ML20087K414
| ML20087K414 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/16/1984 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20087K167 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NUDOCS 8403260157 | |
| Download: ML20087K414 (27) | |
Text
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ANO-1 gjg3260157840316ADOCK 0500o333 p
F 3.1.1. 7 Reactor Coolant System Vents At least one reactor coolant system vent path consisting of at least two valves in series shall be operable at each of the following locations whenever the Reactor Coolant average temperature is above 280'F.
1.
Reactor Vessel head 2.
Pressurizer steam space 3.
Reactor coolant system Hot Leg high points (2 locations)
A.
With one of the above vent paths inoperable, startup and/or power operation may continue provided the inoperable vent path is maintained closed; restore the inoperable vent path to operable status within 30 days, or be in hot 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 shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
B.
With two or more of the above vent paths inoperable, maintain the inoperable vent paths closed and restore at least two vent paths to operabic status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot 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 shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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16b l
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BASES:
The plant is designed to operate with both reactor coolant loops and at least one reactor coolant pump per loop in operati'on, and maintain DNBR above 1.30 during all normal operations and anticipated transients.(1)
Whenever the reactor coolant average temperature is above 280 F, single failure considerations require that two loops be operable.
The decay heat removal system suction piping is designed for 300 F, thus, the system can remove decay heat when the reactor coolant system is below this temperature.(2,3)
One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.(4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.
The code safety valves prevent overpressure for a rod withdrawal accident.(5) The pressurizer code safety valve lif t set point shall be set at 2,500 psig il percent allowance for error and each valve shall be capable of relieving 300,000 lb/h of saturated steam at a pressure not greater than
.3 percent above the set pressure.
The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered.
Inspection and manual actuation of the internal vent valves (1) ensure operability, (2) ensure that the valves are not open during normal operation, and (3) demonstrate that the valves begin to open and are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.
The reactor coolant sents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling.
The operability of at least one reactor coolant system vent path from the reactor vessel head, the reactor coolant system high points, and the pressurizer steam space ensures the capability exists to perform this function.
The valve redundancy of the vent paths serves to minimize the probability of inadvertent actuation and breach of reactor coolant pressure boundary while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
Testing requirements are covered in Section 4.0 for the class 2 valves and Table 4.1-2 for the vent paths.
These are consistent with ASME Section XI and Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Require-ments," 11/80.
REFERENCES l
(1) FSAR, Tables 9-10 and 4-3 through 4-7 (2) FSAR, Section 4.2.5.1 and 9.5.2.3 (3) FSAR, Section 4.2.5.4 (4) FSAR, Section 4.3.10.4 and 4.2.4 (5) FSAR, Section 4.3.7 Amendment No. 21, 56 17
I TABLE 4.1-2 (Continued)
Minimum Equipment Test Frequency Item Test Frequency 17.
RCS Vent Paths Demonstrate At least once per 18 operability by months during cold flow verification shutdown O
t I
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-73b-1
3.5.1.7 The Decay Heat Removal System isolation valve closure setpoints shall be equal to or less than 340 psig for one valve and equal to or less than 400 psig for the second valve in the suction line.
The relief valve setting for the DHR system shall be equal to or less than 450 psig.
3.5.1.8 The degraded voltage monitoring relay settings shall be as follows:
1 The 4.16 KV emergency bus undervoltage relay setpoints shall a.
be >3115 VAC but <3177 VAC.
b.
The 460 V emergency bus undervoltage relay setpoints shall be
>423 VAC but <431 VAC with a time delay setpoint of 8 seconds il second.
3.5.1.9 The following Reactor Trip circuitry shall be operable as indicated:
1.
Reactor trip upon loss of Main Feedwater shall be operable (as determined by Specification 4.1.a. items 2 and 36 of Table 4.1-2) at greater than 5% reactor power.
(May be bypassed up to 10% reactor power.)
2.
Reactor trip upon Turbine Trip shall be operable (as determined by Specification 4.1.a items 2 and 42) at greater than 5% reactor power.
(May be bypassed up to 20% reactor power.)
3.
If the requirements of Specifications 3.5.1.9.1 or 3.5.1.9.2 cannot be met, restore the inoperable trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or bring the plant to a hot shutdown condition.
