ML17340A523
| ML17340A523 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 12/23/1980 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17340A522 | List: |
| References | |
| NUDOCS 8012300334 | |
| Download: ML17340A523 (24) | |
Text
3.3 CONTAINMENT
~A1icabilit~
~0b ective:
Specification:
2.
3.
Applies to the integrity of the containment.
To define the operating status of the containment.
- 1. CONTAINMENT INTEGR ITY a.The containment integrity (as defined in 1.5) shall not be violated unless the reactor is in the cold shutdown condition.
b-The containment integrity shall not be violated when the reactor vessel head is removed unless the reactor is in the refueling shutdown condition.
INTERNAL PRESSURE If the internal pressure exceeds 3 psig or the internal vacuum exceeds 2 psig, the condition shall be corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be brought to hot shutdown-CONTAINMENT ISOLATION VALVES With Keff 0-99,
%%d thermal power excluding decay heat
> 0, and an. average coolant temperature Tavg > 200'F the following conditions shall be met:
The containment isolation valves for Phase A containment isolation, Phase B containment isolation, and Contaianent Ventilation Isolation shall be operable s<ith the isolation times of each power operated or automatic valve within the limits established for testing in accordance with Seciton XI of ASME Boiler and Pressure Vessel Code and applicable Addenda as requried by 10 CFR 50, Section 50.55a(g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i), or the valve is closed.
- 3. 3-1
ii
3.5
.I NSTRUMENTATION
~AIicabii~it:
Ob ective:
Applies to reactor safety featur es and accident monitoring instrumentation systems.
To delineate the conditions of the instrumentation and safety circuits necessary to ensure reactor safety-1.Tables 3.5-1 through 3-5-5 state the minimum instrumentati on operation condi tions.
2.Mith the number of OPERABLE accident monitoring instrumentation channel(s) less than the Total Number of Channels shown in Table 3-5-5, either restore the inoperable channel(s) to OPERABLE status within 7 days, or be'n a condition with K ff <0.99,
%%d thermal power excluding decay heat equal fo zero, and an average coolant temperature Tavg < 350'F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.With the number of OPERABLE accident-monitoring instrumentation channels less than the t)INIMUM CHANNELS OPERABLE requirements of Table 3-5-5, either restore the inoperable channel(s) 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 a condition with K ff < 0.99,
%%d thermal power excluding decay heat equal fo zero, and an average coolant temperature Tavg < 350'F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 3. 5-1
(Qi
~C 1.
TABLE 3.5-2 ENGINEERED SAFETY FEATURES ACTUATION NO.
FUNCTIONAL UN!T 1.
SAFETY INJECTION 1-1 Manual 1.2 High Containment Pressure 1.3 High Differential Pressure between any Steam Line and the Steam Line Header 1.4 Pressurizer Low Pressure*
1.5 High Steam Flow in 2/3'team Lines with Low.
Tava or Low Steam Lsn5 Pressure 2-CONTAINMENT SPRAY MIN.
OPERABLE CHANNELS 1/1 inc in each of 2 lines 2
MIN.
DEGREE OF REDUNDANCY 0'
OPERATOR ACTION IF CONDITIONS OF COL'UMN 1 OR 2 CANNOT BE MET Cold Shutdown Cold Shutdown Cold Shutdown Cold Shutdown Cold Shutdown 2..1 High Containment Pressure 2 per and High-High Containment set Pressure (Coincident) 3.
AUXILIARY FEEDMATER 3.1 Low-Low Steam Generator
,Level 3.2 Loss of Voltage (both 4KV busses) 1/set Col d Shutdown Hot Shutdown Cold 'Shutdown 3.3 Safety Injection 3.4 Trip of both Main Feedwater Pump Breakers
(-'
See I above- )
0 Cold Shutdown This signal may be manually bypassed, when the reactor is shut down and pressure is below 2000 psig.
IQ/
1
NO.
FUNCTIONAL UNIT 1.
High Containment Presure 2.
High-High Contairment, Pressure TABLE 3.5-4 ENGINEERED SAFETY FEATURE SET POINTS CHANNEl ACTION Safety Injection Containment Spray*
Steam Line Isolation*
Containment Isolation*
See No.
1 SET POINT
< 6 psig
< 30 psig 3.
4 ~
Pressurizer Low Pressure High Steam, Line Differenti al Pressure
'(2/3 between any header and any line)
Safety Injection Safety Injection
> 1?15 psig
< 150 psi 5-High Steam Line.Fl ow (2/3 1 ines)
Safety Injection Steam Line Isolation d/p for 3.84xl0 lb/hr, 770 psig, 100K RP d/p for 0.64xl06 lb/hr, 1005 psig, 0$
RP d/p linear with 1st stg. press.,
0 100/o RP
'oincident with:
Low Steam Line Pressure, or Low Tavg 6.
