ML20126H777
| ML20126H777 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/03/1981 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126H782 | List: |
| References | |
| NUDOCS 8104200551 | |
| Download: ML20126H777 (32) | |
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UNITED STATES i
NUCLEAR REGULATORY COMMISSION m
E WASHINGTON, D. C 20555 o
'g IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE 00CXET NO. 50-331 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 License No. CPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light and Power Company, Central Iowa Power Cooperative, and Corn Belt Power Cooperative (the licensees) dated June 17, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in confomity with the aoplication, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) 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 ecmmon 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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-49 is hereby amended to read as follows:
(2) Technical Soecifications j
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 67, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
551 8104240
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This license amendment-is effective as of the date of its issuance.
t FOR THE NUCLEAR REGULATORY COMMISSION
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Thomas A.-Ippolito, Chief Operating Reactors Branch #2.
Division of Licensing
Attachment:
Changes to the Technical l
Specifications Date of Issuance: ~ April 3,.1981 '
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ATTACHMENT TO LICENSE AMEN 0 MENT NO. 67 FACILITY OPERATING LICENSE NO. OPR-49 DOCKET NO. 50-331 i'
Replace the following pages of the Appendix "A" Technical' Specifications with the enclosed pages. The revised pages are identified by Amendment
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number and contain vertical lines indicating the area of change.
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r 1.0-1 1.0-1 i
1.0-2 1.0-2 1.0-3 1.0-3 1.0-4 1.0-4 1
1.0-5
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1.0-6 1.0-6
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1.0-7 1.0-7 1.0-8 1.0-8 x
1.1-2 1.1-2 1.1-3 1.1-3 i
3.2-5.a 3.2-5.a j
3.2-16 3.2-16 3.5-27 3.5-27 l
t 3.6-7 3.6-7 3.6-10 3.6-10 3.6-11 3.6-11 3.6-12 3.6-12 l
t 3.6-13 3.6-13 3.6-14 3.6-14 w
. Remove Replace 6.1-1 6.1-1 6.2-4 6.2-4 6.4-1 6.4-1 6.5-1 6.5-1 6.6-1 6.6-1 6.7-1 6.7-1 6.9-2 6.9-2 6.9-3 6.9-3 6.9-4 6.9-4 Delete the following pages of the Appendix "A" Technical Specifications.
Delete Delete 1.0-9 3.6-10a 1.0-10 3.6-10aa 1.0-11 3.6-10b 1.0-12 3.6-10c 1.0-13 6.9-5 1.0-14 6.9-6 1.0-15 6.9-7 1.0-16 6.9-8 1.2-7 6.9-9 4
3.5-28 6.9-10 3.5-29 6.9-11 3.5-30 6.9-12 3.5-31 6.9-13 3.5-32 6.9-14 w
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DAEC-1 1.0 DEFINITION $
The succeeding frequently used terms are explicitly defined so tnat a uniform interpretation of the specifications may be achieved.
1.
SAFETY LIMIT The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured.
Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.
2.
LIMITING SAr-if SYSTEM SETTING (LSSS)
The limiting safety system settings are settings on instrumentatien wnich in-itiate the automatic protective action at a level such that the safety limits will not be exceeded..These settings take into consideration the instrumen-tation tolerances and the instruments are required to be periodically cali-brated as specified in these Technical Scecifications. Tne limiting safety system setting plus tne tolerance of the instrument as given in the system design control document gives the limiting trip point for operation.
This additional margin nas been establisned so that with proper operation of the instrumentation the safety limits will never be exceeded. The inecuality sign whicn may be given merely signifies the preferred direction of cperational trip setting.
3.
LIMITING CONDITIONS FOR OPERATION (LCO)
The limiting conditions specify the minimum acceptable levels of system per-formance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal sit-uations can be safely controlled.
4.
DELETED j
i knendment No. 67 1.0-1
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DAEC-1 5.
OPERABLE-OPERABILITY 4
A system or component shall be considered operable when it ts capable of performing its intende' d. function in its required manner.
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5 6.
OPERATING
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Operating means that a system or component is performing its intended functions in its' required manner.
7.
IMMEDIATE Immediate means that the required action will be initiated as socn as 1
practicable considering the safe operation of the unit and the importance of the required action.
8.
REACTOR p0WER OPERATION Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and abcve 1% rated power.
9.
