ML20041G341
| ML20041G341 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/04/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Nebraska Public Power District (NPPD) |
| Shared Package | |
| ML20041G342 | List: |
| References | |
| DPR-46-A-077, TAC 42227, TAC 47003, TAC 47004 NUDOCS 8203220100 | |
| Download: ML20041G341 (11) | |
Text
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UNITED STATES
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'i NUCLEAR REGULATORY COMMISSION WASH!NGTON D. C. 20555
's...>...s NEBRASXA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION APENDPENT TO FACILITY OPERATING LICENSE Amendment No. 77 License No. OPR-46 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nebraska Public Power District dated October 21, 1980 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 3.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Ccmmission; C.
There is reasonable assurance (1) that the activities authori:ed 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 amencment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amerJed by changes to the Technical Spec-ifications as indicated in the attachment to this license amer.dment and paragraph 2.C(2) of Facility Operating License No. CPR-46 is hereby amended to read as fo11cws:
(2) Technical Specifications The Technical Specifications contained in Ascendices A and 3, as revised througn Amendment No. 77, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
0203220100 G20304 PDR ADOCK 05000298 P
. 3.
This license amendment is effective as of the date of issuance.
FOR THE fiUCLEAR REGULATORY COMMISSION Comenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Dated: March 4, 1982 l
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ATTACHMENT TO LICENSE AMEfl0 MENT fl0. 77 FACILITY OPERATING LICENSE NO. OPR-46 OOCKET NO. 50-298 Revise the Appendix A Technical Specifications as indicated below. Marginal lines on the revised pages indicate the area of change.
Remove Insert 61 61 62 62 62a 52a 77 77 86 86 98 98 104 104 105 105
d
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TABl.E 3.2.C COtiTit0L ROD WITilDRAWAL BLOCK IriSTRUMEllTATION
!n
'm 15 Minimum Nutther of
's
]S F""CLI""
Trip Level Setting Operable Instrument Channels / Trip System (5) u?
< (0.66W + 42%) FRP _ (2) 2(1)
I APRtltipscale (Flow Blas)
__ 12%
HFLPD 2(1)
.m APlotlipscale (Startup) 2(1) 1 N i
APRM Downscale (9) 1 2.5%
2(1)
APRtl Inoperative (10b)
I
._ (0.66W + 41%) (2)
RiiM lipscale (Flow Bias) i j
1 klift Downscale (9)
> 2.5%
1 RBtl Inoperative (10c)
IltM lipscale (8)
< 108/125 of Full Scale 3(1)
- 1 3(1)
! 'i' 11G1 Downscale (3) (8)
> 2.5%
t 1
3(1) litM De t ec t o r tio t Full In (8) 3(1) 1RH Inoperative (8)
(10a)
_ 1 x 10 Counts /Second 1(1)(6) 5 SRM lipscale (8) 1(1)(6)
SRil Detector tiot Full in (4)(8)
(> 100 cps) 1 1(1)(6) l SRM inoperative (H)
(loa)
_ 10% Difference In Rectre. Flows 1
Flow Blas Coinparator I
J
_ 110% Recirc. Flow Flow hias lipucale/Inop.
SRtl Downscale (8)(7)
> 3 Counts /Second (11) 1(1)(6) 1(12)
SI)V Wa t e r I.e ve l lii gh
< 18 gallons i
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!Mi.'E5 FWt TA3LJ 3.2.C 1.
For the startup and run positions of the itenetor bde Selector S. tit h, the Cou< rol Rod Withdraaal Block Inatrumentation trip system shall be i. :rable f or each function. The SP21 and 1101 block; need not be ope rable in " dun" mode, and the AP;01 (flui biased) and RJ:! ro l Slacks need not be op e r a le in
";tartup" mode. The Control Rod 'lithdrawal diock Instrumentation trip syaten is t one out of "n" trip system, and as such requirea thnt w!y one ins trument channel sp-:ci.f ted in the funct ion coluin maat exceed the Trip L.
el Se t t ing to cau,e a rod b lock.
