ML20199J028
| ML20199J028 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/23/1998 |
| From: | Dick G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20199J034 | List: |
| References | |
| NPF-37-A-098, NPF-66-A-098 NUDOCS 9802050194 | |
| Download: ML20199J028 (41) | |
Text
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UNITE 3 STATES pe g
NUCLEAR REIULATCRY CEMMISSl3N v
WAsHINeToN, D.c. SteeHeM 1
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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50454 4
BYRON STATION. UNIT NO.1 l
AMENDMENT TO FACluTY OPERATING LICENSE 4
4 i
Amendment No. 98 i
License No. NPF-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendm.nt by Commonwealth Edison Company (the licensee) dated May 21,1997, as supplemented b) letters dated November 18,1997, December 3,1997, January 8,1998 and January 13,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the j
Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 4
C.
There is reasonable assurance (i) that the activities authorhed by this amendment can be conducted without endangering the health and safety of the public, and (il, that such activities will be conducted in compliance with the Commission's j
regulations; i.
D.
The issuance of this amendment will not be inimical to the common defense and secunty or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requiromants have been satisfod.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the atta& ment to this license amendment, and paragraph 2.C.(2) of Facility Operating Lloonse No. NPF 37 is hereby amended to read as follows:
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9902050194 980123 PDR ADOCK 0S000454 P
2 (2) Te&nical Specirmations The Technical Specifications contained in Appendix A as revised through Amendment No. 98 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licenses shall operate the facility in accordance with the TecF 1
Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be i
implemented within 60 days.
FOR THE NUCLEAR REGUI.ATORY COMMISSION
)tt Coorg
. Dick, Son Project Manager Project Directorate lil 2 Division of Reactor Projects til/IV Offica of Nuclear Reactor Regulation Atta& ment:
Changes '.o the Technical Specifications Date of Issuance: January 23, 1998 1
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NUCLEAR RECULATCRY COMMISSlaN wAsmeeron,o.c. -
5 COMMONWEALTH EDISON COMPANY l
DOCKET NO. STN Sc.455 By40N STATION. UNIT NO. 2 j
AMENDMF.NT TO FACILITY OPERATING LICENSE Amendment No. 98 Uoonse No. NPF 66 1
1.
The Nuclear Regulatory Commission (the Commission) has found that:
1 A.
The application for amendment by Commonwealth Edison Company (the licensee) dated May 21,1997, as supplemented by letters dated November 18,.
1997, December 3,1997, January 8,1998 and January 13,1998, complies with i
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the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; 4
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in complianos with the e
Commission's regulations; D.
The issuanos of this a7ndment will not be inimical to the common defense and security or to the healt md safety of the public; and E.
The issuance of this Ornendment is in accordance with 10 CFR Part 51 of the '
Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating Uoense No. NPF 66 is hereby amended to read as follows:
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2 (2)
Technical Boecifications The Technical Specifications contained in Appendix A (NUREG 1113), as revised through Amendment No. 98 and revised by Attachment 2 to NPF 66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to Lloonse No. NPF 37, dated February 14,1985, are hereby incorporated into this license. Attachment 2 contains a revision to A7pendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as t,r the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION l
U George Dick, Senior roject Manager Project Directorate lil 2 Division of Reactor Projects -lil/IV Office of Nuclear Reactor Regulation Attacliment:
Changes to the Technical Specifications Date of lasuance: January 23, 1998
1 ATTACHMENT To LICENSE AMENDMENT NOS. 98 AND 98 FACILITY OPERATING LICENSE NOS. NPF 37 AND NPF,-gg i
DOCKET NOS,.6TN 50-454 AND STN 50-455
)
Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identifed by the captioned amendment number and contain marginallines indicating the area of change.
tlamove Paoes insert Pages 4
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Vill Vill XV XV 1-4a 3/4432 3/4 4-32 3/4 4-33 3/4 4-33 3/4 4-34 3/4 4-34 3/4 4-35 3/4 4-35 3/4 4-36 3/4 4-36 3/4437 3/4 4-37 3/4 4-38 3/4 4-39 3'4 4-40a 3/4 4-40b 3/4 4-41 W4442 t
3/4 4-43 B 3/4 4-7 B 3/4 4-7 8 3/4 4-8 B 3/4 4-8 B 3/4 4-9 B 3/4 4-9 B 3/4 4-10 B 3/4 4 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 B 3/4 4-14 B 3/4 4-15 B 3/4 4-16 B 3/4 4-17 6-22b 6-22b t
4
INDEX DEFINITIONS SECTION E6HE 1.0 DEFINITIONS 1.1 ACTI0N........................................................
1-1 1.2 ACTUATION LOGIC TEST..........................................
1-1
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1.3 ANALOG CHANNEL OPERATIONAL TEST...............................
1-1 1.4 AXIAL FLUX DIFFERENCE.........................................
1-1 1.5 CHANNEL CALIBRATION...........................................
1-1 1.6 CHANNEL CHECK.................................................
1-1 1.7 CONTAINMENT INTEGRITY.........................................
1-2
[
1.8 CONTROLLED LEAKAGE............................................
1-2 1.9 CORE ALTERATION...............................................
1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................
1-2 1.11 DOSE EQUIVALENT I-131........................................
1-2a 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................
1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................
1-3 1.14 FREQUENCY N0TATION...........................................
1-3 1.15 IDENTIFIED LEAKAGE...........................................
1-3 1.15.a L,.........................................................,
1-3 1.16 MASTER RELAY TEST............................................
1-3 1.17 MEMBER (S) 0F THE PUBLIC......................................
1-3 1.18 0FFSITE DOSE CALCULATION MANUAL..............................
1A 1.19 OPERABLE - OPERABILITY.......................................
1-4 1.19.a OPERATING LIMITS REP 0.iT.............
1-4 1.20 OPERATIONAL MODE - M0DE......................................
1-4 1.20.a P,..........................................................
1-4 1.21 PHYSICS TESTS................................................
1-4 1.22 PRESSURE BOUNDARY LEAKAGE................................
1-4 1.22.a PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)................
1-4a 1.23 PROCESS CONTROL PR0 GRAM......................................
1-5 1.24 PURGE - PURGING..............................................
1-5 1.25 QUADRANT POWER TILT RATI0....................................
1-5 1.26 RATED THERMAL P0WER..........................................
