ML19257D522
| ML19257D522 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/07/1976 |
| From: | BECHTEL GROUP, INC. |
| To: | |
| Shared Package | |
| ML19257D516 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 H-25994, NUDOCS 8002040522 | |
| Download: ML19257D522 (1) | |
Text
.
K.
Limiting Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent a margin with normal operation lying below these settings.
The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded.
L.
Mode - The reactor mode is established by the mode selector-switch.
The modes include refuel, run, shutdown and startup/ hot standby which are defined as follows:
1.
Refuel Mode - The reactor is in the refuel mode when the mode switch is in the refuel mode position.
When the mode switch is in the refuel position, the refueling interlocks are in service.
2.
Run Mode - In this mode the reactor system pressure is at or above 325 psig and the reactor protection system is energized with APRM l
protection (excluding the 15% high flux trip) and RSM interlocks in service.
3.
Shutdown Mode - The reactor is in the shutdown mode when the reactor mode switch is in the shutdown mode position.
4.
Startup/ Hot Standby - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed when reactor pressure is less than 1000 psig, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod withdrawal interlocks in service.
M.
Operable - A system or component sball be considered operable when it is capable of performing its intended function in its required manner.
N.
Operating - Operating means that a system or component is performing its intended functions in its required manner.
O.
Operating Cvele - Interval between the end of one refueling outage and the end of the nect subsequent refueling outage.
P.
Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:
1.
All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.
2.
At least one door in each air lock is closed and sealed.
l866 213 8002040 g
_3_
COOPI'.R NUCLEAR STATION TABLE 3. 2.11 (PAGE 3)
RESIDUAL llEAT REMOVAL SYSTEM (LPCI MODE) CIRClllTRY REQUIREMENTS Minimum Number of Action Required When Instrument Operable Components Component Operability Instrument I.D.
No.
Setting Limit Per Trip System (1)
Is Not Assured RilR Pump Low Flow Ri!R-dPIS-125 A & B
>2500 gpm 1
A Time Delays RilR-TDR-K4 5, l A& lli 4.25'T'5.75 min.
1 A
RilR Pump Sta rt RilR-TDR-K75A & K70B
- 4. 5 T '5. 5 Sec.
I A
Time Delay RIIR-TDR-K 7511 6 K70A 1 5 sec.
1 A
RilR lleat Exchanger RilR-TDR-K93, A 6 11 1.8 T;2.2 min.
I 11 liypass T.D.
RilR Crosstie Vatve RilR-LMS-2 Valve Not closed (3)
E Position liu s lA 1.ow Volt.
27 X 3/lA Loss of Voltage I
11 Aux. lie l ay Bus IB Low Volt.
27 X 3/lB Loss of Voltage I
11
,g Aux. Relay i
Bus IF Low VoIt.
27 X 1/IF 1.oss of Vo1tage 1
B Aux. Relays 27 X 2/lF Loss of Voltage 1
B liu s 1G 1.ow Volt.
27 X 1/lG Loss of Voltage 1
11 Aux. Relays 27 X 2/lG Loss of Voltage
" Pump Discha rge 1 ine CM-PS-266
>5 psig (3)
D C
CM-PS-270
[15psig
( 3)
D l
Ch Emergency 15uses 27/IF-2, 27/IFA-2 3600 +5%
2 11 O
Undervoltage Relays 27/lG-2, 27/lGil-2 8 sr Ind 12 sec.
2 li g
(degraded voltage) 27/ET-2 time delay 1
11 4
Emergency liuses Loss 27/lF-1, 27/lFA-1, 2900V 15%
of Voltage Relays 27/lG-1, 27/lGB-1, 5 second 11 sec.
27/ET-1 delay 1
11 Emergency 15uses Under-27X7/IF, 27X7/IG, Voltage Relays Timers 27X10/lG 10 second 12 sec.
1 B
NOTES FOR TABLE 3.2.B 1.
When any ECCS system is required to be operable, there shall be two operable trip systems except as noted.
If a requirement of the fourth column is re-duced by one, the indicated action shall be taken.