Low-Low Steam Gener ator Auxiliary Feedwater Level
> 600 psig
> 531 F
> 15$ narrow range 7.
Loss of Voltage (both 4 KV busses) 8.
Safety Injection 9-Trip of both iMain Feedwater Pump Breakers Auxiliary Feedwater Auxiliary Feedwater Auxiliary Feedwater N.
A.
All SI setpoints N A-High and High-High coincident
i~
INSTRUMENT TABLE 3.5-5 ACCIDENT MONITORI NG INSl'RUMENTATION TOTAL NO.
OF CHANNELS MINIMUM CHANNELS OPERABLE 1.
Pressure izer Mater Level Z.
Auxiliary Feedwater Flow Rate 3.
Reactor Coolant System Subcooling Margin Monitor 4.
PORV Position Indicator (Primary Detector) 5.
PORV Block Valve Position Indicator 6.
Safety Valve Position Indicator (Primary Detector)
NOTE:
Not effective until installed, 1/val ve I/val ve 1/valve 1/Yal ve 1/val ve 1/val ve Z per generator 1 per generator For the purpose of this Specification, the pressure and temperature e inputs to the Reactor Coolant System Subcooli ng Margin Monitor are redundant.
Or close the associated block valve and rack out its circuit breaker.
~
0
\\
4
'h
TABLE 4.1-1 MINIMUM FRE UENCIES FOR CHECKS CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS l.a. Nuclear Power Range (Check, Calibrate and Test only applicable above 10$ of rated power.)
b.
Power Distribution Map S(l)
M*(4)
D(2) q*(4)
M(3)
M(1)
- 1) Load vs. flux curve
- 2) Thermal power calculation
- 3) Signal to z T; bistable action (permissive, i.od stop, trips)
- 4) Upper 8 lower detectors for symmetric offset
(+5 to -5K).
- 1) Following initial loading arid p~ior 4o operation above 755 power.
- 2) Once per effective full power month.
- 3) Confirm hot channel factor limits.
2.
Nuclear Intermediate Range S(1)i N.A.
P(2)
- 1) Once/shift up to 50$ R.P.
- 2) Log level;bistable action (permis'sive, rod stop, trip) 3.
Nuclear Source Range S(1)
N-A.
P(2)
- 1) Once/shift when in service.
- 2) Bistable action (alarm, trip) 4.
Reactor Coolant Temperature S~
5.
Reactor Coolant Flow s
6.
Pressurizer Water Level M~.
7.
Pressurizer Pressure S~
8.
4 kv Voltage 5 Frequency N.A.
C 9.
Analog Rod Position Si.
B/W(l)i
- 1) Overtemperature-zT (2)>
- 2) Overpower-
~T Reactor protection circuits only With step counters-
II 0
J, II
TABLE 4.1-.1 SHEET 3 Channel Descri tion Check 24.
Logic Channels 25.
Emer. Portable Survey Instruments 26.
Seismograph N.A.
N.A.
N.A.
23.
Envirormental Radiological Monitors N.A.
Cal ibrate Test A(1)
M(1)
N.A.
Mt N. A-Remarks (1) Flow Make trace.
Test battery (change semi-annual ly) 27.
Auxiliary Feedwater Flow Rate 28.
RCS Subcooling Margin Monitor 29.
PORV Position Indicator (Primary Detector) 30.
PORV Block Valve Position Indicator Mt 31
~
Safety Yal ve Position Indicator Mi 32.
Loss of Voltage (both 4kv busses)
N.A.
N.A N.A N.A.
N.A.
N.A.
N.A.
position and verifying by observation of related parameters R
for AFW actuation at power only R
Check consists of monitoring indicated 33.
Trip of both Main Feedwater Pump Breakers N.A.
N.A.
For AFW actuation at power only
4 0
~
~
TABLE 4.1-1 SHEET 4 Using moveable in-core detector system.
Frequency only Effluent monitors only.
Calibration shall be as specified in 3.9.
S D
W B/W M
Q p
R A
N.A.
Each Shift Daily Weekly Every Two Weeks Monthly Quarterly Prior to each startup if not done previous week Each Refueling Shutdown Annually Not appliable N.A. during cold or refueling shutdowns.
The specified tests, however, shall be performed within one surveillance interval prior to startup.
N.A. during cold or refueling shutdowns.
The specified tests, however, shall be performed within on surveillance interval prior to heatup above 200F.