1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................
1-5 1.28 REPORTABLE EVENT.............................................
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BYRON - UNITS 1 & 2 I
AMENDMENT NO. 98
LIMITING OONDITIONS FOM OPERATION AND SURVEILLANCE RIOUIREMENTS 3ECTION
]El TABLE 3.4-2 REACTOR COOLANT SY3 TEM CHEMISTRY I.IMITS...............
3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS........................................
3/4 4-26 3/4.4.s SPECIFIC ACTIVITY........................................
3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 R2 ACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERBUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
>3#Ci/ GRAM DOSE EQUIVALENT I-131....................
3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM....................................
3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS l
Reactor coolant Syste...................................
3/4 4-32 Pressurizer............................
3/4 4-33 overpressure Protection Systems..........................
3/4 4-34 3/4.4.10 STRUCTURAL INTEGRITY...................
3/4 4-36 3/4.4.11 REACTOR COOLANT SYSTEM VINTS.............................
3/4 4-37 BYRON - UNITS 1 & 2 VIII AMENDMENT NO. 98 1
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BASES SECTION EAGI 3/4.4.5 STEAM GENERATORS..........................................
B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................
B 3/4 4-4 3/4.4.7 CHEMISTRY.................................................
B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY.........................................
B 3/4 4-5 3/4.4.9 PRESSURE /TEMPERATUPE LIMITS...............................
B 3/4 4-7 3/4.4.10 STRUCTURAL INTEGRITY.....................................-
B 3/4 4-10 3/4.4.11 REACTOR VESSEL HEAD VENTS................................
B3/44-10 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS..............................................
B 3/4 5-1 3/4.5.2, 3/4.5.3 and 3/4.5.4 ECCS SUBSYSTEMS.......................-
B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK..............................
B 3/4 5-4
-3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.......................................
B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................
B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES..............................
B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L...................................
B 3/4 6-4 I
v BYRON - UNITS 1 & 2 XV AMENDMENT N0. 98
DEFINITIONS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.22.a The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the pressurizer power operated relief valve (PORV) lift settirgs for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Speci'i'ation i
. 6.9.1.11.
Unit operation within these limits is addressed in LCI,
9.1,
" Pressure / Temperature Limits," and LCO 3.4.9.3, " Overpressure Protection Systems."
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r BYRON - UNITS 1 & 2-1-4a AMENDMENT N0.98
REACTOR COOLANT SYSTEM 3/4.4.5 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1. The Reactor Coolant System temperature and pressure. and heatup and cooldown rates shall be maintained-in accordance with the limits specified'in the.PTLR.
APFLICABILITY: At all times.
ACTION:
-With-any of the limits in the PTLR exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to-determine-the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for con'inued operation or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than200*Fand500psig,respectively,withinthefoTTowing30 hours.
t SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system l
heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.9 1.2 The reactor vessel material irradia*
surveillance specimens 4
shei t be removed and examined, to determine ch in materici properties, as required by 10 CFR Part 50, Appendix H,-in acc.. dance with the schedule in e
the PILR.
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' BYRON - UNITS 1 & 2 3/4 4-32
.TMENDMENT N0.98 y
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REACTOR COOLArf SYSTEM PRESSURIIER 1
LYMYTING CONDITION FOR OPERATION 3.4.9.2 The pressuriser temperature shall be limited to:
'a.
A maxianus heatup est 100*F in any 1-hour period, b.
A maximum cooldown of 200*F in any 1-hour period, and c.
A maximum spray water temperature differential of 320*F.
APPLICABILITY: At all tines.
&GIlQllI With the pressuriser temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressuriser; determir.e that the pressuriser remains acceptable for continued operation or be in at least 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 reduce the pressuriser pressure to less then 500 peig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l SURVEILLANCE REOUIREMENTS 4
4.4.9.2 The pressuriser temperatures shall be determined to be within the limits at least.once.per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the-limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.
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BYRON - UNITS 1 & 2-3/4 4-33 AMENDMENT NO. 98
- REACTOR OGOLANT SYSTEM
,o OVERPRESSURE PROTECTION SYSTEMS fYMITING NetBITION VOR OPERATION J
3.4.9.3 At least two overpressure protection devices shall be OPERABLE, and each device shall be either a.
A residual heat removal (RHR) suction relief valve with a lift setting of less than or equal to 450 peig, or b.
A power operated relief valve (PORV) with a lift setpoint that varies with RCS temperature which does not exceed the 1 Lait established in the PTLR.
APPLICABILITY:
MODES 4, 5, and 6 with the reactor vessel head on.
il, ACTION 6 a.
With one of the two required overpressure protection devices inoperable in MODE 4, restore two overpressure protection devices to OPERABLE status witbin 7 days or depressurise and vent the RCS through at least a 2 square inch vent wituin the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
With one of the two required overpressure protection devices inoperable in MODES S or 6, restore two overpressure protection devices to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
With both of the required overpressure protection devices inoperable, c.
depressurite and vent the RCS through at least a 2 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d.
With the RCS vented per ACTIONS a, b, or c, verify the vent pathway at least once per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open positions otherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
e.
In the event either the PORVs, RHR suction relief valves, or the RCS vents are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the commission pursuant to specification 6.9.2 within.30 days. Tne-report shall describe the circumstances initiating the transient, the effect of the PORVs, RNR suction relief valves, or RCS vents on the transient, and any corrective action necessary to prevent recurrence.
f.
The provisions of Specification 3.0.4 are not applicable.
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d.
BYRON - UNITB 1 & 2 3/4 4-34 AMENDMENT NO. 98
0 SURVEILLANCE REOUIREMENTS i
4.4.9.3.1 Each PORV shall be demonstrated OPERABLE when the PORVs are being used for cold overpressure protection by a.
Performancs of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days when the PORV is required OPERABLE; and b.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least-once per 18 months; and Verifying the PORV isolation valve is open at least once M r c.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.4.9.3.2 Each RNR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:
For RHR suction relief valve RH8708B verify at least once per a.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that valves RH8702A and RH8702B are open, b.
For RHR suction relief valve RH8708A verify at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that valves RH8701A and RH8701B are open.
c.
Testing pursuant to specification 4.0.5.