If the same function is inoperable in nore than one trip system or the fourth column reduced by more than one, action B shall be taken.
Action:
A.
Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If the function is not operable in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, take action B.
B.
Declare the system or component inoperable.
C.
Immediately take action B until power is verified on the trip system.
D.
The high point vent shall be vented weekly upon failure of PS 73A or B, PS 266, PS 268, PS 269, PS 270.
E.
Repair as soon as possible.
It does not directly effect system operations.
2.
In only one trip system.
3.
Not considered in a trip system.
4 Requires one channel from each physical location in the steam line space.
3.
One relay senses each phase of MCC-S and the LO relay is a transfer permissive relay.
O A 0
1866 215 COOPER NUCLEAR STATION TABLE 4.2.B (Page 2)
RHR SYSTEM TEST & CALIBRATION FREQUl:NClES Instrument Item Item 1.D.
No.
Functional Test Freq.
Calibration Freq.
Check Instrumentation 1.
Drywell liigh Pressure PC-PS-101, A,B.C
&D Once/ Month (1)
Once/3 Months None 2.
Reactor Vessel Shroud Level NBI-LITS-73, A & B if l Once/ Month (1)
Once/3 Months once/ Day 3.
Reactor Low Pressure RR-PS-128 A & B once/ Month (1)
Once/3 Months None 4.
Reactor Low Pressure NBI-PS-52 A 6 C Once/ Month (1)
Once/3 Months None N31-PIS-52 B 6 ')
5.
Drywell Press.-Containment PC-PS-119, A,B C 5 D Once/ Month (1)
Once/3 Months None Spray 6.
RllR Pump Discharge Press.
RllR-PS-120, A,B,C 6 D Once/ Month (1)
Once/3 Months None 7.
RHR Pump Discharge Press.
RilR-PS-105, A,B,C & D Once/ Month (1)
Once/3 Months None 8.
RilR Pump Low Flow Switch RilR-dPIS-125 A 6 B Once/ Month (1)
Once 3 Months None b
9.
RilR Pump Start Time Delay Rl!R-TDR-K70, A & B Once/ Month (1)
Once/Oper. Cycle None Y 10.
RHR Pump Start Time Delay RHR-TDR-R75, A 5 B Once/ Month (1)
Once/Oper. Cycle None 11.
RHR lleat Exchanger Bypass T.D.
RHR 'IDR-R93, A & B Once/ Month (1)
Once/Oper. Cycle None 12.
RilR Cross Tie Valve Position RHR-LMS-2 Once/ Month (1)
N.A.
13.
Low Voltage Relays 27 3/lA (7)
None 14.
Low Voltage Relays 27 x 3/lb (7)
None 15.
Low Voltage Relays 21 x 2/lF, 27 X 2/lG (7)
None 16.
Low Voltage Relays 27 X 1/lF, 27 X (1)/lG (7)
None l
17.
Pump Disch. Line Press. Low CM-PS-266, CM-PS-270 Once/3 Months once/3 Months None 18.
Emergency buses Undervol age
.' 7 / l F-2, 27/lFA-2, 27/lG-2, once/ Month Once/18 Months Once/12 hrs.
Relays (Degraded Voltage, 27/lGB-2, 27/ET-2 19.
Emergency Buses Loss of 27/lF-1, 17/lFA-1, 27/lG-1, Once/ Month Once/18 Months Once/12 hrs.
Voltage Relays 27/lGB-1, 17/ET-1 2 6 Emergency Buses Undervoltab!
27X7/lF, 27X7/1G, 27X10/lG Once/ Month once/18 Months None CO Re lays Timers Ch N
3.3 and 4.3 BASES:
(Cont'd) flux.
The requirements of at least 3 counts per second assures that ry transient, should it occur, begins at or above the initial value of 10-8. of rated power used in the analyses of transients cold con-7 ditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal.
A minimum of two operable SRM's are provided as an added conservatism.
5.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation.
Two channels are pro-vided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.
This sytem backs up the operator who withdraws control rods according to written se-quences.
The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod with-drawal errors when this condition exists.
