ML20065N611
| ML20065N611 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/15/1994 |
| From: | West G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065N613 | List: |
| References | |
| NUDOCS 9404270314 | |
| Download: ML20065N611 (17) | |
Text
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/t UNITED STATES
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NUCLEAR REGULATORY COMMISSION (5 jg lf 3
WASHINGTON, D.C. 20555 0001 TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET N0. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.186 License No. NPF-3 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Toledo Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated July 28, 1992, as supplemented on February 17, 1993, 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.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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-3 is hereby amended to read as follows:
9404270314 940415 PDR ADOCK 05000346-P PDR
. (a) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 186, are hereby incorporated in the license.
The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ld.: pnm:w h' h! f i
Garmon West, Asst. Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: April 15, 1994 i
l
)
ATTACHMENT TO LICENSE AMENDMENT NO, 186 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 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 tite area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Insert VIII VIII XIII XIll 3/4 3-12 3/4 3-12 3/4 3-22 3/4 3-22 3/4 9-4 3/4 9-4 3/4_9-9 3/4 9-9 B 3/4 9-1 B 3/4 9-1 B 3/4 9-2 B 3/4 9-2
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................
3/4 9-2 3/4.9.3 DECAY TIME................................................
3/4 9-3 3/4.9.4 CONTAINMEN1 PENETRATIONS..................................3/4 9-4 3/4.9,5 COMMUNICATI0flS............................................
3/4 9-5 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY..........................
3/4 9-6 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING.....................
3/4 9-7 3/4.9.8 DECAY HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels..........................................
3/4 9-8 Low Water Levels..........................................
3/4 9-8a 3/4.9.9 0ELETED...................................................
3/4 9-9 3/4.9.10 WATER LEVEL - REACTOR VESSEL..............................
3/4 9-10 3/4.9.11 STORAGE POOL WATER LEV 2L..................................
3/4 9-11 3/4.9.12 STORAGE POOL VENTILAT10N..................................
3/4 9-12 3/4.9.13 SPENT FUEL POOL FUEL AS3EMBLY ST0 RAGE.....................
3/4 9-13 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS...................................
3/4 10-1 3/4.10.2 PHYSICS TESTS.............................................
3/4 10-2 3/4.10.3 REACTOR COOLANT L00PS.....................................
3/4 10-3
-l 3/4.10.4 SHUTDOWN MARGIN...........................................
3/4 10-4 DAVIS-BESSE, UNIT 1 Vill Amendment No. AE, JAS,186
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREME SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE S a f e ty V a 1 v e s..............................................
3/ 4 7 - 1 A u x i l i a ry Fe edwa t e r Sy s t em.................................
3/ 4 7-4 Condensate Storage Tanks...................................
3/4 7-6 l
A c t i v i ty................................................... 3 / 4 7 - 7 Main Steam Line Isolation Va1ves...........................
3/4 7-9 Motor Driven Feedwater Pump System.....................
3/4 7-12a 3/4.7.2 STE AM GENERATOR PRESSURE / TEMPERATURE LIMITATION.....
3/4 7-13 3/4.7.3 COMPONENT COOLING WATER SYSTEM.............................
3/4 7-14 3/4.7.4 S E RVI C E WATE R S YSTEM.................................
3/4 7-15 3/4.7.5 ULTIMATE HEAT SINK.........................................
3/4 7-16 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATIONSYSTEM..................
3/4 7-17 3/4.7.7 S N U B B E RS...............................................
3/ 4 7-20 3/4.7.8 SEALED SOURCE CONTAMINATION................................
3/4 7-36 3/4.7.9 Deleted 13/4.7.10 Deleted 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0 p e ra t i n g..................................................
3 / 4 8 - 1 S h u t d o w n.................................................
3/ 4 8-5 3/4.8.2 DNSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating..............................
3/4 8-6 A.C. Distribution - Shutdown...............................
3/4 8-7 D. C. Dis tribu tion - Opera ti ng..............................
3/4 8-8 D.C. Distribution - Shutdown...............................
