ML20135C864
| ML20135C864 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/14/1997 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20135C833 | List: |
| References | |
| NUDOCS 9703040346 | |
| Download: ML20135C864 (100) | |
Text
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O ATTACHMENT II FARLEY NUCLEAR PLANT TECHNICAL SPECIFICATIONS CHANGE REQUEST POWER UPRATE FACILITY OPERATING LICENSE CHANGED PAGES FNP Unit 1 FOL NPF-2 Section 2.C (1)
FNP Unit 2 FOL NPF-8 Section 2.C (1) l i
O 9703040346 970214 PDR ADOCK 05000348 P
.=
UCENSE No. NPF-2
.' (5)
Southern Nuclear, pursuant to the Act and 10 CFR l
I Parts 30, 40 and 70, to receive, possess and use in 4
amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and j
(6)
Southern Nuclear, pursuant to the Act and 10 CFR i
Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
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This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and i
50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now i
or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level f//jf" 1
Southern Nuclear is a thorized to operate the I
facility at steady a ate reactor core power levels C)\\
not in excess of -E6&E megawatts (thermal).
Prior to attaining the power level, Alabama Power Company shall complete the preoperational tests, startup tests and other items identified in Attachment 2 to this license in the sequence specified.
Attachment 2 is an integral part of this license.
(2)
Technical Snacifications The Technical Specifications contained in Appendices A and B, as revised througt Amendment No. 94 are hereby incorporated in the license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
~
Farley - Unit 1 Amendment No. 90 l
l L / CENSE ND. NPF - B t
4 (6)
Southern Nuclear, pursuant to the Act and 10 CFR l
l Parts 30, 40 and 70, to possess, but not separate, O
such byproduct and special nuclear materials as may i
be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's 3
regulations set forth in 10 CFR Chapter I and is subject
{
to all applicable provisions of the Act and to the rules, regulations, and orders of the commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
1 (1)
Maximum Power Level i
SouthernNuclearisauthorizedtoo$eratethefacility at reactor co l
I l
nE44 M megawatts thermal.
2771T (2)
Technical Soecifications e
i The Technical Specifications contained in 4
Appendices A and B, as revised through Amendment No.83, are hereby incorporated in the license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
(3)
Initial Test Procram
}()
Alabama Power Company shall conduct the' initial test program (set fcrth in Section 14 of the Final l
Safety Analysis Report as amended) without making any modifications to this program unless such 1
modifications are in accordance with the provisions of 10 CFR Section 50.59.
In addition, Alabama Power Company shall not make any major modifications to this program unless the modifications have been identified and have received prior NRC approval.
Major modifications are defined as:
a.
Elimination of any test identified as essential in Section 14 of the Final Safety Analysis l
Report, as amended; b.
Modification of test objectives, methods or acceptance criteria for any test identified as essential in section 14 of the Final Safety l
Analysis Report, as amended; c.
Performance of any test at a power level different from the level in the described program; and O
Farley - Unit 2 Amenhant I4o. 83
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i ATTACHMENT III s
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I FARLEY NUCLEAR PLANT TECHNICAL SPECIFICATIONS CHANGE REQUEST I
i FNP Unit 1 Technical Specifications Changed Pages List i
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FNP Unit 1 Technical Specifications Marked-up Pages j
FNP Unit i Technical Specifications Typed Pages j
FNP Unit 2 Technical Specifications Changed Pages List i
l FNP Unit 2 Technical Specifications Marked-up Pages i
]
FNP Unit 2 Technical Specifications Tyi,ed Pages 2
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ATTACHMENT III FARLEY NUCLEAR PLANT l
TECHNICAL SPECIFICATIONS CHANGE REQUEST POWER UPRATE
'lNP Unit 1 Technical Specifications Changed Pages List FNP Unit 1 Technical Specifications Marked-up Pages FNP Unit 1 Technical Specifications Typed Pages FNP Unit 2 Technical Specifications Changed Pages List FNP Unit 2 Technical Specifications Marked-up Pages FNP Unit 2 Technical Specifications Typed Pages O
I o
i 1
FNP Unit 1 Technical Specifications Power Uprate Implementation Chanaed Panes Umt 1 Bevision Page 1-6 Replace Page 2-5 Replace Page 2-6 Replace Page 2-7 Replace Page 3/4 2-15 Replace Page 3/4 3-25 Replace Page 3/4 3-27 Replace Page 3/4 3-28 Replace Page 3/4 5-5 Replace Page 3/4 7-9 Replace O
Page B 3/4 2-5 Replace Page B 3/4 6-2 Replace Page B 3/4 7-1 Replace -
Page 5-6 Replace Page 5-8 Replace Page 6-19a Replace Page 6-24 Replace O
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DEFINITIONS ggl ED THERMAL POWER 1'
l.25 the reactor coolant of 4Hf MWtRATED THERMAL POW
'i 2775 REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall b j
the monitored parameter exceeds its trip setpoint at the channel sensor until j
loss of stationary gripper coil voltage.
1 REPORTABLE EVENT i
1.27 A REPORTABLE EVENT shall be any of those conditions specified in Sectio j
50.73 to 10 CFR Part 50.
1 SHUTDOWN MARGIN 1
1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by unic i
the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully i
inserted except for the single rod cluster assembly of highest reactivity worth
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which is assumed to be fully withdrawn.
SOLIDIFICATION t
1.29 This definition deleted.
Refer to the Process Control Program.
e SOURCE CHECK i
1.30 A SOURCE CHECK shall be the qualitative assessment of channel response
{
when the channel sensor is exposed to a radioactive source.
STAccrern TEST RASIS 1.31 A STAGGERED TEST BASIS shall consist of:
A test schedule for n systems, subsystems, trains or other designated a.
components obtained by dividing the specified test interval into n equal subintervals, b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
FARLEY-UNIT 1 1-6 AMDEMDIT NO. 57. 99
. -... -. __.....--. - - -. -._. ~.-
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TABt2 2.2-1 REACTOR TRIP STSTDI INSTRtNtEftfATION TRIP SETPOINTS 5
g FilNCTIONAL INIIT TRIP SETFOINT ALLOVABLE VALUES 1.
Manual Reactor Trip Not Applicable Not Applicable 25.4 2.
Power Range, Neutron Flux Low Setpoint - f 25! of RATED Low Setpoint - $-f6E of RATE TERRNAL POWER TRERNAL F0VER jg4 Eigh Setpoint - S 1991 of RATED Eigh Setpoint - $-H44-of RATED TERRNAL POWER TEERNAL POWER i
S.4 3.
Fever Range, Neutron Flux, i SE of RATED TBERNAL POWER wit f-5dt of RATED THERMAL POWER Eigh Positive Rate a time constant 1 2 seconds with a time constant 2 2 seconds 5.4 4.
Fover Range. Neutron Flux, f 52 of RATED THERMAL POWER with
$ Jydt of RATED THERMAL POVER Nigh Negative Rate a time constant 2 2 seconds with a time constant 2 2 seconds y
5.
Intermediate Range. Neutron 5 25E of RATED TBERMAL F0VER f 30% of RATED THERMAL. POWER Flux 6.
Source Range. Neutron Flux 1 10' counts per second i 1.3 I 10' counts per second 7.
OvertemperatJte af See Note 1 See Note 3 8.
Overpower af See Note 2 See Note 6
/86Z 9.
Pressurizer (ressure--14v 2 1865 psig 2 -19% psig 2388 10.
Pressurizer Pressure--Bigh
$ 2385 psig
$ 4395 psig 11.
Pressurizer Vater
$ 92% of instrument span f 93E of instrument span Level--Bigh Q
29.7 12.
Loss of Flow
/
2 90% of minisue ocasured flov 2-966% of minimum measur ed f low l
5 per loop
- per loop
- 84/D0 em N*
- Minloue measured fIov is 49;f99 gpe per 1oop.
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9 TAar F 2.2-1 f t'nn t I na aa-8 l__
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BRA N TRIP SYsrzsf ThleTar--
N FitMr'TIONAL IBIIT
== TION TsIP ha ruInfTS TRTP SETPOIhrt n
i nr.rna.ramr m varjngs
- 13. Steam Generator Water 24.4 Level--pow-8.ow 2 25% of narrow range instrtament span - each 2 4+rM of narrow range inst ruim sit i
- 14. Deleted steam generator span - each steam generator
- 15. Undervoltage - Reactor Coolant Pumps 2 2680 volts - each bus 2 2640 volts - each bus 6
- 15. Underfrequency Reactor Coolant 2 57.0 Hz - each bus i
Pungs 2 56.9 Hz - each bus
- 17. Turbine Trip A. Low Auto stop 2 45 psig u
Pressu
- 9. Turbing[e 2 43 psig Stop Valve Not Applicable closury Not Applicable
- 18. Safety injection Input i
from RSP Not Applicable t
Not Applicable
- 19. Peactor Coolant Puse i
Not Applicable greaker Position Trip Not Applicable Ie 5
-2 h
t O
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i TABLE 2.2-1 (Continued)
E REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS Qa5 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
-4 20.
Reactor Trip System Interlocks A.
Intermediate Range Neutron
> 1 x 10~
> 6 x 10'II amps amps Flux, P-6 B.
Low Power Reactor Trips P-10 Input 5 10% of RATED 5-H% of RATED l
Block, P-7 THERMAL POWER THERMAL POWER j
P-13 Input
$ 10% Turbine
$ 11% Turbine Impulse Pressure Impulse Pressure Equivalent (Note 4 Equivalent (Note 5) 3D O.4 s
C.
Power Range Neutron Flux, P-8
< -355 of RATED
< -36% of RATED 4
THERMAL POWER THERMAL POWER w
D.
Power Range Neutron Flux, P-10
> 8% of RATED
> -7Er of RATED THERMAL POWER THERMAL POWER E.
Turbine Impulse Chamber, P-13
$ 10% Turbine
$ 11% Turbine Impulse Pressure Impulse Pressure Equivalent (Note 4 Equivalent (Note 5) 50 4 F.
Reactor Trips Block
< 50% of RATED
<-SIE of RATED g
Following Turbine Trip, P-9 THERMAL POWER THERMAL POWER m
y 21.
Reactor Trip Breakers Not Applicable Not Applicable 22.
Automatic Trip Logic Not Applicable Not Applicable M
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TABLE 3.2-1 5;"
DNB PARAMETERS 8
Ey LIMITS PARAMETER 3 Loops in Operation 2 Loops in operation
'520' Indicated Reactor Coolant System T, f 500rFF
(**)
l Indicated Pressurizer Pressure 1
psig*
(**)
l Indicaced Reactor Coolant System 2 207,SS^ gpm***
(**)
l Total Flow Rate 2 M,2D6 W
Y 0;
t E
Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
Values blank pending NRC approval of 2 loop operation.
)
- Value includes a 2.4% flow uncertainty (0.1% feedvater venturi fouling bias included).
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O TABLE 3.3-4
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.iEERED SAFETY FEATURE ACTUATION SYSTEN INSTRtMENTATION TRIP SETPOINTS
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FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES 4
1.
SAFETY INJECTION, TUR8INE TRIP AND FEEDWATER ISOLATION a.
Manual Initiation Not Applicable Not Applicable b.
Automatic Actuation Logic Not Applicable Not Appilcable c.
Containment Pressure--High
~< 4.0 psig
< 4.5 psis-
/847
~
c!.
Pressurizer Pressure--Low 1 1850 psig 14840-psig e.
Differential Pressure i 100 pst i 112 psi Between Steam Lines--High f.
Steam Line Pressure--Low 1 585 psig 1 575 psig I
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Tant,E 1.3-4 f f*rwir landl_
i EMr2]Merarn RAFETY FEATilar ACTilATICBA BYameam
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IMCTatm"*FATItBt TRIP BGrufMFS e
FlatPTinasA L lasIT k
TRIP SETPOIMP ALitN&BLE VALilES y
4.
STEAM LINE ISOLATION l
i a.
Manual Not Applicable I
Not applicable b.
Automatic Actuation Logic Not Applicable Not App!!ca@le i
c.
Contairusent Pressure--
High-Nigh 1 16.2 peig 5 17.5 psig d.
Steam Flow in hso Steam Lines--High, S A function defined Coincident with as follows:
5 A function tn:3 l
Tave--Low-Low A Ap corresponding to defined as follows:
gg,3 40% of full steam A Ap corresponding to -444 t
flow between et and of full steam flow between On and 20% load and then 20% load and th-an a Ap a Ap increasing linearly w
increasing litently to a
a Ap corresponding to to a Ap corresponding to u
/,
!!Gt of full steam 4+h5% of full steam flow at flow at full load full load with T.,g 2Mr 4
i with Tavg 2 543*r
//B.3/
S42.4 I
t s.
Steam Line Pressure--Low 2 585 psig 2 575 psig 5.
1VRBINE TRIP AND FRED NATER r
ISDLATICBf 78.S' i
78.9 a.
steam Generator Water
$ 49;tt of narrow range 5 4Sr54 of narrow range j
j l,
Level--Nigh-High instrument span each instrument span each i
steam generator steam generator H
e6
(
6
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O TABLE 3. 3-4 front /NUED )
ENCINrrarn SAFETY FEATURE ACTUATION SYaiman 11.m mi- -iFATION TRIP SETPO1HIS FUNCTIONAL UNIT TRIP SETPOIMF ALIAMARLE VAIERS 6.
AUXILIARY FEEDNATER Automatic Actua$1on N.A.
a.
Logic N.A.
e 24.S b.
Steam Generator Mater 2 25% of narrow range 2-34r44 of narrow range l l Level--Low-Low instrument span each instrument span each steam steam generator generator c.
Undervoltage - RCP 2 2600 volts 2 2640 volta d.
S.I.
See 1 above la11 SI setpointal i
e.
Trip of Nain Feedwater N.A.
N.A.
Pumps 7.
LOSS OF PONER A
t a.
4.16 kv anergency Bus 2 3255 volts bus 2 3222 volts bus voltage" Undervoltage (Loss of voltage
- u O
Voltaget s 3410 volts bus voltage
- as i
b.
4.16 kw Dnergency sus 2 3675 yotte bus 2 3638 volts bus voltage
- Undervoltage (Degraded voltage 5 3749 volts bus voltage
- i Voltagel 8.
ENGINEERED SAFETY FRA111RE AC'rtlATION SYSTipt INTERLOCKS Ed23 I
- b.
a.
Pressurizer 'P'rees' ure, i
P ll 5 2000 peig 5 2914 pelg 545-545.f Low-LowTayf,P-12 (Increeaing 444*F 5-444*F IDecreasing) 543*F 2 444*F f
c.
Steam Generator Level, P-14 (See 5. abovel 848 0 3
d.
Reactor Trip. P-4 N.A.
N.A.
a agter to eppropriate relay setting sheet calibration requirements.
e
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
By verifying the correct position of each mechanical position stop 1
e.
for the following ECCS throttle valves:
1.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.
2.
At least once per 18 months.
Valve Number I
i CVC-V-8991 A/B/C CVC-V-8989 A/B/C CVC-V-8996 A/B/C CVC-V-8994 A/B/C f.
At least once per 18 months, during shutc own, by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
2.
Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:
a)
Centrifugal charging pump b)
Residual heat removal pump j
g.
By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when ested pursuant to Specification 4.0.5:
gg SID 1.
Centrifugal charging pump
> 2458 14 5 2.
?,
PSID i
h.
Prior to entry into Mode 3 from Mode 4, verify th e n.echanical stops on icw head safety injection valves RHR-HV 603 A/B are intact.
O i
FARLEY-UNIT 1 3/4 5-5 AMENDMENT NO. 26 1
- ~.. - -
PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES i
LIMITING CONDITION FOR OPERATION i
4 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY:
MODES 1, 2 and 3.
1 ACTION:
}
l HOD 1 - With one main steam line isolation valve inoperable, POWER OPERATION a
may continue provided the inoperable valve is restored to OPERABLE p
status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
t 4
MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided the isolation valve is restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i
after entering MODE 2; otherwise, be in HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The provisions of Specification 3.0.4 are not applicable.
[}
SURVEILLANCE REQUIREMENTS
~, -
- 4. 7.1. 5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5-seconds when tested pursuant to Specification 4.0.5.
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FARLEY-UNIT 1 3/4 7-9 AMENDMENT NO. 26 l
- -. -. - ~ -
POWER DISTRIBUTION LIMITp BASES J
3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and J
periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts.
The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 ir provided to allow identification and correction of a i
dropped or misaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on Fg is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0.
J' For purposes of monitoring QUADRANT POWER TILT RATIO when one encore detector is inoperable, the movable incore detectors are used to confirm that the normalised symmetric power distribution is consistent with the QUADRANT POWER 4
TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four synestric thimbles. The two sets of four symmetric l
thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, B-3, B-13, L-T.,
1-11, and N-8.
3/4.2.5 DNE PARAMETERS 2.9 2Po9 77/E4/'EE44E of i
The limits on the DNB relat parameters assure that ch of the parameters ar l
maintained within the normal steady state envelope of o ration assumed in the.
transient and accident analy es.
The limits are consiste with the initial FSAR assumptions and have n analytically demonstrated ade ate to meet the j (%d. 3 esign criterion throug ut each analysed transient.
The dicated T Q
value o 440,4*F is based o the average of two control board re ngsandy n
indication uncertainty of 4v4'F.
The indicated pressure value of 43G6-g is i
based on the average of two control board readings and an inv. ation uru..-e m i -t ef -ee psi. The indicated total RCS flow rate is based on two elbow 24 tap measurements from each loop and an uncertainty of 2.4% flow 0.1 ow is included for feedwater venturi o AS READ M THE PL ANT COMPUTER The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of Tav and pressurise e
g t e con r board readings are sufficient t$ ensure that the parameters are restored within their 1 ewin bined changes and other ex et d r s_iant si operation.