3.5.1.10 The control room ventilation chlorine detection system instrumentation shall be operable & capable of actuating control rnum isolation and filtration systems, with alarm / trip setpoints i
adjusted to actuate at a chlorine concentration of <5 ppm.
3.5.1.11 The Containment High Range Radiation Monitoring instrumentation shall be operable with a minimum measurement range from 1 to 107 rads /hr.
Amendment No. 66, 6Z, 69
-42a-
J Table 3.5.1-1 (Cont'd)
OTHER SAFETY RELATED SYSTEMS 1
2 3
4
.5 No. of Operatoi-Action channels Min.
Min.
if conditions of No. of for sys-operable degree of column 3 or 4 Functional Unit channels tem trip channels run_dancy ca6not be met il.
Containment High Range Radiation Monitoring
- 2 N/A 1
0~
Note 19
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12.
Containment Pressure - High
'f Range 2
N}A'
.1 0'<
' Note 10 r.?
2.
13.
Containment. Water Level -
j
/,
~
Wide Range
'2 N/Ai 1
0 Note 10 r
/
i s
r
~, r-F e
r f'
ff
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- . i e,
d f
1 1
[p Table 3.5.1-1 (Cont'd)
$e Notes Cont'd z
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13.
Channels may be bypassed for not greater than 30 seconds during reactor coolant pump starts.
If the
}S automatic bypass circuit or its alarm circuit is inoperable, the undervoltage protection shall be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, otherwise, Note 14 applies.
y, 14.
With the number of channels less than required, restore the inoperable channels to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdo n 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 us w
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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15.
This trip, function may be bypassed at up to 10% reactor power.
16.
This trip function may be bypassed at up to 20% reactor power.
17.
With no channel operable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the inoperable channels to operable status, or initiate g
and maintain operation of the control room emergency ventilation system in the recirculation mode of g[
operation.
18.
With one channel inoperable, restore the inoperable channel to operable status within 7 days or within the aext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> initiate and maintain operation of the control room emergency ventilation systems in the recirculation mode of operation.
19.
With the number of channels less than required, initiate alternate methods of monitor the containment radiation level within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and (1) either restore the inoperable channel to operable status within 7 days, or (2) prepare and submit a Special Report to the Commissicn pursuant to Specification 6.12.4 within 30 days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.
Table 4.1-1 (Cont'd)
Channel Description Check Test Calibrate Remarks 20.
Reactor Building Spray System NA M(1)
NA (1) Including RB spray pump, spray System Logic Channels valve, and chem. add. valve logic channels.
21.
Reactor Building Spray System Analog Channels a.
Reactor Building Pressure Channels NA M
R r
22.
Pressurizer Temperature Channels S
NA R
23.
Control Rod Absolute Position S(l)
NA R
(1) Compare with Relative Position Indicator.
24.
Control Rod Relative Position S(l)
NA R
(1) Check with Absolute Position Indicator N'
25.
Core Flooding Tanks a.
Pressure Channels S
NA R
b.
Level Channels S
NA R
I 26.
Pressurizer Level Channels S
NA R
27.
Makeup Tank Level Channels D
NA R
28.
Radiation Monitoring Systems other W
M(1)
Q(2)
(1) Check functioning of self-checking than containment high range feature on each detector.
monitors (item 55)
(2) R for those detectors inaccessible during normal operation.
29.
High and Low Pressure Injection NA NA R
Systems:
Flow Channels
Table 4.1-1 (Cont'd)
Channel Description Check Test Calibrate Remarks 47.
EFW Actuation Control Logic NA M
R 48.
EFW Flow Indication R
NA R
49.
RCS subcooling margin monitor D
NA R
50.
Electromatic relief valve flow 0
NA R
monitor 51.
Electromatic relief block valve D
NA R
position indicator 52.
Pressurizer safety valve flow D
NA R
monitur 53.
Pressurizer water level indicator D
NA R
fj 54.
Control Room Chlorine Detector D
M R
T 55.
Containment High Range Radiation D
M R
Monitors 56.
Containment Pressure-High M
NA R
57.
Containment Water Level-Wide Range M
NA R
Note:
S-Each Shift T/W-Twice per Week R-Once every 18 months W-Weekly Q-Quarterly PC-Prior to going Critical if not M-Monthly P-Prior to each startup done within previous 31 days D-Daily if not done previous NA-Not applicable week B/M-Every 2 Months
. = _ _ _ _ - _ _ _ _.
