e.e.......
1.
"BWR Transient Analysis Model Utilizing the RETRAN Proyan."
t TVA.TRSI.01.A.
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3
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A ene ent No. if.125 n
i
1.2 jiSJ 5,8
(
5
_REACTot r00LANT SYS*t'4 INTtCRITY h
The safety Itails for the reactor coolant system pressure have l een selected such that they are below pressures at which it can be I howTL 1
that the integrity of the system is not endangered. Hewever, tl e pressure safety limits are set high enough such that no foresee <
Die l
4 circumstuces can cawse the systen pressure to rise over these limits.
l The pressure safety limits arc arbitrarily selected to be the lowest traestent overpressures allowed by the sprit able codes. ASME notler and Pressure vessel Code. Section 111 and USAS Piping Code. Section 531.1.
The design pressure (1.250 psis) of the reactor vessel is estaMished such that, when the 10-percent allowaace (125 psi) allowed by the ASME l
Botier and Pressure vessel Code section 111 for pressure tranatents is adJed to the design pressure, a transient pressure limit of 1.375 psts 1s established.
i Corrascondingly, the design pressure (1.148 psig for suction and 1.326 l
pssg for discharge) of the reactor rettrculation system piping are such that when the 20-percent allowance (230 and 265 psi) allowed by USAS l
Pipics Code. Section 531.1 for pressure transients are added to the design pressures, transtant pressure limits of 1,378 and 1.391 psig are established.
Thus, the pressure safety limit appli6.able to power operation is estabitsbed at 1.375 psig (the lowest transient overpressure allowed by the vartinent codes). ASME 5:11er and Pressura Vessel Code. Section 111. ard.V5A5 71 ping r
Code. Se:tter. 331.1.
i L The currect cycle's safety analysis concerning the most severe abnormal
/f operational transient resulting directly in a reactor coolant systes pressure increase is given in the reload licenstng submittal l
for the current cycle. The teatter vessel presswre code limit of 1.375 psig i
given in swbsection 4.2 of the safety Analysis report is well above the peak pretsure produced by the overpressut* trat> stent described above.
Thus, the press;te saf aty limit applicable to powe es. cation is well above the peak pressure that can result due to reasonably espected overpressure i
l M1;;her design pressures have been established f or piping within the reactor j
coulant system than for the reactor vossal. These increased design pressures 6
ct"ate a consistent design which assures that. if de the pressure within th. reheter vessel does not exceed 1.375 pais. the pres:4"* within the pietag cannot exceed their respective transtant pressure limits due to
(
st.itic and pump heads.
The safety limit of 1.375 psig actually applies to any pntet tri the reactor ve'.se14 however because of the static water head, the hashest prn*ure I
point will occur at the bottom of the vessel. Because the pressure is not monitored at this p.oint. it cannot be directly determined if this safety ltant has been violated. A.lso because of the potentially vatying head level and flow pressure drops, an equivalent pressure can.not be a priori determined for a i
s 2s A !
A, mend ents No5. 3! It.US
~I t
SURVEILLANCY prQU!PEMENT LIM 2 TING CONDITIONS FOR OPEPat!0N
- s.1 MACTOR 1 ROTEC f o'4 MY9 TEM 4,1 R/. ACT0!! l'!t0Tl CTf 0?! SYtTr.4 k
R.
Th e RPS pow e r se nt t erta s 1.
Tve RPS power eenitoring
[
rysten tantrumentattes channels for each inservice i
shall be deternised operable RP5 NG sets nr alternate scur, e shall hi operable.
At least ease per 6 seethe by perf ormasse of chessel f asetional te st e,
- 1. With one RFS electric pescr Penitoring thanrel for 16 service RF$ M* set or alternate power supply inoperable, restore the l
ineperable channel to t
epcrahle atatun within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> er reerve the associated RFS MC set or alterr. ate pober supply frem service.
2 Tith both RP$ electric poeer sentterieg ehemsels for sa laservice RFS be set er alter-ne te pow er espply taopershle.
testore et leset ese te operable statse witkis 30 s tatte s er r eeev e the sesestated tr$ u0 set er l
al t e r na te pes or supply f r om eerinso.
t 1
A enc ent No. 125 32 i
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REACTOR PROTECTION SYSTEM (SCHAr a NSTRUMENTATION REQUIREMENT
- 2' U
i.
Mio. No. of Opere.ble Inst.
Modos in Which Function C!scenets riust lie Operalite Per Trip Shut-St ar tup /:fot Systemtll 8 7.- ) Rip Function Trip I.evel Sett_ int, doun Refuel (1)
Standt,y _
Run Action (l) 4 IIain *; team Line Isola-tion Valve Closure
$101 Valve Closure X(6) 1.A or 1.C i
j 2
Turbine Ct,at. Valve i
Fast Closure or
>550 psig X(4) 1.A or 1.D Turbine Trip 1
l 4
Turbine Stop Valve i
Closurc
$101 Valve Closure X(4) 1.A or 1.D w
\\
i 2
Turbine First Stage Pressure Permissive not >l54 psig X(18)
X(18)
X(18)
(19)
(PIS-1-81A&B.
PIS-1-91 A&ll)
)
2 Main Steam Line High 3X Normal Full Power X(9)
X(9)
X(9) 1.A or 1.C Radiation (14)
Background (20) 2 Low Scram Pilot Air
{,
Header Pressure
>50 psig X(2)
X(2)
X Y
1.A i
W e
4 i
N0?($ F00 7AP.E 3.1. A 1.
There shall be two operable of tripped trip systems for each function.
If the minir.wm nutder of coerable instrvmont channels per trip system canmot be met for one trip system, trip the inoperable channels or entire trip system within one howr, or, alternatively, take the below listed action for thtt trip function, if the minimum number of operable instrument channels cannot be met by either trip system, the appropriate action listed below (refer to right-hand column of Table) shall be taken. An inoperable channel need not 60 placed in the tripped conditi
- 9 -
where this would cause the trip function to occwr. In these cases, the' incoerable chanrel shall be restored to operable status within two v.vwes, or take the action listed below for that trip function.
r all A.
Initiate insertion of operable ' rods and camolete insertion o operable rods within fewr hours. in refueling mode, suspeno all operations involving core altecations and fwlly insert all operable centrol rods within one hour.
B.
Redsce power level to IRM range and place mode switch in the
' StartsplHot Staneey position within 8 howes.
C.
Reduce twreine toad and close main steem line isolatten valves within 8 hoves.
D.
Redsco po.er to less than }CS of rated.
2.
Scram discharge volLTe high bypass may to used in shutdown or refuel to bvpass scram discharge volu-e scram and scram pilot air header low presswee scram witn coatrol red bloca for rta: tor protection system reset.
3.
CELETID.
4 8ypassed when turbine first stage pressure is less than 154 psig.
5.
IRMs are bypassed when APRFs are onseale and the reactor mode switch is in the rwn position.
6.
The design permits closure of any two lines witnout a scras being initiated.
7.
- en the reactor is swberitical and tre reactor water temperature is less than 21281. only the folio.ing trip functions need to be o;erable:
Q A.
Mode sultch in shutdssi B.
Manwel scram C.
High flwu IRM D.
Scram discharge volw e high level E.
Scram pilot air header low presswee 8.
Not reagired to be operable when primary containment intogelty is not reasired.
9.
Not reewired if all main stearlines are isola'ed.
35 A,mendments Nos. 20.n7 J 25
+
J l
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3 O
k 3e+
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?
Tastt 4.1.4 SfACT9E FEDTICTION ST5ffM (5CXAtt) tie 5T30HENTATIO4 tm8CTI0stAL Tt315,
U 9tDsDef FWCTIONAL TEST rarQIltMCit5 Fat SAFCit Istis. AND Conta0L claculTS*,* {-
' I L ',
)
~
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Crmac_(R Fvaltlaget' reeg '
wret.,4 Fregeese, til-i
- tM e 5.ttch te %=tdows A
flect PeJe Telects try 51.wtJown
, tech Refeettog outege Itseast Scram Al Trip CheeseI,eed Alero.
. torty ) Neethe,',_
i I
. 8; Esti
'., -?
.. g Righ Fles C,
left Che'amet-e' 4 Ale s= (t)
- 0ece f ar Ween thertog 'sefeelte e
f.
- * ** 'eed Refere tech $tattwp e
IneperatI==
C Trip theenet *eed Ales = (4).
.'Osco Fe't Week tNetog 8efuella
'* iJ Sefere tecle Stettop e
s
& FEM l
,'E
[* [ ;.
~
'1 algb Fles (111 scree)
C
'~
TetyOstht$cters(4)
BeforefechStercepsedWeekt 1 inies segetred tag te operable
=
litp.h FIux. (Flow Blased)
.B TripOutput[* Relays (4)~
,,0nce/Ucek Dish rl.a (Fixed Tr tp)
. a l Trty Oerywt selaye (4)
- 0=cclueek 3.
~
t Iroparattee A
Tetp Outret Selare (4)
Once/Weth 2
. i.
i tuecate 3
Tety out,' et matere (t) i oocetWelrh i
fen Blas 5
(6)
{
(I)
,3 }, ',
8tsh Bescter Freeewr.-
R l' Trf y Chsomah ama Alste ( 7)..
.Oscif Imonth (pis-3-nAA e us. c D) l,-
e I Ela'lirTVL{TeeA.b,)
Trip Cheesel es4 *3erm ( 7)
- pece/ Imontis B1 6
=
Trty Cheneet eed Aler= ( 7)
'Bec ef I snonti.
e,scret im veter teest g.,
sig%(tts-3-203 A-0)
B water Levet le Scie. Otecharge Teek
. ?. I IIost Swit c hes A
Trip ChIinpel $nd.tlaren
- OnCc/montli
. * [.
(ts-8s-4s c-r)
B Trip Cidnnel and I.Intan ( 7)
Osicc/ month Electronic !.evel Switches (ts-M-45A. Eb G. 31) tta in St e aan I.Ine titr.le Itaaltat ion t.
I r t p t luiuse.I.in.l Al.irie (4) onec/1 anonths (8)
title 4.1.1
[
REACTOR PROTECTION SYSTER (3CRAMI INSTWCMENTATION f00CTIONAL TESTS MINIMUM FUNCTIO 4AL TEST FREQUENCIES FOR SAFETY INSTR. AND CONTROL CIRCUITS Mlaimun Frequency (3)
Group (2)
Functional Test g
5 Mala Steam Line Isolation Yalve Once/3 Noeths (8)
T Closure A
Trip Channol and Alare Je Turbine Control Valve Fast Once/Nonth (1)
J Closure or Turbine Trip A
Trip Channel and Alarm
~
- Turbine First Stage Pressere U
Permissive B
Trip Channel and Alarm (7)
Every 3 Nonths.
l (PIS-1-81 A&B, PIS-1-91 ALE)
Once/Nonth (1)
Turbine Stop Valve Closure A
Telp Channel and Alarm Once/6 Months Low Scram Pilot Air A
Trip Channel and Alarm Header Pr*ssure I*S 85-35 A1, A2, B1, & B2 w
on 4
e
k*
e i
):0Tr.S_Jn14 TAsLE 4 1 A s
1.
Initially the minimun frequency for the indlested tests shall he nnee I
por conth.'
t, 2.' A description of the three groups is included in the Bases of this specification.
3.; ronctiorisi rests hre not required when the svt ees are not recuireri to he oper.Wie or are oper.itinr. (i.e., aircady tripped).
If tomtn are nissed. they shall be perforced prior en returninr, the systems to an operable statun.
4 This instrunentation is excepted from the instrument channel test definition. This instrueent channel functional test will consist of injectinc a sirulated electrical signal into the measurement channels.
5.
(DELETED) 6 Tre hunctional test of the flow biss natunrk in perfnrned in accordanec with Table 4.2.C.
4
'. - 3 7.- Tunctional tet.t consists of the injection of a simulated signal into
'ths. ele:tronic trip circuitry in place of the sensor signal to verify operability of the trip and ala m functions.