MEASUREMENTS /S HEAT BMANCE The 18 month surveillance of the to al RCS flow ra may be performed by one of two alternate methods, one method is a precision eler " tri performed at the beginning of each fuel cycle. The other method is based on the Ap measurements f rom the cold leg elbow tape, whi are correlated to past precision heat balance measurements. Correlation of the flow indication channels with selected precision loop flow reler" trier fer thi: ::th " i: documented in WCAP-14750. Use of the elbow tap op measurement method removes the requirement for performance of a precision RCS flow ::12:i;.;^.ri; measurement for that cycle. The monthly surveillance of the total RCS low rate is a reverification of the RCS flow requirement using process compu er indications of loop elbow tap measurements that are correlated either to the pre ision RCS flow meheurement or the elbow tap measurement at the beginning of t a fuel cycle.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of gnificant flow degradation using the control board indicators fed by elbow tap asurements.
O HEAT BALANCE Revised by NRC letter FARLEY-UNIT 1 B 3/4 2-5 dated: December 11. 1996
- = - ---
AND FAbM A MS L B EVENT /S S3 P3/&.
THE ANALYSES EEsulT.S
'Y*'***
DEMdNSTRATE THAT T
max j
.M psigd on a event is initial positive pressure of up to 3 psi the maximum containment pressu 1 remain be ow the de q _5J_ps THESE CMTA/NMENT ANALYSES CAllOLAT/DNS /NCLUDE 3/4.6.1.5 ATR TEMPEMA 1
The limitations on containment average air temperature ensure that the j
overall containment average al.r temperature does not exceed the initial temperature condition assumed in the accident anal sie for a IDch or steam fiiTRiCoureo iNrERNRt im oDwo a ar %
1'~ ~ ~ ~
cc'*
c-l 3/4.6.1.6 cONTaTWWFMT muuu-t 1-u am
[ O(([E[ O(($8M8[ d[88[d -
43 3 PS/6.
I This limitation s that the struct al integrity of thw-e..ia nt i
will be maintained compar le to the orig design standards for the lif f
the facility. Structural ity is ired to ensure that the con amont l
will withstand the maximum pressure of 48'peig in the event of a he i
measurement of the containment lift off force, visual examination of te ne, anchorages and exposed interior and estarier surfaces of the containment, and j
the containment leakage tests are sufficient to demonstrate this capabili l
l i
The surveillance requirements for demonstrating the containment's structural integrity are in compliance with the roccamendations of paragrap I
c.1.3 of magulatory Guide 1.35 ' Inservice surveillance of Ungrouted Tendons n
j Prestressed concrete containannt structure," January 1976.
3/4.6.1.7 CONTATNMEN? VBrFTt_1vfGE SYSTEM The 48-inch containment purge supply and exhaust isolation valves are j
required to be closed in MODES above COLD SEUTDOWN since these valves have not been demonstrated capable of closing during a 14ch or steam line break accident. Maintaining these valves closed during plant operations ensures
)
that excessive quantities of radioactive materials will not be released via the containment purge system.
4 l
The use of the contain= ant purge lines is restricted to the S-inch vont supply and exhaust isolation valves to esaure that the site h-ad=*y does guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss-of-
)
coolant accident dur v==*4== --h-mN
^'
l
/NADD/T/dN, STRUCTURAL /NTEBR/TF /S REau/KED To EA6tME THRT THE CDUTA/NMENT MLL MTHSTAND THE JfAf/ MUM FEAK CALCULATED /NTERNAL PRESBURE OF S3 PS/& /N 7NE EVENT OF A MSLB, /NCLUD/N6 AN 1
IN/f/AL PdS/T/VE PRESSURE OF UP TO 3 PS/&. AS PER APPEND // I l
DPT/oN B, THE LOCA PEAK LALLULATED CdNTA/NMENT /NTEENAL J
l PRESSURE DEF/NES THE Pa VALUE FAE THE CoMTA/NMEA/T i
LEAKA&E KATE TEST /N& PROGRAM KEOU/ RED BY SURVE/LLANCE 4. b. /. 2.
ranz.rr-uurr 1 s 3/4 6-2 Amendment No. 122
3/4.7 PLANT SYSTEMS BASES
'~
3/4.7.1 TURBINE CYCLE i
3/4.7.1.1 SAFETY VALVES The OPERASILITY of the main steam line code safety valves ensures that the sec=ndary system pressure will be limited to within 1104 4
(1194 psig) of its design pressure of 1085 pstg during the most severe antteipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 1006 RATED THERMAL PCWER coincident with an assumed loss of condenser heat sink (i.e., no steam
)
bypass to the condenser).
j The specified valve lift settings and relieving capacities are in a==ordance with the re74 rements of Section III of the ASME Boller and i
Pressure Code, 1971 Edition. The total relieving capacity for all jg73
- j3 valves on all of the steam lines is at least 12,984,660 lbs/hr which is 112 percent of the total secondary' steam flow of 11,;;;,";; lbs/hr at 1006 RATED THERNAL POWER.
A m.tnimum of 2 CPERABLE safety alves per steam generator ensures that sufficient relieving capacity av f=r the allowable THERMAL PCWER restriction in Table 3.7-2.
j g
STARTUP and/or POWER CPERATION is allowable with safety va ve inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER 1
required by the reduced reactor trip settings of the Power Range Neutron Flux channels.
The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis, i
!U i
u i
1 i
i j
1 FAALEY-UNIT 1 3 3/4 7 1 AMENDMENT No. 26.112 5
- - _ _ -. -. _ _ ~
i i
O DESIGN FEATURES j
]
5.3 REACTOR CORE fvtL ASSEMBLIES 5.3.1 The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy, zirealoy-4, or ZIRLo" fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO )
fuel material.
2 as 4
(
Limited substitutions of zirconium alloy, siccaloy-4, ZIRLo", or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC-approved codes and j-methods, and shown by tests or analyses to comply with all fuel safety design bases.
A limited number of lead test assemblies that have not completed i
representative testing may be placed in non-limiting core regions.
CONTROL RCD ASSEMBLIES 3
5.3.2 The reactor. core shall contain 40 control rod assemblies.
material shall be silver, indium and cadmium as approved by the NRC.The control j
5.4 REACTOR COOLANT SYSTD(
e j
DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
In accordance with the code requirements specified in Section 5.2 a.
t of the FSAR, with allowance for normal degradation pursuant to j
the applicable Surveillance Requirements, I
l b.
For a pressure of 2485 psig, and I
For a temperature of 650*F, except for the pressurizer which is 680'F.
c.
I VOLUME 9829 1
5.4.2 The total water and steam volume of the reactor coolant system is M M t l
100 cubic feet at a nominal Tavg of M F.
5.5 METEOROLOGICAL TOWER IOCATION 5/o 7. Z I
5.5.1 The meteorological tower shall be located as shown on Figure 5.5-1.
i k
,1 i
i
!O 3
FARLEY-UNIT 1 5-6 AMENDMENT No. 26,52, i
66,79,91,110 d
I
o O
O R
TABLE 5.7-1
(
{
g C0090NENT CVCLIC OR TRAIISIEllT L181ITS l
E q
CYCLIC on 8
TRAltsIENT LletIT DESIGIl CVCLE COMPONENT e
p se Tamusitui Reacter Coelant System heatup cycles at i let*F/hr Heatup le - T"8 from 1 200*F i
200 coolaisun cycles at to >
i
< IS0*F/hr.
CaeTdsum cycle - T,,,,,,,
j 1 558"F to 3 200*F.
200 pressurizer coeldsist cycles r
et i 200*F/hr.
Pressurtner coeldeue cycle L
temperatures from 1 650*F to
$ 299*F.
i SS less of lead cycles, without
> 155 et RATER TIERMAL POWER ts t
tamediate turbine er reacter trip.
k of RATES TIE 8014L PeWER.
T l
as 48 cycles et less er offsite Less of offsite A.C. electrical A.C. electrical power.
ESF Electrical System.
l SS cycles of less of flew in ene Less of only one reacter reacter coolant leep.
coolant pump.
L 400 reacter trip cycles.
1985 to E of AATES TOElpl4L POWER.
i i
14 inadvertent ammillery sprey Spray water temperature differeettal acts.tl.a ce i.
> am F.
- l 54 leen tests.
Pressurized te 1 2485 psig.
j 5 bydrostatic pressure tests.
Pressurized to 1 3888 psig.
l
,5 Seceedary System I steam ilan break.
i areek la a > 6 Sech st.se ise..
E j
@ hydrostatic pressure tests.
Pressurized to > 1356 psig.
/D 1
i ADMINISTRATIVE CONTROLS 5
j 2.
WCAF-lo216-P-h, Rev. lA, " Relaxation of Constant Axial offset Control /
j Fg Surveillance Technical Specification," February 1994 (M Proprietary).
k (Methodology for Specifications 3.2.1 - Axial Flux Difference and 3.2.2 - Beat Flux Bot Channel Factor.)
4
)
.,K w w G26; + ;.,
.s.
2, "T.'.; 1901 '1::;ie Of "::W.;;'.:::: 5:1;;ti;r.
...o
,.,_m w.
6
,an,
,o r
j (Methodology for Specification 3.2.2 - Heat Flux Hot Channel i
Factor.)
l The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, Eccs I
limits, nuclear limits such as shutdown margin, and transient and accident
]
analysis limits) of the safety analysis are met.
I The core OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and
{
Resident Inspector.
l t
)
ANNUAL DIESEL CENERATOR RELTAnftfTY DATA REPORT i
6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually.
q This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.
1 1
)
Insen A to pg 6.19a g
{
3a.
WCAP-12945-P, Vol. I &,1992-1993, " Code Quahfication Document for Best l
Estimate LOCA Analysis". (W Proprietary) l 3b.
Letter, N. 3. Liparuto, (W) to R. C. Jones, (USNRC), " Revision to Westinghouse l
Uncertainty Methodology", NTD-NRC-95-4575, Oct.M; 1995.
Letter, R. C. Jones (USNRC) to N. J. Liparulo (W), "A@ccep i
3c.
i the Topical Repon, WCAP-12945(P), Westinghouse Code Qualification Document for 1
Best Estimate I.oss-of-Coolant Analysis", June 28,1996.
1 t
4
~
FARLEY-UNIT 1 6-19a AMENDMENT No.
i I
1 i
ADMINISTRATIVE CONTROLS j
6.is MAJOR m-rs To mammivr wkzTr mamwT sYmas (Liquid, Gaseous, I
solid) i j
This specification deleted. Refer to the offsite Dose calculation Manual and l
i the Process Control Program.
I 6.16 CONTAIMMENT I.E&EAGE RATE T5 STING FROGRAM
{
A program shall be established to implement the leakage rate testing of j
containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option 5, as modified by approved amamptions. This program shall be in accordance with the guidelines contained in Regulatory guide 2.163,
- performance-Based 1
containment Leak-Test am,' dated September 1995.
te peak calculated % ',Pa [ - t internal pressure for the design basis loss coolant accident, Pa is AS" ig.
43
}k The maximum allowable containment i rate, La, at Pa, is 0.15% of containment air weight ter day.
Leakage rate acceptance criteria ares j
a.
containment overall leakage rate acceptance criterion is s 1.0 L.
During plant startup fallowing testing in accasemaos with this program, the 4
leakage rate acceptance criteria are s 0.60 1. for the combined Type a and c j
tests, and 5 0.75 La for Type A tests; I
l
)
b.
Air lock testing acceptance criteria ares 1
l 1)
Overall air lock leakage rate is 5 0.05 La.when tested at 2 Paa i
j 2)
For each door, leakage rate is s 0.01 La when pressurized to j
2 10 poig.
j The provisions of Specification 4.0.2 do not apply to the-test frequencies specified in the containment Leakage Rate Testing Program.
l The provisions of Specificatierr e.0.3 arr appliaahle to the rm=*ei-+
l Leakage Rate Testing Program.
l 1
i b
]
i j
4 f
1
DEFINITIONS i
RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to i
the reactor coolant of 2775 MWt.
l i
REACTOR TRIP SYSTEM RESPONSE TIME
)
1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from
]
when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in section t
50.73 to 10 CFR Part 50.
SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which i
the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SOLIDIFICATION 1.29 This definition deleted. Refer to the Process Control Program.
j SOURCE CHECK 1.30 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
J EJAGGERED TEST BASIS i
1.31 A STAGGERED TEST BASIS shall consist of:
a a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
i J
J a
a J
FARLEY-UNIT 1 1-6 AMENDMENT NO.
4
O O
O TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Q
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES a
1.
Manual Reactor Trip Not Applicable Not Applicable a
2.
Power Range, Neutron Flux Low Setpoint - s 25% of RATED Low Setpoint - 5 25.4% of RATED l
THERMAL POWER THERMAL POWER High Setpoint - 5 109% of RATED High Setpoint - s 109.4% of RATED l
THERMAL POWER THERMAL POWER 3.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with 5 5.4% of RATED THERMAL POWER l
High Positive Rate a time constant 2 2 seconds with a time constant 2 2 seconds 4.
Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with 5 5.4% of RATED THERMAL POWER l
High Negative Rate a time constant 2 2 seconds with a time constant 2 2 seconds 5.
Intermediate Range, Neutron S 25% of RATED THERMAL POWER S 30% of RATED THERMAL POWER 4
Flux 6.
Source Range, Neutron Flux s 105 counts per second 5 1.3 X 105 counts per second 7.
Overtemperature AT See Note 1 See Note 3 8.
Overpower AT See Note 2 See Note 6 9.
Pressurizer Pressure--Low 2 1865 psig 2 1862 psig l
- 10. Pressurizer Pressure--High 5 2385 psig s 2388 peig l
- 11. Pressurizer Water s 92% of instrument span s 92.4% of instrument span l
h Level--High 5g
- 12. Loss of Flow 2 90% of minimum measured flow 2 89.7%'of minimum measured flow l
g per loop
- per loop
- 8
- Minimum measured flow is 88,100 gpm per loop.
l
n.....- -. - --.
. - _ -. - ~..
. =..
..~.._ ~_--.....
i O
~)
O i
9 TABLE 2.2-1 fContinuedI I
E.
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS c
I d
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES H
- 13. Steam Generator Water 2 25% of narrow range instrument 2 24.6% of narrow range instrument l
Level--Low-Low span - each steam generator span - each steam generator 14.
Deleted
- 15. Undervoltage - Reactor 2 2680 volts - each bus 2 2640 volts - each bus l
Coolant Pumps
- 16. Underfrequency - Reactor 2 57.0 Hz - each bus 2 56.9 Hz - each bus Coolant Pumps
- 17. Turbine Trip A.
Low Auto stop 2 45 psig 2 43 psig l
g 4
Pressure B.
Turbine Stop Valve Not Applicable Not Applicable Closure
- 18. Safety Injection 1.'out Not Applicable Not Applicable I
from ESF i
- 19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip i
a E
5
=
t
r
+ If l
t.
I l
6 l
l l
l 3
)
)
5 5
S s
D ee D
ee D
E p
E rt E
D rt E
U m
T uo T
E uo T
e e
L a
l A
RE ns(
RE AE ns(
RE b
b V
1 W
ie W
RW i e W
a a
1 f O b rt f O O
b rt f O c
c E
oP rP n oP fP rP n oP i
i L
0 u
e o
u e
l l
B 1
%L Tel
%L L
Tel
%L p
p A
A sa A
p p
4
- 4. A
%A sa M
%l v M
M
%l v
- 4. M A
A W
x O
0R 1 ui 0R
- 6. R 1 ui 0R L
6 1E 1 pu 3E 7E 1 pu 5 E t
t L
H mq H
H mq H
o o
A 2
sT 5I E sT 2T sI E sT N
N STN I
Or T
r.
S
)
)
P 4
4 I
s ee ee R
p rt rt T
m D
uo D
uo D
e e
)
T a
ER esN ER DR esN ER l
l d
N N
TE ns(
TE EE ns(
TE b
b n
e O
I 0
AW i e AW TW i e AW a
a u
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c T
P P
rP n P
RP rP n P
i i
i A
T 0
f u
e f
u e
f l
l t
T E
1 oL Tel oL fL Tel oL p
p n
N o
S A
sa A
oA sa A
p p
E C
x
%M
%l v
% M M
%l v
% M A
A M
P U
0R 0 ui 0R
%R 0 ui 0R f
I 1
1 E 1 pu 3 E 8E 1 pu 5 E t
t R
R H
mq H
H mq H
o o
1 O2 T
T 2
sT sI E 5T 2T sI E 5T N
N SN I
2 t
t u
u E
M p
p E
L n
n T
B I
I S
A Y
T 0
3 S
1 1
P P
P I
RT 0
8 1
3 R
1 O
P P
9 T
n P
s C
o P
k r
A c
x x
E o
t s
u u
r R
u p
l l
e p
l e
i F
F b
i r
N r
m r
e T
n n
a kT t
e o
o h
c n
g r
r r
C oe I
s n
o t
t l n r
c a
t u
u e
Bi m
e i
R c
e e
s b
e k
g a
N N
l sr t
a o
e e
u pu s
t R
e e
p i T e
L y
r a
7 g
g m
r S
B p
i6 r-n n
I Tg i
p d -
eP a
a n
eP w
R R
e ri p
r i
i T
r m
o,
n ow r
r,
Pk r
r i
t o T
T T
c ex c
e e
b cl I
i N
r t u wo w
w r
al r
t nl ol o
o u
eo U
o I F LB P
P T
RF o
a t
t m
L c
c o
A a
a t
N e
e u
O R
A B
C D
E F
R A
I i
TCN U
0 1
2 F
2 2
2 Ee5e" 5.
5 a.
i
.. - ~ -..
l TABLE 3.2-1 y
DNB PARAMETERS' Y
j LIMITS l
- 1 PARAMETER 3 Loops in Operation 2 Locos in Operation g
Indicated Reactor Coolant System Tavg 5 580.3*F
(**)
l Indicated Pressurizer Pressure 2 2209 psig*
(**)
l Indicated Reactor Coolant System 2 264,200 gpm***
(**)
l i
Total Flow Rate i
i
?
i r
t t
i
=
N Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per j
z minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
?
t Values blank pending NRC approval of 2 loop operation.
- Value includes a 2.4% flow uncertainty (0.1% feedwater venturi fouling bias included).
i
O O
O TABLE 3.3-4
- g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
=
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 5"
1.
SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION a.
Manual Initiation Not Applicable Not Applicable b.