By utilizing the RPS bypass l og i.e (see n ce 5 below and note 1 of Table 3.1.1) for the Cont rol Rod Uithd c i.:a '
Dlock Ins trumentation, a suf ficient nu.nbe r ef ins t rume nt channel; u li alwayt be ope rabic to provida redundint rod withdra. cal block protect ion.
2.
N is the recircul:ition loop flow in pereant of design. Trip level setting is in pereunt af rated power (23dl :-1Wt).
3.
l'Ci downscale is bypassed when it is on its lowest range.
4 This function in hypassed when the count is > 100 cps and I101 above range 2.
5.
Sy 6 sign one ins trument channel; i.e., one APR21 or I101 per RPS trip sys tem
'ay b.
bypassed.
For the AP!C1's and IRM's, the minimun number of channels specified is that minimuu number requirel in each RPS channel and does not refer to a ninimum number required by the control rod block instrumentation trip function.
By design only eue of two R3:1's or one of four SRM's may be h yp.i u sel.
Fo r t he SPJ1's, the m.'nimua number of channels specified is the minticn num.ber re luired in each of the two cir:uit loops of the Control Rod Block Ins trumentat Lon Trip Sys teu.
For the R3:i' s, the minimun nunber of channels specif tel is the ninimtu number required by the Control Rod Block Instrumentation Trip Systen as a whole (except when a limiting control rod patter 1 exists and the requirements of Specifica tion 3.3.3.5 apply).
6.
LP'I channels A,E,C,G all in rande 8 or higher bypasses SR'1 channels A&C functions.
Iic! ciannals B,F,U,il all in range 8 or higher bypasses S:01 channels BSD functions.
7.
This function is hypass=1 when IR11 is above range 2.
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8.
This f uact ion is by,u,."'. chea the node switch is pined in Run.
9.
Zh a function is onlj m ivo ahen the mole s.; itch is in Run.
This function la n t o c b; ally bypn.4ed /w a the 1:01 instrume it a tion is ope rabic an.! not high.
10.
in op e :ti t ive tric, c produced by the follning f unctions:
a.
SR1 and IRM (1).lede swi tah not in ope rate (2)
Poaer 3 apply volt go low (3) Circuit huarts nat in circuit i
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Amendment No. M, 32, M 77
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i I:0Tdb F0k T.\\nL-; 3.2.C (Continued) i b.
APT!
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(1) !!o.la saitch not in operato (2) L e t., titan 11 LPlet input.s (3) Circult boards not Lt ci.r cu t t 1
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c.
Rii:1 (1)
Mo.i witch not in operate j
j (2) Ctreult boards not in circutt
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(3) P&1 f.iils to null i
(4) Lc i th in required numiba r d LP;C1 inputi tSr rol selected I
11.
During n;! ral unloading /relo.tding, the 3101 couat race sill be below 3 cps for sr.e pa r tal of time.
See Specitication 3.11.B.
12.
!Jith the nur.ha r of 0?EMBLE channel, lens than requird by tite Minimun 3
liu thar of operable las trument Channels /'f rip Sys tem requirene.ns, place the inap n able channel in the tripped condition wit:iin eau hour.
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I Amendment No. 61 77
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- R TABl.E 4.2.C 9
SURVEILI.ANCE REQtlIREMENTS FOR ROD WITilDRAWAL Bl.0CK INSTRtIMENTATION t
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Functiona1 D
Function Test Freq.
Calibration Freq.
Instrument Check s
APRM tipucale (Flow Blas)
(1)
(3)
Once/3 Months once/ Day 3
APRM Upscale (Startup Mode)
(1)
(3)
Once/3 Montiis Once/ Day j
APRh Downscale (1)
(3)
Once/3 Months once/ Day i
APRh inoperative (1)
(3)
N.A.
Once/l)ay RiiM lipscale (Flow Blas)
(1)
(3)
Once/6 Months once/ Day i
niiM Downseale (1)
(3)
Once/6 Months once/ Day Riin.inope rat ive (1)
(3)
N.A.