BYRON - UNITS 1 & 2 3/4 4-35 AMENDHENT NO. 98
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- REACTOR CDOLANT SYSTEM 3/4.4.1Q STRUCTURAL INTEGRITY LYMYTING -ITIOff FOR OPERATION 3.4.10. The structural integrity of AsME Code Class 1,-2, and 3 components i
'shall be maintained in accordance with specification 4.4.10.
J.
APPLICARILITY: All MODES.
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With the structural integrity of any AsME Code Class.1 component (s)-
a.
not conforming to the above requirements,-restore the structural integrity of the.affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant system temperature above 200*F.
b.
With the structural integrity of any AsME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate
.the affected component (s) prior to increasing the Reactor Coolant
-System temperature ambove 200*F.-
S With the structural integrity of any AsME Code Class 3 component (s) c.
l not conforming to-the above requirements, restord the structural integrity of the affacted component (s) to within.its. limit or isolate l
the affected component (s) from service.
d.
The provisions of-specification 3.0.4 are not applicable.
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- SURVEII.I.ANCE REOUIREMENTS 2
4.4.10 In addition to the requirements of specification 4.0.5, each reactor coolant = pump flywheel'shall be inspecte' as follows:
a.
Volumetric examination of the areas of higher stress concer.tration'at the bore and keyvays will be performed each 40 month period during refueling or maintenance shutdowns coinciding with the service inspection schedule as. required by Section-XI of the ASME Code.
b.
Visual examination of.all exposed surfaces will be performed and a surface examination of the bore and keyway surfaces will.be performed whenever the flywheels are removed for maintenance purposes, but not more frequently than once each 10 year interval.
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BYRON - UNITS 1 & 2 3/4 4-36 AMENDMENT No. 98
.r
e REACTOR CDOLANT SYSTEM e
3/4.4.11 m m' TOR COOTk W SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one reactor versel head vent path consisting of two valves in series powered from emergency busses shall be OPERABLE and closed.
APPLICABILITY: MODES 1, 2, 3 and 4.
EZI95:
With the above rea. tor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is main-tained closed with power removed from the valve actuator of all the valves in the inoperable vent paths restore the inoperable vent path to OPERABLE status within 30 days, or, be in NOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SNUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERAELE at least once per 18 months by:
Verifying all manual isolation valves in each vent path are locked in a.
the open position, b.
Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING, and Verifying flow through the reactor vessel head vent paths during c.
venting operations at COLD SHUTDOWN or REFUELING.
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BYRON - UNITS 1 & 2 3/4 4-37 AMENDMENT NO. 98 l
,. REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomenon. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
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3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and tem)erature changes. These Icsis are introduced by startup (heatup) and slutdown (cooldown) operations, power transients, and reactor trips. LCO 3.4.9.1, " Pressure / Temperature Limits,"
limits the pressure and temperature changes during RCS heatup and cooldown to within the design assumptions and the stress limits for cyclic operation.
The PRESSURE AND TEMPERATURE LIMITS REPORT (PTLt-) contains pressure and temperature (P/T) limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of en.6p of reactor coolant temperature.
Each P/T limit curve defines an acceptable region for normal operation.
c The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine tnat operation is within the allowable region.
LCO 3.4.9.1 establishes operating limits that provide a margin to non-ductile failure of the reactor vessel and )iping of the Reactor Coolant Pressure Boundary (RCPB). The vessel is tie component most subject to non-ductile failure, and the LCO limits apply to the entire RCS, except the pressurizer, which has different design characteristics and operating functions.
10 CFR Part 50, Appendix G, requires the establishment of P/T limits for
- specific material fracture toughras requirements of the RCPB materials.
Appendix G of 10 CFR Part 50 requires an adequate margin to non-ductile failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.
It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section % Appendix G.
The neutron embrittlement effect on the material toughness is reflected by increasing the Nil Ductility Reference Temperature (RT
) as exposure to neutron fluence increases.
The actual shift in the RT of the vessel material will be established periodically by removing and evia unting the irradiated reactor vessel material specimens, in accordance with ASTM E 185 and Appendix H of 10 CFR Part 50. The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99, Revision 2.
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vusel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate BYRON - UNITS 1 & 2 B 3/4 4-7 AMENDMENT N0. 98
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- JREA T0R C0OLANT SYSTEM
. BASES ER$SURE/ TEMPERATURE LIMITS (Continued) the most restrictive limit. Across the span of the P/T limit curves,-
differsat locations are more restrictive and, thus, the curves are composites of the most restrictive regions.-
The heatup curve represents a different set of restrictions than the cooldown etwa because the directions of the thermal gradients through the vessel web are reversed. The thermal gradient reversal alters the location of
' the tensile stress between the outer and inner walls during heatup and cooldown, respectively.
The criticality limit curve includes the 10 CFR Part 50, Appendix G, requirement that it be 140*F above the heatup curve or the cooldown curve, and not less than the, minimum pennissible temperature for ISLH testino. However, the criticality curve is not operationally limiting; a more restr' ctive limit exists in LCO 3.1.1.4, " Minimum Temperature for Criticality."-
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in non-ductile failure of the RCPB, possibly _ leading to a nonisolable leak or loss-of-coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of.the RCPB components. The ASME Code.
Section XI, Appendix-E, provides a recommended methodology for evaluating an-
-operating event that causes an excursion outside'the limits.
OVERPRESSURE PROTECTION SYSTEMS The Low Temperature Overpressure Protection'(LTOP) System controls the -
RCS-pressure at low temperatures so the integrity of the RCPB is not compromised by violating the P/T limits-of 10 CFR Part 50, Appendix G.
The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the maximum allowable actuation logic setpoints-for the pressurizer Power-Operated Relief Valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the 10 CFR Part 50, Appendix G, requirements during the-
- MODES in which the LTOP system is necessary.
The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material-toughness decreases and the raterial becomes less resistant to stress at low temperatures. RCS presrure, therefore, is maintained low at-low temperatures and is increased only within the limits specified in the PTLR.
The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only during shutdown; a pressure fluctuation can occur-more quickly than an operator can react-to relieve tne condition.
Exceeding the RCS P/T limits by a significant amount could cause non-ductile failure of the reactor vessel.
LCO 3.4.9.1, Pressure / Temperature Limits," req ~res
-administrative control of RCS pressure and temperature during heatup :id cooldown to prevent exceeding the PTLR' limits.
