ML20031F650
| ML20031F650 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/30/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20031F651 | List: |
| References | |
| TAC-46520, TAC-46521, NUDOCS 8110200261 | |
| Download: ML20031F650 (20) | |
Text
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UfJITED STATES
[ %,... (f i NUCLEAR REGULATORY COMMISSION g
WASHING TON, D. C, 20555 3 J, gb 9, jak s' y
\\..u.< ~ f WISCONSIN ELECTRIC POWER COMPANY
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DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY CPERATING LICENSE Amendment No. 55 License No. DPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated February 4,1981, 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 Chaoter 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.
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR,24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 55, are hereby incorporated in the license.. The licensee shall operate the facility in accordance with the Technical Soecifications.
3.
This license amendment is ef fective at of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l' )
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Robert A. Clark, Chief Operating Reactors Branch d3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: September 30, 1981
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,b, UNITED STATES
[ T,,,9,' 'g NUCLEAR REGULATORY COMMISSION
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v WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301
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POINT BEACH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 60 License No. DPR-27 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated February 4,1981, complies with.the standards and requiremente of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapte r 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.
n
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-27 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 60, are hereby incorporated in the license. The licensee shall operate tne facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR YHE NUCLEAR REGULATORY COMMISSION wt..I i
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: September 30, 1981
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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 55 TO FACILITY OPERATING LICENSE NO. OPR-24 AMENDMENT NO. 60 TO FACILITY OPERATING LICENSE NO. DPR-27 4
00CXET NOS. 50-266 AND 50-301 Revise Appendix A as follows:
Remove Pages Insert Pages 15.3.1.2 15.3.1.2 15.3.1.3 15.3.1.3 15.3.1.3A 15.3.5-1 15.3.5-1 15.3.5-5 15.3.5-5 15.3.5-6 Table 15.3.5-1 Table 15.3.5-1 Table 15.3.5-2 (Continued)
Table 15.3.5-2 (Continued)
Table 15.3.5-3 Table 15.3.5-3 Table 15.3.5-3 (Continued)
Table 15.3.5-3 (Continued)
Table 15.3.5-5 Table 15.4.1-1 (Continued)
Table 15.4.1-1 (Continued)
Table 15.4.1-1 (Continued)
Table 15.4.1-1 (Continued)
Table.15.4.1-2 (Continued)
Table 15.4.1-2 (Continued)
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3.
Pressurizer Safety Valves 1
a.
At least one pressurizer safety valve shall be operable whenever the reactor head is on the vessel.
b.
Both pressuriser safety valves shall be operable whenever the reactor is critical.
4.
Pressurizer Power Operated Relief. Valves (PO RV) and PORV Slock Valves,
a.
Two PORVs and their associated block valves shall be operable.
1.
If a PORV is inoperable, the FORY shall be restered to ar.
operable condition within one hour or the associated block valve shall be closed.
2.
If a PORV block valve is inoperable, the block valve shall be restored to an operable condition within one hour or the block valve shall be closed with power removed from the block valve; otherwise, the unit shall be in the hot shutdown condition within the next six hours.
5.
The pressurizer shall be operable with at least 100 KW of pressurizer heaters available and a water level greater than log and less than 957. during steady state power operation. At least one bank of pressurizer heaters shall be supplied by an emergency bus power supply.
Basis:
Tinen the boron concentration of the reactor coolant system is to be reduced the process must be uniform to prevent sudden reactivity changes in the reactor. Nixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.
The residual heat removal pump will circulate the primary system volue.e-in approximately one half hour. tlc pressurizer is of lii:le concern 15~3.1-2 Unit 1 Amendment No. //, 55 Unit 2-Amendment No. f!, 60
l beccus2 'of tha low prassurizar volumi and bactuss pressurizar boron concentration normally will be higher than that of the rest of the reactor coolant.
