ML19347C971
| ML19347C971 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/23/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19347C972 | List: |
| References | |
| NUDOCS 8103100414 | |
| Download: ML19347C971 (17) | |
Text
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UNITED STATES p(
NUCLEAR REGULATORY COMMISSION
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-K WASHINGTON, D. C. 20555
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NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.68 License No. DPR-46 1.
The Nuclear legulatory Comission (the Comission) has found that:
A.
The submittals by Nebraska Public Power District dated January 30 and October 15, 1980 and January 6,1981, comply 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 th? submittals, 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications
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as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:
(.2) Technical Soecifications Thi. Technical Specifications contained in Appendices A and B, as revised through Amendment No.68, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
8103100'//{.
. 3.
Th.is licens.e amendment is, effective as-of its date of issuance.
FOR THE NUCLEAR REGULATORY COM.""SION 044~
Thomas A.'
Ippolito, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
' Changes to the Technical
. Specifications Dated: February 23, 1981 i
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ATTACHMENT TO LICENSE AMENDMENT NO.68 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Remove the following pages of the Appendix "A" Technical Specifications and replace with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 111 tii 4
4 55 55 60 60 71 71 80 PJ 87 87 102 102 136 136 150 150 174 174 219a 219a 229 229 235a 235a i
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TABLE OF CONTENTS (cont'd)
SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENTS Pace No.
3.14 Fire Detection System 4.14 216b 3,15 Fire Suppression 'Jater System 4.15 216b 3.16 Spray and/or Sprinkler System (Fire Protection) 4.16 216e 3.17 Carbon Dioxide System 4.17 216f 3.18 Fire Hose Stations 4.18 216g 3.19 Fire Barrier Penetration Fire Seals 4.19 216h 5.0 MAJOR DESIGN FEATURES 217 - 218 6.0 ADMINISTRATIVE CONTROLS 219 - 237 6.1 Organization 219 6.2 Review and Audit 220 6.3 Station Operating Procedures 226 6.4 Actions to be Taken in the Event of Occurrences Specified in Section 6.7.2.A.
227 6.5 Actions to be Taken if a Safety Limit is Exceeded 227 6.6 ~ Station Operr. ting Records 228 6.7 Station Reporting Requirements 230
' 6.8 Environmental Qualification 235a 6.9 Systems Integrity Monitoring Program 235a 6.10 It, dine Monitoring Program 235a
-iii-Amendment No. _78, H,yy, 68
_.__.m l
K.
Limiting Safety System Setting (LSSS) - The limiting safety systes settings are settings on instru=entation wcich initia:e :he au:o=a:ic protective
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ac:1on at a level such that the safety limits will not be exceeded. The 4
region between the safety limit and :hese settings represent a margi; with normal operation lying belev these set:ings. The margin has been established so that with proper operation of the instru=entation :he safe:y limits will never be exceeded.
L.
Mode - The reactor mode is established by the sede selec:or-switch. The modes include refuel, run, shu:down and startup/ hot standby which are defined as follows:
1.
Refuel Mode - The reactor is in the refuel mode when :he 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 =odo Obe reactor system pressure is at or above 825 psig and the reac:;-. :ctection system is energi:ed with APRM l,
-protection (excluding
.e 15% high flux : rip) and RSM interlocks in service.
3.
Shutdown Mode - 1he reactor is in the shutdown mode when the reactor made sw1:ch is in the shutdown mode posi: ion.
4 Startup/ Mot Standby - In this sede the reac:or protection scram trips iniciated by the main steam line isolation valve closure are bypassed when reac:or pressure is less than 1000 psig, che low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energi:ed with APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod w1:hdrawal interlocks in service.
M.
Operable - A system or couponent shall be considered operable when it is capable of perfor=ing 1:s intended function in its required manner.
N.
Operating - Operating means that a system or component is performing its intended functions in 1:s required manner.
O.
Operating Cvele'- In erval between the end of one refcaling outage and the end of the next'subsecuen: refueling outage.
T
?.