HOT STANDBY CONDITION t
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Hot standby condition means operation with coolant temperature greater than 2120F, reactor vessel pressure less than 1035 psig and the mode switch in tne Startup/ Hot Standby position.
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- 10. COLD CONDITION
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Reactor coolant temperature equal to or less than 2120F.
- 11. HOT SHUTDOWN t
The reactor is in the shutdown mode and the reactor coolant temperature l
greater than 2120F.
- 12. COLD SHUTDOWN The reactor-is in the shutdown mode, the reactor coolant temperature 8
equal to or less' than 2120F, and the reactor vessel is vented to atmosphere.
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Amendment No. 67 1.0-2
DAEC-1
- 13. MODE OF OPERATION A rea.: tor mode switch selects the proper interlocks for the operational status of the unit. The following are the modes and interlocks provided:
a.
Startup/ Hot Standby Mode - In this mode the reac' tor protection scram trips, initiated by main steam line isolation valve closure, are bypassed when reactor pressure is less than 1035 psig, the reactcc protection system is energized with IRM neutron monitoring system trip, the ApRM 15% high flux trip, and control red with ~
drawal interlocks in service. The lower pressure MSIV closure 880 psig trip is also bypassed. This is intended to imply the Startup/ Hot Standby position of the mode switch, b.
Run Mode - In this mode the reactor vessel pressure is at above 880 psig and the reactor protection system is energized with APRM protection (excluding the 15: high flux trip) and RBM inter-locks in service.
c.
Shutdown Mode - placing the made switch to the shutdown position initiates a reactor scram and power to the control red drives is removed.. After a short time period (about 10 sec), the scram si;nal is removed allowing a scram reset and restoring the normal valve lineup in the control red drive hydraulic system; also, the main steam line isolation scram is bypassed if reactor vessel pressure is below 1035 psig.
d.. Refuel Mode - With the mode switch in the refuel position interlocks are established so that one control red only may be withdrawn when the Source Range Monitor indicates at least 3 cps and the refueling crane is not over the reactor; also, the main steam line isolation scram is bypassed if reactor vessel pressure is below 1035 psig.
If the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.
14.
DESIGN AND RATED POWER Rated power (100% power) refers to operation at a reactor power of 1593 Mwt.
Design power, the power to which the safety analysis applies is 105% of rated steam flow which corresponds to 1658 Mwt.
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i Amendment No. 67 1,0-3
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DAEC-1
- 15. PRIMARY CONTAINMENT INTEGRITY Primary Containment Integrity means that the drywell and pressure suppression chamber are intact and all of the.fellowing conditions are satisfied:
a.
All nonautomatic containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident con-ditions are closed. These valves may be opened to perform necessary operational activities.
b.
At least one door in each airlock is closed and sealed.
c.
All automatic containment isolation valves are operable or deactivated in the isolated position.
d.
All blind flanges and manways are closed.
- 16. SECONDARY CCNTA:NMENT INTEGRITY Secondary containment integrity means that the reactor building is intact and the following conditions are met:
a.
At leatt one door in each access opening is closed.
b.
The standby gas treatment system is operable..
c.
Ali Reactor Building ventilation system autcmatic isolation valves are operable or deactivated in the isolated position.
- 17. OPERATING CYCLE Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
- 18. REFUELING OUTAGE Refueli.ng outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the required survet11ance testing need not be performed uhtil the next regularly scheduled outage.
Amendment No. 67 1.0-4
4 DAEC-1
- 19. ALTERATION OF THE REACTOR CORE The act of moving any component in the region above the core support plate, below the upper. grid and within the shroud. Normal control red movement with the control drive hydraulic system is not defined as a core alteration.
Normal movement of in-core instrumentation and the transversing in-core probe is not defined as a core alteration.
- 20. REACTOR VESSEL PRESSURE Unless otherwise indicated, reactor vessel pressures listed in the Tech-nical Specifications are those measured by the reactor vessel steam space detectors.
- 21. THERMAL PARAME*ERS a.
Minimum Cirtical Power Ratio (MCPR) - The value of critical power ratio (CPR) for that fuel bundle having the lowest CPR.
b.
Critical Power Ratio (CPR) - The ratio of that fuel bundle power which would produce boiling transition to the actual fuel bundle power.
Transition Boiling - Transition boiling means the boiling regime c.
between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.
d.
Deleted Linear Heat Generation Rate - the heat output per unit length of e.
fuel pin.
f.