3/4 8-11 DAVIS-BESSE, UNIT 1 VII AmendmentNo.25,J03J06,J35,)ff,174 SEP 2 21992
INDEX BASES SECTION 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY......................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING.................... B 3/4 9-2 3/4.9.8 COOLANT CIRCULATION...................................... B 3/4 9-2 3/4.9.9 DELETE0.................................................. B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTO STORAGE P00L..................R VESSEL AND
........................ B 3/4 9-2 3/4.9.12 STORAGE P0OL VENTILATION................................. B 3/4 9-3 3/4.9.13 SPENT FUEL POOL FUEL ASSEMBLYSTORAGE.................... B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.... B 3/4 10 3/4.10.2 PHYSICS TESTS............................................ B 3/4 10-1 3/4.10.3 REACTOR COOLANT L00PS.................................... B 3/4 10-1 3/4.10.4 SHUTDOWN MARGIN.......................................... B 3/4 10-1 3/4.11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID HOLDUP TANKS..................................... B 3/4 11-1 3/4.11.2 EXPLOSIVE GAS MIXTURE.................................... B 3/4 11-1 3/4.12 DELETED
\\
DAVIS-BESSE, UNIT 1 XIII Amendment No. $9,JJ9,JJS,J79,186
INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area.................................................
5-1 L ow Po pu l a t i o n Zon e............................................
5-1 S i t e B o u n d a ry..................................................
5-1 l
5.2 CONTAINMENT Configuration..................................................
5-1 1
Des i gn Pres su re a nd Tempera ture................................
5-4
_S. 3 REACTOR CORE l
}
f u e l As s emb l i e s................................................
5-4 C o nt rol Rods...................................................
5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature................................
5-4 V o 1 um e.........................................................
5 - 5 5.5 METEOROLOGICAL TOWER LOCATION..................................
5-5 5.6 FUEL STORAGE C r i t i c a l i ty..................................................... 5-5 Drainage.......................................................
5-5 Capacity.......................................................
5-6 i
5.7 COMPONE NT CYCL IC OR TRANSI ENT L IMIT............................
5-6 DAVIS-BESSE, UNIT 1 XIV Amendment No. 3S 170 MAR 9 1932
TABLE 3.3-3 (Continued)
TABLE NOTATION Trip function may be bypassed in this MODE with RCS pressure below 1800 psig.
Bypass shall be automatically removed when RCS pressure exceeds 1800 psig.
Trip function may be bypassed in this MODE with RCS pressure below 600 psig.
Bypass shall be automatically removed when RCS pressure exceeds 9 0 psig.
One must be in SFAS Channels #1 or #3, the other must be in Channels
- 2 or #4.
This instrumentation, or the containment purge and exhaust system noble gas monitor (with the containment purge and exhaust system in operation), must be OPERABLE during CORE ALTERATIONS or movement of irradiated fuel within containment to meet the requirements of Technical Specification 3.9.4.
When using the containment purge and exhaust system noble gas monitor, SFAS is not required to be OPERABLE in MODE 6.
All functional units may be bypassed for up to one minute when starting each Reactor Coolant Pump or Circulating Water Pump.
When either Decay Heat Isolation Valve is open.
The provisions of Specification 3.0.4 are not applicable.
LCIION STATEMENTS ACTION 10 -
With the number of OPERABLE functional units one less than the Total Number of Units, STARTUP and/or POWER OPERATION may proceed provided both of the following conditions are satisfied:
The inoperable functional unit is placed in the tripped a.
condition within one hour.
For functional unit 4a the sequencer channel shall be placed in the trip by physical removal of the sequencer module, ped condition b.
The Minimum Units OPERABLE requirement is met; however, one additional functional unit may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.1.
ACTION 11 -
With any component in the Output Logic inoperable, trip the associated components within one hour 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION 12 -
With the number of OPERABLE Units one less than the Total Number of Units, restore the inoperable functional unit to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION 13 -
a.
With less than the Minimum Units OPERABLE and reactor coolant pressure 2 438 psig, both Decay Heat Isolation Valves (DH11 and DH12) shall be verified closed.
DAVIS-BESSE, UNIT 1 3/4 3-12 Amendment No. SS 37, 52, JpJ, M, D.9,188
?
- 5 Y'
en
'U TABLE 3.3-3 (Continued)
SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E
Q FUNCTIONAL UNIT TOTAL NO.
UNITS MINIMUM UNITS OF UNITS TO TRIP APPLICABLE w
_ MODES A_CTION 3.
MANUAL ACIllATION SFAS (except Containment a.
Spray and Emergency Sump Recirculation) 2 2
2 b.
Containment Spray 1,2,3,4,6****
12 2
2 2
4.
SEQUENCE LOGIC CHANNELS 1,2,3,4 12 a.
Sequencer 4
2***
3 1,2,3,4 10f b.
Essential Bus Feeder Breaker Trip (90%)
2 1
,k 2*****
1,2,3,4 Diesel Generator Start, e.