Primary Containment Integrity - Primary containment integri:7 =eans
- hat the drywell and pressure suppression chamber are intact and all of the:following conditions are satisf,1ed:
1.
All manual containment isolation valves on lines connec:ed to the' reactor coolant system or containment which are not required to
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Joe open during accident conditicas are closed.
2.
A: least one door in each air lock is closed and sealed.
k 1 Amendment No.'68
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t COOPER Ni!CI. EAR STL.ON Tall 1.E 3.2.11 (PAGE 3) f RESIDilAI. IIEAT.REHOVAl. SYSTEM (I.PCI HODE) CIRC 111TRY REQlllRFJIENTS i
Minimum Number of Action Requi t eil When Instrument Operable Components Component Operalillity Inst rument I.D. No.
Setting I.imit Per Trip Systein (1)
Is Not Assured j
RilR Pump I.ow Flow RIIR-dPIS-125 A 6 B 12500 gpm I
A l
Time Delays RllR-TDR-K45, lA&lB 4.25;Tj.75 min.
I A
I IUIR Pump Start HilR-TDR-K75A & K70ll 4.5_4 5.5 Sec.
I A
2 Time Delay RilR-TDR-K75B & K70A
<.5 sec.
1
.A RilR lleat Exchanger RIIR-TI)R-K93, A & 11
- 1. 8_4_<2. 2 m i n.
I 15 flypass T.ll.
IllIR Crosst le Va lve RilR-l.HS-2 Valve Not closed (3)
E Position llus I A low Vo l t.
27 X 3/lA 1.oss of Voltage I
11 Aux. Relay lius lit 1.ow Vo l t.
27 X 3/Ill I.oss of Voltage I
11 Aux. Ite l.iy g
l Ilus IF low Volt.
27 X.1/lF I.oss of Voltage i
11 Aux. Relays 27 X 2/lF Loss of Voltage I
il ilus 1G 1.ow Volt.
27 X 1/lG I.oss of Vol tage I
11 Aux. Helays 27 X 2/IG 1.oss of Voltage Pump D1ocharge I.Inu CH-PS-266 15 psIg (3)
D CH-PS-270 115 psig (3)
D l
Emergency linses 27/lF-2, 27/lFA-2 3600 15%
2 is lludervoltage Relays 27/lG-2, 27/ lGil-2 8 seconal 12 sec.
2 11 (elegraded voltage)
I 27/ET-2 time delay 1
11 Emergency lluses I.oss 27/lF-l, 27/lFA-1, 2900V 15%
of Voltage Relays 2 7/ lG-1, 2 7 / lGil-1, 5 second il sec.
27/ET-1 delay I
H Emergency lloues tinder-27X7/lF, 27X7/lG, Voltage Itelays Timers 27X10/lG 10 second 12 sec.
I I1 Amendment No. 68 i
I
O.
NOTES FOR TA3LI 3.2.3 1.
'4 hen any ECCS systes is required :o ha operable, there shall be two operable crip systa=s except as noted. If a require =ent of the fourth colu=n is re-duced by one, the indicated action shall be taken.
If the same function is inoperable in = ore than one trip system or che fourth colu=n reduced by more than one, action 3 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 ac: ion 3.
3.
Declare the system or co=ponen: inoperable.
C.
Immediately take action 3 until power is verified on :he : rip systes.
D.
The high point vent shall be vented weekly upon failure of PS 73A or 3, PS 296, PS 268, PS 269, PS 270.
l 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.
a.
Requires one channel from each p'aysical location in the steam line space.
5.
One relay senses each phase of MCC-5 and the LO relay is a ::ansfer permissive relay.
Amendment No. 68.-_
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,1 COOPER NilCl. EAR STATION TAlli.E 4. 2.11 - (Page 2)
HilR SYSTEM TEST & cal.lBRATION FREQllENCIES Instrument 1 tem
-Item I.D. No.
FameLiuna1 Test Freq.
CaIIbrat1on Freq.
Check Inut rument at lon.
1.
Drywell liigh Pressure PC-PS-101, A,li C _.&.D Once/Honth (1)
Once/3 Honths None 2.