LCO 3.4.9.3, " Overpressure Protection Systems," provides RCS overpressure protection by having a minimum coolant input capability and having adequate pressure relief capacity.
Limiting coolant input capability requires all Safety Injection (SI) pumps and all but one charging pump (a centrifugal BYRON - UNITS 1 & 2 B 3/4 4-8 AMENDMENT NO. 98 A
~
.y
',IREACTORCOOLANTSYSTEM BASES i
PRESSURE / TEMPERATURE LIMITS (Continued) 4 i
charging pump)-incapable of injection into the RCS and isolation'of the SI accumulators. The pressure relief capacity requires either two redundant RCS t -
relief valves, or a depressurized RCS and an RCS vent:of sufficient size. One~
RCS relief valve or the open RCS vent is'the overpressure protection device that; acts to terminate an increasing pressure event.
4 With minimum coolant _ input capability, the ability to provide core coolant addition is restricted. The LCO does not require'the makeup control-system deactivated or the SI actuation circuits blocked. Due to the lower
_ pressures in the LTOP MODES and the ex>ected core decay heat levels, the makeup
- system can provide adequate flow via tie makeup-control valve.
If conditions-4 require the use of.more than one centrifugal charging pus) for makeup in-the-event of loss of inventory, then pumps can _be made availa>1e through manual-actions.-
.The LTOP System for pressure relief consists of two PORVs with reduced lift -settings, or. two Residual Heat Removal -(RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and a RCS vent of sufficient size. Two RCS relief valves are required for redundancy. One RCS relief. valve has adequate' relieving capability-to prevent overpressurization for the required coolant input capability.
PORV Reauirements As designed for the LTOP-System, each PORV is signaled to open if the RCS pressure approaches a limit determined by the LTOP actuation logic. The LTOP actuation logic monitors both RC5 temperature and RCS pressure and determines when a condition is approaching the PTLR limits. The wide range RCS temperature indications are auctioneered to select the' lowest temperature-l Signal.-
~
The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temperature. The calculated pressure 5
111mit is then compared with the indicated RCS pressure from a wide range pressure channel.-- If the indicated pressure meets'or exceeds the calculated i_
value, a PORViis signaled to open.
I The:PTLR presents the PORV setpoints ior the LTOP system. The setpoints
- are normally staggered so only one valve opos during a low temperature --
overpressure' transient. ' Having the setpoints of both valves within the limits in the PTLR ensures that the 10 CFR Part 50, Appendix G, limits will not be exceeded in any analyzed event.
i i
When a PORV is opened in an increasing pressure transient, the subsequent L
relief will cause the pressure-increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached
.and the valve is signaled to= close. The pressure continues to decrease below the reset pressure as the valve closes.
[
RHR Suction Relief Valve Reauirements i
During the LTOP MODES, the RHR System is operated for decay-heat removal L
'and low pressure letdown control. Therefore, the RHR suction isolation valves are open'in the piping from the RCS hot legs to the inlets of the RHR pumps.
[
BYRON - UNITS l & 2 B 3/4 4-9 AMENDMENT NO. 98 i
~
. REACTOR COOLANT SYSTEM BASES I
_ PRESSURE / TEMPERATURE LIMITS (Continued)
While these' valves are open, the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.
The RHR suction isolation valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. The RHR suction relief valves are spring loaded, bellows-type relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Class 2 relief valves.
RCS Vent Reauirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in au RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.
4 For an RCS vent to meet the flow capacity requirement, it requires removing a pressurizer safety valve; removing a PORV's internals, and disabling its block valve in the open position; or similarly establishing any comparable vent. The vent path (s) must be-above the level of reactor coolant, so as not to drain the RCS when open.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Sect an XI of the l
ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Components of the Reactor Coolant System were designed to provide access to pennit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Adhnda through Summer 1975.
3/4.4.11 REACTOR VESSEL HEAD VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability exists to perform this function.
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single vent valve pctver supply or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System Vent Systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TM1 Action Plan Requirements," November 1980.
BYRON - UNITS 1 & 2 B 3/4 4-10 AMENDMENT NO. 98
ADMINISTRATIVE CONTROLS RE20RIltlA_REQEMENTS frontinueril OPERATING LIMITS REPORT (Continued) 9.
WCAP-10054-P-a,.* Westinghouse'Small Break ECCS Evaluation Model Using the i
NOTRUMP Code,* dated, higust.1985 (Westinghouse Proprietary).
(Methodology for Specification: AxialFluxDifference,HeatFluxHot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor).
10.
Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and Comed application of the UET methodology addressed in " Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Controls Systems."
The 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 analysit are mat.
ny s.id-cycle revisions or supplements The OPERATING LIMITS REPORT, inclu m
thereto, shall-be provided upon iskunce, for each reind cycle, to the NRC
- Document Control Desk with copics to the Regional Administrator and Resident Inspector.
REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE _ LIMITS REPORT (PTLR) 6.9.1.10 RCS pressure and temperature limits-for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, as well as heatup
- and cooldown rates and pressurizer power-operated relief valve lift settings, shall be established and documented in the PTLR for the following:
- 1) LCO 3.4.9.1, " Pressure / Temperature Limits," and
- 2) LCO 3.4.9.3, " Overpressure Protection Systems."
The analytical methods used to determine the RCS pressure and temperature limits shall be those reviewed and approved by the NRC, specifically those described in NRC letter dated January 21, 1998, "Bron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report."
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence perioo and for any revision or supplement thereto.
i f
i BYRON - UNITS 1 & 2 6-22b AMENDMENT NO. 98
.ps>*%g j+
4 UNITED STATES g
j NUCLEAR REEULATORY COMMISSI2N WASHINGTON, D.C. SceeHe01
\\...../
COMMONVvt-ALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO.1 6MENDMENT TO FACILITY OPERATING LIC5 HEE Amendment No. 89 Ucense No. NPF-72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated May 21,1997, as supplemented by lettes dated November 18, 1997, December 3,1997, January 8,1998 and January 13,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules ar"1 regulations set forth in 10 CFR Chapter I; B.