155 o
Load Shed on Essential Bus (59%)
2 1
E,"
5.
2 INTERLOCK CHANNELS 1,2,3,4 158 m
o e
g Decay Heat Isolation Valve a.
1 1
hh 1
1,2,3 13 f l b.
Pressurizer Heaters k
2 2
2 3******
14 Ll k
~
mti
g 5
TABLE 4.3-2 (Continued)
T SAFETY FEATURES ACTUATION SY_ STEM INSTRUMENTATION SU
- Ur CHANNEL MODES IN WHICH c
_ FUNCTIONAL UNIT CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE
=
CHECK CALIBRATION TEST REQUIRED
~
5.
INTERLOCK CHANNELS Decay Heat Isolation Valve a.
S b.
Pressurizer Heater R
S R
1, 2, 3 3 ##
- See Specification 4.5.2.d.1 t'
4 Y
TABLE NOTATION y
(1)
Manual actuation switches shall be tested at least once per 18 months during shutdown circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNC All other least once per 31 days.
(2) pressure to the appropriate side of the transmitter.The CHANNEL FUNCTIONA acuum or These surveillance requirements in conjunction with those of Section 4 9 4 apply during oyg ALTERATIONS or movement of irradiated fuel within the containment only if using the SFAS a
radiation monitors listed in Table 3.3-3, Items la, 2a, and 3a, in lieu of the containmen area exhaust system noble gas monitor.
rge and
[z When either Decay Heat Isolation Valve is open.
P.*
5
- - - - - - - - - - ^ ' ^ ~ ^ ^
E TABLE 4.3-2 Y
SAFETT FEATURES ACTUATION SYSTEM INSTRUMEtTTATION SURVEI N
{
N."
g CHANNEL MODES IN VHICH q
_ FUNCTIONAL UNIT CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST
~
1.
INSTRUMENT STRINGS REQUIRED
' Containment Radiation - High a.
S R
H b.
Containment Pressure - High S
R 1,2,3,4,68 l.
Containment Pressure - High-High M(Z) c.
.RCS Pressure - Low R
1, 2, 3 S
d.
1 M(2) 1, 2, 3 S
RCS Pressure - Low-Lov R
H e.
S 1,2,3 f.
BUST Level - Low-Low R
M 1,2,3 s
S R
M
[
2.
OUTPtfr LOGIC 1, 2, 3 I
I, Incident Level 01: Containment a.
Isolation S
R b.
Incident Level 82: High Pressure M
1,2,3,4,65 Injection and Starting Diesel Generators S
R M
c.
Incident Level 93: Low Pressure Injection 1,2,3,4 EI d.
Incident Level 54: Containment H
S R
[
Spray 1,2,3,4 S
R e
Incident Level SS: Containment M
e.
1,2,3,4 Sump Recirculation Permissive S
R H
i f
3.
MANUAL ACTUATION 1,2,3,4 sey a.
SFAS (Except Containment Spray NA NA g
and'Energency Sump Recirculation)
M(1) 1,2,3,4,68 b.
-k Containment. Spray NA NA M(1) 1, 2, 3 4.
SEQUENCE LOGIC CHANNELS
)
S NA M
1, 2, 3, 4
~~~
- - - - ~
- - ~ ~ - -
REFUELING OPERATIONS CONTAINMENT PENETRATIONS.
LIMITING CONDITION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:
bolts,The equipment door closed and held in place by a minimum of four a.
b.
A minimum of one door in each airlock closed, and atmosphere to the outside atmosphere shall be either:Each c.
1.
Closed by an isolation valve, blind flange, or manual valve, or 2.
exhaust isolation valve.Be capable of being closed by an OPERABLE i
1
' APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION.
With the requirements of the above specification not sa l
a.
of irradialed fuel in the containment.
3.0.3 are not 6pplicable.
The provisions of Specification b.
With the containment purge and exhaust isolation system inoperable, close each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere.
The provisions of Specification 3.0.3 are not applicable.
c.
SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above re OPERABLE containment purge / quired containment penetrations s to be either in its closed isolated condition or capable of being closed by an and exhaust valve within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during, CORE ALTERATIONS or movement l
irradiated fuel in the containment, by:
3 l
Verifying the penetrations are in their isolated condition, or a.
b.