8, The functional test frequency decreased to once/1 m tme ta reduce CM11engss to relief valves p e r.WRIO 0 7 p, I t a.ii n, y,, j, g,
d 4
39 Amenc.ents rios. 82,105,701,125 I
+
TAILE 4.1.8
{
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNEL; h
j Instrument Channel Croup (1)
Calibration Minimum Frequency (2)
E IRM High Flus C
Comparison to APRM on Controlled Note (4)
Startups (6)
E:'
{[ APRM High Flux Output Signal B
Heat Balance Once every 7 days n)
Flow Blas Signal B
Calibrate Flow Blas Signal (7)
Once/ operating cycle LPRM Signal B
TIP System Traverse (8)
Every 1000 F.frective Full Power Hours High Reactor Pressure Standard Pressure Source Once/18 Months (9)
(PIS-3-77AA. BB, C. D)
B High Drywell Pressure Standard Pressure Source Once/18 Months (9)
(PIS-64-56 A-D)
B
, Reactor Low Water Level Pressure Standard Once/18 Months (9)
(LIS-3-203 A-D)
B High Water Level in Scram Discharge Volume Float Switches (LS-85-45 C-F)
A Calibrated Water Column Once/18 Nonths Electronic Level Switches (LS-85-45 A. B. G. H)
B Calibrated Water Column Once/18 Months (9)
Main Steam Line Isolation Valve Closure A
Note (5)
Note (5)
Main Steam Line High Radiation B
Standard Current Source (3)
Every 3 Months Turbine First Stage Pressure Permissive (PIS-1-81 ALB. PIS-1-91 A&B)
B Standard Pressure Source Once/18 Months (9)
Note (5)
Turbine Stop valve Closure A
Note (5)
Turbine Cont. Valve Fast Closure A
Standard Pressure Source Once/ Operating Cycle en Turbine Trip
~ Low Scram Pilot Air A
Standard Pressure Source Once/18 Months Header Pressure PS 85-35 A1 A7. B1 & B7
D p r t' S Fok T A L'L ' 4.1. 3
!)
1.
A description of three groups is included in the bases of this specification.
2.
Calibrations are not required when the systems are n ot required to be operable or are tripped.
If calibrat Lons are missed, they shall be performed prior to returning the system to an operable status.
3.
Tne current scurce provides an instrument channel alignment.
Calibration using a radiation source shall be made each refueling outage.
4 Required frecuency is initial atartup following each refueling outage.
5.
Pnysical inspection and actuation of these position switches will be performed once per operating cycle.
6.
On controlled startups
, overlap between the IRM's and APRM's will be veritied.
7.
The Flow Bias signal Calibration will consist of calibrating the sensors, flow converters, and signal offset networks during eacn operating cycle.
The instrumentation is an analoq type with redundant flow signals that can be conpared.
The f low comparator trip and upscale will be functionally t.euted accord:nJ to Table 4.2.C to ensure the proper operating during the operating cycle.
Refer to 4.1 Bases for f urther e xplanation of calibration f requency.
8.
A cc pletc tip syste '.raverse calibrates the t?Ru sirnals to the 5.rocess co puter. T5c individual LPR.M meter readitta vill be adjusted as a nini un at the beginning of each operating cycle before reaching 100 powe r.
9.
Calibratien consists of the adjust =ent of the pri=.ary sensor and asseciated components so th'at they correspond within acceptable range and accuracy to known values of the para::eter which the channel noniters. including adjustment of the electronic trip circuitry, so that its output relay changes state at or core conservatively than the analog equivalent of the trip level setting.
s
.t
.s Y
Al endments Nos. U.22.125 O
3.1 BASIS Too reactor protection system automatically initiates a reactor serem tu 1.
Preservi the integrity of the fuel cladding.
2.
Preserve toe Integrity of the reactor coolant system.
Minimi2e the energy which nust be absorbed following a lost of 3.
coolast accidont, and pre.,e ts criticality.
This specification provides % limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when W%en lestrument channels eay be out of service because of maintenance.
necessary, one channel eay be made Inoperable for brief intervals to condset required functional tests and calibrations.
The reactor protection trip system is supplied, via a separate bus, by Its own high inertia, ac Potor generator set. Alternate power is available to either Reactor Protection System bus from an electrical bus that can receive stendby electrical power. The RFs monitoring system provides an isolation betwen non-class lE power supply and the class IE RPS bus. Thl. will ensure th.t failure of a non-class IE reactor protection pc.or supply will not cawse adverse interaction to the class lE Reac*tr Protection System.
The reactor protection system is mada up of two Independent trip syste-s (refer to Section 7.2, FSAR). There are usually four channels provided to monitor each critical parameter, with two channels in each trip The owtputs of the channels in a trip system are comoined in a system.
The logic such that either channel trip ullt trip that trip system.
sirNitaneows tripping of both trip systets will produce a reactor scram.
This system meets tho intent of IEEE. 279 for Nwclear Power Plant Protection Syste-s. The system has a reliability greater than that of a 2 out of 3 system and semenhat less than that of a I out of 2 system.
With the enception of the Average Po,er Range Henitor (APM) channels, the intermediate Range Monitor (IPA) channels, the Main Sten isolatien valve closure and the Turbine Stop Valve closure, each trip system Icgic has one instrument channel. W on the minimum ccedition for operation on the number of operable instrument channels per untripped protection trip system is met or if it cannot be pet and the ef fected protection trip system is placed in a tripped condition, the effectiveness of the protection system is preservedi 1.e., the system can tolerato a single f ailure and still perform its Intended f unction of scra rsing the Three APM instrument channels are provided for each protection reactor.
trip system.
Each protection trip system has one more APRM than is necessary to reet the minimum nweber roowired per channel. This allows the bypassing of one APM por protection trip system for maintencnce, testing or calibration. Additional IM channels have also been provided to allcw for bypassing of one swch channel. The bases fer the scras setting for the IPN, APM, hig% reactor pressure, roactor low water level, M$1V closwre, turbine control valve fast closure and turbine stop valve cicswee are discwssed in Specification 2.1 and 2.2.
42 Amendment No.125
3.1 OsS($
modas, in tne pomor rangs the APM system provides roovired protection.
Ref. Sectien 7.5.7 FSAR. Thus, the IM System is not required in the Rwn The APRl4's ar.d tho IRM's provide adeawate coverage in the startup meca.
and interrediate range.
The high reactor pressure, high dry. ell pressure, reactor low water level, Icw scram pilot air header pressure and scram discharge vol6me hig5 levet scrams are required for Startup and Rwn N das of plant operation. it.ev are, therefore, reewired to be operational for these m des of reactor operation.
The require ent to have the scram functions as indicated in Table 3.1. A cceratie in the Ref wel mode is to as'swre that shif ting to the Refuel mode dweleg reac*cr power operation does net diminish the need for the reactor protection system.
Because of the APM downscale limit of 131 when in the Run mode and hig5 level limit of $ ISS when in the Startup Mode, the transition between the Startup and Run Modes erwst be made with the APPJ4 instr 6 entation Indicating between M and 15% of rated power or a control roJ scram will occur. in addition, the IM system siyst be indicating teles the High Flus setting (12oll2$ofscale)orascramwilloccurwhen in the Startup Mode. For normal operating conditions, these limits provide assw*ance of overlap between the IR*4 systtwn and APRl4 systom so that there are no "gaps" In the power level indications (i.e., the power levet is centinwowsly mnito*cd f ecrn beginning of startup to full power and frcan full power to shutdc.n). Wnen power is being roduced, if a transfer to the Startup enode is trace and the IR.*Ps have not been fully inserted (a male;grational but not irJessible condition) a control rod block irrediately occurs so that reactivity Insertion by control rod withdramal cannot occur.
The Ic. scram pilot air header presswro trip performs the sarm function as the hig% nater level in the scram discharge instru. tent volume for fast fill events in which the high level instrument response time may be inadequato. A fast fill event is postulated for certain degraded control air events in which the scram outlet valvas unscat enougg to allow 5 gsr, per d'ive leakage into the scra.t discharge volw 4 but not enow;S to cowse control rod insertion.
44 Amendment No. U2,125
e e
o g
TA41LL
- k. I. A PRIMAlsY CONTAlleet :df AND la'aCTtsie a+UILDING I'4J!ATION IM4TRO*tt* CAT 88eN 3
- finasun 1:o.
g Ins t r smt:ne M,
le..nr.els c;,_ rat.le
- s Penerks IEMN1N 3J )
Nect inst Tra'_kwe' Settar n
,Actlon.(Il _
ao Instrument Channe t -
2 S id".:m e-we. e ! aeree A or 1
Below t r ap act t ing does the 2
f a and r) tollowing:
4 a.
Ins t a.at e? 8e.setor Sai!Jang peactor tow water tevet (EL:
isoia.aon (LIS-3-203 A-D) b.
I n st a at.1. Fr a mar y Cont.ar.sp nt g
!solatnon c.
Ina t a at es DWS D
tis D
1.
Above trap settang asolates tre tu Instriament Channel -
~ 15 W ay shutdoom coolars; sactaon walves IJO
- 1 Beactor afigh Pressiare of the ano sistem.
U s,
2, b 2
Instrinsent Channel -
t HO# at.tve vessel aero A
1.
Below trip setting initiates Man Stean Line Isc.l.t acn w
y;3 Peactor tow water 14wel "e
(LIS-3-56 A-D)
A or 1.
Above. trip setting Joes tt.~
~
44' J
Instrument Channel -
5 2.5 p sg
~ following:
to (B and El 3D ',, $ * **** ' "' *
- Magla Dr ywell Pressure (6)
(PIS-64-56 A-D) b.
Inst aatec Pr imary Cont.in,ent Isolation Initiates SGTS c.
8 1.
Above trip setting initiates Maac 2
Instrument Channel -
S 3 tires normal rated steam Lane Isolation full power tackgrourmt Nigh andiation Maan Steam Line Tesanel (6) 3 1.
Below trap settang initiates Maan 2 325 psig (el Steam Line isolatnon 2
Instrument Channel -
1 sis Preasure Main Steae Above tr op setting initiates Maan Line (PIS-1-72, 76, 87, 86) 8 f.
i 1001 of rated steam fl<me Steam Lane Isolation 283)
Instrument Cftennel -
Righ Flow Mean Steam Line (Pd15-1-13A-D,25A-D,36A-D,SOA-D) e a
O
e l
t if3tt 3.1.0 tuin1NDrTatton TRAf tagitAtt.5 CE CosGRCLS T32 Cosf AFD CettfatteEWT C00ttmC SYSTEN!
cx
- 1
<D "e
stei e me.
seearn e
~
DantebSe per Arttee Mddl _
Ytte tevel Settlag 7
F= settee o
A
- 1. sete. est,setzte setteeted trCs.
2 f..t,.eet Ch..et -
e + 7e
- ete.e.. e.e t.....
...oer t u.a.
t,,e1
- 1. Nltiplier teleye lettlete BCIC.
E.
{LIS-3-58A-D)
A l
a.. ease t w w.s.r t e.t
~e 47a*ek va vesset asse.
N 2
estr
.at t h.. 1 -
L l
M (LIS-3-58A-D)
- 1. Below tatt eettleg lettletee Cab.
3f&* ehrwe esenet sete.
A 1
Meltipiter relays 1sttlete LpCt.
e 2
Iaett m aat th== a4ector tav Vanes 14 1 (LIS-3-58A-D)
- 3. hittplles.te1*F f ra. C15 tettf etse
~
occidset eigenet (11).
o-m
- 1. Below trip eettlege is coalmetI*a A
i 2(ILS Ru:rwen t th ---1 e 338~ above vessel eerw.
witti arrwett mish prose.ee. te*
a..c t.
t., v.t.r ta t
- I* ' I '" I '"* ' * * '" ' I I' * * * * #
- Y j
steer end CSS er as pep runstes, (LIS-3-58A-D) i tattletes ADS.
j
- 1. Relow tely settlag persteelve for A
lite) to.te m et O smet -
g 344" above wesent sere.
laittattag else.te se ADS.
f s..cser tu v.ter tsvet
- se.t..t.
(LIS-3-184. 185)
- 1. Selow trip settles preenste taedver-3 tene rms.t Cb==1 -
- 311 3/16" above westet sete.
A test eyeesttee of esetstemaet ortet 1*
- e :*e t.* W.se e t. ret (21) case higic) dertog ecstdest emeattlee.
(LIS-3-52,62)
- I* t**B**#~
P **E8I*E P I* I*I',* I',I,,,,,,,g c,'*,,g,,,,,,,,,7 A
2 to as w. n,..,g g,,,I 5 p*IS
,y Da m it uggs rg,,,,,,
Jeting accidaat condialeae.
(PIS-64-58E-H)
O O
3=
3 en LADLE 3.2.5 (feettaved)
Sg stet _
r0
- Sporebte For 2
7e t= 3e it) r cgg'a Tate Levet lettt*R Acttaa S emerb e
- 1. abese tely settlas to eoeleacete. =tth a
sa
~
I***' " *8 N ""*I
, 2.5 potg te* e e ter preeeeve inteletee Ctt.
Derwell esgg reeee.,*
- -le t.tler est e,. salutete mect (PIS-64-58A-D)
- t. wittpiter relay f ee. C11 tattletes C
ecst4ent signet.tt1) 3 Immers eas then et -
4 10"ob..e veeset seg,
a I, se,e arty set tles tripe sec tetsle-a steen pope S eet t e s te. tr t e r Em.e t e
(LIS-3-56A-D)
- 1. abe e arty seastag respe restes.le-A 3
leear==ent N. eel e tt's pet ctea pops
~
Seet ser utgt. res e wre 3
(PIS-3-204A-D) i,aueeart,.ettlgI e-s ate 2h A
s.sta e.