The facility w;ll operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amenument can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
I D
e 2-(2)
Technical Specifcations The Technical Specirmations contained in Appendix A as revised through Amendment No. 89 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifestions and the Environmental Protection Plen.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
.)Mt Georg
. Dick, Seni Project Manager Project Directorate ill 2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation
Attachment:
l Changes to the Technical Specifications Date of lasuance: January 23, 1998 J
ja neg4 k
UNITED STATES.
g NUCLEAR REGULATORY COMMISSION WAsHINeToN D.C. segeHeM
- ...+-
COMMONWEALTH EDISON COMPANY DOCKET No. 8TN 50 457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 89 License No. NPF-77
- 1, The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendmmt by cot.T,0,;;;n"f, Edison Compny (the licensee) dated May 21,1997, as supplemented by letters dated November 18, 1997 December 3,1997, January 8,1998 and January 13,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
Thww is reasonable assurance (i) that the activities authorized by this Amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the i
Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and -
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 vf the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as -
indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amendec; to read as follows:
l
2 (2)
Technical Specificatirn The Technical apeafications contained in Appendix A as revised through Amendment No.- 89 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No, NPF-72, dated JuY 2, 1987, are hereby incorporated into this license, The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date if its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION George Dick, oenior P o oct Manager Project Directorate 1112 Division of Reactor Projects - lil/lV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical
' pecifications Date af issuance: January 23, 1998
ATTACHMENT TO LICENSE AMENDMENT NOS. 89 AND 89 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50 457 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
8mmqvs_P_naal insert Paoes l
I Vill Vill XV XV 1-4a 3/4 4-32 3/4 4-32 3/4 4-33 3/4 4-33 3/4434 3/4 4-34 3/4 4-35 3/4435 l
3/4436 3/4 4-36 l-3/4 4 3/4 4-37 3/4438 3/4439
$4440 3/4 4 40.a 3/4/4-41 3/4 4-42 3/4 4-43 B 3/4 4 7 8 3/4 4-7 83/44-8 8 3/4 4-8 0 3/4 4 B 3/4 4 9 83/44-10 83/44-10.
B 3/4 4-11 '
B 3/4 4-11 B 3/4 4-12 D 3/4 4-13 '
03/44-14 B 3/4 4~15 B 3/4 4-16 B 3/4 4-17 6-22b 6-22b -
a IhDZZ DEFINITIONE D
SECTION R&Eg 1.0.
DEFINITIONS 1.1 ACTION........................................................
1-1 1.2 ACTUATION LOGIC TEST.......,..................................
1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................
1-1 1.4 AXIAL FLUX DIFFERENCE.........................................
1-1 1.5 CHANNEL CALIBRATION...........................................
1-1 1.6 CHANNEL CHECK.................................................
1-1 1.7 CONTAINMENT INTEGRITY.........................................
1-2 1.8 CONTROLLED LEAKAGE............................................
1-2 1.9 CORE ALTERATION...............................................
1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................
1-2 1.11 DOSE EQUIVALENT t-131........................................
1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................
1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................
1-3 1.14 FREQUENCY NOTATION...........................................
1-3 1.15 TDENTIFIED LEAKAGE...........................................
1-3 1.15.4 L,..........................................................
1-3 1.16 MA S T E R R E LAY T E S T............................................
1-3 1.17 nEMBEH(S) OF THE P0BLIC......................................
1-3 1.18 OFFSITE DOSE CALCULATION MANUAL..............................
1-4 1.19 OPERAHLE - OPERABILITY.......................................
1-4 1.19.a OPERATING !.IMITS REPORT.....................................
1-4 1.20 OPERATIONAL MODE - MODE......................................
1-4 1.20.a P,..........................................................
1-4 1.21 PHYSICS TESTS................................................
1-4 1.22 PRESSURE BOUNDARY LEAKAGE.
1-4 l
1.22.a PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)................
1-4a 1.23 PROCESS CONTROL PROGRAM......................................
1-5 1.24 PURGE - PURGING..............................................
1-5 1.25 QUADRANT POWER TILT RATIO....................................
1-5 1.26 RATED THERMAL POWER..........................................
1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................
1-5 1.28 REPORTABLE EVENT.............................................
1-5 BRAIDWOOD - UNITS 1 & 2 I
AMENDMENT NO. 89
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...............
3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS........................................
3/4 4-26
]
3/4.4.s SPECIFIC ACTIVITY........................................
3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
>1#CI/ GRAM DOSE EQUIVALENT I-131....................
3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM....................................
3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...................................
3/4 4-32 Pressur1 er..............................................
3/4 4-33 ove rpres sure Protection Systems..........................
3/4 4-34 3/4.4.10 STRUCTURAL INTEGRITY.....................................
3/4 4-36 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................
3/4 4-37 dRAIDWOOD - UNITS 1 & 2 VIII AMENDMENT NO. 89
mAsEs sECTIoN BEE 3/4.4.5 STEAM oENERAToRS..........................................
B 3/4 4-3 3/4.4.6 REACTOR OOOLANT SYSTEM LEAKAGE............................
B 3/4 4-4 3/4.4.7 cuRMrsTRY.................................................
B 3/4 4-5
]
3/4.4.8 SPECIFIC ACTIVITY.........................................
B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............................
B 3/4 4-7 3/4.4.10 STRUCTURAL INTEGRITY.....................................
B 3/4 4-10 3/4.4.11 REACTOR VESSEL HEAD VENTS................................
B 3/4 4-10 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS..............................................
B 3/4 5-1 3/4.5.2, 3/4.5.3 and 3/4.5.4 ECCS SUBSYSTEMS.......................
B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK..............................
B 3/4 5-4 5O w.e CONTA1Nx m SYS1 ENS 3/4.6.1 PRIMARY CONTAINMENT.......................................
B 3/4 6-1 3/4.6.2 LEPRESSURIEATION AND COOLING SYSTEMS......................
B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES..............................
B 3/4 6-4 3/4.6.4 COMBUSTIBLE CAS CONTROL...................................
B 3/4 6-4 BRAIDWOOD - UNITS 1 & 2 XV AMENDMENT NO. 89
DEFINITIONS i
PRESSURE AND TEMPERATURE LIMITS REPORT fpTLR) 1.22.a The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits including heatup and cooldown rates.
and the pressurizar power operated rellef valve (PORV current reactor vassel fluence period.. These pressure) lift settings for theand temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.11.