Heactor Vessel Shroud I.evel Hill-1.lTS-73, A & 11 il once/Honth (1)
Once/3 Honths onee/ Day 3.
React or 1.uw Pressure HR-PS-128 A & 15 Once/Honth (1)
Once/3 Hunths None 4.
Heactor lew Pressure NBI-PS-52 A & C Once/Honth (1)
Once/3 Honths None N81-PIS-52 11 & D 5.
Drywell Press.-Containment PC-PS-Il9, A II.C & D Once/Honth (1)
Once/3 Honths Nene Spray 6
HilR Pump Discharge Press.
HilR-PS-120, A, B.C & D '
Once/Honth (1)
Once/3 Honths None 7.
HilR Pinup Disclearge Press.
HilR-PS-105, A,II,C & D Once/Hontti (1)
Once/3 Honths None 8.
RilR Pump I.ow Flow Switch HilR-dPIS-125 A & B once/Honth (1)
Once 3 Honths None b
9.
RilR Pump. Start Time Delay HilR-TDR-K70, A & B Once/Honth (1)
Once/Oper., Cycle None
'I 10 HilR Pump St art Time Delay RilR-TDR-K75, A & 11 Once/Honth (1)
Once/Oper. Cycle None 11.
lillR ' llea t Exchanger liypass T.D.
HilR-TDR-K93, A & B Once/Hunth (1)
Once/Oper. Cycle None 12.
EllR Cross ". le Valve Posi tion HilR-l.HS-2 Once/Honth (1)
N.A.
13.
1.uw Voitage Relays 27 X 3/1A (7)
None 14.
l.ow Volinge He?ays 27 X 3/111 (7)
None 15.
I.ow Voltage Relays 27 x 2/IP, 27 X 2/1C (7)
None 16.
l.ow Voltage Relays 27 X 1/lF, 27 X (1)/lG (7)
None l
17.
Pump D!ach, l.ine Press. l.ow CH-PS-266, CH-PS-270 Once/3 Honths once/3 Hontlis None 18.
Emergency buses lindeixoltage 27/lF-2, 27/lFA-2, 27/IG-2, once/Houth once/18 Honths Once/12 hrs.
Helays-(Degraded Voltage) 2 7/IGB-2, 27/hT-2 19.
Emergency linueu 1.oss of 27/lF-1, 17/lFA-1, 27/lG-1, once/Honth once/18 Honths once/12 bra.
Voltage Relays 3
27/ICB-1, 17/ET-1 20 Emergency linses Undervoltage 27X7/lF, 27X7/lG, 27X10/lG Once/Honth once/18 Honths None Relays Timers Amendment No. 68 1
1 C00 Pelt NilCI. EAR STATION TAlil.E te. 2. F
' PRIMARY CottrAINHEttr SHHVEILI.ANCE INSTRUMENTATION TEST AND CAI.lilRATION FREQllENCIES Ins t ruinent inst rinnent 1.11 No.
Callbration Frequency Ins t rinnent Check Heact or Wat er I.evel NBI-l.1-8SA Once/6 Honths Each Shift Hill-l.1-858 once/6 Honths Each Shift Heactor Pressure RFC-PI-90A Unce/6 Honths Each Shift RFC-PI-901s Once/6 Honths Each Shift lirywell Pressure PC-PI-512A Once/6 Honths Each Shift i
PC-PR-512tt once/6 Honths Each ShifL liryweii Teu.perature PC-TR-503 Once/6 Hunths Each ShiIt PC-TI-505 Once/6 Honths Each Shiit Suppression Chau.her PC-TR-21A once/6 Honths Each Shilt 6
Ai r Tenipe ra t'u r e PC-TR-23, Ch. 1&2 Once/6 Hunths Each Shift l
o 8
Suppression Chaniber PC-TR-21B Once/6 Honths Each Shift Wa t e r Teuipe ra t u re PC-TR-22, Ch. 1&2 Once/6 Honths Each Shift l
- Suppression Chaniher PC-1.1-10 Once/6 Honths Each Shift Water f.evel PC-l.R-I l Once/6 !!anths Each Shift PC-1.1 - 12 Once/6 Honths Each Shitt PC-l.I-13 Once/6 llant hs Each Shitt Suppresulun Chauiber PC-Pit-20 Once/6 Honths Each Shift Pressure i
Cont rol Rod Posillon N.A.