N ei -
e 2.s,.t.
see saector pressere totatetse LFC..
Se7 west migh treenere (PIS-64-58A,D)
- 1. abee Etty setting im eeslwedtoe M 2(84)
Isott w at d's.eet-
- e. 2.5 pelg low seector water levet. 4tywelt blh A
Drywell stah reese. e
'""*'****"*'*I******'
(PIS-64-57 A-D) es asa ev., r tag, tais t ene s aDi.
A 1.
3r!sw tr:r ettelar perat es tre (s- :;ent?4 I
Seesr ent N et -
t.30 pets g 31 C*$ end * ~-CI slit s e t ' e. e s t.as.
Seacter 1,w ree e.g.
(PIS-3-74A&B)
(PIS-68-95,96)
- 1. Re:treulettos discherge vetoe A
3 1.earweet N aset -
230 pelg i 13 ec.settoc.
meester t,. tesse.ge (PS-3-74AAB)
(PS-68-95,96) i
e Tant.C j.2.C IWWI:tsMFerTATIon THaT tutflaTF3 pnD MLitCES 39
. - ope e
,,,s Csianneis per FaceetIon T-tp_t.
,t e,e ttnr TeBy F.enetien (5) z O
483) arart tipscale (Flee Stas) 10.6658. 4M (7) et1)
- Mt Upwale (Startup lWel (9)
< t71 M7 etal armt Demnscale (9)
A 3" ep et3) arpet Inope-ate,,
- (tw) y 7t?)
- M Up*cate (Flee 5'as) 10.64* + tc* (7KI O a
21il pre: Dnunwale (9) 7(7)
Pft* Inoperattwe (for) 68 O IBM Upscale (8) 108/175 or retI scale 6(t) tret IHeenscale (1) (8) 3 sit?S er ru'l scale 6(t)
IINt D-tector snot in Starte Posttlon (8)
(ft) f30=)
68 0,
1p*1 Inope-attwe (8) 5 cee# s/==e.
Ift) (6)
- pH Itpwale (8) 18780 181) (%)
$99firran=cate (4) f8) 2 ) co *m'si--a.
B 3(O (4) sm netector not in st=-t o ra=lesen tant 9)
(*t)
(103) 3tt) (4)
$rN I w attva (9) t
- - e-7( O -
Flaa stas Cne pra er gent terre-me,t.,r.et ci.=eaa v
llSt sectacelettee tiene 2( O flee Stas Opscale i
P-8 9 tack trirtc ria t
147 ps'* t cne (1-e,.-..--s...
WJ' p s*
- stat (r".%.f.* a,'l 75or l(I2)
High Water level, la tiest
~s 75 gal.
Seram D(w harge Tank (LS-85 45L) 1l12)
High Water level (a fast $(ram 3 25 gal, Diuharge Tank (LS-85-45M)
e
.s.
s.
T A tti.f.
3., 2. f' LURVEI1.LAlef t: IBGT 84ts*ttfe! AT 18188
. : s,.
a g
ini.e. 3 or Type Indication
..a Notes
~
and panese 3
operable last r u. cat Inst r. ament instrument e Indicator - iss-to
~ su (23 ps re..e net s LI-3-SHA
. et o r u t er t.ev 41 I:r no-
,4, 2
LI-3-58B til on s dicator o-12eo psiv ; ::::
PI-3-74A
,,,,, c, o,,,,,,u,,,
.s 2
- w PI-3-74B
, tji,, (21 (31 pr yu ell F r es's'dfe "8 '
secorder 0-to gsta p he Indicator o.so psia w
XR-se-so a.
.a 8..
s s.s l I s. '* '...
2 L
PI-64-678 Recorder, indicator
.g l (23 0:
u TI-64-52AB dryuelt7,.ponture....
N o-s oo* 6-2 XR-69-50 secorder e-eooor at gis so suppression Crissayr Air s3 t
o I
XR-6g-s2 l
Temper ature
..g.,.
a 8'-
.4 t
3..,, ;,
I a.
. 's f
6V Indicating I
.. control Rod g*estt,lon Lights
.*l
- st..
g spM, IBM, test's 1
(1) (2) (3) tel N/A 1
tseutron Monitoring O to lect power 3 f
l 1
tuA 8
l ia l s t e '.
Alar:s at 15,psig PS-64-67B Drywell cresguge, Alar. tr te.p.
TS-64-57AA Dryueil r,,pe,atore and
> 2eier and i
s in (21 On tel PIS-64-SilA&
. cress re and Tin e rressure > r.t 61 *. *
- 15-64-67A after Jo.inute I
delay 3
Ind1Cator o to Isot III I
i LI-se-2A CAD Tank "A* IAvel (Il 1
Indicator o to 1004 CAD Tank *D* Ievet Lt-84-11A 1
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- ~
N*,*".t".***"$
s
>M 3 tcM {,6-e
w -- g mm -sst! 1.7.?
From and afscr the 'ists t.%4t one of caso parastsers isreduc (1) during sne rue:eeding utrsy daye us. lese eu:n ine.:=ent.at.icn is ec=nas made cpera. ale.
F:cm and af ter the date tsas one cf esse para.sete:s is ses it.dicated in ce censrol roca, censinued operatica is
(:)
per.a.1:sttle durint ce succeeding seven d.tys v.:.'.eas euct l
1.n e tt u.n r.a t a ti en i s s ce ne r o.a d e ope r a rl e.
an[1fone If the reevir"esents et estes (1) and (1) cannet be set, of the indications ca.v.st be restored in (6) hevra, as orderly (3) shutnevs shall be initiated and the reactor shall be in a cold sendition vitain *6 hours.
These surveillance Lass:-1.sente are" ceneidered so W redudans (s) sa 1ssen ot.her.
be:h the acoustic - nitor and the (3) Tres and af ter the date th.2:
~
te=perature indication on any one valve f ails to indieste is the c:ntr:1 ro.-s, cc :tsued c;eratics is ;er:L:sible durint the su::escins thirty days, unless cne of the two =enitoring channels is soccer cade If both the pra.=.2ry and see:ndary indicatics on any $7V ::11 c;crabic.
pi;e is inc; rsble, the torus t==per::vre vill be =enitored et leas:
La observe acy unex;'sined te ;erature in rea:e wSL:n
- nts per sht!:
sight he int.1:stiva o f an c;en 17.V.
(6) A channel consists of S censors, ons f rom cach siternating toru.
a 1
ba y~
Seven sensors nust be opersble for the ch nnel to be operabic.
(7) Wn.ine of thne instruments is inoperable for raore than 7 dcys, in Iten of any cther reper: required by spe:ifiestion 6.7.2, prepare and suhnit s Sper.ial Report to the Co esission pursuant t.i speciftention 6.7.3 within the nex: 7 days outlinin; the a:ti:n taken, the enuxe of in:perability, and the plans and schedule for restoring the system to opersble s:stus.
in the power operation, startup, or hot shu:dm.n (S) With the plant condition and with the nu:ber of operable channels less than the required operable channels, either restore the inoperable channel (s) to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or initiata the preplanned alternate eethod of monitoring the appropriate parameter.
a -.,.
p., s
~~-
Amend ent Nos f3,f3,125 80
TABLE 4.2.A SURVEILLANCE REQUIREMENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION 5l Function Functional Test Calibration Frequency Instrument Check l
55 Instrument Channel -
(1) (27)
Once/18 Months (28)
Once/ day ll Reactor Low Water Level (LIS-3-203A-D)
Instrument Channel -
(1)
Once/3 Months None j;
Reactor High Pressure un Instrument Channel -
(1) (27)
Once/18 Months (28)
'Once/ day Reactor Low Water Level (LIS-3-56A-D)
Instrument Channel -
(1) (27)
Once/18 Months (28)
N/A High Dryw 11 Pressure (PIS-64-56A-D) 3:
Instrument Channel -
(29)
(5)
'Once/ day High Radiation Main Steam Line Tunnel Instrument Channel -
(29) (27)
Once/18 Months (28)
None I.ow Pressure Main Steam Line (PIS-1-72. 76. 82. 86)
Instrument Channel -
(29) (27)
Once/18 Months (28)
Once/ day High Flow Main Steam Line (PdIS-l 13A-D. 25A-D. 36A-D. 50A-D)
Instrument Channel -
(29)
Once/ operating cycle None Main Steam Line Tunnel High Temperature Instrument Channel -
(1)(14)(22)
Once/3 Months Once/ day (0)
Reactor Building Ventilation High Radiation - Reactor Zone
TABLE 4.2.D SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE OR CONTROL THE CSCS -
Function Functional Test Calibration Instrument Check 73 Instrument Channel (1) (27)
Once/18 Months (28)
Once/ day I
Reactor Low Water Level 5
(LIS-3-58A-Dn Instrument Channel (1) (27)
Once/18 Months (28)
Once/ day Reactor Low Water Level
[{
(LIS-3-184 & 185)
Instrument Channel (1) (27)
Once/18 Months (28)
Once/ day Reactor Low Water Level (LIS-3-52 & 62)
Instrument Channel (1) (27)
Once/18 Months (28)
None Reactor Low Water Level (LIS-3-56A-D)
Instrument Channel (1) (27)
Once/18 M6nths (28)
None so Reactor High Pressure (PIS-3-204A-D)
Instrument Channel (1) (27)
Once/18 Nonths t'28 )
None Drywell High Pressure (PIS-64-58E-H)
Irstrument Channel (1) (27)
Once/18 Nonths (28)
None Drywell High Pressure (PIS-64-58A-D)
Instrument Channel (1) (27)
Once/18 Nonths (28)
None Drywell High Pressure (PIS-64-51A-D)
Instrument Channel (1) (27)
Once/18,onths (28)
None M
Reactor Low Pressure (PIS-3-74A&B. PS-3-74A&R)
(PIS-68-95 PS-68-95)
(PIS-68-96 PS-68-96)
l TABLE 4.7.C SURVEILLANCE REQUIREMENTS FOR INSTRUMENTATION THAT INITIATE ROD BLOCKS Function Functional Test Calibration (17)
Instrument Check g-y APRM Upscale (Flow Bias)
(1)
(13)
Once/3 Months Once/ day (8)
APRM Upscale l
(Startup Mode)
(1)
(13)
Once/3 Months Once/ day (8)
P APRM Downscale (1)
(13)
Once/3 Months Once/ day (8)
APRR Inoperative (1)
(13)
N/A Once/ day (8)
RBM Upscale (Flow Bias)
(1)
(13)
Once/6 Months Once/ day (8)
RBM Downscale (1)
(13)
Once/6 Months Once/ day (8)
RBM Inoperative (1)
(13)
N/A Once/ day (8)
JRM Upscale (1) (2) (13)
Once/3 Months Once/ day (8)
~
IRN Downsca'le (1) (?) (13)
Once/3 Months Once/ day (8)
IRN Detector not in (2)
(Once/ operating Once/ operating cycle (17)
N/A Startup Position cycle)
E IRN Inoperative (1) (2)
(13)
N/A N/A SRM Upscale (1) (2)
(13)
Once/3 Months Once/ day (8)
SRM Downscale (1) (2)
(13)
Once/3 Months Once/ day (8)
SRM Detector not in (2)
(Once/ operating Once/ operating cycle (12)
N/A Startup Position cycle)
N/A SRM Inoperative (1) (2)
(13)
N/A N/A Flow Blas Comparator (1) (15)
Once/ operating cycle (20)
N/A Flow Blas Upscale (1) (15)
Once/3 Months N/A N/A Rod Bloca Logic (16)
~
N/A j
Once/3 Months RSCS Restraint (1)
N/A Once/18 Months West Scram Discharge Once/ quarter Tank Water Level High (LS-85-45L)
N/A Once/18 Months Once/ quarter East Scram Discharge Tank Water Level flich (LS-85-45M)
I
o -
TAllLE.,4. 2. r y
illlllMUtt TEST APID CAI.IBRATIO!! FDr.OurtlCY FOR SUli180!!.f.Al3CC IllSTPUfit:ttTATION g
4
'ns'trument check 5
Instrusaent Channel Calibration l'rc<}in g-y,
..Each Shift Once/6 enonths 2
3
- 1) Reactor water Level (LI-3-58A38)
'S
'Each Shift Once/12 mouths
' *M w
- 2) meactor Pressure g;Each Shift (PI-3-74A&B)
Once/6 smonths 3,
D
- 3) Dryuell Pressure
.,# I-* Each Shi f t (PI-64-6711) anel XR-64-50 Once/6 months
_.y 48 Drywell Temperature r
- s-a (II-64-5?Att) and xx 64-50 I'r Each Shift Once/6 sar, nth s 51 Suppression Chamber Air Tesoperature (XR-64-57 )
'i N
Each Shift tA
- 8) Contr61 mod Position
- .- gg
,Each Shift (2)
S) Neutron Monitoring o.