Unit operation within these limits is addressed in LCO 3.4.9.1,
" Pressure / Temperature Limits," and LCO 3.4.9.3, " Overpressure Protection Systense" BRAIDWOOD - UNITS 1 & 2 1-4a AMENDMENT NO. 89 l
. REACTOR COOLANT SYSTEM 3/4.4.g PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM
+
LIMITING CONDITION FOR OPERATION i
L 3.4.9.1 The Reactor Coolant System temperature and pressure, and heatup and L
cooldown rates shall be==intained in accordance with the limits specified in the PTLR.=
APPLICABILITY: At:all times.
ACTION:
With any of the limits'in the PTLR exceeded, restore the temperature and/or j'
pressure to within the limit within 30 minutes;. perform an engineering t
evaluation to determine the effects of the out-of-limit condition on the' structural' integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than200*F-and500psig,respectively,Hwithinthe_folhing30 hours.
SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor' Coolant System temperature and pressure shall be determined to be within the'11mits at least once per 30 minutes during system
'heatup, cooldown, and inservice leak and hydrostatic testing operations._
4.4.9.1.2-The reactor vessel material-irradiation surveillanc.e specimens shall be' removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H,- in accordance with the schedule in the'PTLR.-
M BRAIDWOOD - UNITS 1 L 2 3/4 4-32 AMENDMENT NO. 89
l REACTOR COOLAlt? SERIEN i
-)
PREBBURIEES I turTruc evwnITIon pom opEmATrow 3.4.9.2.The pressuriser temperature shall be limited.tos A maximum heatup of 100*F in any.1-hour period, a.
b.
A maximum cooldcwn of 200*F in any 1-hour period, and c.
A maximum spray water temperature differential of 320*F.
APPLICABILITY: At all times.
ACTION:
With the pressuriser temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutest-perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressuriser; determine that the pressuriser remains acceptable for continued operation or be in at least 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 reduce the pressuriser pressure to less than 500 peig within the followir.g 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVIILLANCE REoDIREMENTE 4.4.9.2 The pressuriser temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.
BRAIDWOOD - UNITS 1 & 2 3/4 4-33 AMENDMENT NO. 89
. REACTOR COOLhMT SYS EM OVERPREASURE PROTECTION SYSTEMS LYMYTING NITIOtf FOR OPERATION 3.4.9.3 At least two overpressure protection devices shc11 be OPERABLE, and each device shall be either:
a.
A residual heet removal (RNR) suction relief valve with a lift setting of less than or equal to 450 peig, or b.
A power operated relief valve (PORV) with a lift setpoint that varies with RCs temperature which doe ~s not-exceed the limit established in the PTLR.
hPPLICARILITY: NOOES 4, 5, and 6 with the reactor vessel head on.
ACTION:
a.
With one of the two required overpressure protection devices inoperable in NODE 4, restore two overpressure protection devices to OPERABLE status within 7 days or depressurise and-vent the RCs through at-least a 2 square inch vent within the next-8 hours.
b.
With one of the two required overpressure protection devices inoperable in MODES 5 or 6, restore two overpressure protection devices to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
With both of the required overpressure protection devices inoperable, c.
depressurize and vent the RCS through at least a 2 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d.
With the RCS vented per ACTION 8 a, b, or c, verify the vent pathway at least once per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; etherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, e.
In the event either the PORVs, RHR suction relief valves or the RCS e
vents are used to mitigate an RCS-pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, RHR suction relief valves, or RCS vents on the transient, and any corrective action necessary to prevent recurrence, f.
The provisions of Specification 3.0.4 are-not applicable.
BRAIDWOOD - UNITS 1 & 2 3/4 4-34 AMENDMENT NO. 89 i
$1&CTOR 000LhMT SYSTEM aumvmitthuca amournew'Nis 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE when the PORVs are being uaed for cold overpressure protection by a.
Performance of an ANALOG CHANNEL OPERATIONAL TEST on-the PORV actenation channel, but excluding valve operation, at least once per 31 days when the POR',' le seguired OPERABLE; and b.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and Verifying the PORV isolation valve jo open at least once per c.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.4.9.3.2 Bach MR suction rslief valve shall be demonstrated OPERABLE when the RNR suction relief valves are being used for cold overpressure protection as follows:
For RHR suction relief valve RH87083 verify at least once per a.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that valves RH8702A and RH87028 are open.
b.
For RHR suction relief valve RH"708A verify at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that vaivaa RH8701A and RH87013 are open, c.
Testing pursuant t3 specification 4.0.5.
J BRAIDWOOD - UNITS 1 & 2 3/4 4-35 AMENDMENT NO. 89 1
i
SEACTea QQQ1 ANT SYSTEM 314. 4.10..... ATRUCTURAL INTEGRITY I
r_rurtrea naamirrow yon opemattow L
3.4.10 The structural integrity wt AsME Code Class le 2, and 3 componente shall be maintained in accordance with specification 4.4.10.
APPLICARILITY All MODES.
ACTION:
a.
With the structural integrity of any AsME Coc.a Class 1 component (s) not conforming to the above requ aements, restore the structurtti i
integrity of the affected componentis) to within its limit or isolate the affseted component (s) prior to increasing the Reactor coolant system temperature above 200*F.
b.
With the structural integrity of any AsMt Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or inclate the affected component (s) prior to increasing the Reactor Coolant system temperature above 200*F.
c.
With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolata the affected component (s) from service.
d.
The provisions of specification 3.0.4 are not applicable, aumvEf t.t_muCE REoUIREMENTS 4.4.10 In addition to the requirements of *>ecification 4.0.5, each reactor coolant pump flywheel shall be inspected as follows:
a.
Volumetric examination of the areas of higher stress concentration at the bore and keyways will be performed each 40 month period during refueling or maintenance shutdowns coinciding with the service inspection schedule as required by section XI of the ASME Code.
b.
Visual examination of all suposed nurfaces will be performed and a surface examination of the bore and keyway surfaces will be perfarmed l
whenever the flywheels are removed for maintenance purposes, but not more frequently than once each 10 year interval.
l BRAIDWOOD - UNITS 1 fi 2 3/4 4-36 AME2 MENT no. 89 l
l
._, _. _. _ _... ~. _
REactea coolant SYSTEM 3/4.4e11...Rancrom QQDLhBT SYSTEM YENTS j
i j
f1MfT1ano N IT1olf FOR OPERATION i
a i
3.4.11 At least one reactor vessel head vent path consisting of two valves in series powered. rom emergency busses shall be OPERABLE and closed.