Part 1 of the specification requires that a sufficient number of reactor coolant pumps be operating to provide core cooling in the event that a loss of flow occurs. The flow provided in each case will keep DNBR well above 1.30 as discussed in FFOSAR Section 14.1.9.
Therefore, cladding damage and release of fissacn products to the reactor coolant will not occur. Heat transfer analyses (1) show that reactor heat equivalent to 10% of rated power can be re.-oved witn natural circulation only; hence, the specified upper limit of 1%
rated power without operating pumps provides a substantial safety f actor.
Each of the pressurizer safety valves is designed to relieve 288,000 lbs.
per hr. of saturated steam at setpoint.
Below 350'F-and 350 psig in the reactor coolant system, the residual heat rereval system can remove decay heat and thereby control system temperature and pressure. If no residual heat is removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. Cne valve therefore provides adequate defense against over-pressurization. Part 1 e(2) permits an orderly reduction in power if a reactor coolant pump is lost during operation between lot and 50% of rated power.
Above 50% power, an automatic reactor trip will occur if either pump is lost.
c/
The pcwer-c-flow ratib will be maintained equal to or less than 1.0 which ensures that the minimum DNB ratio increases at lower flow since the maximum enthalpy rise does not increase above its normal full-flow maximum value. (2)
A PORV is defined as OPEPABLE if leakage past the. valve is less than that allowed.n Specification 15.3.1.D and the PORV has met its most recent channel test as specified in Table 15.4.1-1.
The PORVs operate to relieve, o
in a controlled manner, reacter coolant system pressure increases below Unit 1-Amendment rio. FA, 55 15.3.1-3 Unit 2. Arrendment tto. /9, 60
i 1
the setting of the pressurizer safety valves. These PORVs have remotely operated block valves to provide a positive shutoff capability should a PORV become inoperable.
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The requirement that 100 KW of pressuricer heaters and their associated 4
controls be capable of being supplied electrical power from an emergency
- bus provides assurance that these heaters can be energized during a loss 4
of offsite power condition :: maintain pressure control and natural cir-4 culation at hot standby.
1 Reference 1
(1) TSAR Section 14.1.6 (2) ?S AR Section 7.2. 3 4
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h Unit 1 Amendment No.'55 Unit 2-n.mendment No. 60 15.3.1-3A F
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15.3.5 INSTRUMENTATION SYSTE."
Operational Safety Instrumentation Applicability :
Applies to plant instrumentation systems.
Obje ctive s :,
To provide for automatic initiation of the Engineered Safety Features in the that principal p ccess variable Ibmits are exceeded, and to delineate event the conditions of the plant instrumentation and safety cir:uits necessary to ensure reactor safety.
Soe: fication:
The Engineered Safety Features initiation instrumentation setting A.
limits shall be as stated in Table 15.3.5-1.
For on-line testing or in the event of a sub-system instrumentation 3.
channel failure, plant operation at rated power shall be permitted to continue in accordance with Tables 15.3.5-2 through 15.3.5-4.
In the event the number of channels of a particular sub-system C.
in service falls below the limits given in ene column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirerent shown in Tables 15.3.5-2 through 15.3.5-4, operator Action when minimum operable channels unavailable.
The accident monitoring instrumentation channels in Table 15.3.5-5 D.
shall be operable. In the event the number of channels in a parti-cular sub-system falls below the minimum number of opersble channels given in Column 2, operation ino subsequent operator action shall be in accordance with Column 3.
Basis:
Instrumentation has been provided to sense accident conditions and to initiate operation of the Engineered Safety Features (l).
Unit 1-Amendment No. 55 15.3.5-1 Unit 2-Amendment No. 60
which autcmatically initiates appropriats action to prevent exceeding 1
established limits. Safety is not compromised, howev=:, by continuing opera-tien with certain instrumentation channa'
't of service since provisions were made for this in the plant des:
TL ipecification outlines limiting conditions for operation necesssary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of servtce.