- 10) Drywell Pressure (PS-6 4-67'S)
Once/6 nooiths
' ~
- taA
~-
28A Once/6 months
- 11) Drywell Pressure (PIS-64-SP.A) s.,'
s g
IJA 8
- once/6 eenths
- 12) Drywell Temperature (TS-64 'i?A),
ftA Once/6 anontles
- 13) Timer (IS-64-67A)
One*/ day Once/6 months
- 14) CAD Tank Level
.i.
Onec/ day.
- 15) Contalument Atmosphere !!asalton o Once/6 we.ths E*Ch SI*IIE Once /6 nosalie If.) Dryvell to suppreesten Chanter,
Diffeuntiti Freesiere
TABLE 4.2.F MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION l[
Instrument Channel Calibration Frequency Instrument check a
IF Relief valve Tallpipe NA Once/ month (24) ff Thermocouple Temperature s
!i 18 Acoustic Monitor on Once/ cycle (25)
Once/ month (26)
Relief Valve Tailpipe em 19 High-Reage Primary Containment Once/18 months (30)
Once/ month C'
Radiation Monitors (RR-90-272CD) (RR-90-273CD) 20 Suppression Chamber Water Once/IS months Once/ month-I.evel-Wide Range (LI-64-159A) (XR-64-159) 2.
21 Drywell Pressure-Wide Range Once/18 months Once/ shift (PI-64-160A) (IR-64-159)
?? Suppression Pool Bulk Temperature Once/18 months Once/ shift (TI-64-161) (TR-64-161)
(TI-64-162) (TR-64-162) 23 Nigh Range Caseous Effluent Once/18 mcaths Once/ shift Radiation Monitor (RR-90-322A) l
notts rea tettts e.t.e tunoven 4.f w (Ceastaveel v,te.te iri,ie twasoeneur teeied denni fwasuenei isei u ee.
1..
te gw tt ed by e e s t ien 4.7.3.1.e an d 6. F.C.1.s.
- 13. The tiew" bles ciesstater wt11 )e te sted.be pettant one flew weit in,
- Te s t** (prodwa lag 1/ 3 es tem) and adjweting the test taput:te etteln ee perater red ntosh. The flew tiee wresale with he verified ty eteerving a teset wresels tttp 11the dwetng operetten and wettfled that it w!!! prodwee e red blesh dwetag"the operettag eyele.
13.
Perforw d during operating cycle. Partiene of the letts la checked pote f eetwentir dertal fwnsttes.e1 toets of the twestiene that p' reduce e red blesh.
!!. *This eattbratian eenetste of ve'meetas the tweetten free servise and perforNr. en elastreatt es111tatten of the ebennet.
13.
Fenestenal test la S tatted to the condttten where eetendary tentainment integrity is not regwnted as spectited in settlene 3.7.C.! sed 3.7.C.3.
19.
Fenetteast test le llatted to the t tee where the 5071 to totwited to meet the reiwiremente of eestten t. F. C.1. a.
23.
Ca11)ratten of the se=perator retuttee tre ingwts f ree both retittulation leeps to be laterrwated. thereby remwing the llev btas signal to the A P1M a nd Art a id s e s a min g t hI r e a s t e r. This cattbratten can esty he 1
pe r f ormed dwe t.it an est e g e.
3 a
s t.
- 21. Legis test to 11stted to the Stee where aetwal operetten of the egwipeent y le perwtee ttle.
!!. cme ebennel of e tther thv feette*.ene we fifwettat sene teatter lettdtat Tentitseten A aitan ten maa t t ernal ly s t es aar D e s aun t e t t et tvely byte se ed for a vieled est to esee J !a ha,re f or f.ne t tene t t e et tne and s e11t r e tten.
23.
(Deleted) 26 This instrument check consists of carpering the therucowsle tesdiets for all valves for tenstatsoce and for nonloal tusetted values (not required dwrlag refweling outages).
- 25. tNring each refueling outate oli acoustic uMtsetna thannels shall s
be saltbrated. This calibratten includes verif tsatten of attelefepeter reopense due to neshanical entitation in the victattv el the senser.
2 fe. This instrwnent check consists of co-paring tMe katettownd tileal levels far all wolves for tenststency and for r.or.&nst espected valves (not reivtred d ttat refwettna.utanee).
tio Amendwnt Nos. fl,H eJH eIn e125
1:07ES TCR TA3tES 4.2. A TEE 0t'CM 4.2.R (continued) 21 Tunctional test consises of the injection of a si=ulated signal into i
the electronic trip circuitry in place of the sensor signal to verify l
operability of the trip and alarm functions.
)
Calibration consists of the adjusc=ent of the pri=ary senso and 28 associated ce=Ponents to that they correspond within accept ble ' range and accuracy to known values of the para =eter which the cha nel monitors, includies adjust =ent of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the a=alog equivalent of tha trip level setting.
- 9.
The' funectonal tes t frequency decreased en ecce/3 meness en recure challenxes to relie f valves per IT'*RI.9-073 7. :cem !!.K.J.16.
- 30. Calibration shall consist of an electronic calibration of the cast.nel, not in:1uding the detecter, ter range decades ateve 10 R/br and a one-point source che:k of the detector belev 10 R/hr with an instalici er portable Es==s source.
4 110a A end ent No. 125
- l t;w; s0 C0Ct7:0NS PCR OPteAT!0'J St'RVI!1.IE0t Sto"!ptvts:T5
~~
1.$.H Maintenance of Tilled Discharte Pipe 4.5.M Maintensmee af 7111ed 31sebaree it-e
~
The swetten of the ACIC and HPCI purps g
ena11 he aligned to the cundensate 1.
Every month prior to the testinc cf etwr ge tank. and the pressure suppree=
the AHn5 (LFC1 and Containeent Spray) anon (b.tober heaJ tank shall normally and core sprav s> 6 tem. the discharte be J11sneJ to serve the discharge piping pipinr. of these rstems shall be a
at the KHM.ind C $ pumps. The condensate vented frna the F (gh point and setcr head tank may be used to serve the RXR flow deterstned.
and C5 Jischarte piptnt if the PSC head tana is unavailable. The pressure 2.
Following any period where the LP:t angicat or s on the discharRe cf the RXR or core spray systems have not been and C5 purse shall indicate not less required to be operable. the dis.
t han list ed below.
charge pipLnt of the inoperable sys-Pl 73-70
-8 pois tem small be vented from the htt5 T 1 - 8 48 pelg point prior to the 7eturn of the Pl=74 51 48 pois eystem to service.
71-74 65 48 pets 3.
Whenever the K7CI or RCIC syster ts A v e r a g e_fi sm a r L in e e r We s t Cemeratie" lined up to take auction from the
)f *f condensate storate tar.k. the dto-Po r t e.-
steadv st Ate power operation. the charee pipint cf the K7CI JeJ R;iC
".nte e Av. race Planet LLnent He t Cen.
thall be vented from tee hten ta'et g r a t a.*n R.i t e O'.ATLHC R ) for caen t ype of of the systrs and water flow observeJ t u e. a s.: f unct ion of aversee planar on'.i eonthly beste, careswre anall not exceed the 1Latting silue,hewu in Tables 3.5.!-1.
-2.
4 When the pyRS and the CSS are re.
1 Jt any (Lae Juring operation 1:
quired to be operatic, tne pressure i s d et e r,tned bv normal surveillance that indicators which sonttor the dis-t he I Lr.it tne value f or APLil0R is be Log charas lines shall be ner.itored eaceeded actLon eNA11 he inittsted daily and the pressure recorded.
vttbin 1) einutes to restore operacten to vithin the prescrihad limite. If t he APLHCR le not returned to withis 1.
Ma n t um av e r a t e Pl a n t e L L e s t heat tne PrescribeJ 1Laite within two (2)
Cemerstten An t e (MATLw0t6 heurs. t he f eatter shall be braucht to The MA?LHOR f er eacn type of f wel as a t he CalJ $hutdown condition eithin function of averste planer esposure 36 baurs. Surveillance and corresponding shall be determLned daily during Jetten small continue until reactor reactor operation at221 rated operation is vtthLn the prescribed thermal povar.
Itstts.
J.
LLnear West Cemerattem e te (L"Oli s
J.
Ltwen Heat demeretten e te (1HCR)
The LH;R s
Esrte.g stcacy state power operati:n. the shall be checked daily durir.g linear heat geoerstion rate (LETR) of reactor fuel operati:n atA23: rated a sv r od Ln any f uel assembly at acy ther:41 power.
asial location shall not exceed 13.4 kV/f t.
If at anv t Lae during operation it is determined by normal surveillante that the limitina value for LHOR is be Lng esecedeJ. ection shall be LAttiated withLa
- 1) ennutes to restore operation to within the prencribed limits.
if the L11~1 to not returneJ t o wi t h LA tne tr e scribec 1Latt e withta two (2) hours the reactor sNall be brousSt to the Cold $hutdown condition vttnin 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> s.
Surveillance and corresponding actics sha;l centinue until rsJctor operatico is vtthin the prescribed 1Latts.
!!9 i
Mercent Nos. U.H M 6
.n -
e
' SU?'!!ILt. A!!CT '?OU:'T"!?!TS L:";7:f!i CC':D:7:0!:*, Fn' CPIP A7:t'!
4.5 00'E At:0 C0"T A!':"E' 7 35 CORF AC CC'
- A!'N!?:-
'CnCLInn SYSTE"S COOL!'!G SY57E"5 L. 5. Y.
Hint-u-C-itical Powe 3 5.r Minir.um C-itiest Pewa-
.; Patte Ut FE t ptt:- 'f*Cr u z MOFR sh.111 to deter-ined daily Tne einimum eettical-power ratto....~1 i
during resctor powe-eperstica (10PR) 4.s a fun: tion of stran tine a91 cera flows shall be equal
,1s't 2 295 Psted thermal power and to er reenter than shown in following any change in powea level e-distribution that figure 3,5.x-1 nultiplied by the would cause ege-atien with a ng shoun in figure 3 5.2, where:
licitina, control red patteen e*
T: O caTave TP, whicheve-is as des'erined in the tases for
% A 23 gaester Sne:ification S.?.
I A30.90 see (Spe:ification 3 3 0.1 2.
The M*F?[ [151t shall be deter-stran tine limit to 20s
. f.ined for each fuel type 8X!,
insertien fro-full withdaarn)
-6XSR. PSX!R, fren Tigure 3 5.K-1 respectively using:
q
,.l',(0.053) her, d)
'd ?s3.71C.1.65
'd
- a. T 0.0 prior to initial n
seaan time measurenents for
~ '
{ t.
the cycle performed in T ve 6
S s.
accordance with m
Specification 4 3.C.i.
n a nunter of surveillan:e rod tests
- h. ifas defined in Specification perforned to date in cycle (in-ciudinr. 200 test).
3 5.R rollowint the conclusion of ea:n stran
- is s:ran ti-e to 200 inse.-tion t-:-
tire surve'.11ance test no fully witnd.-awn of the ith re retui-ad ey Specification 4.3.C.1 and 4.3 0.2.
N totsi nunter of active rods The determinatics of the ceasured in Specificatica 4 3.C.1 limit nust be cor.plete!
at 300 with 72 h urs of each scran tine surveillance requires If at any time during steady state operation it is determined by nornal by Spe:ifiestien 4 3 0.
surveilla-:e that the limitine, value re-M0PR is being exceeded, a: tion shg11 he initiated within 1$ nir,utes to restore opeastion to within the pres:=ibei limits. If the steady state MCPP. is not retuaned to within the prest ited linits within two (2) hours, the reactor shall be nrour,ht to the Cold Snutdown condition w1 thin 3/> % curs. su-veillance and corresponiine, 4: tic 9 sh11) continue until reacto-ope-atica. is within the prescribed 11 nits.
it*
A end ent Nos, f 7.25.125 O
l
\\
l f
)
\\
The posk e l'a d d i n g temperature following a pontuinted
,oss-of-c ool s e t eocident is p r im a r ily a fgattton of the avera is heat seseration reto of all the rods of a fset assembly at.e ny, enssi losetien and la caly dependent secondarily on tj ie red to rod pooer di st ribe t t oa eithis as aasembly.