Fraction of Limiting Power Density (FLPD) - The fraction of limiting power density is the ratto of the linear heat generation rate (LHGR) extsting at a given location to the design LHGR for that bundle type.
g.~'
Maximum Fraction of Limiting Power Density (MFLPD) - The maximum fraction of limiting power density is the highest value existing in the core of the fraction of limiting power density (FLPD).
h.
Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core thermal power to rated thermal power of 1593 MWth.
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-Amendment No. 67 1.0-5
DAEC-1 1
22.
INSTRUMENTATION Instrument Calibration - An instrument calibration means the a.
verification or adjustment of an instrument signal output so
.that it corresponds, with acceptable range, 'and accuracy, to a known value(s) of the parameter which the instrument monitors.
The acceptable range and accuracy of an instrument and its set-point are given in the system design control document and these setpoints are used in the Technical Specifications.
b.
Channel - A channel is an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combined in logic.
Instrument Functional Test - An instrument functional test means c.
the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action.
d.
Instrument Check - An instrument check is qualitative determination of acceptable operability by observation of instrument benavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.
Logic System Functional Test - A logic system functional test e.
means a test of all relays and contacts of a logic circuit to insure all components are operable per design intent.. Where practicable, action will go to completion; i.e., pumps will be started and valves operated, f.
Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may requfre one or more tnstrument channel trip signals related i
to one or more plant parameters in order to initiate trip system
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action.
Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.
g.
protection Acticn - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.
Amendment No. 67 1.0-6
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Instrumentation - Continued h.
protective. Function - A system protective action which results from the protective action of the channels monitoring a partic-ular plant condition.
1.
Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.
L Logic - A logic is an arrangment of relays, contacts, and other components that produces a decision output.
1)
Initiating - A logic that receives signals frcm channeks and produces decision outputs to the actuation logic.
- 2) Actuation - A logic that receives signals (either from initiating logic or' channels) and produces decision out-puts to accomplisn a protective action.
k.
Primary Source Signal - The first signal, which by plant design, should initiate a reacter scram for the subject abnormal occurrence (see FSAR Subsection 14.5).
- 23. FUNCTIONAL TESTS A functional test is the manual operation of initiatien of a system, sub-system, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray _ pump to verify that it runs and that it pumps the required volume of water).
24.
SHUTOOWN The reactor is in a shutdown condition wnen the reacter mode switch is in the shutdown mode position and no core alterations are being performed.
- 25. ENGINEERED SAFEGUARD An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.
Amendment No. 67 1.0-7 l
1 DAEC-1
- 26. SURVEILLANCE FRE00ENCY
- Periodic surveillance tests, checks, calibrations and examinations shall be performed within the specified suveillance intervals. These intervals may be adjusted plus or minus 25%. The operating cycle interval as per-taining to instrument and electrical surveillance shall never exceed 15 months.
In cases where the elapsed interval has exceeded 100% of the specified interval, the next surveillance interval shall commence at the end of the original specified interval.
27.
FIRE SUPPRESSION WATER SYSTEM A fire suppression water system shall consist of a water scurce, pumos, and distribution piping with associated sectionali:ing control or isolation valves.
Such valves include yard hydrant curb valves, the first valve ahead of the water flow alann device on each sprinkler, hose standpipe or deluge system riser.
i Amendment No. 67 1.0-8 m-
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DAEC-1 SAFETY LIMIT l
LIMITING SAFETY SYSTEM SETTING C.
Power Transients Where:
S = Setting in percent of rated power (1,593 MWt)
To ensure that the Safety Limits established in Specification W = Recirculation loop flow 1.1.A and 1.1.B are not exceeded, in percent of rated flow, each required scram shall be Rated recirculation loop initiated by its primary source flow is that recirculation signal. A Safety Limit shall loop fley which corresponds be assumed to be exceeded when to 49x100 lb/hr core ficw.
scram is accomplished by a means other than the Primary.
urce Signal.
For a MFLPD greater than FRP, the f
t D.
With irradiated fuel in the RP reactor vessel, the water level S 5 (0.56 W + 54)
MFLP0 shall not be. less than 12 in.
above the too of the normal active fuel :ene. Top of
. NOTE: These settings assume the active fuel zone is de-operation within the basic fined to be 344.5 incnes thermal design cri aria.
above vessel :ero (see These criteria are LHGR$
Bases 3.2).