Alinos: all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. Exceptiens are backup channels such as reactor cociar.: pump breakers. The removal of one trip channel on process control equipment is acecmplished by placing that channel bistable in a tripped moder e.g., a two-out-of-three circuit beccmes a ene-out-of-two circuit. The source and tntermediate range nuclear instrumenta-tion aystem channels are not intentionally placed in a tripped mode since these are One-out-of-two trips, therefore the trips tre bypassed during testing.
Testing of the NIS power range channel requires bypassing the Dropped Rod protection frem NIS, for the channel being tested. However, the Rod Position 4
System still provides the dropped-rod protection. Testing does not trip the sys-em cr.less a trip condition exists in a concurrent channel.
I The operability of the accident monitoring instrumentation ensures that t
sufficient information is available in selected plant parameters to monitor and assess these variables during and following in accident. The PORV block valves have local, exterral indication of whether the block valve is open or shut.
j If necessary, this local indication can be visually verified during a contain-ment entry inspection to verify the block valve is shut.
If the process computer, which provides the reactor coolant system subcooling i
3 margin monitor, becomes inoperable, subcooling will be monitored by means of a backup plotter method or manually using contrcl board instrumentation and a saturation curve.
Unit 1-Amendment No. 55 Unit 2-Amendment No. 60 15.3.5-5
Paference (1) FSAR - Section 7.5 (2) FSAR - Section 14.3 (3) FS/ A - Section 14.2.5 i
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Unit 1-Amendment tio. 55 15.3.5-6 Unit 2-Arrendment flo. 60
TABLI: 15.3.5-1 1 t4GItJI:1 lusD S API.~lT. Fl:ATURES It4 ETI ATIOtl IIISTI(Urti:rlT Sl?I*rltlG LIMIT:i i
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Fut1CTIONAL UNIT Cil Atititg.
SETTING LIMIT 1
liigh Containment Pressure (lii)
Safety Injection 1 6 psig a
2 liigh Containment Pressure (Ili-lii)
- a. Containment Spray 1 30 psig
- b. Steam I.ine Isolation of Iloth Lines 1 20 psig 3
Pressurizer low Pressure Safety Injection *
> 1715 psig 4
Iow Steam Line Pressure safety Injection *
> 500 psig Leail Tirin* Constant
> 12 secorkts Lag Time Constant
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1 2 seconils 5
liigh Steam Flow in a Stem Line Ste.un i.ine Isolation of d/p corresponding to Coincident with Safety Injection and Af fected Line 10.66 x 106 lb/hr at Low T 1005 psig c
ayg n a..
c+ ct
> 540*P ru -
0 6
IligS-high Steam Flow in a Steam I.ine luolation
< d/p corresponding to MS Steam Line. Coincident with of Af fected I.ine 4 x 106 lb/hr at nu
- o. ct safety Injection H06 psig nD cu to oo et rt nz 7
Low-low Steam Generator Water Auxiliary Peedweter
> 5% of narrow range
.o o Level Initiation instrument ww MN 0
Undervoltage on 4 KV Ilussen Auxiliary Feedwater
> 75% of normal
- =
ww Initiation voltaire
.N.N cr <n o *n
- Initiates alt,o containment isolation, feedwa t e r line inolat ion and ;t arting of all containment fans.
d/p means differential pressure Page 1 of 2
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TABLE 15.3.5-1 (Continued) i l
NO.
FUNCTIONAL UNIT Cll ',NNE L SETTING LIMIT i
9.
Degraded Voltage (4.16 KV)
Disconnection of affected
> 3675 volts f,2%
i.
bus from offsite power Time delay:
13.6 sec + 5%
1 at 0,-95% of voltage settin 10.
Loss of Voltage h
a.
4.16KV Disconnection of affected
- a. 2450 volts f; 3%
bus from of fsite powet Time delay:
0.3 sec.
i Start Diesel f.'5% at 0 volts 1.2 sec. + 5% at a
90% of voltage setting l
Load shedding
- b. 256 volts j; 3%
b.