I i5 I%
E
' lp
'H J
v 4.
4.10 AUXILIARY FEEDWATER SYSTEM
~A1i bilge
~0b'ective:
Applies to periodi'c testing requirements of the auxiliary feedwater system.+
To verify the operability of the auxiliary feedwater system and its ability to respond properly when required-l.
Each turbine-driven auxiliary feedwater pump shall be started at intervals not greater than one month; run for 15 minutes and a flow rate of 600 gpm established to the steam generators.
2.
The auxiliary feedwater di scharge valves shall be tested by operator action during pump tests.
3.
Steam supply and turbine pressure valves shall be
. tested during pump tests.
4.
These tests shall be considered satisfactory if control panel indication and visual observation, of equipment demonstrate that all components have operated properly.
5.
At least once per 18 months:
a.
Verify that each automatic val',ve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.
b.
Verify that each auxiliary feedwater pump receives a start signal as designed automatically upon receipt of each auxiliary feedwater actuation test signal.
N.A. during cold or refueling shutdowns (only for the Unit at cold or refueling shutdown).
The specified tests, however, shall be performed within one surveillance interval prior to starting the turbine.
- 4. 10-1
0 ik 4
'4
TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITIONP LICENSE CATEGORY EQUAL IF ICATIONS One or Two Units Operating All Units Shutdown SRO*
RO Non-L'icensed Auxiliary Operators Shift Technical Advisor None Required This position may be filled by one of the SRO's
- above, provided the individual meets the qualification requirements of 6.3.1 Includes the.licensed Senior Reactor Operator serving as Shift Supervisor.
Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervi si ng the movement of any component within the reactor presssure vessel with the vessel head removed and fuel in the vessel.
Operating is defined as Keff > 0.99, X thermal power excluding decay heat greater than or equal to zero, and an average coolant temperature Tavg
> 200'F.
Shift crew composition may be one less than the minimum requirements for a
period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew. members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
0
'4 LI I a: s.i i~.iJ is:
". ev si
> ))
~ "sa.
> ~
~
"> sa
)
"l)>sliJ I
il j I
l>
'~la~
t II~ IKJ I
C, i e
~
~IP ~l
>I' I
~ et
~
~ I
~
fC 4 3 Ilt
~ 113 Ig,Qts~gwl ash
~ < ~ 'i< i c.
~ ll.
~
!i'm,l>> ",ih l
d.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
e.
ALL CORE ALTERATIONS shall be directly supervised by either a
licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
f.
At least three (3) persons shall be maintained on site at all times for Fire Emergency response.
This excludes two (2) members of the shift crew.
6.3 FACILITY STAFF UALIFICATIONS 6.3.1 6.3.2 6.3.2.1.
Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and'analysis of the plant for transients and accidents-HEALTH PHYSICS SUPERVISOR UAL'IFICATIONS The Health Physics Supervisor at the time of appointment to the position, shall, except as indicated below, meet the following:
1.
He shall have a bachelor's degree or equivalent in a science or engineering
- subject, including some formal training in radiation protection.
2.
He shall have five years of professional experience in.applied radiation protection; where a master's degree in a related field
- is equivalent to one year experience and a doctor'.s degree in a related field is equivalent to two years of experience.
- 3. Of his five years of experience,'hree years shall be in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered at Turkey Point Plant.
6.3.2.2 6.4 When the. Health Physics Supervisor does not meet the above requirements, compensatory action shall be taken which the Plant Nuclear Safety Committee determi nes',
and the NRC Office of Nuclear Reactor Regulation concurs that the action meets the intent of
,Specification 6.3.2.1.
TRAINING 6.4.1
'A retaining and replacement program for the facility staff shall be maintained under the direction. of the Training Supervisor and shall meet or exceed the requirements and reccmmendations of Section 5.5,.
.and Appendix A to 10 CFR Part 55.
6-5
0 P
N
STATE OF PLORXDA
)
)
COUNTY OP DADE
)
ss.
Robert E. Uhrig, being first duly sworn, deposes and says:
That he is a Vice President of Florida Power a Light Company, the Lic nsee herein; That he has executed the foregoing document; that the state-ments made in this said document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.
Robert E. Uhrig Subscribed and sworn to before me this l9 Pz NOTARY PUBLlC, in and for the county of Dade, State of Plorida
/PRY PVQUC STATE Of FLOP'DA tt LARC>-
OQQJ}RSSfON EXPIRES
+VMST 24'i&81 (tIIyP PPV. {rqqgPi qONSlRQ AGNQf
ii V
r<
/,
lw P