&PFLICARILITY: MODES 1, 2, 3 and 4.
&EZ1 gel i
With the above reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is anain-i tained closed with power removed from the valve actua+-** of all the valves j
in the inoperabio vent paths restore the inoperable
. path to OPERABLE status within 30 days, or, be in NOT STANDBY within
..sure and in COLD 4
l SNUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l 5
SURVEILLANCE REQUIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:
]
a.
Verifying all manual isolation valves in each vent-path are locked in 1
the open position, b,
cycling each valve in the vent path through at lu st one complete l
cycle of full travel from the control room during COLD SHUTDOWN or REFUELING, and i
c.
Verifying flow through the reactor vessel head vent paths during venting operations at COLD SHUTDOWN or REFUELING.
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- BRAIDWooD - UNITS 1 a 2 3/4 4-37 AMENDMENT Mo. 89 l
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REAC10R C00LANT $Y$ TEN RAMS SPECIFICACTIVITY(Continued) take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomenon. A reduction-in frequency of isotopic anslyses following power changes may be permissible if
- justified by the data obtained.
3/4.4.9 PRESSURE /TEMPEl@TURE LIMITS 3
~
l All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. -These loads are introduced by reactor trips. )LC0 3.4.g.1and shutdown (cocidown) operations, power transients, and startup (heatup
" Pressure /Temperatt.re Limits," limits the pressure i-j and temperature changes durIng RCS heatup and cooldown to within the design j
-assumptions and the stress-limits for cyclic operation.
)
The PRES $URE AND TEMPERATURE LIMITS REPORT PTLR contains pressure and temperature (P/T) limit curves for heatup, cooldo(wn, )nservice leak and i
hydrostatic (ISLH) testing, and data' for the maximum rate of change of reactor coolant - temperature.
Each P/T limit curve defines an acceptable region for normal operation.
The usual use of the curves it operational guidance during heatup or cooldown maneuvering, when pressure and temperature Indications are monitored and compared to the applicable curve to determine that operation is within the
-allowable region.
i B'
LC0 3.4.g.1 establishes operating limits that provide.a margin to non-ductile failure of the reactor vc:::I
.d piping of the Reactor Coolant Pressure l.
Boundary (RCPB).
The vessel is the component most subject to non-ductile-failure, and the LCO limits apply to the entire RCS, except the pressurizer, which has different design characteristics and operating functions, 2'
f 10 CFR-Part 50, Appendix G, requires the establishment of P/T limits for specific material-frtcture toughness requirements of the RCP8 materials.
Appendix 6 of 10 CFk Part 50 requires an adequate margin to non-ductile failure during. normal operation, anticipated operational occurrences, and system e
hydrostatic tests.
It s>andates the use of the American Society of Mechanical Engineers (ASME) Code,Section XI,1.ppendix G.
l I
The neutron embrittlement effect on the material toughness is reflected by increasing the Nil Ductility Reference Temperature (RT
) as exposure to neutron fluence incrsases.
4 The actual shift in the RT of the vessel material. will be established periodically by removing and ev"aYuntinithe irradiated reactor vessel material
-specimens, in accordance with ASTM E 185 and_ Appendix H of 10 CFR Part 50. The e0 operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99,
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Revision 2.
BRAIDWOOD - UNITS 1-& 2 B 3/4 4-7 AMENDMENT NO. 89
- JtEACTOR COOLANT SYSTEM IAE PRESSURE /TEMPERATURELIMITS(Continued)
The P/T limit cutves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, ar.d temperature rate of change, one location within the reactor vessel will dictate the moat restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive and, thus, the curves are composites of the most restrictive regions.
The heatup curve represents a different set of restrictions than the cooldown curve because t1e directions of the thernal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls during heatup and cooldown, respectively.
The criticality limit curve includes the 10 CFR Part 50, Appendix G.
requirement that it be ;t 40'F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4, " Minimum Temperature for Cr'ticality."
The consequenca of violating the LCO limits is that the RCS has been operated under cond' ions that can result in non-ductile failure of the RCPB, possibly leading to a nonisclable leak or loss-of-coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code,Section XI, Appendix E, provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
OVERPRESSURE PROTECTION SYSTEMS The Low Temperature Overpressure Protection (LTOP) System controls the RCS pressure at low temperatures so the integrity of the RCPB is not compromised by violating the P/T limits of 10 CFR Part $0, Appendix G.
The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the maxirum allowable actuation logic setpoints for tha pressurizer Power-Operated Relief Valves (PORVs) and the maximum RCS pressure for the 6xisting RCS cold leg temperature during cooldown, shutdown, and heatup to meet the 10 CER Part 50, Appendix G, requirements during the MODES in which the LTOP system is necessary.
The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and the material becomes less resistant to stress at low temperatures.
RCS pressure, therefore, is maintained low at low temperatures and is increased only within the limits specified in the PTLR.
3 The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only during shutdown; a pressure fluctuation een occur more quickly than an operator can react to relieve the condition.
Exceeding the RCS P/T limits by a significant amount could cause non-ductile failure of the reactor vessel.
LCO 3.4.9.1, " Pressure / Temperature Limits," requires administrative control of RCS pressure ad temperature during heuup and cooldown to prevent exceeding the PTLR.tmits.
BRAIDWOOD - UNITS 1 & 2 8 3/4 4-8 AMENDMENT NO. 89
,' E&GES COOLANT SYSTEM 4ASEE PRES $URE/ TEMPERATURE LIMITS (Continued)
LCO 3.4.g.3 *0verpressure Protection Systems," provides RCS overpressure protection by havIng a minimum coolant input capability and having adcquate pressure relief capacity.
Limiting coolant input capability requires all Safety Injection ($1 pumps and all but one charging pump a centrifugal charging pus.p) incapable of njection into the RCS and i olation o the $1 accumulators. The pressure-relief capacity requires either u o redundant RCS relief valves, or a depressurized RCS and an RCS vent of sufficient size. One RCS relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.
With minimum coolant insut capability, the ability to provide core coolant
-addition is restricted.- The.00 does not require the makeup control system deactivated or the 51 actuation circuits blocked. Due to the lower pressures in the LiOP MODES ahi the expected core decay heat levels, the makeup system can l-provide adequate flow via the makeup control valve.