Verifying that with the containment purge and exhaust system in operation and the containment pu ge and exhaust system noble gas monitor ca,pable of providing a hi h radiation signal to the control i
room, that after initiation of th high radiation signal the control room, orcontainment purge and exhaust isolation valves can be clo, sed f Containment Purge and Exhaust Isolation test signalIf usin exhaustisolationvalveautomaticallyactuatestoilsisolationeach purge and position.
DAVIS-BESSE, UNIT 1 3/4 9-4 Amendment No.186
I gFUElfNG OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 c
The reactor shall be suberitical for at lea APPLICABILITY:
ours.
During movement of irradiated fuel in the react pressure vessel.
or ACTION:
With the reactor subcritical for less than 72 h operations involving movement of irrad vessel.
ours, suspend all e reactor pressure p cable.
SURVEILLANCE REQUIREMENTS 4.9.3 least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by verification of the date and ti critical for at prior to movement of irradiated fuel in the reactore of subcritic pressure vessel.
I IMIS-BESSE, UNIT 1 ti i
3/4 9 3
I Text for this page was Deleted 1
)
DAVIS-BESSE, UNIT'l 3/4 9-9 Amendment No.186
.l 1
' REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL
\\
j LIMITING CONDITION FOR OPERATION 3.9.10 As a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel.
APPLICABILITY:
within the reactor pressure vessel while in MODE 6.During mo ACTION:
all operation involving movement of fuel assemblie within the reactor pressure vessel.
are not applicable.
The provisions of Specification 3.0.3 SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minim required depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during movement of fuel assemblies or control rods within the reactor pressure vessel.
s"
, DAVIS-BESSE, UNIT 1 3/4 9 10
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION
- 1) the reactor will remain subtritical during CORE ALTERATION volumes having direct access to the reactor vessel. uniform boron and 2) a These limitations are consistent with the initial conditions assumed for the boron dilutio in the accident analysis.
3/4.9.2 INSTRUMENTATION redundant monitoring capability is available to detect chang reactivity condition of the core.
3/4.9.3 DECAY TIME irradiated fuel assemblies in the reactor pressure vessel ens sufficient time has elapsed to allow the radioactive decay of the short lived fission products.
This decay time is consistent with the assumptions used in the safety analyses.
3/4.9.4 CONTAINMENT PENETRATIONS ensure that a release of radioactive material within containm restricted from leakage to the environment.
The OPERABILITY and closure requirements are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
With the containment purge and exhau:t system in operation gas monitor will effectively automatically contain the a high down the containment purge system supply and exhaust fans and closing their inlet and outlet dampers.
then manually close the containment purge and exhaust isolation valv Therefore the uncontrolled release of radioactive material from the containmen,t to the environment will be restricted.
containment isolation signal on high radiationLikewise, use of the SFAS release of radioactive material from the containment to the environment. restricting 3/4.9.5 COMMUNICATIONS station personnel can be promptly informed of significant cha facility status or core reactivity condition during CORE ALTEMTIONS.
DAVIS-BESSE, UNIT 1 B 3/4 9-1 Amendment No.186
~
REFUELING OPERATIONS BASES 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY j
The OPERABILITY requirements of the hoist bridges used for movement of fuel assemblies ensures that:
1 of control rods and fuel ass)emblies, 2) each hoist has sufficient loadfuel h capacity to lift a fuel element and 3 the core internals and pressure vessel engaged during lifting operations. g fo)rce in the event they are inad areprotectedfromexcessiveliftin 314.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel pool ensures that in the event this load is dropped 1 assembly in a fai will be limited to that contained in a single fuel ass)embly, and J2) anythe activity possible distortion of fuel in the storage racks will not result 'n a critical array.
This assum accident analyses.ption is consistent with the activity release assumed in the 3/49.B COOLANT CIRCULATIQN The requirement that at least one decay heat removal loop be in operation ensures that (he) water in the reactor pressure vessel below 140*F a 1
and maintain t during the REFUELING MODE and 2 sufficient coolant circulation iss required maintained through the rea,ctor c(or)e to minimize the effect of a boron dil incident and prevent boron stratification.
of water above the core ensures that a single failure of the o loop will not result in a complete loss of decay heat removal capability.
With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling.
Thus in the event of a failure of the operating DHR loop ade emergency procedures to cool the c, ore.quate time is p,rovided to initiate 3/4.9.9 CONTAINMENT PU.JLGE AND EXHAUST ISOLATION SYSTEM Deleted 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
consistent with the assumptions of the saTety analysis.The minimum water depth is DAVIS-BESSE, UNIT 1 B 3/4 9-2 Amendment No. 3S,135,186 t
.