Stace espected local vartstione la power distriksiton withis e inel eesembly aflect the calcelated peak stad temperetste by lese thaa
- 20*F retettve to the peak temperature for a typleal feel desago. the limit on the everage linear heat generstlos rate is sufficient to assure that sticolated temperaturea ar6 eithin the 10 CFR 50 Appeadts E limit.
The limit t as v al ue for hAPLWGR is shown in Tab'les 3. 5.1 - 1.
-2.
The smaly se a sapporting these listilag wantee is presented la Reference t.
w 163e Amend ent No. 25.125
s.
3.f..*. L1 eaa West Ce** stten Wate itF081 Ints specstteataca assares tna; tr.e 11 rear meat remeratics rate in ar.y r:c is less than tne Cetign linear meat generati n af fuel pellet censificati:n is portJiated.
f,,
\\}
- hel Lh03-sna:1 he enecket Cairy during reec:ce caerat!'n at g 215
- cser to deter ine if fael turnus, er cente
- 1 red sove ent has cause changes in ;&wer distra:utica. Fce '.MOR to to a 11:1 ting va4ue teacw 2.'5 rated tror-a1 Ocver, tr. A fatter veuad P. ave to te less than 0.241 vnitt is pre: Auded ty a car.siderr,tle sargin VMen f..'.
Asying any per= ssitie :...*rol red pattern.
3.f.X. wi- -t vn te*.t ::: 7:ve- *at t e f y:1 At core ther:1; ::we 4eveas aess snan er etuna 1: 25% tne reactor vi;1 te c;trating at 11NL:wn recir:aastics pur.; steet an$ the a Cerat:r void centent will be very :s;i. Ter a:4 desigrated cents 1 r:d patterns vnien -ay be entleyte at inis scint c:erating ;4 ant ex;erience anc tner a hytraulic I
acaaysis ir.Gittted inat the resuAting M703 vaage is in excess of require e.ta by a constderatas narg.n. Witn this &cv void content, any inadvertent tire flew tecrease. was enay ;te:e c erata:n in a mere conserative nede reJative to M0?P.
Tre :a;47 re at-e.ert fer caaca ating M0?R a:cve 215 rated tner-a
- cwer is suffic er.t since ;
- ver distratutten sn f ts are very sacw wnen Mere have not teen s;gn f t: ant ; ser cc centrea red enanges. The reiwirt ent f P cal J14 ting M0?! wnen a aimating ::nts:4 rei ;attern is a ;reatnet ensures trat M0?R vta; te kn:wn relacwing a cr.ange in ;;wer er : wer sr.a:e.
(regar:Assa cf :agnatate) t r.a t c va pia:e c;erati:n at a tner 4; 11:it.
3. !. '. a? '4 Set ::P. t s C;eration is ::nstrained to a =ax;:u: LMOA of 13 * = (J/ f t.
Inis limit is rea:ned unen c:re :axt: : fracti:n of li:: ting power density (OMILPO) equa;s 10.
Ter tat case vnere CMTLPD excee:s tne fract1Jn of ratet inerlai ; ver, c;eration is ;ertitted Only at less than 100.pertent rated ;cwer ang enly v;ta AFA9 s:rL: setting: as re:u;-ed by s:stif.:ati:n 3 3.L.1
- ne 3 rga g,;; 3,;;;ng gng 7;; ;;; a ;f;p se tant are a
- ;;ste: :: ensu. e tat t no cc::ir.at;:n of OMI' ?* ant ???
e v;;; in:rtase tne LHOR transLint sea < te*/:n: that a;1:ve ty tr.e 1-;e r:e r.: 'astic strais. l'=;t.
A 6.htur time perice to act.Leve in;s ::ngit;:n *3 ustifie s.n=e tne attiti:na; =argir. gaine: ty a
tne set::.r. :;;st:ent is a::ve an: bey:nc nat ensure: :y ine safety analys;s.
I i
O 15:
Amend ent Nos. A6.53.f 7.27.12 A.125
..-+
te i
F
}
1& ALE 3.3.1-1 p.et'.t:t vtP.st': /YtLACE N'.4AA f.tt'es' tt Fuel 'twees: 78D B 28!.t.,
QL' lD+
and SDR3284L oerste r:ues Iesee.ee met.w:n
- ** 18 / t '
(wit )
j 2N 11.t l
8.0 10 11.1 3.000 11.s I
1 10.1:0
- 12. it i
l 15.003 12.0 4
- 10. ::
11.8
]
!).003 11.2 I
10.1:1 10.A k
al.:::
10.0 i
i no.c:a 9,g t
l Table 3.$.!- 2 KLFL1lCR VI.RSU$ AVERAOC FLANAA EXFOSUM l
i j
IWSI T)Tt'8s t IIDMR d h$18 I
Ave r.w e l't.w.it i
tr;o wre PAFLM:t (wwJ/t)
(W/f t)
(
4 i
2C0 11.5
(
i 1,0C0 11.6 l
$.000 11.9 s
i 10.000 12.1 l
2 l$.cs0 12.1 261. C 33 12.0 L
23.000 II.h i
i 30,000 11.2 l
ss.co:
10.5 l
i 6c.:ce 10.5 r
63.Cee 10.0 L
)
i 171 Aren M nt No.,85 125
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.....$...........4.........f........ !
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MCPR :.i..1:s Dr P3 x BR/S X SR/ Q'T)+
i Arendment No. 28 125
'" 1 9 i **
de 6
' =
1 3.6/4.6 54tr9:
Empettence in teltef valve operetton shows that a testing of
): percent of the valves per year is adequate to detect failures or delsttorJttons. The rettet valves are benchtested eve ry acceed operatin; cycle to ensure that their set points are with6n ttie
- I percent tolerance. The rettet valves are tested in place once per operattet cycle to establish that they vsll open and pass steam.
Th e cegut/#*ents estseltshed above apply when the nuclear system can be preamwrised above amenent condittene. The se s equi rerents are applica ble at nwslear avsten pressures below normal operatteg pressures because etnormal eperattenal tranatents could possibly start at these concittens owth that eventi.nl everpressure rettef would be needed. However, these transtents are much less severe, in terwa of pressure, than those startang at rated senettiens. The valves need not be functional when the vessel head is rtreved, since the nuclear syster cannot be pressurtsed, efr TPN!rs 1.
?.wclear 5) sten Fressure Rettel System (BrNP TS AA subsection 6.4) 2.
Arend-ent *: to response t o AIO Qu e s t io n 4. 2 o f ",e t teb e r 6, 19 71.
3.
'Treteetten Asannst CNe rpres sure" (ASKI Bet te r and Pressure Ve ssel Cede. Sectnen 111. Article 9) 4 3rcsms Terry Nwelcar Flant Design Defittency Report.. Target Rock Safety Rettef Valves, trans-Atted by J. E. C111 eland to F. E. Kruest, AJgust 29, 1973.
S.
Ceeetic telcas.*sel A plication. Licenstet Tcpical L
Report, N! !-26Ull-F-A, and Acdenda.
I 3.6.t!&,0.t
.' e t * -? s Fal wre of a jet pump natale esserbly ho!!down reementsm. nettle asse-ily e n s e.$ r rtser, vowld intresse the crosseterttenal !!cv area for 41cws:en fiellownnr t he de sign taats double
- ended line break, also, fatlure of the
.I t i t s tre w..< e ! J e l i ci..a t t the einability to refleed the core to t.t*t' rts
- het, e trwel tollvvi..g a rectrew1Jtton line brean.
Therefore. !! a taa4.re ertw: red, repatre ews be Pede.
T*** Jeteetton tecsnteve to as follevs. k's th the two re s t raula tice p.*r s balar.ces in spee: to vnt5tn * $ percent, the flev rates tr. both rettttwla*
tien lects utl! te vertised Iy control room aonatortnq instrarents. !! t.t e I
tse !!rv rate valaes da not dif f er by tore than 10 percent, raser anc scatle asse**17 integrity has bete vert!1ed.
Amend ent 53,125
a b
~
Y J
<t>
Thats 3.7.A s
PptMApt OD9erAlueeDef 150LATaost VALvts Q
o 3
sensoer of Power Maaleem Acelan se opesated valves ogesating stormal Init s at ing 2
)
valve Idest.A f tcat ion Isheasd cettmerd 76ee 8sec.3 Poettien signal o
creep
~
u 1
senta steem11me testat sen walwee e
e 3(T
- ?
714 Autitary 3 otter to P !'
12.-L1 Auztitary letter to R:20
- .3 2?A 513 7Jppreeston Cnseer. cole tir.es l.1.? T.
R:3 Cuppresaton Chs. ster 3r ple,tne L ) 27/.
R:3 06spreeston CneWer Osmote Lines
!.3 00 R:3 Suppression Ch.aber Oswale Lines
-11?
Nll:C Turbtac Is.houst t; ?"
7.*:C Vacuu:. Psp tischcrge
'l l * ' s 3C:0 Turbine Enhewet 71 N RO C Vacuwe Pump Discherne 73 73
- DC
- Tur'atne E.shaust 13.*L RMI Turbine I.ahous'. Drain 73.N3 HPCI TurbLt.e Eshoust 73 609 HMI E.shaust Orsin 76 722 E.9
%.5?
Core Strey to Aut11er/ Dotter U.Pl Core Spre/ to Austitary 3eller Core Sprey to Aut11ery Botier
- 62 is end ent No. !!,125
/
\\
- AU 3.*.T
"?.OWIT 0 0*:7AOZ?C 130*.A7*0!; V/1*.*EO LT!. TID ".!
ilAC. LEAL *.L SEIS;* C CIAO: 1 LI:03 t%e
, Velve Identifteettes
-k.$3 N3 LM: Diseharge
-L.!L N3
-L.5-FJC Suppressisa Chsaber Sort't
-k.V N2 Sappression Chester Opray
-L.C4 FJ7 Drywell Oprey
-1. 41 NG Drywell Spray
- k.i -
N3 LM: Olscharge
- 1../.S Mm LIC Etsenerse
-:.. t M C Suppression case er S;ra/
4.";
Nm Suppreselon Chatter Oprar
-'..'t.
N3 Orywell sprey "L."
M G Orywell Spray l
l 5.i}
Core Sprav Discher,te Core Spray Disc.orl:e l
!.!)
Core Spray Dtsenerse 5.F Core Spray Dis:aarce l
l l
l l
l 263 Amendrent No. !!.125 l
l l
a.
Lw
- 6 c Jee nt Q rea*vett 3.I 18 t' tM"M BrcMoo Terry wait 2 le loested at Browne Ferry hw: lear Plant else em pressety emes by the United States and la evetedt of the Tva. The ette shall teaense of apprestaatel, ILO setet en the meeth ehete of Wheeler Late at feaaessee alver **lle 296 la Lleestone Cowatt. Alabama. The staleve dietames free the ownslet of the setendary esatalament bwlldtag to the bewade t y of t he estivs tem astes se definee in 10 CTA 100.3 e411 he 4,000 feet.
3.2 RCACCR l
A.
The reactor core may contain 764 fuel assemblics consatting of 4 QUAD + denenstratier, assemblies.
8ms asseenlies having 63 (wel rods each, and 8a8R and P8x8R assemblies havins 62 fuel roda each.
3.
The reactor core shall contain 185 crucifore-shaped coctrol rods. The control material shall be boron carbade poveer (3 C) compacted to approntaately 70 percent of 'theoretzt.41 4
density.
(
),)
ALACTSe wilitt.
Tise tiastet vessel shall be as deetttbed in fatte 4.7 2 el the PSAR. The apelltable destan tedes shall be es described in hble 4.2 1 of the FI AA.
e 3.4
,C0wTat et47 0
A.
The pelattpal destaa parseeters for the primary seatelneemt shall be se given in table S.).1 of the flat. The ePollsatte destta sedse shall be se desettbed la Seattom 1.2 el the TSO.
8.
The seseada ry esat aineemt shall be as de.t*1 bed in last&ea S.) of the flat.
C.
Penetrattene te the primety contalament sad piplme passlag j
theewsh outh penettettens shall be deettned in etterdente with the standarde set forth in leettom S.2.3.4 el the Tlat.
3.5 FUtt SE>g A.
The streasement of fuel in the me.*twel steesse fattitty shall be owem thatle!!!.feeety a
esast eleas to les s thea o.9o.ae fie.ee.
. sh.a o.n cseette. Ic.:.i r$ati.
+
33:
Amendment Nos. 35 s4,58st!,125
6.0 Ao"!N!STRAT3t CONTPCLS s.
source tests Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.
C.
Seacial Re eo rt s (in writing to the Director of Regaonal CZfice of Inspection and Enforcement).'
1.
Reports on the f ollowing areas shall be submitted as noted:
a.
Secondary containment 4.7.C Within 90 Leak Rate Testing (5) days of compiegion of each test.
b.
Fatigue Usage 6.6 Annual Evaluation operating Report C.