Automatic Actuation Logic Not Applicable Not Applicable c.
Containment Pressure--High s 4.0 psig 5 4.5 psig d.
Pressurizer Pressure--Low 2 1850 psig 2 1847 peig l
e.
Differential Pressure s 100 psi s 112 poi Between Steam Lines--High w1 w
f.
Steam Line Pressure--Low 2 585 psig 2 575 peig 5
5 il n
5
...m m.
m.
- m..
m
n - ~
., _. ~....~
-..~ ~.......=... ~ --.~.~... -. -.
. ~ ~.. -.
O
~
i TABLE 3.3-4 (Continued) t ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E
E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES Ic 5
4.
STEAM LINE ISOLATION i
4 s
l a.
Manual Not Applicable Not Applicable b.
Automatic Actuation Not Applicable Not Applicable Logic 4
c.
Containment Pressure--
s 16.2 psig s 17.5 peig
[
High-High j
d.
Steam Flow in Two Steam s A function defined 5 A function defined as
[
Lines--Figh, Coincidant as follows: A Ap follower A Ap with Tavg-- Low-Low corresponding to 40%
corresponding to 40.3% of l
I w
of full steam flow full steam flow between i
i between 0% and 20%
0% and 20% load and w
load and then a Ap then a Ap increasing O
increasing linearly linearly to a Ap l
to a Ap corresponding corresponding to 110.3%
l
- a 110% of full steam of full steam flow at flow at full load with full load with Tavg 2 543*F Tavg 2 542.6*F l
e.
Steam Line Pressure--Low 2 585 psig 2 575 peig t
i 5.
TURBINE TRIP AND FEED WATER ISOLATION a.
Steam Generator Water s 78.5% of narrow range s 78.9% of narrow range l
h Level--High-High instrument span each instrument span each y
steam generator steam generator i
5=
H
=
?
t
--,--nn--
O O
O TABLE 3.3-4 (Continued) l M
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS w
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
?g 6.
Automatic Actuation Logic N.A.
N.A.
b.
Steam Generator Water 2 25% of narrow range 2 24.6% of narrow range l
Level--Low-Low instrument span each instrument span each steam generator steam generator c.
Undervoltage - RCP 2 2680 volts 2 2640 volts d.
S.I.
See 1 above (All SI setpoints) e.
Trip of Nain Feedwater N.A.
N.A.
Pumps M.
7.
LOSS OF POWER a.
4.16 kV Emergency Bus 2 3255 volts bus voltage
- 2 3222 volts bus y
Undervoltage (Loss of voltage
- u*
Voltage) 5 3418 volts bus voltage
- b.
4.16 kV Emergency Bus 2 3675 volts bus voltage
- 2 3638 volts bus Undervoltage (Degraded voltage
- Voltage) s 3749 volts bus voltage
- 8.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS a.
Pressurizer Pressure, P-11 5 2000 peig 5 2003 psig l
[
b.
Low-Low Tavg, P-12 R
(Increasing) 545'F 5 545.4'F (Decreasing) 543*F 2 542.6*F Ng c.
Steam Generator Level, P-14 (See 5. above) d.
Reactor Trip, P-4 N.A.
N.A.
- Refer to appropriate relay setting sheet calibration requirements.
= _ - __
I EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4 By verifying the correct position of each mechanical position.stop e.
for the following ECCS throttle valves:
1.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.
2 i
2.
At least once per 18 months.
Valve Number CVC-V-8991 A/B/C
]
CVC-V-8989 A/B/C CVC-V-8996 A/B/C CVC-V-8994 A/B/C f.
At least once per 18 months, during shutdown, by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
2.
Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:
O(s/
a)
Centrifugal charging pump b)
Residual heat removal pump g.
By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to specification 4.0.5:
1.
Centrifugal charging pump 2 2323 paid l
2.
Residual heat removal pump 2 145 psid l
h.
Prior to entry into Mode 3 from Mode 4, verify that the mec?t:aical stops on low head safety injection valve RHR-HV 603 A/B are intact.
N FARLEY-UTIIT 1 3/4 5-5 AMENDMENT NO.
._._m PLANT fYSTEMS MAIN S?EAM LINE ISOLATION VALES
(
LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
I ACTION:
MODE 1 With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is j
restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 6 q
hours.
j' MODES 2 - With one main steam line isolation valve inoperable, and 3 subsequent operation in MODES 2 or 3 may proceed provided the isolation valve is restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 2; otherwise, be in HOT STAllDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 i
hours.
The provisions of Specification 3.0.4 are not applicable.
i SURVEILLANCE REQUIREMENTS i
l 4.7.1.5 Each main steam line isolation valve shall be demonstrated 4
OPERABLE by verifying full clasure within 7 seconds when tested pursuant to l
Specification 4.0.5.
J d
i l
i
!O FARLEY-UNIT 1 3/4 7-9 AMENDMENT NO.
1
POWER DISTRIBUTION LIMITS i
l BASES f-'s (v) 3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts.
The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on F g is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0.
For purposes of monitoring QUADRANT POWER T.ILT RATIO when one excore detector 1s inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-ll, H-3, H-13, L-5, L-11, and N-8.
3/4.2.5 DNB PARAMETERS i
The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial l
I '}
FSAR assumptions and have been analytically demonstrated adequate to meet the k'~'/
DNB design criterion throughout each analyzed transient. The indicated T value of 580.3*F is based on the average of two control board readings and 8n l
indication uncertainty of 2.9*F.
The indicated pressure value of 2209 psig is based on the average of two control board readings and an indication uncertainty of 24 psi.
The indicated total RCS flow rate is based on the average of two elbow tap measurements from each loop as read on the plant computer and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T and pressurizer pressure through the control av board readings are sufficient tS ensure that the parameters are restored within their limits following load changes and other expected transient operation.
The 18 month surveillance of the total RCS flow rate may be performed by one of two alternate methods. One method is a precision heat balance performed at the l
beginning of each fuel cycle. The other method is based on the op measurements from the cold leg elbow taps, which are correlated to past precision heat balance measurements.
Correlation of the flow indication channels with selected precision loop flow measurements is documented in WCAP-14750.
Use of the elbow l
tap Ap measurement method removes the requirement for performance of a precision RCS flow heat balance measurement for that cycle. The monthly surveillance of l
the total RCS flow rate is a reverification of the RCS flow requirement using process computer indications of loop elbow tap measurements that are correlated either to the precision RCS flow measurement or the elbow tap measurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators fed by elbow tap measurements.
x-FARLEY-UNIT 1 B 3/4 2-5 AMENDMENT NO.
CONTAINMENT SYSTEMS i
,_s f\\~ ')
BASES The maximum peak calculated containment internal pressure obtained from a LOCA event is 43 psig and from a MSLB event is 53 psig. These containment analyses calculations include an initial positive pressure of up to 3 psig.
The analyses results demonstrate that the maximum containment pressure will remain below the design limit of 54 psig.
3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial 1
temperature condition assumed in the accident analysis for a LOCA or steam line break accident.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is required to ensure that the containment will withstand the maximum peak calculated internal pressure of 43 psig in the event of a LOCA, including an initial positive pressure of up to 3 psig.
In addition, structural integrity is required to ensure that the containment will withstand the maximum peak calculated internal pressure of 53 psig in the
/)
event of a MSLB, including an initial positive pressure of up to 3 psig. As
(_ /
per Appendix J, Option B, the LOCA peak calculated containment internal pressure defines the Pa value for the Containment Leakage Rate Testing Program required by Surveillance 4.6.1.2.
The measurement of the containment lift off 3
force, visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the containment leakage tests are sufficient to demonstrate this capability.
The surveillance requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of paragraph C.l.3 of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structure," January 1976.
3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be closed in MODES above COLD SHUTDOWN since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system.
The use of the containment purge lines is restricted to the 8-inch vent supply and exhaust isolation valves to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss-of-g-w.
coolant accident during venting operations.
v)
F FARLEY-UNIT 1 B 3/4 6-2
m 3/4.7 PLANT SYSTEMS e-BASES 3/4.7.1 TURBINE CYCLE I
3/4.7.1.1 SAFETY VALVES I
The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient..The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is at least 12,984,660 lbs/hr whic!. is 105.8 percent of the total secondary steam flow of 12,270,000 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the the reduced reactor trip settings of the Power Range Neutron Flux channels.
The reactor trip setpoint reductions are consistent with the assumptions used
\\~
in the accident analysis.
i l
l
\\
FARLEY-UNIT 1 B 3/4 7-1 AMENDMENT NO.
DESIGN FEATURES l
5.3 REACTOR CORE (O
V FUEL ASSEMBLIES 5.3.1 The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy, zircaloy-4, or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.
Limited substitutions of zirconium alloy, zircaloy-4, ZIRLO*, or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
GONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 control rod assemblies. The control material shall be silver, indium and cadmium as approved by the NRC.
5.4 REACTOR COOLANT S1" STEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
/
a k- /
In accordance with the code requirements specified in Section 5.2 a.
of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 650*F, except for the pressurizer which is 680'F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9829 1 100 cubic feet at a nominal T of 567.2*F.
l avg 5.5 HETEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.5-1.
m (O
1 FARLEY-UNIT 1 5-6 AMENDMENT NO.
O s
TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS
=
h Reactor Coolant System 200.heatup cycles at s 100*F/hr Heatup cycle - Tavg from 5 200*F and 200 cooldown cycles at to 2 550*F.
w
< 100*F/hr.
Cooldown cycle - Tavg from l
2 550*F to s 200*F.
200 pressurizer cooldown cycles Pressurizer cooldown cycle at 5 200*F/hr.
temperatures from 2 650*F to 5 200*F.
80 loss of load cycles, without 2 15% of RATED THERMAL POWER immediate turbine or reactor trip.
to 0% of RATED THERMAL POWER.
i 40 cycles of loss of offsite Loss of offsite A.C.
A.C. electrical power.
electrical ESF Electrical
[
System.
80 cycles of loss of flow in one Loss of only one reactor i
reactor coolant loop.
coolant pump.
400 reactor trip cycles.
100% to 0% of RATED THERMAL f
POWER.
L i
10 inadvertent auxiliary spray Spray water temperature actuation cycles.
differential > 320*F.
1 50 leak tests.
Pressurized to 2 2485 psig.
5 hydrostatic pressure tests.
Pressurized to 2 3100 peig.
Secondary System I steam line break.
Break in a > 6 inch steam line.
I 4
10 hydrostatic pressure tests.
Pressurized to 2 1356 psig.
l z
?
1 i
ADMINISTRATIVE CONTROLS g\\
{
i 2.
WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset Control /
Fg Surveillance Technical Specification," February 1994 (W Proprietary).
(Methodology for Specifications 3.2.1 - Axial Flux Difference and 3.2.2 - Heat Flux Hot Channel Factor.)
$a 3a. WCAP-12945-P, Vol. I-IV, 1992-1993, " Code Qualification Document for Best Estimate LOCA Analysis."
(W Proprietary) 3b. Letter, N. J.
Liparulo, (W) to R. C. Jones, (USNRC), " Revision to Westinghouse Uncertainty Methodology," NTD-NRC-95-4575, Oct. 13, 1995.
3c. Letter, R. C. Jones (USNRC) to N. J. Liparulo (W), " Acceptance for Referencing of the Topical Report, WCAP-12945(P), Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis,"
j i
June 28, 1996.
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
I
)
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or
(_ /
supplements thereto, shall be provided upon issuance, for each relosd cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually.
This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.
/T
(
i
%J FARLEY-UNIT 1 6-19a AMENDMENT NO.
~...
. ~.
l ADMINISTRATIVE CONTROLS o
6.15 MAJOR CHANGES TO RADIOACTIVE WAS7;g TREATMENT SYSTEME (Liquid, Gaseous, Solid)
This specification deleted.
Refer to the Offsite Dose Calculation Manual and the Process Control Program.
6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of j
containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option i
B, as modified by approved exemptions. This program shall be in accordance
)
with the guidelines contained in Regulatory guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
I The peak calculated containment internal pressure for the design basis loss of
}
coolant accident, Pa, is 43 psig.
l 4
The maximum allowable containment leakage rate, La, at Pa, is 0.15% of containment air weight per day.
Leakage rate acceptance criteria are Containment overall leakage rate acceptance criterion is s 1.0 La.
a.
During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and C i
tests, and s 0.75 La for Type A tests; I
I b.
Air lock testing acceptance criteria are:
s 1)
Overall air lock leakage rate is s 0.05 La when tested at 2 Pa.
{
i l
2)
For each door, leakage rate is s 0.01 La when pressurized to 2 10 psig.
I The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
f The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.
i FARLEf-UNIT 1 6-24 AMENDMENT NO.
t J
O FNP Unit 2 Technical Specifications Power Uprate Implementation
~
Channed Panes 4
Umt 2 Revision l
J' Page 1-6 Replace l,
Page 2-5 Replace Page 2-6 Replace Page 2-7 Replace Page 3/4 2-15 Replace Page 3/4 3-25 Replace i
Page 3/4 3-27 Replace I
Page 3/4 3-28 Replace Page 3/4 4-12b Replace Page 3/4 5-5 Replace j
4 Page 3/4 7-9 Replace
]
Page B 3/4 2-5 Replace Page B 3/4 4-3b Replace Page B 3/4 6-2 Replace Page B 3/4 7-1 Replace Page 5-6 Replace Page 5-8 Replace Page 6-19a Replace Page 6-24 Replace I
O
]
i I
DEFINITIONS RATED THERMAL POWER.
i
\\
1.25 the ieactor coolant of 465e MWt. RATED THERMAL P f
2775 i
REACTOR TRIP SYSTEM RESPONSE TI
'4 1.26 the monitored parameter exceeds its trip setpoi loss of stationary gripper coil voltage.
i' REPORTABLE EVENT 1.27 50.73 to 10 CFR Part 50.A REPORTA8LE EVENT shall be any of those j
l SHUTDOWN MARGIN 1.28 4
the reactor is subcritical or would be subcritical fro I
i assuming all full length rod cluster assemblies (shutdown and contro j
inserted except for the single rod cluster assembly of highest reactiv
(
which is assumed to be fully withdrawn.
i O SOLIDIFICATION 1.29 This definition deleted.
Refer to the Process Control Program.
i i
SOURCE CHECK j
1.30 when the channel sensor is exposed to a radioactive sour j
i;.
STAccFRED TEST RASIS i
j 1.31 A STAGGERED TEST BASIS shall consist of:
I A test schedule for n systems, subsystems, trains or other designated 4
a.
components obtained by dividing the specified test interval into n i
equal subintervals, b.
i The testing of one system, subsystem, train or other designated i
component at the beginning of each subinterval.
1 j
FARLEY-UNIT 2 1-6 AMEEMENT NO. 49. 91 l
9 L
4
+ a +1b<5 n! me,
a i
i 4
i
=
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s E
2 l2 1
a a
+
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m E
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ng m.
g 9, m 3-E5 2 5
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j FARLET - INFIT 2 7-5 AMIElerr m. 79, 85 O
_m.
t i
i I
L 5
TABLE 2.2-1 iContinuedt j
P i
Q E'
REACTOR TRIP SYSTEM IRISTRIEREhrfATI(BI TRIP SETPQINTS y
FUNCTIONAL UNIT TRIP SETPOINT ALIAMhBLE VAIAIES 24.fo u
- 13. Steam Generator Water 2 25% of narrow range 2-34,44 of narrow range inst rument l
Level--Low-Low instrument span - each span - each steam generator l
- 14. Deleted i
- 15. Undervoltage - Reactor 2 2680 volts - each bus 2 2640 volts - each bus Coolant Pusps j
- 16. Underfrequency 2 57.0 Hz - each bus 2 56.9 Hs - each bus f
Pumps
- 17. Turbine Trip i
i A. Low Auto Stop 2 45 psig 2 43 psig g
e Pressure B. Turbine Stop Valve Not Applicable Not Applicable l
Closure i
- 18. Safety Injection Input Not Applicable Hot Applicable
[
trom ESF
[
- 19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip f
l 5
O
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8 I
_ _.. _ _ _ _ _ _ ~
O O
O S#
TABLE 2.2-1 (Continued)
Q g
REAC10R TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS a
N FINICTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES 20.
Reactor Trip System. Interlocks 0
~II I
A.
Intermediate Range Neutron 1 1 x 10 amps 6 x IO amps Flux, P-6 8.
Low Power Reactor Trips P-10 input
< IDE of RATED
< -HE of RATED Block, P-7 THERMAL POWER THERMAL POWER P-13 Input
< 10E Turbine
< 11% Turbine Impulse Pressure Tapuise Pressure Equivalent (Note 4 Equivalent (Note 5)
C.
Power Range Neutron Flux, P-8 of RATED of RATED m4 THERMAL POWER THERMAL POWER 7, /o D.
Power hange Neutron Flux, P-10
> 81 cf RATED
> -7E of RATED THERMAL POWER THERMAL POWER E.
Turbine Impulse Chamber, P-13
< 10E Turbine
< 11% Turbir:e Tapulse Pressure Tapulse Pressure Equivalent (Note 4)
Equivalent (Note 5)
F.
Reactor Trips Block
< 50% of RATED
-HK of RATED following Turbine Trip, P-9 THERMAL POWER THERMAL POWER 21.
Raactor Trip Breakers Not Appilcable Not Applicable f
- 22. Automatic Trip Logic Not Appilcable Not Applicable
~. - -
- _ _ _.. - ~.
3 o
O (o
m TABLE 3.2-1 r.;
DNB PARAMETERS 8
i LIMITS U
PARAMETER 3 Loops in Operatto 2 Loops in Operation I
5 50.3 Indicated Reactor Coolant System T f -589r?"F
(**)
l Indicated Pressurizer Pressure Z2D 2-N05-psig*
(**)
l Indicated Reactor Coolant System Total Flow Rate 1-257.S00 gpa***
(**)
l 264,20D W
I n
C; i
l Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per I
minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
o:x h
Values blank pending NRC approval of 2 loop operation.
r.