18.5 KW/ft (7x7 array) or 13.4 KW/ft (8x8 array) and MCPR ?. values as indicated in Table 3.12-2 times X,
f where Kf is defined by Figure 3.12-1.
Therfore,
at full power, operation is not allowed with MFLP0 greater than unity even if the scram setting is reduced.
If it is detennined that either of these design criteria is being vio-lated during operation, action 4
must be taken immediately to return to cperation within these criteria.
2.
APRM High Flux Scram When in. the REFUEL or STARTUP and HOT STAND 8Y M00E. The APRM scram shall be set at less than or equal to 15 percent of rated power.
i Amendment No.tS7 1.1-2
DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 3.
APRM Rod Block when in Run Mode.
For operation with MFLPD less than or equal to FRP the APRM Control Rod Block setpoint shall be as shown on Fig. 2.1-1 and shall be:
5 g (0.66 W + 42)
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The definitions used above for the APRM scram trip apply.
For a MFLPD greater than FRP, the APRM Control Rod Block setpoint shall be:
S 1(0.66 W + 42) gh[fD 4.
IRM - The IRM scram shall be sat at less than or equal to 120/125 of full scale.
B.
Sct m ai;6 Iso-2514.5 lation oc inches reactor low above watar level vessel zero
(+12" on level instru-ments)
C.
Scram - turbine
< 10 percent stop valve valve closure closure D.
Turbine control valve fast closure shall occur within i
30 milliseconds of the start of turbine control valve fast
- closure, 1
1.1-3 Amendment No. 67 i
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IAulE 3.2-A f
INSTRUHfNIATION lilAT INIIIAIES l'illHAltY CONIAINHlHi IS01 AIlON (Continued)
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. f Minienum No.
Namdser of o of Operable Ins trtament a Instramment Channels Valve Groups t
P Cliannels Per Provided by Operated by IrlpSyster(1)
Instrument Trip Level Setting ik! sign Signal Action (2) i OI Reactor Cleanieli 130f 3
S f)
Area Amhient liigli Temp.
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Heactor Cleanup A 14"r
- 3 S
p Area liifferential liigh Temp.
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2 loss of Main 6 1 0 ist 119 Vacinam 4
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3, Condensor Vaunas nL
- Note: The actual set point shall be a 14 l-above the 100% power operation andaient temperature conditions as determined by DAEC Plant Test Procedure, f
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TABLE 3.2-C g Minimum No.
of Operable R Instrument R
Channels per Number of A Trip System Instnment Trip Level Setting Prv.vided by Design
~ Action Instrument Channels 2
2 APRMUpscale(FlowBlased) 4(0,66 W t 42) (3-hg)(2) 6 inst. Clannels (1) i 2
APRM Upscale (Not in Run Mode) 5 12 1pdicated on scale 6 Inst, Channels (1) 1 2
APRM Downscale 2 5 indicated on scale 6 Inst. Channels (1) 1 (7)
Rod Block Monttor s(0.66 9 + 39) (-g-$g)(2) 2 Inst. Channels (1)
(FlowBlased)
F 1 (7)
Rod Block Monitor 2 S Indicated on scale 2 Inst. Channels (1) 7 Downscale 2
IRMDownscale(3) g S/128i full scale 6 Inst. Channels (1)~
2 IRM Detector not in (8) l Startup Position 6 Inst. Channels (1) 2 IRM Upscale s 108/125 6 Inst. Channels (1) 2(S) ggegg in (4) 4 Inst. Channels (1) 2(5)(6)
SRM Upscale
$10 counts /sec.
4 Inst Channels (1) 5 s
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DAEC-1
,4.5 BASES Core and Containment Cooling Systems Surveillance Frequencies The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judge-ment and practicality. The core cooling systems have not been designed to be fully testable during operation. Fer example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable.
Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment ecoling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently.
The pumps and motor operated injection valves are also tested each month to assure their operability. A simulated automatic actuation test ence each cycle combined with frequent tests of the pumps and injection valves is deemed to be adequate testing of these systems.
When components and subsystems are out-of-service, overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining equipment. The degree of operability to be demonstrated depends on the nature of the reason for the cut-of-service equipment.
For routine out-of-service periods caused by pre-ventative maintenance, etc., the cump and valve operability checks will be performed to demonstrate operability of the remaining components.