Once/ Day IRtl ;lpacale (1)
(2)
(3)
Once/3 Months Once/ Day IRM Downscale (1)
(2)
(3)
Once/3 Months once/ Day IRM Detector Not Full In (2)
(Once/oper.
Once/Oper. Cycle (10)
Once/ Day ating cycle)
Nm
- I 1RM luoperative (1)
(2)
(3)
N.A.
N.A.
l SRM lipscale (1)
(2)
(3)
Once/3 Months Once/ Day i
SRM Downscale (1)
(2)
(3)
Once/3 Months once/ Day SRH Detect or Not Full In (2)
(Once/oper-Once/Oper. Cycle (10)
N.A.
ating cycle)
SRit inoperative (1)
(2)
(3)
N.A.
N.A.
Flow Bias Comparator (1)
(8)
Once/Oper. Cycle N.A.
Flo.' lilas lipscale (1)
(8)
Once/3 Months N.A.
i Rod lilock I. ogle (9)
N.A.
N.A.
l RSci Rod Group C Bypass (1)
(11)
Once/3 Months N.A.
SDV lii gli Wa ter f.evel Quarterly once/Oper. Cycle N.A.
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e 3.2 3ASES.(cont'd.)
prevention of critical heat flux in a local region of the core, for a single rod withdrawal error from a limiting control rod pattern.
The IRM rod block function provides local as well as gross core protection.
The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.
A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion and thus, control rod cotion is prevented.
The downscale trips are set at 2.5 indicated on scale.
i The flow co=parator and scram discharge volu=e high level co=ponents have only one logic channel and are not required for safety.
The SDV high level rod block does provide adequate ti=e to determine the cause of the level increase and take corrective action prior to automatic scram.
The refueling interlocks also operate one logic channel, and are required for a safety only when the mode switch is in the refueling position.
The effective emergency core cooling for small pipe breaks, the EPCI system,
=ust function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time.
The automatic pressure relief function is provided as a backup to the HPCI in ti.e event the HPCI does not operate.
The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation.
The trip settings given in the specification are adequa:e to assure the above criteria are =et.
The specificatibn preserves the effectiveness of the systes during periods of maintenance, testing, or calibration, and also minimizes the risk of inadver-1 cent operation; i.e., only one instru=ent channel out of service.
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Two air ejector off-gas =cnitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line.
Isolation is initiated when both instruments reach their high : rip point or one has an upscale trip and the other a downscale trip. There is a fifteen minute delay accounted for by the 30-=1nute holdup time of the off-gas before it is reached to the stack.
r Both instru=ents are required for trip but the instruments are so designed that any instrument failure gives a downscale trip.
The trip setting c: 1.0 ci/sec l
(prior to 30 min. delay) provides an i= proved capabili:y to detect fuel pin cladding f ailures to allow prevention of serious degradation of fuel pin cladding integrity which might result f rom plant opera: ion with a misoriented or =isloaded fuel assa=bly.
This limit is = ore restrictive than 0.39 ci/see noble gas release the air ejectors (after 30 min. delay) which was used as the source term rate at for an accident analysis of the aug=ented of f-gas system. Using the.39 ci/sec the maxi =um off-si:e total body dose would be less than the.5 rem limit.
source term, Two radiation =enitors are provided which iniziate the Reactor Building Isolation j
function and operation of the standby gas creatment syste=. The trip is actuated by one hi-hi or two downscale indications.
Amendment No. 30, $1 77
-e e.
SURVEILLANCE REOUIREMENTS 7
LIMITING CONDITIONS FOR OPERATION
- 4. 3.C (Cont ' d. )
- 3. 3.C (Cont' d. )
3.
The maximum scram insertion time for 90% insertion of any operable control j
rod shall not exceed 7.00 seconds.
D.
Reactivity Anomalies I
D.
Reactivity Anomalies During the startup test program and At a specific steady state base condi-startup following refueling outages, I
tion of the reactor actual control rod the critical rod configurations will inventory will be periodically com-be compared to the expected configura-pared to a normalized computer pre-cions at selected operating conditions.