O
- Value includes a 2.4% flow uncertainty (0.1% feedvater venturi fouling bias included).
l "w
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l
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f L,
i
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O 7
4-8 S
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e L
l l
A b
b V
a a
g S
c c
g g
i T
E i i i
s i
i N
L l
l s
p s
s I
B p
p p
p p
O A
p p
9 P
W A
A 4
2 5
T O
5 9
1 7
E L
t t
4
}
1 5
S L
o o
A N
N 1 1 i 3
P I
R T
NO I
T A
T N
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T e
e S
T l
l N
N b
b I
I a
a g
O c
c g i g
4 N
P i
i i
s i
i E
T l
l s
p s
s 3
T E
p p
p p
p S
S p
p 0
3 Y
A A
S P
0 5
0 5
8 0
8 E
I t
t 4
1 1
5 L
N R
o o
B O
T N
N 1 1 i 1
A I
O T
T A
U T
C A
ERU T
D A
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A F
h P
c g
w h
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i i
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r-S N
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r us e
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u u
se r
D 8N n i s
s sn u
E RO o
t s
s ei s
R UI i
a e
e rL s
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u r
r P
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t P
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r N
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NO t
A t
r ae I
G OS i
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it e
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n i
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en L
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L TE A
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S a
b c
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1
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=5* N 1Y
O O
O TABLE 3.3-4 icontinuedi ENGINFFRED SAFETY FEATURE AC'IilATION SYSTEN INSTRIM M ATION TRIP SETPOINTS u
e FUNCTIONAL UNIT TRIP SETPOINT ALILEABLE VALUES Olj 4.
STEAM LINE ISOLATION u
a.
Manual Not Applicable Not applicable b.
Automatic Actuation Not Applicable Not Applicable Logic c.
Containment Pressure--
5 16.2 psie S 17.5 psig High-High d.
Steam Flow in Two S A function defined 5 A function defined fd.3 Steam Lines--High, as follows:
defined as follows:
Coincident with A Ap corresponding to A Ap corresponding to -444 Tavg--Low-Low 40% of full steam of full steam flow between flow between On and 04 and 20% load and then 20% load and then a Ap a Ap increasing linearly increasing linearly to to a Ap corresponding to a Ap corresponding to M of full steam flow at a
w 110% of full steam full load with T vg 2 W *F a
h flow at full load
//4.3 gjg g 3
with Tavg 2 54 3*F 4
e.
Steam Line Pressure--Low 2 585 psig 2 575 psig 5.
'IURBINE TRIP AND FEED WATER l
ISOMTION 7g, yg_
a.
Steam Generator Water S-99;M of narrow range S 96:54 of narrow range l
Level--High-High instrument span each instrument span each i
steam generator steam generator k
5 M
O O
O t
TABLE 3.3-4 floNT/NDED1 Et*,GINEERED SAEL mi dIiATI h a INSTRIM4ENTATION TRIP SETPOIt[l'S 9
FUNr"PINR UNIT TRIP SETPOINT ALI4MARLE VALUES 6.
AUXILIARY FEEDWATER-k H"
a.
Automatic Actuation N.A.
N.A.
Logic 24 lo u
b.
Steam Generator Water 2 25% of narrow range 2 -ibh-34 of narrow range l
Level--Low-Low instrument span each instrument Span each steam steam generator generator c.
Undervoltage - RCP 2 2680 volta 2 2640 volts d.
S.I.
See 1 above fall SI setpoints) e.
Trip of Nain Feedwater N.A.
N.A.
Pumps l
7.
LOSS OF POWER U
a.
4.16 kv Bnergency Bus 2 3255 volts bus 2 3222 volts bus voltage
- t.a Undervoltage (Loss of voltage
- 5 3418 volts bus voltage
- O Voltage) 93 b.
4.16 kV Bnergency Bus 2 3675 volts bus 2 3638 volts bus voltage
- Undervoltage (Degraded voltage
- 5 3749 volts bus voltage
- i Voltage) 8.
ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INTERLOCKS ggg)3 Ib.
a.
Pressurizer Pressure, P-11 5 2000 psig 5 4944 psig 66*I Low-LowTavf,P-12 (Increasing 544*F S-547'r (Decreasing) 543*F 2 3.ir46*P
,b c.
Steam Generator Level, P-14 (See 5. abovel OYE I" h
d.
Reactor Trip, P-4 N.A.
N.A.
Refer to appropriate relay setting sheet alibration requirements.
~ _ - - -
l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 10.
Preservice Inupection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish e baseline condition of the tubing.
This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11.
r* Distance is the distance of the expanded portion of a tube which provides a sufficient length of undegraded tube expansion to resist pullout of the tube fro
/50 tubesheet. The r* distance is equal to ibr&+ inches plus allowance for eddy current uncertainty measurement and is measured down from the top of the tube sheet or the bottom of the roll transition, whichever is lower in elevation. The allowance for eddy current uncertainty is documented in the steam generator eddy current inspection procedure.
12.
r* Tube is a tube:
a) with degradation equal to or greater than 40% below the r* distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal O
wall thickness within the F* distance, and c) that remains inservice.
13.
Tube Expansion is that portion of a tube which has been l
increased in diameter by a rollfn; process such that no crevice exists between the outside diameter of the tube and the hole in the tubesheet. Tube expansion also refers to that portion of a sleeve which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the sleeve and the parent steam generator tube.
14.
Tube Support Plate Repair Limit is used for the l
disposition of an alloy 600 steam generator tube for continued service that is experiencing predeminately axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair lindt is based on maintaining steam generator tube serviceability as described below:
EARLEY-UNIT 2 3/4 4-12b AMENDMENT NO.
l l
EMERGENCY C0ki COOLING SYSTEMS
(]
SURVEILLANCE REQUIREMENTS (Continued) kJ l
e.
i By verifying the correct position of each mechanical position stop for the following ECCS throttle valves:
1.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems
{
t are required to be OPERABLE.
2.
At least once per 18 months.
Valve Number l
CVC-V-8991 A/B/C CVC-V-8989 A/8/C CVC-V-8996 A/B/C CVC-V-8994 A/B/C f.
At least once per 18 months, during shutdown, by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.
2.
Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:
a)
Centrifugal charging pump b)
Residual heat removal pump g.
By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:
82 g) 1.
Centrifugal charging pump 1 2".
- i, 2.
Residual heat removal pump 113.pP h.
Prior to entry into Mode 3 from Mode 4, verify t e mechanical stops on low head safety injection valves RHR-HV 603 A/B are intact.
l l
O
,A LCV-eNI1 2 2/4 e-,
l l
j PLANT SYSTEMS i
MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION
}
3.7.1.5 Each main steam line isolation valve shall be OPERA 8LE.
{
APPLICABILITY: MODES 1, 2 and 3.
i i
ACTION:
M00h1-Withonemainsteamlineisolationvalveinoperable,POWEROPERATION f
may continue provided the inoperable valve is restored to OPERABLE
),
status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided the isolation valve is restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 2; otherwise, be in H0T STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in NOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS O
4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERA 8LE by verifying full closure within t seconcs when tested pursuant to Specification 4.0.5.
7 FARLEY-UNIT 2 3/4 7-9 Amendment No. 13
POWER DISTRIBUTION LIMITS f
l BASES j (
3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protecticn with x-y plane power tilts.
The two hour time >\\1owance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on Fg is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
The two sets of four symmetrie thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11 nd N-8.
l E'9 29 E Mb Ib'E Y i
3/4.2.5 DNB PARAMETERS The limits on the DNB related rameters assure that sa a of the parameters ar maintained within the normal a eady state envelope of oper ion assumed in t transient and accident analys s.
The limits are consistent th the initia gg, y assumptions and have be n analytically demonstrated adequ
.e to meet e
l DFB
~ n eriterion throug ut each analyzed transient. The in cated T value of 99er4#F is based o the average of two control board read 2n gs a indication uncertainty of The indicated pressure value of +995 ig is based on the average of two control board readings and an indication uncertain *y O' 20 pai.
The indicated total RCS flow rate is based onftwo elbow 24 t*P **asurements from each looptand an uncertaint of 2 low is 4
included for feedwater venturi to AS READ ON THE PL ANT C2H1PDTER The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T and pressu
.s p.e.
u e
n to ay$ ensure that the parameters are restored board readings are sufficient t within their limits load changes and other ex ted transient w
operation.
~
MEAUREMENTS /S HEM BALANCE The 18 month T1 e
total RCS flow ormed by one of two alternato methods. One method is a precision-seleri--tri performed at the beginning of each fuel ycle. The other method is based on the op measurements from the cold leg alb taps, which are correlated to past precision heat balance measurements.
Correla ion of the flow indication channels with selected precision loop flow --'^"'--' '--
documented in WCAP-14750.
Use of the elbow tap op measurement method removes the raquirement for performance of a precision RCS flow ;;1cria;;ric measurement for that cycle. The monthly surveillance of the to al RCS flow rate is a reverification of the RCS flow requirement using pro ess computer indications of loop elbow tap measurements that are correlated eithe to the precision RCS flow measurement or the elbow tap measurement at the b ginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative ve ification of significant flow degradation using the control board indicators f $ b elbow tap measurements.
HEAT BALANEE Revised by NRC letter FARLEY-UNIT 2 B 3/4 2-5 dated: December 11. 1996
REACTOR COOLANT SYSTEM RASES
{
Indicationoftubedegradationofanytypeincludingacomplete 5
2.
guillotine break in the tube between the bottom of the upper l
joint and the top of the lower roll expansion does not require that the tube be removed from service.
3.
The tube plugging limit continues to apply to the portion of i
the tube in the entire upper joint region and in the lower roll expansion. As noted above, the sleeve plugging limit applies to these areas also.
j 4.
The tube plugging limit continues to apply to that portion of the tube above the top of ths upper joint.
- b. Laser Welded 1.
Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed fram service.
i 3.
At the weld joint, degradation must be evaluated in both the sleeve and tube.
4.
In a joint with more than one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the limit of the sleeve inspection.
5.
The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.
F* tubes do not have to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded. The F* distance is equal to HEdb& inches plus allowance for eddy current uncertainty measurement and is shasured down from the top of the tubesheet or the bottom of the roll tram sition, whichever is lower in elevation.
/. h l) i Steam generator tube inspections of operating plants have demonstrated the capability to reliably detset wastage type degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to 10 CFR 50.73 prior to resumption of plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision to the Technical Specifications, if necessary.
FARLEY-UNIT 2 B 3/4 4-3b AMENDMENT NO.
AND ff0M A MSLB EDENT /S 53 PS/4Y thE ANAL YSES RESULTS l
DEAUNSTRATE THAT I
rifwENT SYRrEME BASES
[AS002A O k
j T
mamisman ned from a event is l
4e'peig a m an initial po ve pressure of up to 3 poi he maximus j
6 containment press wi1 Q ammin belo pw er u_ _
e.
.THESE [DNTA/NMENT ANALYSEG CALLULAT/dNS /NCLUDE 3/4.s.1.s ara 1
w m
4
}
The limitations on containment average air temperature ensure that the overall containment average air temperature does not esteed the initial
]
temperature condition assume pd analysis for a IDCA r to 1Lae kreak accident.
,ygjgg,yg pg,y,y,
,y, j
af4.s.i.s comrarnwE=T -
e intramrry PRESSufE OT OF T4 5 p5/4.
m-1 This limitation ensurne that the al integrity of the ainment i
will be maintained comparable to the or design standards for t life of j
the facility. Structural 4ntegrity is ired to ensure that the e tainment I
will withstand the maximunfpressure of Ae'peig in the event of a iThe visual====1== tion of *==da==, anchorages aud -,:::d interior and este or sur' aces of the containment, and the cone =4a-ant leakage tents, along vi h j
the data obtained from Unit 1 +=adan surveillance, are sufficient to j
demonstrate this capability.
t i
The surveillance requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of paragra C.1.3 of Regulatory Guide 1.35 "Inservise surveillance of Ungrouted Tendons in Prestressed concrete Containment Structures," January 1976.
3/4.6.1.7 CONTAIMMENT YENTIt.ATION SYSTEM l
The 48-inch containment purge supply and enhaast isolation valves are required to be closed in MODES above COLD MEUTDOWN since these valves have not l
beer. ownonstrated oapable of closing during a Zack or steam lime break j
accideat. Maintaining these valves closed during plant operations ensures j
that excessive quantities of radioactive materials will not be released via the containment purge system.
I j
The use of the con *=i====t purge lines is restricted to the 8-inch vent j
supply and==h===t isolation valves to ensure that the site boundary dose i
guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss-of-coolant accident during vent rations.
~
\\
/N ADDIT /oN, STRUCTURAL /NTE6K/TV /S REau/KED To ENSURE THAT THE i
CONTNNNIENT W/LL W/THSTAND THE A%UMDM PEAKCALLULATED /NTERNAL PRESSURE l
0F 53 P3/s /N THE EVENT OFA MSLB, /NLLOD/N4 AN/H/TML PDS/TNE PRESSDEE l
OF UP To 3 PS/6. AS PER APPEND /t7, OPT /DN B, THE LDCA Fik
{
CAL [ULATED &LWTA/NMEMT/NTERNAL PRESSURE del /'NE$ THE Pa WILUE FAR i
THE [DNTA/NMENT LEAKA&E RATE TES77N& FAD & RAM RE00/KED BY SURVE/LLANCE 4. fo. /. Z.
2 i
FARLEY-UNIT 2 3 3/4 6-2 Amendment No. 114 I
I
_. ~.,
3/4.? PLANT SYSTEMS BASES I
3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVE 5 The OPERA 3ILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 1106 psign of its cesign pressure of 1085 psig during the most severe 11194 anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERhAL POWER coincident with en assumes loss of condenser heat sink ti.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the A5MI lo11er ans Pressure Code, 1971 Edition.
7g 7* jg valves on all of the steam lines is at leastThe total relieving capacity for all 12,984,660 lbs/hr which is
-EHF percent of the total secondary steam flow of 11,013,;;; lbs/hr at 4
100% RATED THERMAL POWER.
A minimum of 2 CPERABLE safety lves p!'.
i steam generator ensures that sufficient relieving capacity i er vst for the allowable THERMAL POWER restriction in Table 3.7-2.
jg 7g,gg 3
STARTUP and/or POWER OPERATION is allowable with safety val inoperable within the liaAtations of the ACTION requiraments on the basis of the reduction in secondary systna steam flow and THERMAL POWER 4
required by the reduced reactor trip settings of the Power Range Neutron j
T1ux channels.
The reactor trip setpoint reductions are consistant with the assumptions used in the accident analysis.
4 1
O l
P rARLEY-UNIT 2 3 3/4 7-1 AMENDMINT NO. 103
((
_ _. ~. _ _ _ _ _.... _... _ _. _ _... _.. _...... _ _ _ _ _ _...... _, _. _ _ _ _,
DESIGN FEATURES _
5.3 REACTOR CORE FUEL ASSEMBLIES 3.3.1 The reactor shall contata 157 fuel assentlies.
initial campesities of natural er slightly enrichedconsist of a fuel material.
uel rods with aa uranium disside (Uot) as Limited substitutions of 21 sonium alley, stacaley-4 fuel red eenfigurations, may be used. steel filler gods for fuel reds, la Fuel assentlies shall be limited to theseeeve fuel designs that have been analysed with applicable NRC methods, and shown by tests er analysee to comply with all fu l
-approved codes sad bases.
A limited auster of lead test assentlies that have not safety design e
representative testiaq say be placed la nea-11mittag core tegio completed ns.
CONTROL ROD ASSEMBLIES S.3.2 material shall he silver, indium and cadmium as appro s.
The centrol S. 4 REACTOR COOLANT SYsians e NRC.
DESIGN PRESSURE AND TT.,GanAivas S.4.1. The roaster coelaat system is designed and shall be meistained In asserdasse with the modo requireamste specified la secti a.
et the tsar, with allowmase for asamm1 degradation pursuant to on S.2 the applicable surveillance Aequirements, b.
Per a pressure of 2485 psig, and Per a temperature of GSO'F, eacept for the pressuriser which i c.
VOLUME s 600*F.
S.4.2 9829 100 cubic feet at a a-m aatThe total water and steam welume of the remet 7,,, og w.
s m is 4345.t S.s xsfeOnormeAL 2_ A mas 5/o 7. 2 S.S.1 The meteorologieal tower shall be located as shoma en Figure S S 1 rAnLEY-UNIT 2 S-s AMuunMENT NO. 43,56,34, 101
o O
O t
I i
- BI TABLE 5.7-1 J,
COMPONENT CYCLIC OR TRANSIENT LIMITS 1
Ey CYCLIC OR t
COMPONENT DESIGN CYCLE TRANSIENT LIMIT m
OR TRANSIENT Reactor Coolant System 2
heatup cycles at i 100*F/hr i
and 00 cooldown cycles at Heatup cgcle - T,8 from 5 200*F l
to > 550 F.
< 100*F/hr.
CooTdown cycle - T ave from 1 550*F to 1 200*F.
200 pressurizer cooldown cycles at i 200*F/hr.
Pressurizer cooldown cycle temperatures from 1 650*F to
-< 200*F.
80 loss of load cycles, without
> 15% of RATED THERMAL POWER to r
immediate turbine or reactor trip.
D% of RATED THERMAL POWER.
on 40 cycles of loss of offsite toss of offsite A.C. electrical A.C. electrical power.
ESF Electrical System.
80 cycles of loss,of flow in one Loss of only one reactor reactor coolant loop.
coolant pump.
400 reactor trip cycles.
100% to 0% of RATED THERMAL POWER.
10 inadvertent auxiliary spray Spray water temperature differential actuation cycles.
> 320*F.
50 leak tests.
Pressurized to 1 2485 psig.
5 hydrostatic pressure tests.
Pressurized to 1 3100 psig.
Secondary System I steam line break.
Break in a > 6 inch steam line.
-fr-hydrostatic pressure tests.
Pressurized to 1 1356 psig.
/0
ADMINISTRATIVE CONTROLS 2.
WCAP-10216-P-A, Rev. lA, " Relaxation of Constant Axial offset control /
Fg Surveillance Technical Specification," February 1994 (M Froprietary).