480 V Time delay: 0.75 sec.
+5% at 0 volts 3.5 sec. _+ 15% at 907. of i a
voltage setting I
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.au CL CL aa
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an tn j
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5
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TABLF. 15.3 0 2 (Cont'd) 1 2
3 4
5 NO. OP filta.
MItilMUM PEluilSS Illt.1 OPEllATOR ACTION NO.OF CilAtit1El.S OPEllABLE DEGitEE OP llY l' ASS IF CONDITIONS OP' NO.
FUNCr10tJAb UNIT Cll ANill:IS TO Cil Atit1EIS i<l:11Uf 4DANCY cot 3 DITI ONS COIAJMN 3 OR 4 CAtitKrl' DE MET TitIP 11.
-2 2
1 Maintain <50s of rated gx2we r 12.
Steam Flow - Feed' Water Flow 2/ loop 1/ loop 1/ loop 1/ loop Maintain hut sinstdown mismatch 13.
Lo.Lo Steam Generator 3/ loop 2/ loop 2/ loop 1/ loop flaintain hot sluitdown Water Level 14.
UndervoLtage 4 KV Dus 2/ bus 1/ bus I/ Inns M.tintain hot tehutdown (both buses) 15.
Underfrequency 4 KV Bus 2/ bus 1/ bus 1/ bus it.iintain hot t;histdown (both buses) '
When block condition exists, maintain normal operation.
f1OTE 1:
I F.' P.
Pull Power
=
Not Applicable
- One addit.ional channel nuy l>e taken out of service for zero power gihysics testienJ.
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Unit 1-Amendment No. M, 55 Unit 2-Amendment No. $@, 60 a
TAllLI: 15.3.5-3 EMEltGEtaCY COOLitlG i
1 2
3 4
5 OPERATOR ACTION
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NO. OF Mill.
Mill.
PEltMISS IllLE IF CONDITIONS OF NO. OF CII ANilCI.S OPE 14AllLE DMGl41:E OF I)YPASS COLUMN 3 OR 4 11 0.
FUtiCTIO!JAL UNIT CilANilMLS TO THIP Cil AfillEl.S 141:DUflDAf1CE COf1DITIONS cat 4NOT BE MET 1.
SAFETY ItJJ ECTIOil
'e a.
Manual 2
1 1
1 Hot Shutdown ***
b.
liigh. Containment Pressure 3
2 2
1 Ilot Shut down * *
- Primary Pressure is Steam Generator Low Steam Pressure / Loop 3
2 2
1 Less than 11100 psig flot S hu t down * *
- c.
d.
Pressurizer Low Pressure 3
2 2
1 Primary Pressure is f.ess than 1I100 psig Ilot Shutdown'
- 2.
COtITAINMi'NT SPHAY Ilot Shutdown ***
a.
Manual 2
2 2
b.
Ili-Ili Containment Pressure 2 sets 2 of 3 2 per 1/r.u t slot Shnt down* * *
(Containment Spray) of 3 in each set.
set 3.
AUXILIARY FEEDWATER Low-low S* cam Generator Water a.
Ilot Shutdown * * *
- i. Start Motor Driven Pump 3/ steam gen.
2/eit.her gen.
2/ steam 1
gen.
ii. Start Turbine Driven 3/ steam gen.
2/ bot.h gens.
2/ steam 1
Ilot Shutdown * "
800-Pump b.
Trip of both flain Ilot Shutdown * *
- Feedpumps starts motor driven pumps 2/ pump 1/ pump 1/ pump 1
I' Undervoltcoe on 4KV c.
isusses st. arts Turbine Ilot Shutdown ***
driven pumti 2/ bus 1/ bus 1/ bus Unit 1-Amendment No. 38, M, SS l
page 1 of 2 Itn i t A Am..ndment flo. E0, B7. 60 t
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TADL1: 15.1.5-3 EMERGENCY COOLIf1G 1
2 3
4 5
OPERATOR ACTION FJO. OF MIN.