. Relief Vaive Tailpipe 3.2.T Vithin 30 days las t rumen t a t ion af ter inoperabiitty I
of therr.or.oup;o and acoustic monitor on one valve.
d.
Seismic Instrummtation
- 3. 2.J. 3 Within 10 Jeys Inope rabili ty af ter 30 df s of incperabiltt.*
e.
Meteorological Monitorins 3.2.I.2 Vithin 10 days Imatrumentaticu af ter 7 days of Isoperability inoperability f.
Primary Containment 4.7.A.2 vignia'90 e37, Intestated Laak llate of co27,, tion et Testing each test.
High-Range Primary Containment 3.2.T Within 7 days Radiation Monitors after 7 days of.
inoperability Hign-Range Gaseous Ef fluent 3 2.F Witnin 7 days Radiation Monitor after 7 days of inoperability D.
Special Report (in writing to the Director of Regional Office of Inspection and Enforcement)
Data shall be retrieved rec all seis=ic instru=ents actuated during a seis=ie event and analyzed to deter =ine the magnitude of the vibratory ground motion.
A Special Report shall be submitted within 10 days after the event describing the =agnitude, frequency spectru=, and resultant effect upon plant features 1:Portant to safety.
356 Ameneent N0s. y,y.125
1
/
%o, UNITED STATES NUCLE AR REGULATORY COMMISSION
~'
w ASHING TON, D. C. 20$55 SAFETY EVA1.UATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.125 TO FACILITY OPERATING LICENSE N(. OPR-52 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-260
1.0 INTRODUCTION
By letter dated August 23, 1984 (TVA BFNP TS-199), as supplemented September 4 and November 13,1984. April 3. May 8. June 27. November 20 and December 30, 1985, and April 29, 1986, the Tennessee Valley Authority (the licensee or TVA) requested an amendment to Facility Operating License No. OPR-52 for the Browns Ferry Nuclear Plant. Unit 2.
The proposed amendment would change the Technical Specifications (TS) of the operating license to:
(1) modify the core physics, thermal and hydraulic limits to be consistent with the reanalyses associated with replacing about one-third of the core during the Cycle 6 core reload outage and (2) reflect changes in various specifications as a result of plar.t modifications performed during the outage.
In addition. TVA has updated the TS pages involved and made administrative corrections.
The areas involved in the amendment are as follows:
A.
Core related changes B.
Changes related to torus modifications I
C.
Miscellaneous plant modifications 1.
Reactor protection system (RPS) modification 2.
Scram discharge instrument volume 3.
Analog trip system 2
4 Scram permission pressure switches
[
5.
Drywell temperature and pressure 6.
TMI Action plan items (NUREG-0737)
}
7.
Testable penetrations 8.
Redundant air supply to the drywell 9.
Demineralizec' water isolation valve
- 10. Residual heat removal (RHR) head spray D.
Administrative changes j
I 1
2 2.0 EVALUATION _
A.
Core related changes TVA made application to amend the Technical Specifications of Browns Ferry Nuclear Plant Unit 2.
The changes were required, in part, in order to permit the reloading and operation of Unit 2 for Cycle 6.
In support of the application TVA submitted a Reload Licensing Report (Reference 1).
The staff has reviewed this document and prepared the following evaluation of those aspects of thq application pertaining to the reload.
Reload Description For Cycle 6, 300 irradiated fuel assemblies will be removed from the core and replaced by 296 General Electric P8X8R assemblies and 4 Westinghouse designed QUA0 + demonstration assemblies.
In addition, the reload analysis has been performed by TVA, with the exception of the LOCA analysis which has been done by General Electric. The demonstration program has been described and analyses performed on the effect of the QUA0 + assemblies on the core parameters by Westinghouse Nuclear Energy System, the manufacturer of the assemblies.
TVA has submitted a report.
WCAp.10507, "QUAD + Demonstration Assembly Report" (Reference 2) for the description of the program and its effects. The use of increased core flow is planned for Cycle 6.
Analyses were performed for both 100 percent and 105 percent of rated flow and the most conservative results were used in determining the operating limits.
Fuel Mechanical Design The P8X8R assemblies to be loaded into the core are identical to those inserted in Cycle 5.
They are standard General Electric BWR fuel assemblies which are described in the GESTAR document (Reference 3) and we conclude that no further review of these assemblies is required.
The mechanical That design of the four QUAD + assemblies is described in Reference 2.
document also describes the fuel rod design analysis.
The acceptability of these analyses for Lead Test Assemblies is the subject of a separate evaluation (Attached). That evaluation concludes that the QUAD +
assemblies may use the various fuel rod design criteria of the P8X8R fuel on an interim basis for the Lead Test Assemblies.
Nuclear Design This reload is the first one performed for Unit 2 by the licensee.
The analysis methods used by TVA are described in References 4, 5 and 6.
These reports have been reviewed and approved by the staff for use in such The results of the analyses are reported in Reference 1.
The analyses.
shutdown margin is calculated to be 1.0 percent reactivity change at the point in the cycle at which it is a minimum. This value exceeds the Technical Specification requirement of 0.38 percent snd is acceptable.
The standby Liquid Control System provides a shutdonn margin of 1.8 percent reactivity change with a boren concentration of 600 ppm boren.
This is an acceptable value.
Reactivity coefficients are not ised in the performance of transients by TVA. However, a void coefficient is obtained in the process of collapsing from 3-0 to 1-0 cross-sections.
his value is in the range of those customarily obtained for BWR reload cores and is acceptable.
The effect of the presence of the four Quad + assemblies on the neutronic behavior of the core is discussed in Reference 2, which is the subject of a separate evaluation (Attached). That evaluation concludes that the presence of tha four QUAD + assemblies has a negligible effect on core neutronics.
TVA has performed cycle specific analyses and concurs with the conclusions of the Westirighouse report. We conclude that the nuclear design and analysis of the Cycle 6 core are acceptable.
Thermal-Hydraulic Design The thermal-hydraulic analysis of the Browns Ferry Unit 2 Cycle 6 reload has been reviewed to determine whether acceptable themal-hydraulic limits have been met, whether acceptable analytical methods were used and whether the core exnibits thermal-hydraulic stability.
Safety Limit MCPR The GEXL Critical Heat Flux Correlation is used to obtain the value of the safety limit MCPR. This correlation has been previously used for Browns Ferry Unit 2 and continues to be acceptable. The value of 1.07 for the safety limit MCPR is generic for BWR reloads and is acceptable.
Operating Limit MCPR The procedures and techniques used to obtain the value of the operating limit MCPR are described in Reference 7 which has been reviewed and approved by the staff. The anticipated transients are analyzed to detemine that which yields the largest reduction in CPR.
Th:t value is then added to the safety limit value (1.07) to obtain the operating limit MCPR.
For the pressurization events both Option A and Option B limits are obtained. The e
results were calculated for the PSXSR fuel. The QUAD + fuel will be leaded into non-limiting core locations and monitored to the same operating MCPR limits.
Operation at 105 Percent of Rated Flow The licensee proposes to operate at core f1)w rates up to 105 percent of rated flow for Cycle 6.
Such operation has been approved for Cycle 5 in Browns Ferry Unit 2 and it continues to be acceptable for Cycle 6.
Analysis of Cycle 6 operation has taken into account such operation.
4 Core Thermal-Hydraulic Stability TVA uses a computerized model for analysis of boiling water r ' actor (BWR) stability for Cycle 6 of Browns Ferry Unit 2 The analysis e del. is based on the LAPUR computer code and is applicable to both core and ' channel hydrodynamic stability.
It is the same model which was used for the analysis of the previously approved Browns Ferry Unit 3 Cycle 6 reload.
The model proposed by TVA has been under review by the staff. The safety evaluation of this model has not yet been issued but the review has progressed sufficiently for the staff to approve the TVA analysis of Cycle 6 of Browns Ferry Unit 2 for the following reasons.
1.
The only significant change in fuel loading between Cycle 6 of Browns Ferry Unit 2 and the previously approved and currently operating Cycle 5 of Unit 2, is the addition of the four QUAD + demonstration assemblies.
The stability characteristics of these assemblies were reviewed separately (see next section) and found acceptable.
2.
The decay ratio as calculated by the TVA model for Cycle 6 of Browns Ferry Unit 2 is.71. Which is lower than the calculated decay ratio (.73) of the previously approved Cycle 6 of Browns Ferry Unit 3.
3.
The TVA model does a good job in predicting the results of the Peach Bottom Thermal-Hydraulic Stability Tests.
Presence of QUAD + Assemblies The thermal-hydraulic performance of the QUAD + assembites is discussed in Reference 2.
The evaluation of that reference (Attached) concludes that use of QUAD + bundles as demonstration assemblies is acceptable provided that the guidelines of Section 4.1 of Reference 2 are followed and that a cycle specific analysis shows at least a margin of 20 percent in power a
between the QUAD + assembly and the lead assembly at full power and flow conditions.
TVA has confirmed that the guidelines were followed and performed analyses to show that a 27 percent power margin exists for Cycle 6.
The staff asked Westinghouse to show that the stability characteristics of the QUAD + assemblies are acceptable for inclusion in the Browns Ferry Unit 3 Cycle 6 core.
The results of Westinghouse's analytical evaluation which qualifies the QUAD + stability margin is presented in Reference 2.
The focus of this evaluation is on individual channel stability since the small number of QUAD + demonstration assemblies in the core will not have any significant impact on the core average parameters and hence not affect overall core stability.
The Westinghcruse analysis show the QUAD +
assemblies to have an additional margin of 0.15 in decay ratio when compared to the P8X8R fuel already in the core.
The Westinghouse evaluation used parametric analyses based on published data to quantify the relative stability margin of the QUAD + deconstration assembly compared to the P8XSR fuel and did not perform detailed stability calculations for the QUAD + assembly itself.
____ _ __ The staff revieked the analysis performed by Westinghouse in Reference 2 and has found it to be a reasonable method for approximacing tP e stability margin for the QUAD + assembly. While the staff finds that sut h an approach is acceptable for the limited number (4) of QUAD + assemblies in the core it is very approximate and considerably more detailed calculations would be required to justify a full reload of QUAD + assemblies. We conclude that the thermal-hydraulic design and analysis for Browns Ferry Unit 2 Cycle 6 are acceptable.
Transient and Accident Analysis Core-widepressurizationtransientswereanalyzedwiththeTVA-REiRAN (Reference 7) code which has been reviewed and approved by the staff.
The two conditions cited in the review use of the COMETHE-I!! J code and approval of the parent RETRAN code, has been satisfied.
Use of TVA-RETRAN is therefore acceptable.
The nonpressurization events were analyzed with the three dimensional core simulator code (Reference 5) since these are either steady state events or very slow transients.
The limiting pressurization transient is the Lead Rejection Without Bypass and the limiting nonpressurization events are the Loss of Feedwater heater and Mislocated Bundle Error. Since the replace-ment fuel is identical to some of the fuel already present in the core, reanalysis of the LOCA event was not required.
Reference 2 presents analyses to show that the W.PLHGR limits for the P80RB284L assemblies can be conservatively applied to the QUA0 + assemblies.
The rod drop accident analysis was performed with the methodology described in Reference 8.
This methodology was approved for use in the Cycle 6 reload analysis for Browns Ferry Unit 3 and is acceptable for Unit 2.
The result of the analysis for Cycle 6 of Browns Ferry Unit 2 is 152 calories per gram peak fuel enthalpy.
This value meets our acceptance criterion of 280 calories per gram for this event and is acceptable.
Technical Specification Changes Scram Permissive Pressurt Switches at 1055 PSIG Current Technical Specifications require the main steam line isolation valve closure and the turbine condenser low vacuum scram functions to be operable in the refuel, startup/ standby, and run modes.
However, these trips are bypassed in the refuel and startup/ standby modes unless the reactor pressure is greater than 1055 psig.
Since the core is orotected by a high pressure trip at 1055 psig in all modes the two scram functions serve no useful purpose in the refuel and startup/ hot standby modes.
TVA proposes to delete the requirement for operability of the scram functions in those modes and to remove the bypass function. As a result of our review of this area of operation, we agree that these scram requirements acco plish no useful purpose in these modes.
We conclude that the proposed Technical Specification change is acceptable.
MCPR-MAPLHGR Specifications The operating limit MCPR as a function of average scram time. T
,has been altered to account for the Cycle 6 reload.
The proposed curve i,Fi 3.5.K-1) is consistent with the value given in the reload report (gure Reference
- 1) and is acceptable.
The MAPLHGR tables have been revised by deleting those for fuel types no longer present in the core and consolidating the data into two tables.
3.5.!-1 and 3.5.I-2.
No changes have been made in the MAPLHGR values.