(Methodology for specifications 3.2.1 - Axial Fluz Difference and
- 3.2.2 - Beat Flux Rot Channel Factor.)
w
/
m.nF-1G2 7,'.4.,
2, "The 1T01 '";.;i ; Cf " ;tinghee;; :celueti;r.
": f:1 ";ing * *2" "':f:, " ".=:t !?S? (M ";:.ricte._y).
j A
u (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
ANNUAL.DIEEEL GENEFATOR RELTaBILITY DATA REDOAT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for eagh diesel generator shall be submitted to the NRC annually.
This report shall cont 11n the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.
O l
I Insert A to pg 6.19a 3a.
WCAP-12945-P, Vol. I-K, 1992-1993, " Code Qualification Document for Best Estimate LOCA sis". @ Proprietary) 3b.
Letter, N. J. Lipamlo, (W) to R. C. Jones, (USNRC), " Revision to Westinghouse Uncertainty Methodology", NTD-NRC-95-4575, Oct.J5.1995.
3c.
Ietter, R. C. Jones (USNRC) to N. J. Lipuulo (W), " Acceptance or Referencing of the Topical Report, WCAP-12945(P), Westinghouse Code Quahfication Document for Best Estimate Loss-of-Coolant Analysis", June 28,1996.
O FARLEY-UNIT 2 6-19a AMENDMENT No.
j j
k
- ~ ~ _ -. ~.
- - ~. _. _
i 1
j l
1 1
i ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGE 5_TD_EADIDACTIYZ WASTE TREATMENT SYSTEMS (Liquid, Gaseous, 4
solid)
This specification deleted. Refer to the offsite Dose calculation Manual and the Process control Program.
1 6.1% pgNTAINMENT LEAEAGE RATE TESTING FROGRAM i
l A program shall be established to implement the leakage rate testing of l
containment.1 required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, option j
B, as modified by approved esemptiona. This program shall be in accordance with the guidelines contained in Regulatory guide 1.163, " Performance-Based' Containment Leak-Test am," dated september 1995.
The peak calculat at internal pressure for the design basis loss of coolant accident, M is.et paig.
j 4.3 The maximum allowable containment rate, La, at Pa, is 0.15% of containment air weight por day.
Leakage rate acceptance criteria ares Containment overall leakage rate acceptance critation is 5 1.0 La.'
a.
During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 La for the contrined Type B and C testa, and S 0.75 La for Type A testes b.
Air lock testing acceptance critairia are:
1) overall air lock leakage rate is s 0.05 La when tested at 2 Pa.
2)
For each door, leakage rate' is 5 0.01 La when pressurised to 2 10 psig.
The previsions cf specification 4.0.2 do not apply to the teet frequencies specified in the containment Leakage Rate Testing Program.
The provisions of specification 4.0.3 are applicahia.to the containment Leakage Rate Testing Program.
FARLEY-UNIT 2 6-24 AMENDMENT NO.jj4
4 r
I DEFINITIONS RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to I
the reactor coolant of 2775 MWt.
l i
REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
i REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
l SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
U'
/
SOLIDIFICATION 1.29 This definition deleted. Refer to the Process Control Program.
SOURCE CHECK 1
1.30 A SOURCE CHECK shall be the qualitative assessment of channel response l
when the channel sensor is exposed to a radioactive source.
STAGGERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist of a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
)
O FARLEY-UNIT 2 16 AHENDMENT NO.
l l
l
O
)
O(
\\
TABLE 2.2-1
8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Q
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES e
1.
Manual Reactor Trip Not Applicable Not Applicable
+1 M
2.
Power Range, Neutron Flux Low Setpoint - 5 25% of RATED Low Setpoint - s 25.4% of RATED l
THERMAL POWER THERMAL POWER High Setpoint - s 109% of RATED High setpoint - s 109.4% of RATED l
THERMAL POWER THERMAL POWER 3.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with 5 5.4% of RATED THERMAL POWER l
High Positive Rate a time constant 2 2 seconds with a time constant 2 2 seconde j
4.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 5.4% of RATED THERMAL POWER l
High Negative Rate a time constant 2 2 seconds with a time constant 2 2 seconds 5.
Intermediate Range, Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER Flux 4
6.
Source Range, Neutron Flux s 105 counts per necond 5 1.3 X 105 counts per second 7.
Overtemperature AT See Note 1 See Note 3 8.
Overpower AT See Note 2 See Note 6 9.
Pressurizer Pressure--Low 2 1865 psig 2 1862 psig l
10.
Pressurizer Pressure--High 5 2385 psig s 2388 psig l
5 11.
Pressurizer Water s 92% of instrument span s 92.4% of instrument span l
f Level--High E
- 12. Loss of Flow 2 90% of minimum measured flow 2 89.7% of minimum measured flow l
g g
per loop
- per loop
- 5
- Minimum measured flow is 88,100 gpm per loop.
l
--= -
~
O O
O L
I'8 TABLE 2.2-1 fContinued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g
e FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES H
- 13. Steam Generator Water 2 25% of narrow range instrument 2 24.6% of narrow range instrument l
M Level--Low-Low span - each steam generator span - each steam generator
- 14. Deleted
- 15. Undervoltage - Reactor 2 2680 volts - each bus 2 2640 volts - each bus Coolant Pumps
- 16. Underfrequency - Reactor 2 57.0 Hz - each bus 2 56.9 Hz - each bus Coolant Pumps 17.
Turbine Trip A.
Low Auto Stop 2 45 peig 2 43 psig 4
Pressure B.
Turbine Stop Valve Not Applicable Not Applicable i
Closure 18.
Safety Injection Input Not Applicable Not Applicable from ESF
- 19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip s
M 5
i E
O O
O TABLE 2.2-1 iContinued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS N
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES a
3 20.
Reactor Trip System Interlocks a
A.
Intermediate Range Neutron 2 1 x 10-10 amps 2 6 x 10-11 amps j
Flux, P-6 i
B.
Low Power Reactor Trips P-10 Input s 10% of RATED
$ 10.4% of RATED l
Block, P-7 THERMAL POWER THERMAL POWER P-13 Input s lot Turbine s lit Turbine Impulse Pressure Impulse Pressure
[
Equivalent (Note 4)
Equivalent (Note 5)
C.
Power Range Neutron Flux, P-8 5 30% of RATED s 30.4% of RATED l
f THERMAL POWER THERMAL POWER M
5 D.
Power Range Neutron Flux, P-10 2 8% of RATED 2 7.6% of RATED l
[
THERMAL POWER THERMAL POWKR E.
Turbine Impulse Chamber, P-13 s lot Turbine s lit Turbine Impulse Pressure Impulse Pressure Equivalent (Note 4)
Equivalent (Note 5)
F.
Reactor Trips Block 5 50% of RATED s 50.4% of RATED l
f Following Turbine Trip, P-9 THERMAL POWER THERMAL POWER 21.
Reactor Trip Breakers Not Applicable Not Applicable 22.
Automatic Trip Logic Not Applicable Not Applicable m
O E
5 5
+
1 I
TABLE 3.2-1
- v DNB PARAMETERS 2
=
b LIMITS 5
PARAMETER 3 Loops in Operation 2 Loops in Operation Indicated Reactor Coolant System Tavg 5 580.3'F
(**)
l Indicated Prc. set 'izer Pressure 2 2209 poig*
(**)
l f
Indicated Reacto Coolant System 2 264,200 gpm***
(**)
l Total F?.ow Rate I
I w
I l
i l
s E
5 8
Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per
[
minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
i Values blank pending NRC approval of 2 loop operation.
t
- Value includes a 2.4% flow uncertainty (0.1% feedwater venturi fouling bias included).
O TABLE 3.3-4
- e ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUnrav1ATION TRIP SETPOINTS
=
FUNCTIONAL UNIT TRIP SETPOINT ALLONABLB VALUES i
1.
SAFETY INJECTION, TURBINE TRIP AND FEEDNATER ISOLATION I
a.
Manual Initiation Not Applicable Not Applicable
[
^
r b.
Automatic Actuation Logic Not Applicable Not Applicable i
c.
Containment Pressure--High 5 4.0 psig s 4.5 psig d.
Pressurizer Pressure--Low 2 1850 peig 2 1847 pelg l
e.
Differential Pressure s 100 poi s 112 poi Between Steam Lines--High U
i f.
Steam Line Pressure--Low 2 585 psig 2 575 psig Y
l' i
f b
P i
E 5
5 x
+
f 5
i
TABLE 3.3-4 fContinued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOJJfI ALLOWABLE VALUES E
4.
STEAM LINE ISOLATION y
a.
Manual Not Applicable Not Applicable b.
Automatic Actuation Not Applicable Not Applicable Logic c.
Containment Pressure--
S 16.2 psig S 17.5 peig High-High d.
Steam Flow in Two Steam S A function defined S A function defined as Lines--High, Coincident as follows: A Ap follows: A Ap with Tavg-- Low-Low corresponding to 40%
correcponding to 40.3% of l
g of full steam flow full steam flow between a
between 0% and 20%
0% and 20% load and y
load and then a op then a Ap increasing w
increasing linearly linearly to a Ap to a Ap corresponding corresponding to 110.3%
l to 110% of full steam of full steam flow at flow at full load with full load with Tavg 2 543*F Tavg 2 542.6*F l
e.
Steam Line Pressure-Low 2 585 peig 2 575 psig 5.
TURBINE TRIP AND FEED WATER ISOLATION a.
Steam Generator Water S 78.5% of narrow range S 78.9% of narrow range l
Level--High-High inetrument span each instrument span each l
5 steam generator steam generator R
5 5
__________.._____mm,_
u m
TABLE 3.3-4 (Continued) l i
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONI.L UNIT TRIP SETPOINT BLLfEELE. VALUES m
i L.
6.
l 5
4 a.
Automatic Actuation Logic N.A.
N.A.
l u
1 b.
Steam Generator water 2 25% of narrow range 2 24.6% of narrow range l
}
Level--Low-Low instrument span each instrument span each steam generator steam generator i
i c.
Undervoltage - RCP 2 2680 volts 2 2640 volts I
d.
S.I.
See 1 above (All SI l
setpoints) l e.
Trip of Main Feedwater N.A.
N.A.
[
Pumps w
7.
LOSS OF POWER N
t b
I a.
4.16 kV Emergency Bus 2 3255 volts bus voltage
- 2 3222 volts bus j
s Undervoltage (Loss of voltage
- Voltage) 5 3418 volts bus i
voltage *
[
b.
4.16 kV Emergency Bus 2 3675 volts bus voltage
- 2 3638 volto bus Undervoltage (Degraded voltage
- Voltage)
$ 3749 volts bus voltage *
[
8.
ENGINEERED SAFETY FEATURE i
ACTUATION SYSTEM INTERLOCKS
[
t a.
Pressurizer Pressure, P-ll 5 2000 peig 5 2003 psig l
b.
Low-Low Tavg, P-12 f
(Increasing) 545'F 5 545.4*F f
g (Decreasing) 543*F 2 542.6'F c.
Steam Generator Level, P-14 (See 5.'above) e i
d.
Reactor Trip, P-4 N.A.
N.A.
- Refer to appropriate relay setting sheet calibration requirements.
[
r i
i
__---_---w______.__------,--_----._-----___.__--____--___.a_
--.----2
_,-----_s--___
u--wee
-~aw s-- w--v swno-r wa--www e n-m
I REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i 10.
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by addy current techniques prior to service to establish a baseline a
condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
11.
F* Distance is the distance of the expanded portion of a tube which provides a sufficient length of undegraded tube expansion to resist pullout of the tube from the tubesheet.
The F* distance is equal to 1.60 inches plus allowance for l
eddy current uncertainty measurement and is measured down from the top of the tube sheet or the bottom of the roll transition, whichever is lower in elevation. The allowance for eddy current uncertainty is documented in the steam generator eddy current inspection procedure.
]
12.
F* Tube is a tuber a) with degradation equal to or greater than 40% below the F* distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains inservice.
13.
Tube Expansion is that portion of a tube which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the hole in the tubesheet. Tube expansion also refers to that portion of a sleeve which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the sleeve and the parent steam generator tube.
14.
Tube Supoort Plate Reoair Limit is used for the disposition of an alloy 600 steam generator tube for 1
continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined 4
within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:
1 a
FARLEY-UNIT 2 3/4 4-12b AMENDMENT NO.
i EMERGENCY CORE COOLING SYSTEMS k
SURVEILI.ANCE REQUIREMENTS (Continued)
By verifying the correct position of each mechanical position stop e.
for the following ECCS throttle valves:
1.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS i
subsystems are required to be OPERABLE.
i 2.
At least once per 18 months.
Valve Number CVC-V-8991 A/B/C 1
CVC-V-8989 A/B/C CVC-V-8996 A/B/C
}
CVC-V-8994 A/B/C f.
At least once per 18 months, during shutdown, by:
j
^.
1.
Verifying that each automatic valve in the flow path actuates to its correct position on a safety injectioa test signal.
O 2.
Verifying that each of the following pumps start automatically upon receipt of a safety injection test signals a)
Centrifugal charging pump b)
Residual heat removal pump g.
By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to specification 4.0.5:
1.
Centrifugal charging pump 2 2323 paid l
2.
Residual heat removal pump 2 145 psid l
h.
Prior to entry into Mode 3 from Mode 4, verify that the mechanical stops on low head safety injection valve RHR-HV 603 A/B are intact.
FARLEY-UNIT 2 3/4 5-5 AMENDMENT NO.
PLANT SYSTEMS k
MAIN STEAM LINE ISOLATION VALES LIMITING CONDITION FOR OPERATION i
3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
MODE 1 With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 2 With one main steam line isolation valve incperable, and 3 subsequent operation in MODES 2 or 3 may proceed provided the isolation valve is restored to OPERABLE status or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 2; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The provisions of Specification 3.0.4 are not applicable.
)
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 7 seconds when tested pursuant to l
Specification 4.0.5.
FARLEY-UNIT 2 3/4 7-9 AMENDMENT NO.
29'#ER DISTRIBUTION LIMITS t
BASES 3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts.
The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the maximum allowed power by 3 percent for each preent of tilt in excess of 1.0.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-ll, and N-8.
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to meet the DNB design criterion.throughout each analyzed transient. The indicated T valueof580.3*Fisbasedontheaverageoftwocontrolboardreadingsand"En indication uncertainty of 2.9*F.
The indicated pressure value of-2209 psig is based on the average of two control board readings and an indication uncertainty of 24 psi.
The indicated total RCS flow rate is based on the average of two elbow tap measurements from each loop as read on the plant computer and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).
l The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T boardreadingsaresufficientt$andpressurizerpressurethroughthecontrol av ensure that the parameters are restored i
within their limits following load changes and other expected transient I
operation.
The 18 month surveillance of the total RCS flow rate may be performed by one of two alternate methods. One method is a precision heat balance performed at the l
beginning of each fuel cycle. The other method is based on the op measurements from the cold leg elbow taps, which are correlated to past precision heat balance measurements. Correlation of the flow indication channels with selected p;ecision loop flow measurements is documented in WCAP-14750. Use of the elbow l
tap Ap measurement method removes the requirement for performance of a precision RCS flow heat balance measurement for that cycle. The monthly surveillance of l
the total RCS flow rate is a reverification of the RCS flow requirement using procons computer indications of loop elbow tap measurements that are correlated either to the precision RCS flow measurement or the elbow tap measurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators i
fed by elbow tap measurements.
FARLEY-UNIT 2 8 3/4 2-5 AMENDMENT NO.
..._.-_._._m.m._-
i 1
J i
Indication of tube degradation of any type including a complete f
guillotine break in the tube between the bottom of the upper joint
{
and the top of the-lower roll expansion does not require that the j
tube be removed from service.
3.
The tube plugging limit continues to apply to the portion of the i
tube in the entire upper joint region and,in the lower roll
[
expansion. As noted above, the sleeve plugging limit applies to j
these areas also.
4.
The tube plugging limit continues to apply to that portion of the
{
tube above the top of the upper joint.
i l
- b. Laser Welded 1.
Indications of degradation in the length of the sleeve between the j
weld joints must be evaluated against the sleeve plugging limit.
2.
Indication of tube degradation of any type including a complete i
break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed from service.
i
{
3.
At'the weld joint, degradation must be evaluated in both the l
sleeve and tube.
4.
In a joint with more than one weld, the weld closest to the end of i
the sleeve represents the joint to be inspected and the limit of j
i the sleeve inspection.
1 j
5.
The tube plugging limit continues to apply to the portion of the
}
tube above the upper weld joint and below the lower weld joint.
F* tubes do not have to be plugged or repaired provided the remainder of the
)
tube within the tubesheet that is'above the F* distance is not degraded. The j
F* distance is equal to 1.60 inches plus allowance for eddy current l
j
}
uncertainty measurement and_is measured down from the top of the tubesheet or j
the bottom of the roll transition, whichever is lower in elevation.
l j
Steam generator tube inspections of operating plants have demonstrated the j
j capability to reliably detect wastage type degradation that has renetrated 20%
{
of the original tube wall thickness.
t l
Whenever the results of any steam generator tubing inservice inspection fall into category c-3, these results will be reported to the commission pursuant i
to 10 CFR 50.73_ prior to resumption of plant operation, such cases will be considered by the commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional addy-current inspection, and revision to the Technical Specifications, if l
necessary.
1 4
l-k 1
FARLEY-UNIT 2 B 3/4 4-3b AMENDMENT NO.
1 f
I j
_ _ _. _. _ _ _m
___m.
E t
A f
4 CONTAINMENT SYSTEMS l
BASES i
The maximum peak calculated containment internal pressure obtained from j
a LOCA event is 43 peig and from a MSLB event is 53 peig. These containment analyses calculations include an initial positive pressure of up to 3'psig.
The analyses results demonstrate that the maximum containment pressure will remain below the design limit of 54 psig.
i 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the j
overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA or steam
}
line break accident.
f I
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY 4
l S
This limitation ensures that the structural integrity of the containment i
will be maintained comparable to the original design standards for the life of l
the facility.
Structural integrity is required to ensure that the containment will withstand the maximum peak calculated internal pressure of 43 psig in the 1
event of a LOCA, including an initial positive pressure of up to 3 psig.