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.
6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.
FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2.
Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
6.2.2.1 Administrative controls shall be established to limit the amount of overtime worked by plant staff performing safety related functions.
These administrative controls shall be in accordance with the guidance provided by the NRC Policy Statement on working hours (Generic letter No. 82-12).
6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum
+
qualifications of ANSI N18.1-1971 for comparable position, except for (1) the Health Physics Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) 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 response and analysis of the plant for transients and accidents.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A training program for fire protection training shall be maintained and shall meet or exceed the requirements of Section 27 of the NFPA Code-1975 with the exception of irequency of training which shall be held at least quarterly.
-117-
6.5 REVIEW AND AUDIT 6.5.1 Plant Safety Committee (PSC) Function 6.5.1.1 The Plant Safety Committee shall function to advise the General Manager on all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The Plant Safety Committee shall be composed of the:
(See Page 121)
Amendment No. 16,30,3#,37
-117a-
- 7,50,59,64
a.
The facility shall be placed in at least hot shutdown within one hour.
b.
The Nuclear Regulatory Commission shall be notified and a report submitted pursuant to the requironents of 10 CFR 50.36 and l
Specification 6.12.3.1.
6.8 PROCEDURES
]
6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
a.
The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.
b.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
e.
Emergency Plan implementation.
f.
Fire Protection Program implementation.
g.
New and spent fuel storage.
h.
Post accident sampling.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the PSC and approved by the ANO General Manager prior to implementation and reviewed period.cally as set forth in administrative procedures.
6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two members of the plant staff, at least one of whom holds a Senior Reactor Operator's License of the unit affected.
c.
The change is documented, reviewed by the PSC and approved by the ANO General Manager within 14 days of implementation.
Amendment No. 16, 30, 34, 37, 76, 82
-127-
l 6.12.4 Unique Reporting Requirements Unique reports cover inspections, tests, maintenance, and special reports l
that are appropriate to assure safe operation of the plant.
The frequency 1
and content of these reporte are determined on an individual case basis and designated in these Technical Specifications.
Unique reports shall be i
submitted in writing to the appropriate Regional Office within 90 days of the completfon of the tests, inspections and maintenance unless indicated 2
otherwise within the referenced specification.
i The subjects of unique reports shall include:
I (a) Tendon surveillance.
(Specification 4.4.2) 1 i
(b) Inoperable containment radiation monitors (Specification 3.5.1.11) i i
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Amendment No. 9, 82
-146-i
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l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION........
3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTDOWN.............................
3/4 4-3 3/4.4.3 SAFETY VALVES - OPERATING............................
3/4 4-4 3/4.4.4 PRESSURIZER..........................................
3/4 4-5 3/4.4.5 STEAM GENERATORS.....................................
3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................
3/4 4-13 Reactor Coolant System Leakage.......................
3/4 4-14 j
3/4.4.7 CHEMISTRY............................................
3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY....................................
3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...............................
3/4 4-22 Pressurizer..........................................
3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components................
3/4 4-26 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.........................
3/4 4-27 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS...............................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 300*F.......................
3/4 5-3 I
3/4.5.3 ECCS SUBSYSTEMS - T,yg < 300*F.......................
3/4 5-6 3/4.5.4 REFUELING WATER TANK.................................
3/4 5-7 ARKANSAS - UNIT 2 V
Amendment No. 29
INDEX i
BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4,4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION........
B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES.............................
B 3/4 4-1 3/4.4.4 PRESSURIZER..........................................
B 3/4 4-2 3/4.4.5 STEAM GENERATORS.....................................
B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.......................
B 3/4 4-3 3/4.4.7 C!!EMISTRY............................................
B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY....................................
B 3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..........................
B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGRITY.................................
B 3/4 4-11 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.........................
B 3/4 4-11 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS...............................
B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS........,....................
B 3/4 5-1 3/4.5.4 REFUELING WATER TANK (RWT)...........................
B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT..................................
B 3/4 6-1 t
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 CONTROL..............................
B 3/4 6-4 ARKANSAS - UNIT 2 X
Amendment No. 29
E55 d*
TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION z
MINIMUM CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RA3GE ACTION 1.