A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit (i.e., MCPR = 1.07, and LHGR = as defined in 1.0.A.4).
During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.
It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other person-nel qualified to perform this function may be designated by the station superintendent.
C.
Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e.,
to prevent the MCPR from becoming less than the safety limit.
The limiting power transient is defined in Reference 3.
Analysis of this transient shows that the negative reactivity rates resulting from the scram provide the required protection, and MCPR remains greater than the safety limit.
On an early BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter.
The design of the present control rod drive (Model CRDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e.,
it can no longer interfere with scram performance, even if ccmpletely blocked.
1866 217
-102-
TABLE 3.7.4 (Page 21 PRIMARY CONTAINMENT TESTABli ISOLATION VALVES TEST PEN. NO.
VALVE NUMBERS MEDIA X-39A RHR-MO-26A and RHR-M0-31A, Drywell Spray Header Supply Air X-39B RHR-MO-26B and RHR-M0-31B, Drywell Sp ray Header Supply Air X-41 RRV-740AV and RRV-741AV, Reactor Water Sample Line Air X-42 SLC-12CV and SLC-13CV, Standby Liquid Control Air X-205 PC-233MV and PC-237AV, Purge and Vent Supply to Torus Air X-205 PC-13CV and PC-243AV, Torus Vacuum Relief Air X-205 PC-14CV and PC-244AV, Torus Vacuum Relief Air X-205 MV-1303 and MV-1304, ACAD Supply to Torus Air X-210A RCIC-MO-27 and RCIC-13CV, RCIC Minimum Flow Line Air X-210A RHR-M0-21A, RRR to Torus Air X-210A RHR-MO-16 A, RHR-10CV, and RHR-12CV, RHR Minimum Flow Line Air X-210B RHR-M0-21B, RHR to Torus Air X-210B HPCI-17CV and HPCI-M0-25, HPCI Minimum Flew Line Air X-210B RRR-MO-16B, RHR-11CV, and RHR-13CV, RHR Minimum Flow Line Air X-210A and 211A RRR-M0-34A, RHR-MO-38A, and RHR-MO-39A, RHR to Torus Air X-210B and 211B RHR-MO-34B, RHR-M0-38B, and RHR-MO-39B, RHR to Torus Air X-212 RCIC-15CV and RCIC-37, RCIC Turbine Exhaust Air X-214 HPCI-15CV and HPCI-44, HPCI Turbine Exhaust Air X-214 HPCI-A0-70 and HPCI-AO-71, HPCI Turbine Exhaust Drain Air X-214 RHR-MO-166A and RHR-MO-167A RHR Heat Exch. Vent Air X-214 RRR-MO-166B and RHR-MO-167B RHR Heat Exch. Vent Air X-220 PC-230MV and PC-245AV, Purge and Vent Exhaust from Torus Air X-221 RCIC-12CV and RCIC-42, RCIC Vacuum Line Air X-222 HPCI-30 and HPCI-16CV, HPCI Turbine Drain
. Air
-174-
I6 4--.--*--
g 15 0 15 t -
- -- --* - - ' - - ~ ~ * ' -
- - ~ ~
D 15 -
2- - -
- 14.7 -
2
'.t 4 9.-. _-
. _sg... _. _...
u 1g
- 4 3.-
g
., 3 5 _ _ - _ ____
7 wo O-g3
\\
]y
______.r__-._____.___--.
wg
+_--3.___.-.---+p yg r __._
1_-----g
Md/t) 3 (x10 )
Eigure 3.11-1.1
>Iaximum Average Planar Linear Heat Generation Rate versus Exposure with LPCI >!odification and Bypass Flow Holes Plugged, Initial Core Fuel Type 3.