In j
addition, structural intsgrity is required to ensure that the containment will l
withstand the maximum peak calculated internal pressure of 53 psig in the event of a MSLB, including an initial positive pressure of up to 3 psig. As
[
per Appendix J, Option B, the LOCA peak calculated containment internal i
pressure defines the Pa value for the containment Leakage Rate Testing Program 1
required by Surveillance 4.6.1.2.
The visual examination of tendons,.
l anchorages and exposed interior and exterior surfaces of the containment, and 1
the containment leakage tests along with the data obtained from Unit 1 tendon surveillance, are sufficient to demonstrate this capability.
4 I
j The surveillance requirements for demonstrating the containment's j
structural integrity are in compliance with the recommendations of paragraph
{
C.1.3 of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed. concrete containment Structure," January 1976.
4 I
l 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM 4
The 48-inch containment purge supply and exhaust isolation valves are required to be closeo in MODES above COLD SHUTDOWN since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive meterials will not be released via I
the containment purge system.
l d
{
The use of the containment purge lines is restricted to the 8-inch vent l
supply and. exhaust isolation valves to ensure that the site boundary dose
]
guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss-of-4 coolant accident during venting operations.
)
i 4
FARLEY-UNIT 2 B 3/4 6-2 4
l 3/4.7 PLANT SYSTEMS BASES f
(
i 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 peig) of its design pressure of 1085 poig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 1004 RATED THERMAL POWER coincident with an assumod loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in I
accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is at least 12,984,660 lbs/hr which is 105.8 percent of the total secondary steam flow of 12,270,000 lbs/hr at 1001 RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that 1
j sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of l
the reduction in secondary system steam flow and THERMAL POWER required by the the reduced reactar trip settings of the Power Range Neutron Flux channels.
O,'
The reactor tri: tutpoint reductions are consistent with the assumptions used in the acciden* *ialysis.
0 4
l 4
1 FARLEY-UNIT 2 B 3/4 7-1 AMENDMENT NO.
DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy, zircaloy-4, or ZIRLO" fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) is fuel material.
Limitad substitutions of zirconium alloy, sircaloy-4, ZIRLO*, or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 control rod assemblies. The control material shall be silver, indium and cadmium as approved by the NRC.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
In accordance with the code requirements specified in Section 5.2 a.
of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 650*F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9829 1 100 cubic feet at a nominal T of 567.2*F.
l avg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.5-1.
FARLEY-UNIT 2 5-6 AMENDMENT NO.
O TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS M
COMPONENT CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT
=
Reactor Coolant System 200 heatup cycles at f 100'F/hr Heatup cycle - Tavg from s 200'F 5
and 200 cooldown cycles at to 2 550'F.
< 100'F/hr.
Cooldown cycle - Tavg from 2 550*F to S 200*F.
200 pressurizer cooldown cycles Pressurizer cooldown cycle at 5 200'F/hr.
temperatures from 2 650'F to S 200'F.'
I 80 loss of load cycles, without 2 15% of RATED THERMAL POWER immediate turbine or reactor trip.
to 0% of RATED THERMAL POWER.
40 cycles of loss of offsite Loss of offsite A.C.
A.C. electrical power.
electrical ESF Electrical System.
m t
I 80 cycles of loss of flow in one Loss of only one reactor reactor coolant loop.
coolant pump.
400 reactor trip cycles.
100% to 0% of RATED THERMAL POWER.
10 inadvertent auxiliary spray Spray water temperature actuation cycles.
differential > 320'F.
i 50 leak tests.
Pressurized to 2 2485 poig.
I 5 hydrostatic pressure tests.
Pressurized to 2 3100 peig.
5x secondary System 1 steam line break.
Break in a > 6 inch steam E
line.
e 5
10 hydrostatic pressure tests.
Pressurized to 2 1356 psig.
l i
t v
-~.
l a
j ADMINISTRATIVE CONTROLS 2.
WCAP-10216-P-A, Rev. IA, " Relaxation of Constant Axial Offset Control /
Fg Surveillance Technical Specification," February 1994 (W Proprietary).
a (Methodology for Specifications 3.2.1 - Axial Flux Difference and 3.2.2 - Heat Flux Hot Channel Factor.)
4 j
3a. WCAP-12945-P, Vol. I-IV, 1992-1993, " Code Qualification Document for Best Estimate LOCA Analysis."
(W Proprietary) i 3b. Letter, N. J. Liparulo, (W) to R. C. Jones, (USNRC), " Revision to Westinghouse Uncertainty Methodology," NTD-NRC-95-4575, Oct. 13, 1995.
3c. Letter, R. C. Jones (USNRC) to N. J. Liparulo (W), " Acceptance for Referencing of the Topical Report, WCAP-12945(P), Westinghouse Code Qualification Document for Best Estimate Loss-of- & lant Analysis,"
3 June 28, 1996.
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be datermined so that all applicable limits 4
(e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident 1
analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or
[
supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and I
Resident Inspector.
l ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT 1
}
6.9.1.12 The number of tests (valid or invalid) and the number of failures I
to start on demand for each diesel generator shall be submitted to the NRC
}
annually. This report shall contain the information identified in Regula'.ory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.
)
O FARLEY-UNIT 2 6-19a AMENDMENT NO.
ADMINISTRATIVE CONTROLS O
6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEEE (Liquid, Gaseous, Solid)
This specification deleted. Refer to the Offsite Dose Calculation Manual and the Process Control Program.
6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 43 psig.
l The maximum allowable containment leakage rate, La, at Pa, is 0.15% of containment air weight per day.
Leakage rate acceptance criteria ares Containment overall leakage rate acceptance criterion is s 1.0 La.
a.
During plant startup following testing in accordance with this program, the C
leakage rate acceptance criteria are s 0.60 La for the combined Type B and C tests, and 5 0.75 La for Type A tests; b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is 5 0.05 La when tested at 2 Pa-j 2)
For each door, leakage rate is 5 0.01 La when pressurized to j
2 10 psig.
The provisions of Specification 4.0.2 do not apply to the test frequern s s specified in the containment Leakage Rate Testing Program.
The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program, i
i O
FARLEY-UNIT 2 6-24 AMENDMENT NO.
E1hae amu 3. ed 4
.A.hea-4 4 h.J--AA-p4 4 84p4.-*
.h.Jbl,4.$
4.,,4.nhMee-4.h4 J Lhh w Wh e-14 6
-Jh'.-41-$-A4
'"clar M4 h e + A 4-4 etothe
-eme weAa4 4 mm-4 J+m4Ch-4=aMe.J a u A ad..
t a
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I i
l i.
4 4
(
l ATTACHMENT IV 4
3 FARLEY NUCLEAR PLANT TECHNICAL SPECIFICATIONS CHANGE REQUEST POWER UPRATE 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION O
O
t f
s Joseph M. Farley Nuclear Plant Units 1 and 2
[
Power Uprate SIGNIFICANT HAZARDS EVALUATION 2
a l
INTRODUCTION & BACKGROUND The Farley Power Uprate Project has been structured causirtent with the methodology established in WCAP-10263, "A Review Plan for Upratmg the Liomsed Power of a PWR Power Plant," dated 1983.
Since its submittal to the NRC, the methodology has been used successfully as a basis for power uprate projects on over twenty pressurized water reactor (PWR) units.
The methodology in WCAP-10263 established the ground rules and criteria for power uprate pam,
including the broad categories that must be addressed, such as NSSS performance parameters, dugn transients, systems, components, accidents and nuclear fuel, as well as the interfaces between the NSSS and the Balance of Plant (BOP) fluid systems Inherent in this methodology are key points that promote correctness, consistency, and licensability. The key points inchule the use of well-defined analysis input assumptions / parameter values, use of currently approved analytical techmques (e.g.,==aAadalogies and computer codes), and use of currently applicable licensing criteria and standards Southem Nuclear Operating Company (SNC) has completed a comprehensive engineering review program for Farley Nuclear Plant (FNP) to increase licensed power from 2652 MWt to 2775 MWt that included a reanalysis or evaluation of all LOCA, non-LOCA, mass and energy re!use, and dose analyses. All acceptance criteria continue to be met. All major NSSS cu...pcs.ci.ts (e.g., Reactor Vessel, Pressurizer, RCPs, Steam Generators, etc.) have been assessed with respect to hanadiag conditions expected for power operation at the uprated power. In all cases operation has been found acceptable. Major systems and sub-systems (e.g., safety injection, auxiliary feedwater, RHR, turbine generator, etc.) have been reviewed and acceptable performance has been verified for their normal operation and, as applicable, for their safety-related functions. All reactor trip and ESF actuation setpomts have been es a==I. and the proposed
{
setpoint modifications will assure adequate protection is afforded for all design bases events. Control j
systems including rod control, pressurizer level and pressure control, and steam generator level control have been evaluated and been found acceptable for operation at the uprated conditions. In addition, a range ofTm (567.2'F to 577.2'F) has been included in the uprate program to afford FNP flexibility in plant operation and fuel design.
To meet the criteria of 10 CFR 50.46 for LBLOCA, best estimate analyses were performed using the i
approved Westinghouse Best Estimate model W COBRA / FRAC. The SBLOCA continues to be analyzed using the current methods The technical bases for these proposed changes to the FNP Units I and 2 Operating Licenses (NPF-2 and NPF-8) and Technical Specifications are contained in the Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report, WCAP 14723, arid BOP Licensmg Report.
i i
l i i
PROPOSED CHANGES l
The proposed amendment will change the maximum reactor core power level for facility operation from I
2652 MWt to 2775 MWt in the Facility Operating Licences (FOL). In addition, the proposed amendment involves the following Technical Specifications (T/S) changes. The defined rated thermal power for Farley will be changed to 2775 MWt. The DNB parameters for RCS T Pressurizer Pressure, and RCS flow i
have been modified. The revised RCS flowrate is 264,200 gpm. These changes are consistent with FNP's assumed thermal design flowrate (258,000 gpm) and FNP specific RTS/ESFAS procedures and 1
mstrumentation.
l Reactor trip system interlock setpoint for power range neutron flux (P-8) and ESF actuation trip setpoint l
l for steam generator water high-high level for turbine trip and feedwater isolation (P-14), and ESF actuation J
system interlock for low-low Tm(P-12) (increasing) have been modified to reflect analytical results.
j An evaluation of additional teactor trip system and ESF actuation system safety analysis limits and trip setpoints results in changes to ti.c a!!vwable values for several functions. These functions will be discussed amilisted later in this document.
The comprehensive power uprate reanalyses and evaluations have resulted in improved operational i
margins. Based on the results ofnew contamment analyses, the maximum peak calculated contamment 4
internal pressure for a LOCA event will be revised from 48 to 43 psig. lne main steamline isolation valve 1
closure time requirement will be revised from 5 to 7 seconds This will aid in the functional testing of the i I i \\
valve. To account for possible Centrifugal Charging Pump (CCP) or Residual Heat Removal (RHR) pump degradation, reduced ECCS flows have been found acceptable. The correspondmg surveillance 1
requirement will be modified to reflect these changes. Also, an increase from 5 to 10 in the number of i
secondary systems hydrostatic tests (Table 5.7-1) has been analytically verified to be acceptable. For Unit 2, the steam generator F* distance will be revised.
Changes to the plant design features and administrative controls are also included. These changes are a i
revision to the RCS fluid volume contamed in Section 5.4 and the addition of the NRC approved references for best estimate LOCA (BELOCA) listed in Section 6.9.1.
l i{
The following 4 groups of proposed changes will each be evaluated with respect to the criteria of 10 CFR 50.92. Listed below are the groupings.
I.
DEFINITION, DESIGN FEATURE & ADMINISTRATIVE CONTROL CHANGES l
a.
Section 1.25 Rated Thermal Power Definition b.
Section 5.4 RCS Fluid Volume c.
Section 6.9.1.11 COLR References i
11.
DNB PARAMETERS j '
Table 3.2-1 &
DNB Parameters Bases 3/4.2.5 i
w a
pwrupl4. doc 2
RMMOE 2/13S7 i
f
,, _ ~..
i 4
i i
3
)
i i
III.
MISCELLANEOUS OPERATIONAL AND MARGIN ENHANCEMENT CHANGES i
a.
Section 3/4.5.2 ECCS Subsystem - T, > 350'F o.
Bases 3/4.6.1.4 Containment Systems & Leakage Rate Testing Program Bases 3/4.6.1.6 &
}
Section 6.16 f
c.
Bases 3/4.7.1.1 Main Steam Safety Valve (MSSV) i d.
Section 3/4.7.1.5 Main Steam Isolaton Valve (MSIV)
I e.
Table 5.7-1 Component Cyclic or Transient Limits f.
Section 3/4.4.6 &
Steam Gmerators (Unit 2 only)
Bases 3/4.4.6 l
i IV.
ALLOWABLE VALUES AND TRIP SETPOINTS FOR REACTOR TRIP SYSTEM AND ESFAS
}
a.
Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints i
b.
Table 3.3-4 Engineered Safety Feature Actuation System Instrumentation Trip l
SetPoints i
}
10 CFR 50.92 EVALUATIONS I.
DEFINITION, DESIGN FEATURE & ADMINISTRATIVE CONTROL CHANGES a.
Section 1.25 Defmition of reactor core rated thermal power is changed from 2652 to 2775 MWt.
4 t
b.
Section 5.4 Revise RCS fluid volume from 9723 to 9829 ft' at a T, = 567.2*F.
l c.
Section 6.9.1.11 BELOCA methodology is added to COLR references i
l De change to the reactor core rated thermal power definition from 2652 MWt to 2775 MWt is based on the FNP uprated power value. His change is also applicable to the mammum power level defined in FOL Section 2.C(l). The change in RCS fluid volume contamed in the design feature section of the Technical Specifications reflects the current calculated total water and steam volume at 567.2'F for the RCS. Dese revised values are included i the supporting uprating evaluations and analyses; therefore, these changes are acceptable. In i
addition, the COLR references will include the NRC approved Westinghouse BELOCA methodology which was used for the uprated FNP LBLOCA analysis.
}
Based on the information presented above and the analyses and evaluations performed for the proposed power upratmg, the following conclusions can be reached with respect to 10 CFR 50.92.
4 1.
The proposed changes to the rated thermal power definition, RCS fluid volume, and COLR references do not increase the probability or consequences of an accident previously evaluated in the FSAR. He comprehensive analytical efforts performed to support the proposed upratmg included a review and evaluation of all components and systems (including interface systems and control systems) that could be affected by this change. The revised power uprate i
value and RCS fluid volume were inputs to applicable safety analyses. All systems will
_ function as designed, and all performance requirements for these systems have been evaluated
(
and found acceptable. None of these proposed changes directly initiate any accident; therefore, the probability of an accident has not increased All dose consequences have been analyzed or j
evaluated with respect to these parameters, and all acceptance criteria continue to be met.
l Therefore, the consequences of an accident previously evaluated in the FSAR have not increased.
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The proposed changes do not create the possibility of a new or diffen:nt kind of accident than j
any accident already evaluated in the FSAR. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed changes. 'Ihe proposed technical specificaten changes have no adverse effects on any safety-related system and do not i
challenge the performmce or integrity of any safety-related system. Therefore, the possibility of a new or different kind of accident is not created C
2
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3.
The proposed operating license and technical specification changes do not involve a significant reduction in a margin of safety. All analyses supportmg the proposed power uprate reflect the
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RCS fluid volume and rated thermal power values. The use of NRC approved BELOCA
=*hadalogy must be referenced since BELOCA will now be the LBLOCA analysis licensing i
basis for FNP. All acceptance criteria (including LOCA peak clad temperature, DNB criteria, l
contamment temperature and pressure, and dose limits) continue to be met. 'Iherefore, the proposed changes do not involve a significant reduction in the margin of safety.
1 I
Based on the above information and on the analyses performed to support the proposed uprating, these i
proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
11.
DNB PARAMETERS CHANGES t
a.
Table 3.2-1 &
DNB parameters Bases 3/4.2.5 Indicated RCS T.,
580.3'F Indicated Pressurizer Pressure 2209 psig Indicated RCS Total Flow Rate 264,200 gpm "Ihese values reflect the input values for certain non-LOCA accident analyses. All accident analysis acceptance criteria (including DNBR, containment response for pressure and temperature, and dose consequences) have i
been met, and therefore, these values are acceptable. The appropriate Bases have been revised to reflect these proposed changes based on the FNP specific calculation assumptions /results.
The DNB parameters (Indicated RCS T.,, Indicated Pressurizer Pressure and Indicated RCS Total Flow Rate) 1 are calculated based on the most limiting analytical values plus indication uncertainties. All values reflect FNP specific analytical values used to verify the adequacy of FNP safe operation at the proposed uprated power level conditions. In addition, the instrument uncertainty associated with FNP specific calibration practices and equipment have been included in the calculation of the final indicated values. The flow measurement uncertainty (2.4% in footnote) also includes 0.1% uncertainty for possible venturi fouling. WCAP-12771, Rev.
1, " Westinghouse Revised Thermal Design Procedure instrument Uncertainty Methodology for Alabama Power Farley Nuclear Plant Units I and 2 (Uprating to 2785 MWt NSSS Power)," includes the resised calculations for uprate.
Based on the information presented above and the analyses and evaluations performed to support the proposed uprating, the following conclusions can be reached with respect to 10 CFR 50.92.
peptuoc 4
IUMMGE 2/13S7
i 1.
The proposed technical specification changes for DNB pas.T.cters do not involve a significant l
increase in the probability or consequences of an accident previously evaluated in the FNP FSAR. The mechanical design features associated wM VANTAGE 5 fuel and the improved l
A~lalogie: (such as Revised 'Ihermal Design Procedure) provide capability for relaxation ofanalgical input parameters such that increased DNBR margin can be generated without violation of any acceptance criteria. The indicated DNB puameters bound the analytical values used to support the proposed uprating. In each case, the appropriate design and acceptance criteria are met. All performance requirements for my system or component have been evaluated and support the revised analysis assumptions. Overall plant integrity is not reduced. Furthermore, the parameter changes are associated with features used.as limits or mitigators to assumed accident scenanos and r,re not accident initiators lherefore, the probability of an accident has not significantly increased "Ihe radiological consequences of accidents previously evaluated in the FSAR have been assessed due to the proposed technical specification changes. Evaluations have confirmed that the doses remain within previously approved acceptable limits as well as those defmed by 10 CFR 100. Therefore, the radiological consequences to the public resulting from any accident previously evaluated in the FSAR has not significantly increased.