AREA MONITORS a.
Spent Fuel Pool Area Monitor 1
1 1.5x10-2 R/hr 10 101 R/hr 13 b.
Containment High Range 1
1, 2, 3 & 4 N/A 1
- 107 R/hr 18 2.
PROCESS MONITORS a.
Containment i.
Gaseous Activity u,);
a)
Purge & Exhaust Isolation 1
ALL MODES 1 2 x background 10 - 108 cpm 16 u,
A3 b)
RCS Leakage Detection 1
1, 2, 3 & 4 Not Applicable 10 - 108 cpm 14 o'
11.
Particulate Activity a)
RCS Leakage Detection 1
1, 2, 3 & 4 Not Applicable 10 - 108 cpm la b.
Control Room Ventilation Intake Duct Monitor 1
ALL MODES i 2 x background 10 - 108 cpm 17 0 With. fuel in the spent fuel pool or building.
m.
r TABLE 3.3-6 (Continued)
TABLE NOTATION ACTION 13 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 14 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.
ACTION 16 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.
ACTION 17 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the control room emer-gency ventilation system in the recirculation mode of operation.
ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, initiate alter-nate methods of monitoring the containment radiation level within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and (1) either restore the inoperable Channel to OPERABLE status within 7 days or (2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.
ARKANSAS - UNIT 2 3/4 3-26
N 55 TABLE 4.3-3 h
RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
m h3 CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE t
INSTRUMENT CHECK CALIBRATION TEST REQUIRED 1.
AREA MONITORS a.
Spent Fuel Pool Area Monitor S
R H
b.
Containment High Range S
R****
M 1, 2, 3, & 4 2.
PROCESS MONITORS a.
Containment us i.
Gaseous Activity 30 a)
Purge & Exhaust y
Isolation R
ALL MODES no b)
RCS Leakage Detection S
R M
1, 2, 3, & 4 ii.
Particulate Activity a)
RCS Leakage Detection S
R M
1, 2, 3, & 4 b.
Control Room Ventilation Intake Duct Monitor S
R M
ALL MODES
- With fuel in the spent fuel pool or building.
- Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to initiating containment purge operations and 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 containment purge operations.
- Within 31 days prior to initiating containment purge operations and at least once per 31 days during containment purge operations.
- Acceptable criteria for calibration are provided in Table II.F.1-3 of NUREG-0737.
n 5
Si 5;
TABLE 3.3-10 C
.g POST-ACCIDENT MONITORING INSTRUMENTATION N
MINIMUM CHANNELS INSTRUMENT OPERABLE 1.
Containment Pret sure (Normal Design Range) 2 2.
Containment Pres.sure (High Range) 1 3.
PressurizerPre)sure 2
4.
Pressurizer Watt r Level 2
5.
Steam Generator Pressure 2/ steam generator 6.
Steam Generator' Water Level 2/ steam generator h
7.
Refueling Water Tank Water Level 2
8.
Containment Watdr Level - Wide Range 1
1 9.
Emergency Feedwater Flow Rate 1/ steam generator 10.
Reactor Coolant System Subcooling Margin Monitor 1
11.
Pressurizer Safety Valve Acoustic Position Indication 1
12.
PressurizerSafdtyValveTailPipeTemperature 1
e 8n O
I o
N953 i
TABLE 4.3-10 t
C}
POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS N
CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION 1.
Containment Pressure (Normal Design Range)'
M R
2.
Containment Pressure (Wide Range).
M R
3.
Pressurizer Pressure M
R 4.
Pressurizer Water Level M
R 5.
Steam Generator Pressure M
R 6.
Steam Generator Water Level M
R 7.
Refueling Water Tank Water Level M
R l
8.
Containment Water Level - Wide Range M
R 9.
Emergency Feedwater Flow Rate M
R 10.
Reactor Coolant System Subcooling Margin Monitor M
R 11.
Pressurizer Safety Valve Acoustic Position Indication M
R 12.
Pressurizer Safety Valve Tail Pipe M
R a
.if t:
i i
Peactor Coolant System Vent Limiting _Cgpditions for Qperation r
3.4.11 Reactor Coolant System Vents At least one reactor coolant system vent path consisting of at least two valves in series shall be OPERABLE at each of the following locations:
A)
Reactor Vessel Head B)
Pressurizer Steam Space (RCS High Point Vents)
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
a.