Mill.
PERMISSI BIE IF CONDITIONS OF NO. GP CIIANilEIII OPERADLE DEGREE OP HYPASS COLUMN 3 OR 4 CllANNELS TO THIP Cll AtJNEl.S I<EDUNDANCY COf1DITIONS CANNUP BE MI?F 10.
FUNCTIONAL UNIT I
3.
AUXILI ARY FEEDWATER(Continued)
S.I.
INITIATING FUNCTIONS AND RFQUIREMENT AS IN 1. AIYJVE I-Safety In]ection Signal Starts Motor Driven Pumps 4.
SAFETY RELATED ELECTRICAL. BUSES (4.16 KV) 3/ bus 2/ bus 2/ bus 1/ bus Degraded Voltagc a.
(4.16 KV) 2/ bus 1/ bus 1/ bus 1
b.
Loos of Voltage (480 V) 3/ bus 2/ bus 2/ bus 1
Hot Shutdown "**
Loss of Voltage c.
- - Must activate 2 switches simultaneously.
shutdown, the unit shall be in cold
- - If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reaching hot shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the event causing the unit shutdown.
i d diesel generator is operating
- Normal operation provided both diesel generators are available, and the acsoc ate If minimum conditions are not met uithin 7 days, the affected 00**
l and providing power to the affected safeguards bus.
unit shall be placed in hot shutdown.
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l Unit 1-Amendment No. 38, (/, 55 page 2 of 2 Unit 2-Amendment No. 50, 52. 60 t
TABLE 15.3.5-5 IllSTitututt3T OPEl(ATIt4G cot 3DITIOrJS l '"It filDICATIOtt I
2 flINIMUM 3
11 0. OP OPUIMUIE OPUlmTOR ACTION IP CONDITIOt4S 11 0.
FUNCTIONAL UNIT CllAtJNELS CilAtlNEL OF COLUMt1 2 CANNOT BE MET 1.
PoltV Posit. ion Indicator 1/ Valve 1/ Valve If the operability of the Poltv position indicator cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, shut the e
associated Poltv Block Valve.
3.
POltV Ulock Valve Position Indicator 1/ Valve 1/ Valve If the operability of the PGitV Block Valve Position Indicator cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, shut I
and vet if y the Hlock Valve shut. by direct observa-l tion or declare the Block Valve inaoperable.
3.
Safety Valve Position Indicator 1/ Valve 1/ Valve if the operability of the Safety Valve Position l
Indicator cannot he restored within seven days, he in at least flot !;hutdown within the next 12 hcurs.
4.
Iteactor Coolant System Subcooling 1
1 If the operability ot' a subcooling monit.or cannot be restored or a backup monitor made functional within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, he in at least Ilot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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If the operability of the auxiliary feedwater flow l
l S.
Auxiliary Peedwater Flow Rate
- 1 1
rate indicator cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, l
be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(
6.
Control Rod ttisalignment as Monitored 1
1 Ing individual rod posit. ions once/hr., af ter a load change >10% or a f t er '30 inches o f cont rol rod mot ion, by On-Line Computer l
installed combination of auxiliary feedwater pump discharge ilov indir,ators and auxiliary
- Applies to presentgly I
I feedwater flow to steam generator indicators.
Unit 1-Amendment No. 55 Unit 2-Amendment No. 60
TAllIE 15.4.1-1 (COrlTI tauf:D)
Channel Description -
Check calibrate Test itema rks 24.
Contaisunent Pressure S
R M**
Narrow range containment pressure
(~1.0, 63 psiy exeIuded) 25.
S team Generator Pr essure S***
R Ma**
2u.
Turbine First. Stage Pressure S**
R Ma*
27.
Emergency Plan Radiation M
R M
Instruments 38.
Environmental -lonitors M
N.A.
fl. A.
29.
Overpressure Mitigating S
R 30.
PORV Position Indicator S
R R
I 31.
PORV Dlock Valve Position Q
R ti. A.