The values for the P80RB284L type are to be used for the QUA0 + fuel.
Such use is justified in Reference 2 for demonstration assemblies and is acceptable.
Reference in Bases At various locations, the Technical Specification Bases have been revised to reflect the fact that the safety analyses were performed by TVA. These revisions are acceptable.
Based on the review described above, we conclude that Browns Ferry Unit 2 may be loaded and operated for Cycle 6.
This includes the presence of four QUAD +
bundles as lead test assemblies.
This conclusion is based on the following:
1.
The safety analyses have been performed by previously approved methods and procedures, except for those directly relating to the dwonstration assemblies.
2.
The use of the demonstration assemblies has been approved (see Attached evaluation) subject to certain conditions.
These conditions have been met for Browns Ferry 2 Cycle 6.
3.
The Cycle 6 core meets all the staff's acceptance criteria.
B.
Changes Related to Torus Hodifications One of the changes to the TS is to revise the tables that list the surveillance instrumentation associated with the suppression pool bulk temperature. This modification provides an improved torus temperature monitoring system which consists of 16 sensors. This will provide a more accurate indication of the torus water bulk temperature as required by NUREG-0661 and will replace the suppression chamber water temperature instruments presently listed in the TS.
This change has been previously approved for Unit 3 by Amendment ho. 78 dated August 27. 1984 The change to the TS are necessary follow up actions essential to the implementation of this improvement.
The changes to the TS place operability and calibration requirements on the new temperature monitoring system.
Since these are new instruments, the surveillance requirements are not presently in the TS.
-7 We have reviewed t,his proposed change and find it consistent with NRC guidance and it is, therefore, acceptable.
C.
Miscellaneous plant modifications L
1.
Reactor Protection System (RPS) Modifications.
By letter dated August 7, IS78, the Commission advised TVA that during review of Hatch Unit 2 the staff had identified certain deficiencies in the design of the voltage regulator system of the motor generator sets which supply (g).
TVA was required to evaluate the RP$ pow (RPS).
power to the reactor protection system Pursuant to 10 CFR 50.54 er supply for Browns Ferry 1, 2 and 3 in light of the information set forth in our letter. By letter dated September 24, 1980, the staff informed TVA (and most other BWRs) that "we have determined that modifications should be performed to provide fully redundant Class IE protection at the interface of non-Class IE power supplies and RPS."
The staff also advised TVA that "we have found that the conceptual design proposed by the General Electric Company and the installed modification on Hatch are acceptable solutions to our concern." By letter dated December 4,1980. TVA committed to install the required modifications.
By letters dated October 30, 1981 and July 28, 1982 NRC sent TVA model Technical Specifications for elec,tric power monitoring of the RPS design and modifications.
By letter dated June 27, 1985, the staff approved the TVA proposed design modifications to the RPS power supply system.
During the current outage of Unit 2. the RPS is being modified to provide a fully redundant Class IE protection at the interface of the non Class IE power supplies and the RPS.
This will ensure that failure of a non Class IE reactor protection power supply will not cause adverse interaction to the Class IE reactor protection system.
The Technical Specifications are being revised similar to the model TS provided to TVA to reflect the limiting conditions for operation and surveillance requirements associated with the RPS modifications.
Page 42 is being modified to add a description of these s tions in the Bases.
lased on our Safety Evaluation dated June 27, 1985, and the TS submttted.
we find the proposed amendment acceptable.
2.
Scram discharge instrument volume The scram discharge instrument volumes (SDIVs) were modified to address Unit No. 3 in June 1980(hy the partial rod insertion everton Browns Ferry inadequacies identified 0
The modifications of interest to this Safety Evaluation involve replacing the scram discharge tank's ficat devices
- 0) Briefly, an undetected accumulation of water in the SDV reduced the available free volume for discharge of scram water which inhibited insertion of the control rods.
The level detection system utilized float type instruments and an inspection of the instruments turned up several floats that had been damaged.
It could only be concluded that the floats had been subjected to harmful hydrodynamic forces.
8-with new electronic level instruments.
These instruments will initiate a scram on high level.
Tables 4.1.A and 4.1.8 were revised to reflect changes to the rpqu' ired surveillance testing on the two electronic level switches. The acceptability of the changes to the surveillance testing will be addressed in Section C-3 of this SE.
Based on our review, we conclude that the proposed modifications to the Technical Specifications in the instrumentation and controls area are acceptable.
The basis for our det'ermination is that the modifications are consistent.with the staff guidelines as stated in the BWR Scram Discharge Safety Evaluation Report, dated December 1,1980.
In addition, these proposed modifications have been previously approved for Browns Ferry Unit 1. Amendment No. 93.
'3.
Analog trip system The analog transmitter trip system (ATTS) is a new design for portions of the system instrumentation of the Reactor Protective Syrtem (RPS) of Boiling Water Reactors.
It was developed by the General Electric Company (GEland is being supplied as original equipment in later built BWRs (e.g., BWR 6).
GE developed the ATTS to offset operating disadvantages of the digital sensor switches of the original safety system instrumentation. The principal objective of the ATTS is to improve sensor intelligence and reliability while enhancing testing procedures.
The design was adapted to Browns Ferry Unit 2 to replace the existing mechanical switches that sense drywell and reactor pressures with analog r
loops and to modify the reacter water level indication loops to improve the reliability, accuracy and response tire of the instrumentation. Change in design basis, protective function, redundancy, trip peint, and logic would not be involved or modified as a result of the ecuiorent changes.
Basically, the licensee is proposing to replace Barton, Barksdale, Static
- Ring, and Yarway instruments with Rosemount analog pressure transmitters and RosGmcunt analog trip units.
Along with the system enhancement offered by the new electronic instrumentation, the licensee proposed to extend the maximum calibration interval to "once an operating cycle." This was based on the high reliability of the analog instrumentation systems.
The various calibration intervals (not the same as functional test intervals) being used at the plant are:
1 Once every 7 days 2
Once every 3 months 3
Once every 6 months
- 4) Once every 18 months
- 5) Once each refueling outage The channel calibration once per operating cycle is less conservative than the present requirement for calibrations of some systems once every 18 months.
It has come to our attentian that the duration of an operating cycle may not be adequately defined. Mid-cycle shutdown may occur such that an operating cycle may be extended well beyond the 18-month period which has been previously considered to be the longest operating cycle. The operating cycle time is dependent on the reload fuel design, which can vary between 12 and 18 months.
The primary factor in setting the calibration intervals is the drift of the transmitters and trip units. The total loop accuracy and the total loop drift are added to obtain the trip setpoint.
In many cases, the manufacturer's specifications only provide drift values for 6 to 12 month intervals.
These drift values must now be extrapolated linearly to provide for 18 months or longer calibration intervals.
Based on the above information, we concluded that the Technical Specification changes extending the calibration frequencies to "once/ operating cycle" are acceptable if these calibration frequencies / intervals are limited to 18 months maximum.
This limitation of once/ operating cycle not to exceld 18 months for calibration intervals applies to the analog pressure transmitters and analog alarm units only and not to the mechanical pressure switches 3
and their associated alann units.
By letter dated April 29, 1986 TVA submitted sJpplement 3 to the amendment request dated August 23, 1984, which made the change from once per operating cycle to a minimum frequency of once per 18 months. Based on that supplement and our review we conclude that the proposed modifications are acceptable.
4.
Scram pemissive pressure switches This has been covered in Section A above.
5.
Drywell temperature and pressure The drywell temperature and pressure surveillance instrumentation is being upgraded this outage to provide qualified, more reliable instrumentation.
The TS. Tables 3.2 F and 4.2.F. have been revised to reflect new instrument numbers for the new upgraded drywell temperature and pressure instrumentation. The surveillance requirements recain the same. We have reviewed the proposed changes and based on our review find them acceptable.
_ 6.
TMI Action plan items (NUREG-0737)
In November 1980, the staff issued NUREG-0737, "Clarification o TM1 Action Plan Requirements," which included all TM! Action Plan items ap roved b/
the Commission for implementation at nuclear power reactors.
NUREG-0737 identifies those items for which Technical Specifications are required. A number of items which require Technical Specifications were scheduled for implementation af ter December 31, 1981.
The staff provided guidance on the scope of Technical Specifications for all of these items in Generic Letter 83-36. Generic Letter 83-3f as issued to all Boiling Water Reactor licensees on November 1,1983.
In this Generic Letter, the staff requested 1icensees to:
a.
review their fact 11ty's Technical Specifications to determine if they were consistent with the guidance provided in the Generic Letter, and b.
submit an application for a license amendment where deviations or absence of Technical Specifications were found.
By letter dated August 23, 1984, as supplemented. TVA responsed to Generic Letter 83-36 by submitting Technical Specification change request for Browns Ferry Unit 2.
This evaluation covers the following TM! Action Plan items:
Noble Gas Effluent Monitor (!!.F.1.1)
The itcensee has supplemented the existing nornal range monitors to provide noble gas monitoring in accordance with TM! Action Plan Item
! !. F.1.1.
The propostd Technical Specifications for Noble Gas Effluent Monitor are conaistent with the guidelines provided in Generic Letter 83-36.
TherefG e. we conclude that the TSs for Item
!!.F.1.1 are acceptable.
Sampling and Analysis of Plant Effluents (!!.F.1.2)
The guidante provided by Generic Letter 83-36 requested that an administrative program should be established, implemeated and maintained to ensure the capability to collect and analyze or measure representative samples of radioactive todines and particulates in plant gaseous effluents durthg and following an accident.
The licensee has proposed TSs that are included with the TSs for Surveillance Instrumentation. The proposed TSs for sampling and analysis of plant effluents meet the intent of our guidan,e.
Therefore, the proposed TSs are acceptable.
Drpell High-Range Radiation Monttor (!!.F.1.3)
The licensee has installed two drywell radiation monitors in Browns Ferry Unit 2 that are consistent with the guidance of 1MI Action Plan Item
!!.F.1.3.
Generic Letter 83-36 provided guidance for limiting conditions for operation and surveillance requirements for these monitors.
The Itcensee proposed TSs that are consistent with the guidance provided in Generic Letter 83-36.
Therefore, we conclude that the proposed TSs for Item !!.F.1.3 are acceptable.
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Orw ell Pressure Monitor (!!.F.1.4)
Browns Ferry Unit 2 has been provided with two wide range char 1els for monitoring drpell pressure following an accident.
The licenste has proposed T$s that are consistent with the guidelines contained in Generic Letter 83-36.
Therefore, we conclude that the proposed TSs for drywell pressure monitors are acceptable.
4 Suppression Pool Water Level Monitor (!!.F.1.5)
The suppression pool water level monitors at Browns Ferry Unit 2 provides the capability required by TN! Action Plan Item !!.F.1.5.
The proposed TSs contain limiting conditions of operation and surveillance requirements that are consistent with the guidance conttined in Generic Letter 83 36.
Therefore, we conclude that the proposed T5s for
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suppression pool water level monitors are acceptable, i
7.
Testable Penetrations 2
Modifications are being made to the flange side of 14 containtient isolation 4
valves which cannot be isolated from primary containment to be tested.
This modification will provide two gaskets with a pressure tap between the i
gaskets to allow the flange to be leak tested. Operability of the valve will not be affected by this modification.
Fourteen new testable penetrations resulted and they were added to the table of testable penetrations with double o ring seals (Table 3.7.8).
New surveillance requirements are also being added.
This change was previously approved for Unit 3 by Amendment No. 78 dated August 27, 1984 Several editorial changes were also made to this table.
They include revising the identification name on several penetrations, adding a penetration that was tested but was inadvertently lef t out of the table and removing penetration j
X-213A which no longer exists. These changes are purely administrative.
1 Other minor corrections to this table were also made.
Penetration X-35G was J
listeo in this table for "T.I.P Orives" and is being vised to reflect that it is a "Spare." The drpell head is being added to is table.
It was inadvertently not listed, but was included in the surveillance program. We
(
have reviewed the proposed changes and find that the changes bring Table 3.7.8 into conformance with 10 CFR 50 Appendix J for all testable penetrations with double o-ring, and are acceptable.
8.
Redundant Air Supply to tt.e Orp ell This proposed change was removed by supplement 2 to the amendment request dated Dectriber 30, 1935.
9.
Demineralized Water Isolation Valve I
The TSs are revised to delt
'nry containment isolation valve 2-1143 of the demineralized water system
-ve holated the demineralized water line to the torus ring header, to longer used, so the valve will be removed and the line cap',
Mlated functions will be adversely was previously approved for Unit 3 affected by disconnectin;
'4 by Amendment No. 78 dateo.
_ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ We have reviewed 'this change and find that the TS change replacing the valve by a cap that will not leak is acceptable.
- 10. Residual Heat Removal (RHR) Head Spray Two isolation valves on the residual heat removal head spray line were recoved from Unit 2.