2.
The proposed technical specification changes do not create the possibility of a new or different kind of accident from any previously evaluated in the FS AR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the revised DNB parameters The revised analytical assumptions have no adverse effect and do not challenge the performance of any other safety-related system This has been verified in WCAP 12771, Rev.1. Therefore, the possibility of a new or different kind of accident is at created 3.
The proposed technical specification changes do not involve a significant reduction in the margin of safety. The margin of safety for fuel-related parameters (such as DNB and Kw/ft) are defined in the Bases to the Technical Specifications. The uncertainties associated with the proposed DNB parameter changes are included in the core safety limits. Performance of analyses and evaluations with the reactor core safety limits defined by RTDP have confkswd that the operatmg envelope defined by the Technical Specifications continues to be bounded by the revised analytical basis, which in no case exceeds the acceptance limits. Therefore, the margin of safety provided by the analyses in accordance with these acceptance limits is not reduced.
Based upon the preceding information, it has been determined that the proposed changes to the DNB parameters do not involve a significant hazards consideration as defined in 10 CFR 50.92.
III.
MISCELLANEOUS OPERATION AND MARGIN ENHANCEMENT CHANGES a.
Section 3/4.5.2 For the ECCS subsystem with T, > 350'F, modifications are proposed to surveillance requirements for the recirculation flow differential pressures for CCP (change 2458 psig to 2323 psid) and i
RHR pump (change 136 psig to 145 psid). FNP specific pump performance and margins, including allowances for possible pump degradation, have resulted in revised ECCS performance flowrates l
1hese flowrates have been verified to be acceptable from pump, system, and LOCA and non-LOCA analysis perspectives.
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i RU b.
Bases 3/4.6.1.4 For the Contamment Systems internal pressure and structural Bases 3/4.6.1.6 &
integrity Bases, revise the maximurn peak calculated contamment Section 6.16 internal pressure for a LOCA event from 48 to 43 psig and add the peak calculated contamment pressure for the MSLB event of 53 psig.
Clarify that the analyses results are based on an initial pressure of 3 i
psig and that the LOCA results defme the P. value for the Contamment I >=hy Rate Testing Program. These revisions are supported by the contamment analyses performed for uprate.
L c.
Bases 3/4.7.1.1 For the MSSV technical specifice. tion Bases, revise percentage of steam flow at 100% RTP from 112 to 105.8% and total waaary steam flow at 100% RTP from 11,613,849 to 12,270,000 lbs/hr. The increase in primary side power as a result of the uprating results in increases in total steam flow. The revised Bases reflect the new plant
]
full power steam flow values resulting from the uprating changes.
d.
Section 3/4.7.1.'
The MSIV closure time requirement is revised from 5 to 7 seconds i
h analyses (such as main steamline break and mass and energy releases) that model main steamline isolation have verified that this i
increase in stroke time is acceptable. This operational enhancement will provide margin to MSIV test acceptance criteria.
e.
Table 5.7-1
'lhe component cyclic or transient limits for secondary system hydrostatic pressure is changed from 5 to 10. The evaluations and analyses associated with the revised parameters resulting from the proposed uprate operation have allowed relaxation of the previous limit of 5. The results of the design transients and component evaluations confirm the acceptability of this change.
f.
Section 3/4.4.6 &
The Unit 2 Steam Generator F* distance is changed from 1.54 to Bases 3/4.4.6 1.60 inches. This revision is necessary to account for increased primary to secondary differential pressure.
As a result of the comprehensive uprate program, the opportunity was afforded for certain FNP parameter relaxations so that operational margins can be gained. Each of the effected technical specification parameters will be addressed.
h surveillance requirement for the centrifugal changing pumps (CCP) and the Residual Heat Removal (RHR) pumps has been modified to reflect FNP pump parameters, including a flow degradation allowance to allow for possible pump degradation. The revised flowrates were verified acceptable for all analyses that are sensitive to these parameters. These analyses include LBLOCA, SBLOCA, main steamline break, and MSLB/LOCA mass and energy release. In all cases, acceptance criteria continue to be met including LOCA PCT, MSLB DNBR, and containment integrity. 'lhe revised flow rates have also been assessed versus putnp and system performance The results of that assessment are that the design criteria for the pump and systems continue to be met.
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. O ne Bases changes for contamment internal pressure response to LOCA and MSLB events reflect the results of N.J FNP contamment analyses performed for the uprated operating conditions. He analyses demonstrate that the I
contamment remains within the design pressure limit for accident conditions, including an initial positive pressure of 3 psig. He change to the P. value in the FNP contamment Leakage Rate Testing Program is l
consistent with the new LOCA analysis results and 10 CFR 50, Appendix J, Option B.
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He Bases change for MSSV steam flow at 100% RTP has been modified based on the new calculated steam l
flows that have resulted from the proposed uprated conditions. %e change is made to ensure that the Bases accurately reflect the plant and analytical conditions.
4 An increase in the MSIV closure time from 5 to 7 seconds has been found acceptable. Results of analyses, 3
such as main steamline break and mass and energy releases, have verified that this increase isjustified. All acceptance criteria continue to be met. This increase will afford FNP some margin to the acceptance criteria i
j for MSIV stroke time requirements.
t NSSS major components (i.e., Reactor Vessel, Steam Generator, Pressurizer, etc.) and systems have been i
evaluated or reanalyzed for the design and cyclic transients resulting from the proposed uprating condition. All
]
i major components and systems have been found acceptable for the design and cycle transients. Included in the i
transients is an increase from 5 to 10 in==i y system hydrostatic pressure transients. This will afford additional operational flexibility while providing assurance of continued safe operation of FNP.
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De F* steam generator tube plugging criterion is licensed at FNP Unit 2. Primary to secorxiary pressure i
differential across the steam generator tubes is used to calculate the specific F* value. He FNP Unit 2 F*
I distance has been revised for boundmg plant conditions associated with upratmg, T., wmdow, and steam l
{
generator tube plugging of 20%. These boundmg inputs were applied to the NRC approved F* methodology s
1 for Farley Unit 2 to develop the revised value of 1.60 inches.
Based on the proceeding information and on the analyses and evaluations performed to support the proposed j
uprating, the following conclusions with respect to 10 CFR 50.92 can be reached.
l l
1.
The proposed changes do not increase the probability or consequences of an accident previously evaluated in the FSAR. Explicit modeling of these parameters is included in the 4
j uprate analyses and evaluations. He comprehensive analytical effort performed to support the i
proposed uprating has included a review and evaluation of all components and systems (including interface systems and control systems) that could be affected by this change. In addition LOCA and non-LOCA analyses and evaluations have verified that all acceptance l
criteria continue to be met. All systems will function as Ng-i None of these proposed
]
changes can directly initiate any accidents; therefore, the probability of an accident has not j
been increased All dose consequences have been analyzed or evaluated with respect to these parameters, and all acceptance criteria continue to be met. Therefore, the consequences of an 1
j accident previously evaluated in the FSAR have not increased.
i 2.
The proposed changes do not create the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed changes. He proposed j
. technical specification changes have no adverse effects on any safety-related system and do not challenge the performance or integrity of any safety-related system. Derefore, the possibility of a new or different kind of accident is not created.
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3.
The proposed technical specification changes do not involve a significant reduction in a margin of safety, All analyses supporting the proposed power uprate reflect these proposed values.
I All acceptance criteria (including LOCA peak clad temperature, DNB criteria, containment temperature and pressure, and dose limits) continue to be met. 'Iherefore, the proposed changes do not involve a significant reduction in the margin of safety.
i Based on the above information and on the analyses performed to support the proposed uprating, the proposed
{
changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
j IV.
ALLOWABLE VALUES AND TRIP SETPOINTS FOR REACTOR TRIP SYSTEM AND ESFAS I
i Table 2.2-1 Reactor Trip System k&nsw.tation Trip Setpoints a.
Allowable Values:
1 Power Range, Neutron Flux, Low Power Range, Neutron Flux, High Power Range, Neutron Flux, High Positive Rate Power Range, Neutron Flux, High Negative Rate Pressurimr Pressure, Low
)
Pressurimr Pressure, High Pressurimr Water Level, High Loss of Flow (includmg footnote for MMF)
O Steam Generator Waur Level, Low-low Reactor Trip System Interlock / Low Power Reactor Trips Block, P-7 (P-10 input)
Reactor Trip System Interlock / Power Range Neutron Flux, P-10 Reactor Trip System Interlock / Reactor Trips Block Following Turbine Trip, P-9 Trip Setpoint and Allowable Va!ae:
RTS Interlock / Power Range Neutron Flux, P-8 b.
Table 3.3-4 Engineered Safety Features Actuation System Instrumentation Trip Setpoints Allowable Values:
SI, Turbine and Feedwater Isolation /Pressuriar Pressure, Low Steamline Isolation / Steam Flow in Two Steam Lines, High Auxiliary Feedwater/ Steam Generator Water Level, Iow-Low ESFAS Interlocks /Pressurimr Pressure, P-ll i
Trip Setpoints and Allowable Values:
Turbine Trip and Feedwater Isolation Ste.un Generator Water Level, High-High (P-14)
ESF Actuation System Interlocks /lew-lew Tavg, P12 (Decreasing Low-Low Tavg Trip Setpoint Unchanged) pwrupl4. doc 8
RJM/MGE 2/13,97
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The changes in trip setpoints listed above have resulted from the analyses performed to support FNP operation at the proposed uprated power. Setpoint uncertainty calculations confirm the acceptability of the setpoints. The results of the analyses conclude that since all acceptance criteria continue to be met, the proposed values are acceptable, r
he allowable value changes listed above have been modified to reflect the results of updated setpoint i
calculations based on FNP specific uncertainties, cali5ation practices, calibration equipment, installed i
hardware and procedures. Dese allowable valec: were calcuhted using the Westinghouse setpoint i
methodology consistent with that recently used for the FNP OTAT and OPAT margin improvement j
program that was reviewcxi and approved by the NRC. This methodology includes refinements which j
better reflect plant calibration practices and equipment performance Based on the information presented above and the analyses and evaluations performed for the proposed power uprating, the following conclusions can I c reached with respect to 10 CFR 50.92.
1.
The proposed changes do not increase the probability or consequences of an accident previously evaluated in the FSAR. He comprehensive engineering effort performed to support the proposed uprating has included evaluations or reanalysis of all accident analyses including all dose related events. Setpoint calculations have verified acceptability of the proposed setpoints and allowable value changes. All systems will function as designed, and all performance requirements on these systems have been verified to be acceptable. Neither allowable values nor the setpoints initiate any accident; therefore, the q
probability of an accident has not been increased. All dose consequences have been Q
analyzed or evaluated with respect to these parameters, and all acceptance criteria continue to be met. Therefore the consequences of an accident previously evaluated in the FSAR have not increa-ed.
2.
The proposed setpoints and allowable value changes do not create the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed changes. He proposed technical specification changes have no adverse effects on any safety-related system and do not challenge the performance ofintegrity of any safety-related system. He specified trip setpoints associated with the respective RTS and ESFAS functions ensure all accident analyses criteria continue to be met. Derefore, the possibility of a new or different kind of accident is not created.
3.
De proposed technical specification changes do not involve a significant reduction in a margin of safety. All analyses supporting the proposed power uprate reflect these proposed values. Setpoint calculations demonstrate that margin exists between the setpoint and the corresponding safety analysis limits. The calculations are based on FNP instrumentation and calibration / functional test methods and include allowances for uprated power conditions. All acceptance criteria (including LOCA peak clad temperature, DNB criteria, containment temperature and pressure, and dose limits) continue to be met herefore, the proposed changes do not involve a significant reduction in the margin of safety.
Based on the previous information and on the analyses performed, the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
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ATTACHMENTIV 1
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.l FARLEY NUCLEAR PLANT i
l TECHNICAL SPECIFICATIONS CHANGE REQUEST 4
10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION 3
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Joseph M. Fadey Nuclear Plant Units I and 2 1
Power Uprate SIGNIFICANT HAZARDS EVALUATION INTRODUCTION & BACKGROUND The Farley Power Uprate Project has been structured consistent with the 'aar6ialogy established in WCAP-10263, "A Review Plan for Upratmg the Licensed Power of a PWR Power Plant," dated 1983.
Since its submittal to the NRC, the methodology has been used successfully as a basis for power uprate projects on over twenty pressurized water reactor (PWR) units.
The methodology in WCAP-10263 established the ground rules and cniteria for power uprate projects, includmg the broad categories that must be addressed, such as NSSS performance parameters, design transients, systems, w.-==, accidents and nuclear fuel, as well as the interfaces between the NSSS and the Ralance of Plant (BOP) fluid systems Inherent in this ~6ialogy are key points that promote correctness, consistency, and licensability. "Ihe key points include the use of well-defined analysis input assumptions / parameter values, use of currently approved analytical techniques (e.g., methodologies and computer codes), and use of currently applicable licensing criteria and standards.
Southern Nuclear Operatmg Company (SNC) has completed a comprehensive engmeering review program for Farley Nuclear Plant (FNP) to increase licensed power from 2652 MWt to 2775 MWt that included a reanalysis or evaluation of all LOCA, non-LOCA, mass and energy release, and dose analyses. All acceptance criteria continue to be met. All major NSSS components (e.g., Reactor Vessel, Pressurizer, RCPs, Steam Generators, etc.) have been== M with respect to bounding conditions expected for power operation at the uprated power. In all cases operation has been found acceptable. Major systems and sub-systems (e.g., safety injection, auxiliary feedwater, RHR, turbine generator, etc.) have been reviewed and ec=*2ble performance has been verified for their normal operation and, as applicable, for their safety-related functions. All reactor trip and ESF actuation setpoints have been===~=~t and the proposed setpomt modifications will assure adequate protection is afforded for all design bases events. Control systems including rod control, pressurizer level and pressure control, and steam generator level control have been evaluated and been found sc=*2ble for operation at the uprated conditions. In addition, a range of T., (567.2*F to 577.2*F) has been included in the uprate program to afford FNP flexibility in plant operation and fuel des:gn.
To meet the criteria of 10 CFR 50.46 for LBLOCA, best estunate analyses were performed using the approved W~dag6>se Best Estimate model E COBRAffRAC. The SBLOCA continues to be analyzed using the current 'aa*Wis.
'Ihe technical bases for these proposed changes to the FNP Units 1 and 2 Operating Licenses (NPF-2 and NPF-8) are =deaed in the Farley Nuclear Plant Units 1 and 2 Power Uprate Project NSSS Licensing Report, WCAP 14723, and BOP Licensing Report.
O
O PROPOSED CHANGES De proposed amendment will change the maxunum reactor core power level for facility operation from 2652 MWt to 2775 MWt in the Facility Operating Licenses (FOL). In addition, the proposed aW=ent involves the following Technical Specifications (T/S) changes. De defmed rated thermal power for Farley will be changed to 2775MWt. De DNB parameters for RCS T,, Pressurizer Pressure, and RCS flow have been modified. The revised RCS flowrate is 264,200 gpm. Dese changes are consistent with FNP's assumed thermal design flowrate (258,000 gpm) and FNP specific RTS/ESFAS procedures and mstrumentation Reactor trip system interlock setpoint for power range neutron flux (P-8) and ESF actuation trip a-tpoint for steam generator water high-high level for turbine trip and feedwater isolation (P-14), and ESF actuation system interlock for low-low T,(P-12) (increasing) have been modified to reflect analytical results.
An evaluation of additional reactor trip system and ESF actuation system safety analysis limits and trip setpomts results in changes to the allowable values for several functions. Dese functions will be discussed and listed later in this document.
De comprehensive power uprate reanalyses and evaluations have resulted in improved operational margins. Based on the results of new containment analyses, the maximum peak calculated contamment internal pressure for a LOCA event will be revised from 48 to 43 psig. He main steamline isolation valve Q
closure time requirement will be revised from 5 to 7 seconds his will aid in the functional testmg of the V
valve. To account for possible Centrifugal Charging Pump (CCP) or Residual Heat Removal (RHR) pump degradation, reduced ECCS flows have been found acceptable. De corresponding surveillance requirement will be modified to reflect these changes. Also, an increase from 5 to 10 in the number of secondary systems hydrostatic tests (Table 5.7-1) has been analytically verified to be acceptable. For Unit 2 the steam generator F* distance will be revised.
Two adnunistrative changes are also included. Dese changes are a revision to the RCS volume contamed in Section 5.4, and the addition of the NRC approved references for best estimate LOCA (BELOCA) listed in Section 6.9.1.
De followmg 4 groups of proposed changes will each be evaluated with respect to the criteria of 10 CFR 50.92. Listed below are the groupings.
i 1.
DEFINITION AND ADMINISTRATIVE CHANGES a.
Section 1.25 Rated Dermal Power Definition b.
Section 5.4 RCS Volume c.
Section 6.9.1.11 COLR References II.
DNB PARAMETERS Table 3.2-1 &
DNB Parameters Bases 3/4.2.5 popluoc 2
IUWMGE 101/97
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l III.
MISCELLANEOUS OPERATIONAL AND MARGIN ENHANCEMENT CHANGES a.
Section 3/4.5.2 ECCS Subsystem - T., > 350 'T i
b.
Bases 3/4.6.1.4 Contamment Systems & Leakage Rate Testing Program l
j Bases 3/4.6.1.6 &
l Secten 6.16 2
c.
Bases 3/4.7.1.1 Main Steam Safety Valve (MSSV)
)
d.
Section 3/4.7.1.5 Main Steam Isolaten Valve (MSIV) l l
c.
Table 5.7-1 0-- Cyclic or Transient Limits i
f.
Section 3/4.4.6 &
Steam Generators (Unit 2 only)
)
Bases 3/4.4.6 IV.