Indicator 32.
Safety valve Position Indicator
.M R
N.A.
33.
PORV Operability N.A.
R M
Performance of a channel functional test but excluding valve operation.
34.
Subcooling Marr;in Monitor M
R N.A.
35.
Undervoltage on 4rv Bus N.A.
R Ma*
For Auxiliary feedwater Pump Initiation 36.
Auxiliary Feedwater Flow Hate Sec Remark?
R N.A.
Flow Rate laulication will be checked at each unit startup.uul shutdown 37.
Degraded 4.16 KV Voltage S
R Loss of Voltage (4.16'KV)
S R
M**
38.
.a.
b.
Loss of Voltage (4.16 KV)
S R
M**
19.
4160 V. Frequency N.A.
R IJ. A.
Unit 1-Amendment No. 38, M 55 Unit 2-Amendment No. $@, $5, 60
TABLE 15.4.1-1 (CotiTIllUED) w
'S - Cach Shift M - Monthly P - Prior to each startup if not dono previous week.
n - Daily W
Weekly 11 - 1:ach lie fueling Shutdown (Hut not to exceed 20 months).
Q - Quarterly 13. A. - tJot applicable.
D/W - Diweekly l
tiot required during periods of refneling shutdown, but must be per formed prior to st.ar ting up i f it has not been performed during the previous surveillance period.
- tiot required during periods of refueling shutdown if steam generator vessel temperature is greater than 70*P.
i
- When used for the overpressure mitigating system each PoltV shall be denonstreated operable by:
a.
Performance of a channel functional test on the PoltV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which tne PoltV is reipaired operable acid at least once per 31 days tnereafter when the PoltV is required operable.
b.
Testing valve operatlon in accordance with the inservice test requirements of the ASME Holler and Pressure Vessel Code,Section IX.
Unit 1-Arnendment No. 35, #, 55 Unit 2-Amendment No. 50, 55, 60
+
o
TAN.E 15.4.1-2 (COMTINUEDI Test Frecuency 4
14.
Refueling System Interlocks Functioning Each refueling shutdown c
15.
Service Water System Functioning Each refueling shutdown
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16.
Primary System Leakage Evaluate Monthly (6) 17.
Diesel Fuel Supply Fuel' inventory Daily 15.
Turbine Step and Functioning Monthly (6)
Governor Valves 9
19.
Lcw Pressure Turbine Visual and magnetic Every five years Rot:r Inspecticn (5) particle or liquid penetrant 20.
Sorte Acid System Storage Tank Daily Temperatuce 21.
Scri Acid System Visual cbservation Daily of piping te=peratures (all >145'F) 22.
Scric Acid Piping Heat Electrical circuit Menthly Tracing operability 23.
PORV Block Valves
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Complete Valve Cycle Quarterly (6) 24.
Integrity of Post Accident Evaluata Yearly Recovery Systems Cutside Centainment l
-(l)
A radiochemical analysis for this purpose shall consist of.a quantative measure-ment of each radienuclide with half life of '30 minutes such that at least 95%
of total activity of prinary coolant is accounted'for.
(2)
I determination will b; started when the gross activity analysis of a filtered j
sample indicates j>10 ac/c and will be redetermined if the pr: mary coolant gross radioactivity of a filtered saeple increases by 'more than 10 tc/c:.
(3)
Drop. tests shall be conducted at rated reactor coolant flow.
Rods shall be-l dr:pped under both ecid and hot conditions, but cold drop tests need not be timed.
(4)
Drop tests will be conducted in the hot condition for rods on which maintenance was performed.
(5)
As' accessible'without disassembly of rotor.
(6)
Not re, quired during periods of refueling shutdown.
(7)
At least once per week during periods of refueling shutdown.
(9).
At least' three ti=es. per week (with maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples) during periods of refueling shutdown.
Unit 1-Amendment No. 32, #5, 55 Unit 2-Amendment No. 50, 53, 60
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