H.A.
Each Shift Neutrun Honitorlog (APRH)
H.A.
Once/ Week Each Shift Torus to Drywell PC-drit-20 Once/6 Hunths Each Shift Diiferenttal Poessure suppresston Chasuber/
PC-PR-20/513 (2)
Once/6 Honths Each ShifL Drywell Pressure (AP)
Amendment No. 68 1
3.2 BASES (cont'd.)
Trip settings of <100 mr/hr for the monitors in the ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas eteat-ment system operation so that none of the activity released during the re-fueling accident leaves the Reactor 3u11 ding via the normal ventilation path but rather all the activity is processed by the standby gas treatment system.
Flow transmitters are used to record the flow of liquid from the dryvell sumps. An air sampling system is also provided to detect leakage inside the primary containment.
For each parameter monitored, as listed in Table 3.2.F. there are two (2) channels ef instrumentation. By comparing readings between the two (2) channels, a near continuous survelliance of instrument performance is available.
bla. Any deviation in readings will initiate an early recalibration, there-by maintaining the quality of the instrument readings.
The recirculation pump trip has been added as a means of ilmiting the con-sequences of the unlikely occurrence of a failure to scram during an antici-pated transient. The response of the plant to this postulated event falls within the envelope of study events given in General Electric Company Topical Report, NEDO-10349, dated March,1971.
The liquid radvaste monitor assures that all liquid discharged to the discharge canal does not cxceed the limits of Section 2.4.1.b of the Environmental Technical Specifications. Upon sensing a high discharge level, an isolation signal is generated which closes of radwaste discharge valve. T5e set point is adjustable to compensate for variable isotopic discharges and dilution flow rates.
The main control room ventilation isolation is provided by a detector monitoring the intake of the control room ventilation rystem. Automatic isolation of the normal supply and exhaust and the activation of the emergency filter system is provided by the radiation detector trip function at the predetermined trip level.
The mechanical vacuum pump isolation prevents the exhausting of radioactive gas thru the 1 minute holdup line upon receipt of a main steam line high radiation signal.
The operability of the reactor water level instrumentation in Tables 3/4.2.F ensures that sufficient information is available to monitor and assess accident situations.
Amendment No. 68 b
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3.3 and 4.3 3ASES:
(Cont'd) flux. The require =ents of a: least 3 counts per second assures that any :ransient, shculd i: occur, begins at or above the initial value of 10-3% of raced pcwer used in :he analyses of transients cold cen-dicions. One operable SEM channel vould be adequate to scnitor the approach to criticalitj using hemogeneous pa:: erns of scattered centrol rod w1:hdrawal. A ninimum of two operable SRM's are provided as an added conservatism.
5.
The Rod Block Monitor (R3M) is designed to au:c=a:1cally prevent f us.1 damage in the event of erroneous rod w1:hdrawal fr:s loca:1ons of high power density during high pewer level eperation. Two channels are pro-vided, and one of these may be bypassed frem the console for maintenance and/or testing. Tripping of one of :he channels will block erroneous rod withdrawal socn enough to prevent fuel damage. This sytem backs up the operator who wi:hdraws centrol rods according :o wri::en se-quences. The specified restrictions w1:h one channel out of service ccuservatively assure that fuel draage vill not occur due to red vi:h-drawal errors when this condition exists.
A limiting centrol rod pattern is a pat:ern which results in the core being on a thermal hydraulic 11=1: (i.e., MC?R = 1.07, and LEGR = as defined in 1.0.A.4).
During use of such pa: terns, i: is judged : hat tes:ing of :he REM system prior to vi:hdrawal of such reds :o assure 1:s operability will assure that i= proper withdrawal does not occur.