With less than one vent path from each of the locations OPERABLE, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path (s) is maintained closed; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT 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 SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With both vent paths A and B above inoperable, restore at o.
least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT 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 SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ERVEILLANCE RE00lEEMENTS 4.4.11 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by verifying flow through the reactor coolant vent system vent patas during COLD SHUTDOWN.
ARKANSAS - UNIT 2 3/4 4-27 l
REACTOR COOLANT SYSTEM BASES 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 Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).
ASME Code Class 1 components of the reactor coolant system were de-signed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition and Addenda through Summer 1973.
ASME Code Class 2 and 3 components were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition and Addenda through Winter 1972.
3/4.4.11 REACTOR COOLANT SYSTEM VENTS The reactor coolant vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling.
The OPERABILITY of at least one vent path from the reactor vessel head and the reactor coolant system high point ensures the capability exists to perform this function.
The valve redundancy of the vent paths serves to minimize the probability of inadvertent actuation and breach of reactor coolant pressure boundary while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
Testing requirements are consistent with ASME Section XI for Class 2 valves (see 3/4.4.10 above) and Item II.B.1 of NUREG-0737,
" Clarification of TMI Action Plan Requirements," November 1980.
ARNAKSAS - UNIT 2 B 3/4 4-11
i 6.0 ADMIN 111RATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.
6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure G.2-1.
FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:
a.
Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
b.
At least one licensed Operator shall be in the control room when fuel is in the reactor.
c.
At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
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 responsibility during this operation.
f.
A site Fire Brigade of at least 5 members shall be maintained onsite at all times.
The Fire Brigade anall not include 3 members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.
g.
Administrative controls shall be established to limit the' amount of overtime worked by plant staff performing safety related functions.
These administrative controls shall be in accordance with the guidance provided by the NRC Policy Statement on working hours (Generic Letter No. 82-12).
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ARKANSAS - UNIT 2 6-1
ADtWili1EATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a.
The unit shall be placed in at least HOT STANDBY within one hour.
b.
The Safety Limit violation shall be reported to the Commission, the Vice President, Nuclear Operations and to the SRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the PSC.
This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
d.
The Safety Limit Violation Report shall be submitted to the Commission, the SRC and the Vice-President, Nuclear Operations within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
a.
The applicable procedures recommended in Appendix "A" of Reg.
Guide 1.33, Revision 2, February 1978.
b.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
e.
Emergency Plan implementation.
t.
Fire Protection Program implementation.
g.
Modification of Core Protection Calculator (CPC) Addressable Constants.
NOTE:
Modification of the CPC addressable constants based on information cbtained through the Plant Computer - CPC data link shall not be made without prior approval of the Plant Safety Committee.
h.
New and spent fuel storage.
i.
Post accident sampling.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the PSC and approved by the ANO General Manager prior to implementation and reviewed periodically as set forth in administrative procedures.
ARKANSAS - UNIT 2 6-13 Amendment No. 5,77,24,25,43,52
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r ADMINISTRATIVE CONTROLS occurrence of the event.
The written report shall include, as a minimum, a completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances i
surrounding the event.
a.
Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those i
established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
I b.
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation, c.
Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems, d.
Abnormal degradation of systems or other those specified in 6.9.1.8.c above designed to contain radioactive material resulting from the fission process.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Administrator of the Regional Office within the time period specified for each report.
These reports shall be submitted covering the activities identffied below pursuant to the requirement of the applicable reference specification:
a.
ECCS Actuation, Specifications 3.5.2 and 3.5.3.
b.
Inoperable Seismic Ponitoring Instrumentation, Specification 3.3.3.3.
Inoperable Meteorological Monitoring Instrumentation, Specifica-c.
tion 3.3.3.4.
d.
Seismic event analysis, Specification 4.3.3.3.2.
e.
Inoperable Fire Detection
's trumentation, Specification 3.3.3.8.
f.
Inoperable Fire Suppression Systems, Specifications 3.7.10.1 and 3.7.10.2.
g.
Primary coolant specific activity, Specification 3.4.8.
h.
Inoperable Containment radiation monitors, Specification 3.3.3.1.
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