16 15 2~_ #5 6 3
15,4 :
a v
6 4 6 - ----- - -
x 1$
~-_._ g g.___
~~ ~ !3.9
,4 ga E-
_ _ _h -
. g
._._ _ __ __\\_ _,; 3.3___
.______.____.._...._____r.___,______,
gg 13 a-
___.-____._.._.._t
_._____q._________
+
w
_____.._,___x qg 6
12-
\\ _u_s.. _ y _ _._ _ _ _.. _ _ _,_
?.
xs
' _ _. _.h yy qg
.___._._____.g z
._.___._......j II g
z a
0 5
!O 15 20 25 30 IS 40 PLANAR AVERAGE EXPOSURE (>Md/t) 3 (x10 )
Figure 3.11-1.2 diaximum Average Planar Linear Heat Generation Rate versus Exposure with LPCI >!odification and Bypass Holes Plugged, Initial Core Fuel Type 2.
1866 219 211
e COOPER NUCLEAR STATION TABLE 4.2.F PRIMARY CONTAINMENT Sl?RVEILLANCE INSTRUMENTATION TEST AND CALIBRATION FREQUENCIES Instrument Instrument I.D. No.
Calibration Frequency Instrument Check Reactor Water Level NBI-LI-85A Once/6 Months Each Shift NBI-LI-85B Once/6 Months Each Shift Reactor Pressure RFC-PI-90A Once_/6 Months Each Shift RFC-PI-90B Once/6 Months Each Shift Drywell Pressure PC-PI-512A Once/6 Months Each Shift PC-PR-512B once/6 Menths Each ShifL Drywell Temperature PC-TR-503 Once/6 Months Each Shift PC-T1-505 Once/6 Months Each Shift Suppression Chamber PC-TR-21A Once/6 Months Each Shift j,
Air Temperature PC-TR-23, Ch. 1&2 Once/6 Months Each Shift l
o t
Suppression Chamber PC-TR-21B Once/6 Months Each Shift Water Temperature PC-TR-22, Ch. I62 Once/6 Months Each Shift l
Suppression Chamber PC-LI-10 Once/6 Months Each Shift Water Level PC-LR-11 Once/6 Months Each Shift PC-LI-12 Once/6 Months Each Shift PC-LI-13 Once/6 Months Each Shift Suppression Chamber PC-PR-20 Once/6 Months Each Shift.
Pressure Control Rod Position N.A.
N.A.
Each Shift Neutron Monitoring (APRM)
N.A.
Once/ Week Each Shift CD Torus to Drywell Ch PC-dPR-20 Once/6 Months Each Shift Differential Pressure O
Suppression Chamber /
PC-PR-20/513 (2)
Once/6 Months Each Shift Drywell Pressure (AP) g
6.6.2.G (Cont'd) usage evaluation per the ASME Boiler and Pressure Vessel Code l
Section III was performed for the conditions defined in the design specificaton.
The locations to be monitored shall be:
a.
The feedwater nozzles b.
The shell at or near the waterline c.
The flange studs 2.
Monitoring, Recording, Evaluating, and Reporting a.
Operational transients that occur during plant operations will, at least annually, be reviewed and compared to the transient l
conditions defined in the component stress report for the locations listed in 1 above, and used as a basis for the existing fatigue analysis.
b.
The number of transients which are comparable to or more severe than the transient evaluated in the stress report Code fatigue usage calculations will be recorded in an operating log book.
For those transients which are more severe, available data, such as the metal and fluid temperatures, pressures, flow rates, and other conditions will be recorded in the log book.
c.
The number of transient events that exceed the design specification quantity and the number of transient events with a severity greater than that included in the existing Code fatigue usage calculations shall be added. When this sum exceeds the predicated number of design condition events by twenty-five2, a fatigue usage evaluation of such events will be performed for the affected portion of the RCPB.
H.
Records of individual plant staff members showing qualifications, training and retraining.
6.6.3 Records and logs relating to the following items shall be kept for two years.
A.
The test results, in units of microcuries, for leak tests of sources performed pursuant to Specification 3.8.A.
B.
Records of annual physical inventories verifying accountability of the sources on record.
1866 221 1.
See paragraph N-415.2, ASME Section III, 1965 Edition.
^
The Code rules permit exclusion of twenty-five (25) stress cycles from secondary stress and fatigue usage evaluation.
(See paragrapns N-412(t)(3) and N-417.10(f) of the Summer 1968 Addenda to ASME Section III, 1968 Edition.)
-229-