I: is the respensibility of the Reac:or Engineer :o identify these limiting patterns and the designated rods either unen the patterns are ini:1 ally established or as they develop due :o the occurrence of inoperable control rods in other than li=1 ting patterns. Other person-nel qualified to perform :his function =ay be designated by the station superintendent.
C.
Scram Insertion Ti es The control red system is designed :o bring the reactor suberi:ical at a race fasc.enough to prevent. fuel damage; i.e.,
- o prevent the MCPR f ca becoming less :han the safe:y 1131:. The lisi:ing pcwer transien: is defined in Reference 3.
Analysis of this ::ansiene shcws cha: che negative reactivity races resul:ing f res :he scram provide the required protec:1on, and MCPR remains greater than the safety 11mic.
On an early 3WR, some degradation of control rod scram performance occurred
-during plant s:artup and was determined to be caused by particulate material (peobably cons:ruction debris) plugging an. internal control red drive fil:ar.
The design of :he present centrol red drive (Model CRD31443) is gressly i= proved by the relocation of the fil:e to a loca:1cn out of.:he scram drive
-pa:h; i.e.,
it can no 1:nger in:erfere with scram performance, even if ec=pletely blocked.
Amendment No. 68
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LIMITI1;G CO:OITIO:IS FOR OPERATIC:I SURVEILLA::CE REQUIRCIE ITS 3.6.D Safety and Relief Valves 4.6.D Safety and Relief Valves 1.
During reactor power operating condi-1.
Approximately half of the safety valves tions and prior to reactor startup and relief valves shall be checked or f rom a Cold Condition, or whenever replaced with bench checked valves reactor coolant pressure is greater once per operating cycle. All valves than atmospheric and temperature will be tested every two cycles.
greater than 212 F, all three safety valves and the safety modes of all The set point of the safety valves relief valves shall be operable, ex-shall be as specified in Specification cept as specified in 3.6.D.2.
2.2.
2.
2.
At least one of the relief valves shall be disassembled and inspected each re-a.
From and af ter the date that the fueling outage.
safety valve function of one relief valve is made or founa to be inope ra-J.
The integrity of the relief safety valve ble, continued reactor operation is bellows on any three stage valve permissible only during the succeeding shall be continuously =onitored.
thirty days unless such valve function is sooner made operable.
4 The operability of the bellows monitoring system shall be demonstrated once every b.
From and af ter the date that the safety three months when three stage valves valve function of two relief valves is are ins talled.
made or found to be inoperable, con-tinued reactor operation is permissible 5.
Once per operating cycle, with the only during the succeeding seven days reactor pressure > 100 psig, each relief unless such valve function is sooner valve shall be' manually opened until made operable.
the main turbine bypass valves have closed to compensate for relief valve 3.
If Specification 3.6.D.1 is not met, opening.
an orderly shutdown shall be initiated 6.
and the reactor coolant pressure shall a.
Operability of the relief valve position be reduced to a cool shutdown condi-indicating pressure switches and the tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
safety valve position indicating thersocouples shall be demonstrated 4.
From and af ter the date that position once per operating cycle.
indication on any one relief valve is made or found to be inoperable, contin-b.
An Instrument Check of the safety and ued reactor operation is permissible relief valve position indicating devices only during the succeeding thirty days shall be performed monthly.
unless such valve position indication is sooner uade operable.
Amendment No.68
-136-
3.6.D & 4.6.D BASES (cont'd.)
The relief and safety valves are bench tested every second operating cycle to ensure that their set points are within the + 1 percent tolerance. Additionally, once per operating cycle, each relief valve is tested manually with reactor pressure above 100 psig to demonstrate its ability to pass steam.
The requirements established above apply when the nuclear system can be pres-surized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. Ilowever, these transients are much less severe, in terms of pressure, than starting at rated conditions.
The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
The position indicating pressure switchea for the relief valves and the thermo-couples for the safety valves serve as a diagnostic aid to the operator in the event of a safety / relief valve failure.
If position indication is lost, alternate means are available to the operator to determine if a safety valve is leaking.
E.