If conditions require the use of more than one centrifugal charging pump for makeup in the event of loss of inventory, then pumps can h made available through manual actions.
The LTOP Systas for-pressure relief consists of two PORVs with reduced lift settings, or two Residual Heat Removal (RHR) suction relief valves, or one PORY and one RHR suction relief valve, or a depressurized RCS and a RCS vent of sufficient size. Two RCS relief valves are required for redundancy. One RCS relief valve has adequate relieving capability to prevent ovarpressurization for the required coolant input capability.
PORV Renuirements As designed for the LT0P System, each PORV is signaled to open if the RCS pressure approaches a limit determined by the LTOP actuation logic. The LTOP actuation logic monitors both RCS temperature and RCS pressure and deteuines when a condition is approaching the PTLR limits. The wide range RCS temperature indications are auctioneered to. select the lowest temperature signal.
The lowest temperature signal is processed through a function generator l
that calculates a pressure limit for that temperature. The calculated pressure limit is then com pressure channel. pared with the indicated RCS pressure from a wide range If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.
The PTLR presents the PORV setpoints for the LTOP system. The setpoints are normally staggered so only one valve opens during a low temperature overpressare transient. Having the setroints of both valves within the limits in the PTLR ensures that the 10 CFR Part 50, Appendix G, limits will not be exceeded in any analyzed event.
c When a PORY is opened in an increasi.ng pressure transient, the subsequent relief will_ cause the pressure increase to slow and reverse. As the PORY _-
releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is' signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-g AMENDMENT NO. 89
ptEACTOR QQQL&MT SYSTEM mamma PREASURE/ TEMPERATURE LIMITS (Continued) num Suetion Ralief valva Ramuir-- ats During the LTOP N00Es, the RNR system is operated for decay heat removal and low pressure letdown control. Therefore, the RNR suction isolation valves are open in the piping from the RCs hot lege to the inlets of the RNR pumps.
While these valves are open, the RNR suction relief valves are exposed to the RCs and are able to relieve pressure transients in the RCs.
The RNR tuction isolation valves must be open to make the RNA suction relief valves OPERABLE for RCs overpressure mitigation. The RHR suction relief valves are spring loaded, bellows-type relief valves with pressure tolerances and accumulation limite established by section III of the American society of Nechanical Engineers (AsME) Boiler and Pressure Vessel Code for Class 2 relief valves.
RCS Vant_Raouirements Once the RCs is depressurised, a vent exposed to the containment atmosphere will maintair. the RCS at containment ambient-pressure in an RCs overpressure transient, if the relieving requirements of the transient do not excSed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.
For an RCs vent to meet the flow capacity requirement, it requires removing a pressuriser safety valves removing a PORV's internals, and disabling its block valve in the open positions or similarly estaolishing any comparable vent.
The vent path (s) atot be above the level of reactor coolant, so as not to drain the RCs when open.
i 3/4.4.10 STRUCTURAL INTECRITY The inservice inspection and testing programs for ASME Code Class., 4
.and 3 componente ensure that the structural integrity and operational readiteas of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with section XI of the ASME Boiler and Pressure vessel Code and applicable Addenda as required by 1: JFR Part 50.55a(g) except where specific written relief has been granted by the Commission purusant to 10 CFR Part 50.55a(g)(6)(1).
Components of the Reactor Coolant system were designed to provide access to permit inservice inspections in accordance with section XI of the AsME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through summer 1975.
3/A.4.11 REACTOR VESSEL HEAD..YENTS Reactor Coolant systum vents are provided to exhaust noncondensible gases and/or steam from the Reactor coolant system that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel houd vent path ensures the spability exists to perform this function.
DRAIDWOOD - UNITS 1 & 2 2 3/4 4-10 AMENDMENT NO. 89
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REACTOR COO 1 ANT SYST4 maats 3/4.4.11 RFACTOR VESSEL HEAD VENTE (Continued)
The valve redundancy of the Reactor Coolant system vent paths serves to minimise.the probability of inadvertent or irreversible actuation while ensuring that a single vent valve power supply or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor coolant system Vent'8s'stoms are consistent with the requirements of Item 11.3.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," Now d er 1980.
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4 BRAIDWOOD - UNITS 1 & 2 8 3/4 4-11 AMENDMENT NO. 89
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REPORTIMO REQUIREMENTS fCentinuedI i
QPERATIMO LIMITS REPORT fContinuedi 2
9, WCAP-10054-P3A, " Westinghouse small Break BCCS Evaluation Model Using the NOTRUNP Code," dated August 1985 (Westinghouse Proprietary).
(Nethodology for specification: Axial Flux Difference, Neat Flux Mot Chmnel Factor, and Nuclear Enthalpy Riog Hot Channel Factor).
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I i
10.
Consd letter from D. Saccomando to the office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that i
documents applicable sections of WCAP-11992/11993 and Cosed application of the UET methodology addressed in " Additional Information Regarding l
Application for Amendment to Pacility Operating Licensee-Reactivity Controls systems."
The operating limits shall be determined so that all applicable limite (e.g.,
4 fuel thermal-mechanical limits, core thermal-hydraulic limite, 3CCs limits, nuclear limits such as sNUTDoWN MARGIN, and transient and accident analysis l
limite) of the safety analysis are met.
The OPERATImo LIMITS REPORT, including any mid-cycle revisions or bupplements 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.
- Nar40R Nf hM SYSTEM fRCB) PREBEURE AND TEMPERATURE LIMITS REPORT iPTLRe 6.9.1.10 RCs pressure and temperature limits for heatup, cooldown, low temperature operation, criticality,_and hydrostatic testing, as well as heatup and cooldown rates'and pressuriser power-operated relief valve lift settings, shall be established-and documented in the PTLR for the following;
- 1) LCO 3.4.9.1, " Pressure / Temperature Limits," and 2; Loo 3.4.9.3, " overpressure Protection systems."
The analytical methods used to determine the RCs pressure.ad temperature limits shall be those reviewed and aoproved by the NRC, specifically those described in NRC letter dated January 21, 1998, "Syron Station, Units 1 and 2, and traitawood station, Units 1 and 2, Peceptance for Referencing of Pressure Temperature Limits Report."
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any ravision or supplement thereto.
mRAIDWoop - UNITS 1 s 2 6-22b AxrNDurut No. 89
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