ALLOWABLE VALUES /fRIP SETPOINTS TABLE 2.2-1 AND 3.3-4 FOR REACTOR TRIP AND ESFAS.
1 a.
Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints b.
Tabic 3.3-4 hgW Safety Feature Actuation System Instrumentation Trip Setpoints 10 CFR 50.92 EVALUATIONS I.
DEFINITION AND ADMINISTRATIVE CHANGES a.
Section 1.25 Defmition of reactor core rated thermal power is changed from 2652 O
to 2775 MWt.
b.
Section 5.4 Revise RCS volume from 9723 to 9829 A' at a T., = 567.2 *F.
c.
Section 6.9.1.11 BELOCA =thmblogy is added to COLR references These changes are admmistrative and editorial in nature. The change to the defmition of reactor core rated thermal power from 2652 MWt to 2775 MWt is based on the FNP uprated power value. This change is also applicable to the mammum power level defined in FOL Section 2.C(l). The change in RCS volume contamed in the feature section of the Technical Specifications reflects the current calculated volume for the FNP Units.
These revised values are included in the supportmg uprating evaluations and analyses; therefore, these changes are acceptable. In addition, the COLR references will include the NRC approved Westmghouse BELOCA methodology which was used for the uprated FNP LBLOCA analysis.
Based on the information presented above and the analyses and evaluations performed for the proposed power upratung, the followmg conclusions can be reached with respect to 10 CFR 50.92.
1.
The proposed changes to the definition of rated thermal power, RCS volume and BELOCA methodology do not increase the probability or consequences of an accident previously evaluated in the FSAR. The comprehensive analytical efforts performed to support the proposed upratmg inicluded a review and evaluation of all components and systems (including lief.m systems and control systems) that could be affected by this change. The revised power uprate value and RCS volume were inputs to applicable safety analyses. All systems will function as designed, and all performance requirements for these systems have been evaluated and found acceptable. None of these proposed changes directly initiate any accident; O
therefore, the probability of an accident has not increased All dose consequences have been analyzed or evaluated with respect to these parameters, and all acceptance criteria continue to be met. Therefore, the consequences of an accident previously evaluated in the FSAR have not increased p-pl4.4=
3 IUWMGE 10187
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2.
"Ihe proposed changes do not create the possibility of a new or different kind of accident than i
any accident already evaluated in the FSAR. No new accident scenarios, failure Maakm<
l or limiting single failures are introduced as a result of the proposed changes. The proposed j
j technical specification changes have no adverse effects on any safety-related system and do not i
l challenge the performance or integrity of any safety-related system lherefore, the possibility
]
of a new or different kind of accident is not created j
4 3.
The proposed operatmg license and technical specification changes do not involve a significant reducten in a margm of safety. All analyses supporting the proposed power uprate reflect the i
l RCS volume and rated thermal power values. 'Ihe use of NRC approved BELOCA
-*kadalogy must be referenced since BELOCA will now be the LBLOCA analysis licensing
}
basis for FNP. All =%+== criteria (including LOCA peak clad temperature, DNB criteria, j
contamment temperature and pressure, and dose limits) continue to be met. Therefore, the j
proposed changes do not involve a significant reduction in the margin of safety, i
l Based on the above information and on the analyses performed to support the proposed uprating, these proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
11.
DNB PARAMETERS CHANGES 1
a.
Table 3.2-1 &
DNB parameters Bases 3/4.2.5
- (
Indicated RCS T, 580.3*F Indicated Pressurizer Pressure 2209 pig l
~
Indicated RCS Total Flow Rate 264,2 30 gpm i
l These values reflect the input values for certain non-LOCA accident taalyses. All accident analysis acceptance criteria (including DNBR, contamment response for pressure and temperature, and dose consequences) have i
been met, and therefore, these values are acceptable. The appropriate Bases have been revised to reflect these proposed changes based on the FNP specific calculaten assumptons/results.
l 1he DNB par.. Ass (1~6 W RCS T., Indicated Pressurizer Pressure and Indicated RCS Total Flow Rate) are calculated based on the most limiting analytical values plus indication uncertainties. All values reflect FNP i
specific analytical values used to verify the adequacy of FNP safe operation at the proposed uprated power level v=h=. In addition, the uncertainty associated with FNP specific procedures, calibration practices, i
j and equipment have been included in the calculation of the fmal indicated values. The flow measurement l
uncertamty (2.4% in footnote) also includes 0.1% uncertainty for possible venturi fouling. WCAP-12771, Rev.
4 h
l 1,"Wea:Wse Revised Thermal Design Procedure Instrument Uncertainty MM adalagy for Alabama Power Farley Nuclear Plant Units 1 and 2 (Upratmg to 2785 MWt NSSS Power)," includes the revised calculanons for uprate.
j j
Based on the information presented above and the analyses and evaluations performed to support the proposed upratmg, the followmg conclusions can be reached with respect to 10 CFR 50.92.
3 O 4
RJWMOE 1/31/97 l
perupl4Anc 4
1.
The proposed technical specification changes for DNB pai..mers do not involve a significant increase in the probability or consequences of an accident previously evaluated in the FNP FSAR. The mechanical design features associated with VANTAGE 5 fuel and the improved
=whadalogies (such as Revised Thermal Design Procedure) provide capability for relaxation of analytical input parameters such that incmased DNBR margin can be generated without violation of any acceptance criteria. The indicated DNB parameters bound the analytical values used to support the proposed uprating in each case, the appropriate design and acceptance criteria are met. All performance requirements for any system or -:=g _=r have been evaluated and support the revised analysis assumptions. Overall plant integrity is not reduced. Furthermore, the parameter changes are associated with features used as limits or mitigators to assumed accident scenarios and are not accident initiators. Therefore, the l
probability of an accident has not signi6cantly increased The radiological consequences of accidents previously evaluated in the FSAR have been assessed doe to the proposed tecnnical specification changes. Evaluations have confirmed that the doses remain within previously approved arraanble limits as well as those defined by 10 CFR 100. Therefore, the radiological consequences to the public resulting from any accident previously evaluated in the FSAR has not significantly increased 2.
The proposed technical specification changes do not create the possibility of a new or different kind of accident from any previously evaluated in the FSAR. No new accident scenarios, failure mechanisms, or limiting single failures are ' troduced as a result of the revised DNB m
("
pases. The revised analytical assumptions have no adverse effect and do not challenge the performance of any other safety-related system This has been verified in WCAP 12771, Rev.1. Therefore, the possibility of a new or different kind of accident is not created 3.
The proposed *~ hair =1 specification changes do not involve a significant reduction in the margin of safety. The margm of safety for fuel-related parameters (such as DNB and Kw/ft) are defined in the Bases to the Technical Specifications. The uncertainties===~4=W with the proposed DNB parameter changes are included in the core safety limits. Performance of analyses and evaluations with the reactor core safety limits defined by RTDP have confirmed that the operatmg envelope defined by the Technical Specifications continues to be bounded by the revised analytical basis, which in no case exceeds the -:W = = limits. 'Iherefore, the margin of safety provided by the analyses in accordance with these acceptance limits is not reduced.
Based upon the preceding information, it has been deternuned thr.t the proposed changes to the DNB parameters do not involve a significant hazards consideration as.icfined in 10 CFR 5092.
Ill.
MISCELLANEOUS OPERATION AND MARGIN ENHANCEMENT CHANGES a.
Section 3/4.5.2 For the ECCS subsystem with T., > 350 *F, modifications are proposed to surveillance requirements for the recirculation flew differential pressures for CCP (change 2453 psig to 2323 p6d) and RHR pump (change 136 psig to 145 psid). FNP specific pump performance and margins, including allowances for possible pump O
degradation, have resulted in revised ECCS performance flowrates.
.l These flowrates have been verified to be acceptable from pump, system, and LOCA and non-LOCA analysis perspectives.
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Bases 3/4.6.1.4 For the Contamment Systems internal pressure and structural i
Bases 3/4.6.1.6 &
integrity Bases, revise the maximum peak calculated contamment Section 6.16 internal pressure for a LOCA event from 48 to 43 psig and add the peak calculated contamment pressure for the MSLB event of 53 psig.
1 Clarify that the analyses results are based on an initial pressure of 3 psig and that the LOCA results define the P. value for the j
Contamment I M==a Rate Testing Program. These revisions are j
supported by the contamment analyses performed for uprate.
f c.
Bases 3/4.7.1.1 For the MSSV technical specification Bases, revise percentage of steam flow at 100% RTP from i12 to 105.8% and total==d='y steam flow at 100% RTP from 11,613,849 to 12,270,000 lbs/hr. The increase in primary side power as a result of the uprating results in increases in total steam flow. The revised Bases reflect the new plant
?
full power steam flow values resulting from the uprating changes j
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Section 3/4.7.1.5 The MSIV closure tune requirement is revised from 5 to 7 seconds.
4
'Ihe analyses (such as main steamline break and mass and energy l
releases) that model main steamline isolation have verified that this increase in stroke time is =eT2ble. This operational enhancement j
will provide margin to MSIV test s-:-3 ; = criteria.
t e.
Table 5.7-1
'Ihe component cyclic or transient limits for secondary system hydrostatic pressure is changed from 5 to 10. The evaluations and j
analyses associated with the revised parameters resulting from the i
proposed uprate operation have allowed relaxation of the previous l
limit of 5. The results of the design transients and sq+ =t i
evaluations confirm the acceptability of this change.
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Section 3/4.4.6 &
1he Unit 2 Steam Generator F* distance is changed from 1.54 to i.
Bases 3/4.4.6 1.60 inches 1his revision is necessary to account for increased primary to==d= y differential pressure.
i As a result of the comprehensive uprate program, the opportunity was afforded for certain FNP parameter i
relaxations so that operational margins can be gained. Each of the effected technical specification parameters j
will be addressed.
The surveillance requirement for the centrifugal changing pumps (CCP) and the Residual Heat Removal (RHR) pumps has been modified to reflect FNP pump parameters, including a flow degradation allowance to allow for possible pump degradation. The revised flowrates were verified acceptable for all analyses that are sensitive to i
these parameters. These analyses include LBLOCA, SBLOCA, nmin steamline break, and MSLB/LOCA mass and mergy release. In all cases, arvm criteria continue to be met including LOCA PCT, MSLB DNBR, and contamment integrity. The revised flow rates have also been assessed versus pump and system i
perfor: nance The results of that assessment are that the design criteria for the pump and systems contmue to j
be met.
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j De Bases changes for containment internal pressure response to LOCA and MSLB events reflect the results of j
FNP matmament analyses performed for the uprated operatmg conditions. De analyses A maaerate that the i
contamment remams within the design pressure limit for accident conditions, including an initial positive pressure of 3 psig. De change to the P. value in the FNP contamment leakage Rate Testing Program is j
consistent with the new LOCA analysis results and 10 CFR 50, Appendix J, Option B.
I The Bases change for MSSV steam flow at 100% RTP has been modified based on the new calculated steam J
flows that have resulted from the propored uprated conditions. The change is made to ensure that the Bases j
accurately reflect the plant and analytical conditions.
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An increase in the MSIV closure time from 5 to 7 seconds has been found acceptable. Results of analyses, such as main steamline break and mass and energy releases, have verified that this increase is justified. All i
i acceptance criteria continue to be met. This increase will afford FNP some margin to the acceptance criteria j
for MSIV stroke time requirements.
NSSS major components (i.e., Reactor Vessel, Steam Generator, Pressurizer, etc.) and systems have been j
evaluated or reanalyzed for the design and cyclic transients resulting from the proposed uprating condition. All major components and systems have been found acceptable for the design and cycle transients. Included in the transients is an increase from 5 to 10 in aad= y system hydrostatic pressure transients. His will afford j
additional operational flexibility while providing assurance of continued safe operation of FNP.
De F* steam generator tube plugging criterion is licensed at FNP Unit 2. Pnmary to secondary pressure l
differential across the steam generator tubes is used to calculate the specific F* value. De FNP Unit 2 F*
distance has been revised for haaaAag plant conditions associaed with uprating, T., wmdow, and steam i
generator tube plugging of 20% Dese boundmg inputs were applied to the NRC approved F* methdology j
for Farley Unit 2 to develop the revised value of 1.60 inches l
Based on the proceeding information and on the analyses and.. valuations performed to support the proposed j
upratmg, the following conclusions with respect to 10 CFR 50.92 can be reached.
l 1.
The proposed changes do not increase the probability or consequences of an accident j
previously evaluated in the FS AR. Explicit modeling of these parameters is included in the upate analyses and evaluations ne comprehensive analytical effort performed to support the proposed upratmg has included a review and evaluation of all components and systems (including interface systems and control systems) that could be affected by this change. In l
addition LOCA and non-LOCA analyses and evaluations have verified that all acceptance l
criteria continue to be met. All systems will function as designed. None of these proposed j
changes can directly initiate any accidents; therefore, the probability of an accident has not i
been increased. All dose consequences have been analyzed or evaluated with respect to these parameters, and all acceptance criteria continue to be met. Therefore, the consequences of an accident previously evaluated in the FSAR have not increased 2.
The proposed changes do not create the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new accident scenarios, failure =~haaisms i
or limiting single failures are introduced as a result of the proposed changes. The proposed j
technical specification changes have no adverse effects on any safety-related system and do not j
challenge the performance or integrity of any safety-related system. Therefore, the possibility of a new or different kind of accident is not created k.
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t, 3.
De proposed technical specification changes do not involve a significant reduction in a margin of safety. All analyses supporting the proposed power uprate reflect these proposed values.
All ecca-criteria (including LOCA peak clad temperature, DNB criteria, contamment 4
temperature and pressure, and dose limits) continue to be met. Acrefore, the proposed I
changes do not involve a significant reduction in the margin of safety.
2
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Based on the above information and on the analyses performed to support the proposed uprating, the proposed i
changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
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IV.
ALLOWABLE VALUES AND TRIP SETPOINTS FOR REACTOR TRIP AND ESFAS.
a.
Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints i
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Allowable Values:
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Power Range, Neutron Flux, Low Power Range, Neutron Flux, High i
Power Range, Neutron Flux, High Positive Rate Power Range, Neutron Flus High Negative Rate Pressurizer Pressure, low Pressurizer Pressure, High Pressurizer Water level, High j
Loss of Flow (including footnote for MMF) j Steam Generator Water Level, Ism-Low j
Reactor Trip System Interlock / Low Power Reactor Trips Block, P-7 (P-10 input)
Reactor Trip System Interlock / Power Range Neutron Flux, P-10 Reactor Trip System Interlock / Reactor Trips Block Following Turbine Trip, P-9 Trip Setpoint and Allowable Value:
i RTS Interlock / Power Range Neutron Flux, P-8 5
b.
Table 3.3-4 Engineered Safety Features Actuation System instrumentation Trip Setpoints l
l Allowable Values:
SI, Turbine and Feedwater Isolation /Ps essurizer Pressure, Low Steamline Isolation / Steam Flow in Two Steam Lines, High 1
Auxiliary Feedwater/ Steam Generator Water level, low-Iow I
ESFAS Interlocks / Pressurizer Pressure, P-ll l
Trip Setpomts and Allowable Values:
Turbine Trip and Feedwater Isolation Steam Generator Water level, High-High I
(P-14)
ESF Actuation System Interlocks /1.aw-low Tavg, P12 (Decreasing low-low
)
Tavg Trip Setpoint Unchanged) 8 IUWMGE 1/31/97 j
pwrupt u ne
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i The changes in trip setpoints listed above have resulted from the analyses performed to support FNP operation at the proposed uprated power. Setpoint uncertainty calculations confirm the acceptability of the 4
l setpoints. The results of the analyses conclude that since all acceptance criteria continue to be met, the 4
proposed values are acceptable.
I The allowable value changes listed above have been modified to reflect the results of updated setpoint calculations based on FNP specific uncertainties, calibration practices, calibration equipment, installed i
hardware and procedures. These allowable values were calculated using the Westinghouse setpoint i
{
Madalogy consistent with that recently used for the FNP OTAT and OPAT margin improvement program that was reviewed and approved by the NRC. This methodology includes refinements which 4
better reflect plant calibration practices and equipment performance i
Based on the information presented above and the analyses and evaluations performed for the proposed power upratmg, the followmg conclusions can be reached with respect to 10 CFR 50.92.
1.
The proposed changes do not increase the probability or consequences of an accident previously evaluated in the FSAR. The cue.pd.casive engineering effort performed to support the propose.d uprating has included evaluations or reanalysis of all accident analyses including all dose related events. Setpoint calculations have verified acceptability of the proposed setpoints and allowable value changes. All systems will function as designed, and all performance requirements on these systems have been verified to be acceptable. Neither allowable values nor the setpoints initiate any accident; therefore, the probability of an accident has not been increased All dose consequences have been
(
analyzed or evaluated with respect to these parameters, and all acceptance criteria contmue to be met. Therefore the consequences of an accident previously evaluated in the FSAR have not increased 2.
The proposed setpoints and allowable value changes do not create the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed changes. The proposed technical specification changes have no adverse effects on any safety-related system and do not challenge the performance ofintegrity of any safety-related system. The specified trip setpoints associated with the respective RTS and ESFAS functions ensure all accident analyses criteria continue to be met. Therefore, the possibility of a new or different kind of accident is not created 3.
The proposed technical specification changes do not involve a significant reduction in a margm of safety. All analyses supporting the proposed power uprate reflect these proposed values. Setpoint calculations demonstrate that margin exists between the setpoint and the correspondmg safety analysis limits. The calculations are based on FNP mstrumentation and calibration / functional test methods and include allowances for uprated power conditions. All acceptance criteria (including LOCA peak clad terrperature, DNB criteria, contamment temperature and pressure, and dose limits) continue to be met.
Therefore, the proposed changes do not involve a significant reduction in the margin of safety.
Based on the previous information and on the analyses preformed, the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
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RJM/MGE 101/97
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6 ATTACHMENT V j
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4 FARLEY NUCLEAR PLANT Jl i
TECHNICAL SPECIFICATIONS CHANGE REQUEST j
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POWER UPRATE 4-NSSS LICENSING REPORT
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