Jet Pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase n.e cross-sectional flow area for blowdown following the design bases double-ended line break. Therefore, if a failure occurs, repairs must be made.
The detection technique is as follows. With the two recirculation pumps balanced in speed to within + 5%, the flow rates in both recirculation loops will be verified by Control Room monitoring instruments. If the two flow rate values do not differ by more than 10%, riser and nozzle assembly integrity has been verified.
If they do differ by 10% or more, the core flow rate measured by the jet pump dif fuser differencial pressure system must be checked agains t the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10% or more (with the derived value higher) dif fuser measure-ments will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the plant shut down for repairs. If tha potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115% to 120% for a single nozzle failure).
If the two loops are balanced in flow at the sama pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.
In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive dif ferential pressure but the net effect would be a slight decrease (3% to 6%) in the total core flow measured.
This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump dif fuser dif ferential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of a Amendment No.68
-150-
- % M E9
~ ~.-.. _
e TABLE 3.7.4 (Page 2)
PRIMARY CONTAIW. INT TESTA 3LE ISOLATION VALVES TEST PEN. NO.
VALVE NDf3ERS MEDIA I',
X-39A RHR-MO-26A and RHR-MO-31A, Drywell Spray Header Supply Air X-39B RHR-MO-263 and RHR-MO-313 Drywell Spray Header Supply Air
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.X-41 RRV-740AV and RRV-741AV, Reactor 'Jacer Sanple Line Air X-42 SLC-12C7 and ELC-13C7, Standby Liquid Control Air X-205 PC-233MV and PC-237AV, Purge and Vent Supply to Torus Air i
X-205 PC-13C7 and PC-243AV, Torus Vacuum Kelief 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-13C7, RCIC Minimum Flow.ine Air X-210A RHR-MO-21A, RHR to Torus Air X-210A RHR-MO-16A, RHR-10CV, and RHR-12CV,' RER Minimum Flow Line Air X-210B RHR-MO-213, RHR to Torus Air X-210B HPCI-17C7 and HPCI-MO-25, HPCI Minimum Flow Line Air X-2103 RHR-MO-163, RHR-llc 7, and RHR-13CV, RHR Minimum Flow Line Air X-210A and 211A RER-MO-34A, RHR-MO-38A, and RHR-MO-39A, RER to Torus Air X-2103 and 211B
- RHR-MO-343, RHR-MO-383, and RHR-MO-393,- RHR to Torus Air X-212 RCIC-15C7 and RCIC-37,-RCIC Turbine-Exhaust Air X-214 HPCI-15CV and HPCI-44, HPCI Turbine Exhaust Air X-214 HPCI-AC-70 and HPCI-AO-71, HPCI Turbine Exhaust Drain Air
_X-214
_ RHR-MO-166A and RER-MO-167A RER Heat Exch. Vent Air X-214
' RER-MO-166B and RER-MO-1673 PSR Heat Exch. Vent -
- Ad :
1X-220-PC-230MV and PC-245AV, ? urge and Vent Exhaust from Torus Air X-221' z RCIC-12C7 and RCIC-42, RCIC Vacuum Line Air '
- X-202 HPCI-50-and HPCI-16C7, HPCI Turbine Drain Air Amendment No. 68 171-7 4
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G.
A Fire Brigade of at least 5 members shall be maintained at all times. This excludes the 3 members of the minimum shif t crew necessary for safe shutdowns, and other personnel required for other essential functions during a fire emergency. Three fire Brigade members shall be from the Operations Department and 2 support members may be from other departments inclusive of Security personnel.
Fire Brigade composition may be less than the minimum requirements for a period cf 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 accoccodate unexpected absence of Fire Brigade members provided i= mediate action is taken to restore the Fire Brigade to within the minimum requirements.
H.
In order to perform the function of accident assessment an engineer f rom the normal plant engineering staf f shall be assigned to each shif t during reactor operation.
If the lack of qualified engineers necessitates, an additional senior reactor operator assigned to each shif t may substitute in the performance of the accident assessment funct ion. This requirement is ef fective until January 1,1981.