The head spray line was removed and the penetration capped. The TS are being revised to remove these valves from the table of valves to be tested.
The change deletes prinry containment isolation valves 74-77 and 74-78 of the RHR system head spray,from Tables 3.7.A and 3.7.F.
The removal of the head spray line is part of the Intergranular Stress Corrosion Cracking Study being done on Browns Ferry. No safety related functions will be adversely affected by disconnecting this line.
We have reviewed this change and find it acceptable.
D.
Administrative Changes Several administrative changes are being made to the Technical Specifications, These include revising the Table of Contents to reflect the change discussed d
above, and miscellaneous editorial changes such as to delete obsolete references, char.ge bases to reflect the changes to the Technical Specifications, correct page numbers, correct typographical errors, etc. The surveillance requirements for the personnel air lock is being changed to be consistent with the surveillance for Units 1 and 3.
The proposed change includes deletion of the reference to safety valves in conjunction with relief valves.
The safety i
valves with unpiped discharge have been removed and replaced with relief valves.
3.0 ENVIRONMENTAL CON 510'iRAT!0NS 1
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted a ia as defined in 10 CFR Part 20 and changes surveillance requirements.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant ha:ards i
consideration and there has been no public coment on such finding.
)
Accordingly the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Parsuant to 10 CFR $1.22(b), no 4
environmental impact statement or environmental assessment need be prepared in connection with the issuance of the Omendment.
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4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
I (1) there is reasonable assurance that the health and safety of.the.
j will not be endangered by operation in the proposed manner, and'(2) public a
such activities will be conducted in compliance with the Commission's regulations.
I and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Attachment:
Evaluation Principal Contributors:
W. Brooks, G. Schwenk. J. Mauk C. Patel, and M. Grotenhais Dated: August 19, 1986 l
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R ferences l
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1.
Browns Ferry Nuclear Plant Reload Licensing Report, Unit 2, Cycle 6; TVA-RLR 002, July, 1994, as supplemented.
2.
L. T. Mayhue,
- QUAD + Demonstration Asse.mbly Report" WCAP-1050' (Proprietary), March, 1984.
3.
GESTAR !!. ' General Electric Standtrd Application for Reactor Fuel",
NE00 2d.011-A-4, January, 1982.
4.
B. L. Darnell, et. al', "Methods for the Lattice Physics Analysis of LWR's," TVA TR78-02A, April, 1978.
5.
- 5. L. Forkner, et al, "Three Dimensional Core Simulator Methods".
TVA-TR78-03A, January, 1979, 6.
"Verification of TVA Steady State BWR Physics Methods", TVA TR79 01A, January, 1979.
7.
"8WR Transient' Analysis Model Utilizing the RETRAN Program". TVA TRB1-01, Decembe r,1981.
8.
Browns Ferry Nuclear Plant Reload Licensing Report, Unit 3, Cycle 6; TVA-RLR-001, January 1934.
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ATTACHMENT EVAt,Q_2 /. RELATING TO TOPICAL REPORT WCAP-10507 QUA0 + OEMONSTRATION ASSEMBLY REPORT
1.0 INTRODUCTION
WestinghouseNuclearEnergysystershaspreparedareport,WCAP10507l"QUAD-Demonstration Assembly Report" and submitted it to the NRC staff for information.
Since TVA has referenced tnis report in its application for the Cycle 6 reload of Browns Ferry Unit 2, the staf f has performed a "mini-review" of the report to evaluate the impact of including four of the QUAD + assemblies in the core as Lead Test Assemblies (LTAs).
All aspects of the assembly performance are evaluated except that of thermal-hydraulic stability.
That aspect is the subject of a separate evaluation.
The evaluation follows.
2.0 EVALUATION The QUAD + assembly has been designed to be a reload bundle for BWR/3 through RWR/6 cores with either "C" or "0" lattice designs.
It is intended to provide reduction in fuel cycle costs along with increased thermal margins.
Care has been taken to make the QUAD + assembly compatible with currently used BWR bundles, pa-ticularly the P8mBR design.
Details of the design of the QUAD +
assembly are held to be proprietary information by Westinghouse.
The report also includes a set of constraints to be used when inserting QUAD +
assemblies into a core as lead test assemblies (LTAs).
These include:
1.
The QUAD + demonstration assembly will not become a lead assembly during normal operation.
2.
The QUAD + demonstration assembly will not become limiting under trar.sient conditions.
3.
One QUAD + demonstration assembly should be placed 4djacent to a Local Power Range Monitor (LPAH) string.
4.
QUAD + demonstration assemblies should be loaded quarter-core symmetric.
a
5.
QUAD + demonstration assemblies ' vill not be loaded less than one row away from the analytically determined potential dropped rod.
6.
QUAD + assemblies should preferably not be loaded next to control rods which are inserted in the power range of operation during the first cycle.
2.1 Fuel Mechanical Desian The QUAD + assembly is designed to have the same length as the standard BWR assembly but has slightly larger lateral dimensions.
The QUAD + channel design has improved creep resistance compare ( to the standard design which ensures that an adequate gap between assemblies is maintained throughout core residence time to permit unhampered control rod movement.
The upper and lower end fittings of the QUAD + design interface with the core internals in the same manner as those of
- e standard design.
The QUAD + assembly contains more fuel rods than the standard assembly.
Each rod is smaller in diameter than the standard rod and is surrounded by Zircalley cladding which has been specially treated to improve corrosion resistance.
Six-inch blankets of natural uranium are provided at the top and bottom of the fuel stack and gadolinia is used in selected rods to improve radial power distribution and to control assembly reactivity.
Top and bottom structures are designed to be compatible with the core internals.
Grid spacers have been designed for low flow resistance and improved thermal performance.
Fuel rod integrity is assured by evaluation to design criteria which prevent excessive fuel temperatures, excessive internal rod gas pressures due to fission gas release, clad flattening, fatigue, corrosion above clad material removal limits, and excessive cladding stresses and strains during normal operation and anticipated transients.
The Westinghouse PAD fuel performance code was used for the analyses.
This code has been approved for use with PWR fuel and we find its use for QUAD + fuel acceptable for lead test assemblies.
This conclusion is based on the fact that large margins will be maintained between safety limits and expected fuel duty for the LTAs.
The design evaluations show that the QUAD + fuel meets all the design criteria with margin.
2.2 Nuclear Design The nuclear design of the QUAD + assemblies is described in the report.
The assemblies were designed to be as nearly the same as the P8x8R replacement fuel as feasible.
The assembly design and comparison calculations were performed with the PHOENIX and POLCA codes.
These codes have not been formally reviewed by the staff but information has been provided by Westinghouse to show that the PHOENIX assembly code gives results consistent with their standard design methods.
The POLCA code is sufficiently similar to the Westinghouse PALADON code to persuit the conclusion that the 3-D comparisons are acceptable, particularly since the QUAD + assembly are located in non-limiting positions.
Comparisons were made between the two assemblies for:
' assembly reactivity (K,. vs exposure)
.g.
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- local peaking factor
- void coefficient
- moderator temperature coefficient
- Doppler toefficient
' cold rodded and unrodded reactivity
- rod worth as a function of void content
' delayed neutron fraction and prompt neutron lifetime.
These calculations demonstrated that the QUAD + assembly characteristies were similar of those of the P8x8R assembly it is designed to replace, or were conservative with respect to it.
Three dimensional calculations were perforced with a QUAD + assembly replacing a standard assembly to confirm that such replacement has no significant effect on core behavior.
The QUAD + assembly has a slightly flatter end of-cycle axial power distribution than the standard assembly due to a smaller void coefficient in the former.
LPRM readings near the QUA0 + assembly were within 1 to 3 percent of those for a standard assembly assembly - well within the LPRM uncertainty. We conclude that $4bstitution of four QUAD + assemblies for four standard assemblies will have negligible effect l
on the neutronic behavior of the core.
2.3 Thermal Hydraulic Analysis Acceptability of the therral hydrau' sic design is based on hydraulic compatibility of the QUAD + desipe. with the 8x8R standard design and on acceptable CPR performance.
It is claimed that flow tests have shown that virtually identical pressure drops exist across the two bundle types at rated core flow and power conditions, but no data are presented.
Outer bypass flows and in channel flows are also the same for the assembly types.
Hydraulic compatibility is thereby assured.
The CPR performance of the QUA0 + assembly is calculated with the AA 74 correlation developed by ASEA-ATOM for an 8x8 fuel assembly.
This use is supported by the observation that the improved spacer l
grid design results in extra CPR margin for the QUAD + assembly. The use of the GExL safety limit value of 1.07 for the QUA0 + assembly (used with the AA 74 correlation) is supported by the fact that the convoluted uncertainties of the parameters used in the CPR evaluation are essentially the same for the two correlations.
However, the form of the two correlitions is different and the conclusion that a limit of 1.07 applies to both may not be valid.
Finally the GExL correlation will be used for the QUAD + demonstration assemblies when operating in the reactor.
The two correlations have been compared for a number of plant operating conditions and shown to give similar results.
In order to obtain additional margin to CPR limits the guidelines listed in Section 1 above are designed to provide a 10-20 percent margin in power between the QUAD + assemblies and the leading assembly under normal cperating core conditions.
2.4 Transient and Accident Analyses 2.4.1 Core-Wide Transients The consequences of core-wide transients depend upon core-wide neutronics parameters, which are not altered significantly by the presence of the four 3
l QUAD + assemblies.
Thus the core response is not altered but tSe transient response of the assemblies themselves must be considered.
For slow transients, such as loss of feedwater heater, the change in CPR for the QUAD + ssembly is essentially the same*as that for the P8x8R assembly.
The rapid tra sients, such as loed rejection without bypass, result in larger MCPR change for the QUAD + fuel relative to the standard fet1.
For a typical such trar isnt the change in CPR of a QUAD + bundle could be as great as 8 percent lar er than that for the standard bundle.
As indicated in Section 4 above a eargin of 10 to 20 percent is provided by following the guidelines given in Section 1.
In view of the increased change in CPR during transients and the uncertainties in the applicability of the GEXL correlation to the QUAD + assembly we conclude that the generic mergin of 10 to 20 percent is not sufficient.
We will therefore require cycle specific calculation. to assure that a margin of at least 20 percent is present.
2.4.2 Dropped Red The QUAD + assemblies will be p1tted in the core in positions at least one row away from the rod shown by analysis to have the greatest worth in the startup regime where the consequer.ces of the rod drop accident are significant.
The QUAD + assembly will thus not be limiting for this event.
2.4.3 Rod Withdra.a1 Error The rod werths at power are smaller for QUA0 + assemblies than for standard ones.
In addition the QUAD + assemblies will be loaded into non limiting locations.
The intent of the demonstration progran is to have the QUAD +
assemblies in non rodded locations at power.
For these reasons the presence of the QUAD + assemblies will not affect '.he rod withdra.a1 error analysis.
2.4.4 Fuel Misleadina Event The mislocction and miserientation of QUAD + assembly has been analyzed.
Since it has been designed to have essentially the same reactivity as the corresponding P8x8R assembly the analysis for the latter assembly is applicable.
Thi flatter enrichment distribution factor of the QUAD + assembly result in smaller changes in LHGR and CPR for miserientation events than with the corresponding P8x8R assembly.
2.4.5 Loss of Coolant Accident (LotA)
The QUA0 + assembly has several features which tend to mitigate the consequences of the loss of coolant event when compared to the equivalent P8x8R assecbly.
These include isproved radiation heat transfer charact, eristics and a thinner enannel which is wore ensity quenched.
The lower plate design tends to delay the voiding of the assembly leaoing to ar. extended film bniling period.
For the same fuel bundle power, the linear heat generation rate in the fuel is lower.
These reactors tend to reduce the pe A cladding temperature in a LOCA coroared to the equivalent PSx8R assembly.
Thus it may be concluded that, the 4
LOCA analysis performed for a core loaded with standard assemblies will be applicable to QUAD + fuel and that MAPLHGR limits obtained for the equivalent P8x8R assembly say be conservatively applied tc the QUAD + assembly.
- 3. 0 CONCLUSIONS Based on the review which is described above we conc 1LJe that WCAP- 050T presents sufficient information to support the use of up tc four QUA0 + bundles as demonstration assemblies in 8WR/3 through BWR/6 cores provided tha++
1.
The guidelines presented in Section 4.1.2 of VCAP 10507 are adhered to, and 1
2.
Cycle specific analyses are performed to show that a margin of at least 20 percent in pow:r exists between
. the QUAD + assembly and the Itad assembly when the core is operating at full power, full flow conditions.
Any more extensive leading of QUA0 + assemblies into SWRs will be subject te review in considerably greater depth than is described in this evaluation.
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