3 diction of the inventory. If the These comparisons will be used as base difference between observed and pre-data for reactivity monitoring during dicted rod inventory reaches the subsequent power operation through-equivalent of 1" ik reactivity, the out the fuel cycle. At specific power reactor will be shut down until the Operating conditions, the critical rod cause has been determined and correc-configuration will be compared to the tive actions have been taken as configuration expected based upon ap-appropriate, propriately corrected past data. This 1
comparison will be made at least every E.
Recirculation Pumps full power month.
A recirculation pu=p shall not be started while the reactor is in natural circulation flow and reactor power is greater than 1% of rated ther=al power.
t F.
If Specifications 3.3.A through D I
above cannot be met, an orderly I
shutdown shall be initiated and the reactor shall be in the Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C.
1.
The scram discharge volume (SDV) vent and drain valves shall be cycled and verified open at least once every 31 days and prior to reactor start-up.
2.
The SDV vent and drain valves shall be verified to close within 30 sec onds after receipt of a signal for control rod scram cnce per refueling l cvele.
i 3.
SDV vent and drain valve operabil-icy shall be verified following any -sintenance or modification to I
any portion (electrical or mechan-ical) of the SDV which may affect the operation of the vent and drain valves.
Amendment No. 32 77 93
3.3 and 4.3 BASES:
(Cott'd)
The occurrence of scram times within the li=its, but significantly longer than the average, should be viewed as an indication of a syste=atic problem with control rod drives.
In the analytical treatment of the transients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and start of motion of the control rods.
This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. Approximately the first 90 milliseconds of each of these time intervals result from the sensor and circuit delays; at this point, the pilot scram solenoid deenergizes. Approximately 120 milliseconds later, the control rod =otion is esti= aced to actually begin. However, 200 milliseconds is conservatively assumed for this ti=e interval in the transient analyses and this is also included in the allowable scram insertion times of Specification 3.3.C.
The time to deenergize the pilot valve scram solenoid is measured during the calibration tests required by Spec 4.1.
D.
Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.
The magni-tude of this excess reactivity may be inferred from the critical rod con-figuration. As fuel burnup progresses, anomalous behavior in the excess reactivity =ay be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state.
Power operating base conditions provide the most sensitive and directly inter-pretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.
t Requiring a reactivity co=parison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds l
1% ak. Deviations in core reactivity greater than 1% ak are not expected and require thorough evaluation. One percent reactivity limit is con-sidered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.
1 E.
Recirculation Pu=os Until analyses a-e submitted for review and approval by the NRC which prove that recirculation pump startup from natural circulation does not cause a reactivity insertion transient in excess of the most severe coolant flow increase currently analyzed, Specification 3.3.E prevents starting racirculation pu=ps while the reactor is in natural circulation above 1%
of rated thermal power.
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Amendment No. 62 77
_to;_
3.3 and 4.3 B ASES :
(Cont'd)
G.
Scram Discharge Volume To ensure the Scram Discharge Volume (SDV) does not fill with water, the vent and drain valves shall be verified open at least once every 31 days.
This is to preclude establishing a water inventory, which if sufficiently large, could result in slow scram times or only a partial control rod insertion.
The vent and drain valves shut on a scram signal thus providing a contained volume (SDV) capable of receiving the full volume of water discharged by the control rod drives at any reactor vessel pressure.
Following a scram the SDV is discharged into the reactor building drain system.
REFERENCES 1.
NEDO-10527, " Rod Drop Accident Analysis for Large Boiling Water Reactors," Paone, Stirn & Woolley, 3-72, Class I.
2.
NED0-10427, Supple =ent 1, " Rod Drop Accident Analysis for Large Soiling Water Reactors," Stirn, Paone & Yound, 7-72, Class I.
3.
" Supplemental Reload Licensing Submittal for Cooper Nuclear Station Unit 1,"
(most current approved submittal).
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l Amendment Nom 77 m