6.1.4 The minimum qualifications, training, replacement training, and retraining of plant personnel at the ti=e of fuel loading or appointment to the active position shall meet the requirements as described in the American National Standards Institute U-18.1-1971,
" Selection and Training of Personnel for Nuclear Power Plants".
The Assistant to Station Superintendent qualifications shall comply with Section 4.2 of ANSI-N18.1-1971. The Chemistry and Health Physics Supervisor shall meet or exceed the qualifications of Regulatory Guice 1.8, Sept. 1975; personnel qualification equivalency as stated in the Regulatory Guide may be proposed in selected cases. The mint =na frequency of the retraining program shall be every two years. The training program shall be under the direction of a designated me=ber of the plant staff.
A.
A training program for the fire brigade vill be maintained under the direction of the plant training coordinator and shall meet or exceed the requirements of Section 27 of the NFPA Code 1970, except for Fire Brigade training sessions which shall be held at least quarterly.
The training program requirements vill be provided by a quali-fied fire protection engineer, thru the Risk Manager.
Amendment No. 68
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6.6.2.G (Con:'d) usage evaluation per the ASME Soiler and ?: essure 'lessel Code l
Section III was performed for the condi:icns defined in the design specificaten. The loca:icns :o be =eni:ored shall be:
a.
The feedwa:er no::les b.
The shell at or near the wa:erline 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 ce= pared to the transient l
conditices defined in the cc=ponent stress repor: for the locations listed in 1 above, and used as a basis for :he existing f atigue
- analysis, b.
The nu=ber of transients which are co= parable to or more severe than the transien: evaluated in :he stress repor: Code fatigue usage calculations will be recorded in an operating log book.
For those ::ansients which are = ore severe, available data, such as the metal and fluid te=pera:ures, pressures, flew races, and other conditiens will be recorded in the icg back.
c.
The number of ::ansient even:s tha: exceed :he design specifiestion quantity and the nu=ber of transient events with a severity greater than that included in :he existing Code fatigue usage calculations shall be added. When this sum exceeds :he predica:ed number of 2
design condi:icn events by twenty-five, a fa:igue usage evaluation of such events will be perfor=ed for the af fected por:1cn of :he RC?B.
'd.
Records of individual plan staff se=bers shewing qualificaticas,
- sining and retraining.
6.6.3
-Recor-ds and logs rela:ing :2 :he follcwing 1: ems shall be kept for evo years.
A.
The test resul:s, in units of mic ccuries, for leak :ests of scurces performed pursuan: :o Specifica:1on 3.3.A.
3.
Records of annual physical inventories verifying accountability of the sources on. record.
1.
See paragraph.N-415. 2. ASME See: ion III, 1965 Edi:1cn.
2.
The Code rules per=1; exclusica of :wenty-five (25) s::ess cycles f:ce secondary
~
s::ess and fatigue usage evaluation.
(See paragraphs N-412(:)(3) and N-417.10(f) of :he Su==e 1963 Addenda-:o ASME See:1cn !!!, 1963 Edi:icn.)
Amendment No. 68
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- D.
6.8 Environctntal Oualification A.
By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:
Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equip =ent in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interi:n Staff Position on Environnental Qualification of Safety-Related Electrical Equipment", December 1979.
Copies of these documents are attached to Order for Modification of License DPR-46 dated October 24, 1980.
B.
By no later than December 1,1980, complete aad audit 1ble records must be available and maintained at a central locatloa which describe the environmental qualification method used for all safety-ralated electrical equipment in suf ficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588.
Thereafter, such records should be updated and maintained current as equipment is replaced, further tasted, or otherwise further qualified.
6.9 Systems Integrity Monitoring Program A program shall be established to reduce leakage from systems outside the primary contaianent that would or could contain highly radioactive fluids during a serious accident to as low as practical levels. This program shall include provisions establishing preventive maintenance and periodic visual inspection requirements, and leak testing requiremente for each system at a frequency not to exceed refueling cycle intervals.
[
6.10 Io' dine Monitoring Program A program shall be established to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident l
conditions. This program shall include training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.
.. u L Amendment No. 68
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