However, if a failure due to a design deficiency, caused the outage, then the demonstration of operability should be thorough enough to assure that a generic problem does not exist. For example, if an out-of service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the operability checks.
4 Redundant operable components are subjected to increased testing during equipment out-of-service times. This adds further conservatism and increases assurance that adequate cooling is available should the need arise.
The RHR valve power bus is not instrumented. For this reason surveillance requirements require once per shift observation and verification of lights and instrumentation operability.
t 3,5-27 Amendment No. 67 4
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DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS b.
The indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10%.
c.
The diffuser to lower plenum differential pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than 10%.
2.
Whenever there is recirculation flow with the reactor in the Start-up or Run mode, and one recircu-lation pump is operating, the diffuser to lower plenum differ-ential pressure shall be checked daily and the differential press-ure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential press-cres in that loop by more than 10%.
F.
Jet Pumo Flow Mismatch F.
Jet Pumo Flcw Mismatch 1.
When both recirculation pumps 1.
Recirculation pump speeds shall are in steady state operation, be checked and logged at least the speed of the faster pump -
once per day.
may not exceed 122% of the speed of the slower pump when core power is 80% or more of rated power or 135% of the speed of the slower pump when core power is below 80% of rated power.
t 2.
From and after the date that one i
recirculation loop is made or found to be inoperable for any reason, continued reactor oper-ation is permissible only during the succeeding-24 hours unless such loop is sooner made oper-able.
If this requirement cannot be met, an orderly shut--
sown shall be initiated and the reactor shall be in a cold shutdown. condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Amendment No. 67 3,5-7 y
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s LIMITDIG C0tBITIONS FOR' OPERATI0tf SURVEIIIANCE REQUIRDENTS l'w H.
Shock Suppressors (Snubbers)
H.
Shock Suppressors (Snubbers) f 1.
During all modes of opera-The following surveillance re-tion, except Cold Shutdown quirements apply to all hydrau-j and Refusi, all safety re-lic snubbers listed in Tables laced snubbers listed in 4.6-3 and 4.6-4 :
1 i
Tables 4.6-3 and 4.6-4 shall be operable, except as noted 1.
All hydraulic snubbers whose in 3.6.H.2 through 3.6.H.4 seat material has been below.
demonstrated by operacing I
experience, lab testing or 2.
Pros and after,the time that analysis to be compatible a snubber is determined to with the operating environ-be inoperable, conched re-ment shall be visually in-actor operation ~is permissi-spected. This inspection ble only during the succeed-shall include, but not nec-ing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the essarily be limited to, in-sunbber is sooner made oper-spection of the hydraulic j
. able or replaced.
fluid reservoir, fluid con-nections and linkage con-3.
If the requirements of nections to the. piping an,d 3.6.H.1 and 3.6.H.2 cannot anchor to verify snubber be met, an orderly shutdown operability in accordance 1
shall be initiated and the with the following schedule:
i reactor shall be in a cold l
(
shutdown condition within 36 Number of Snubbers-i hours.
Found Inoperable l
During Inspection Next Required
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4.
If a snubber is determined or During Inspec-Inspection i
tar be inoperable while the tien. Interval Interval reactor is in the shutdown or refuel mode, the snubber 0
18 months
~257.
shall be made operable or 1
12 months t 25%
j replaced prior to reactor 2
6 months 25 %
startup.
3, 4 124 days 25%
i 5,6,7 62 days e 25%
5.
Sachbers may be added to 28 31 days
- 25%
safety related systems with-out prior License AmenAn*Et The required inspection in-3' to Tables 4.6-3 or 4.6-4 pro-terval shall not be length-vided that a revision to ened more than one step at Table 4.6-3 or 4.6-4 is in-a time.
cluded with the next License Amendment request Snubbers are categorized Li2 i
two groupe, " accessible and i
inaccessible," based on their accessibility for in-l!
spection during reactor operation. 'These two groups eill be inspected independ-l ently according to the above schedule.
o 3.6-10 Amendment No. 67 p.. _
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r LIMITING CONDITIONS FOR OPERATI N SURVEIUANCE RE'QUTREMENTS 2.
All hydraulic snubbers
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whose seal materials are other than ethylene propy-lene or other material that has been demonstrated to be compatible with the operating enviranraant shall be visually inspected for operability every 31 days.
3.
The initial inspection.
t shall be performed within six months t 25% from the data of issuance of these specifications. For the purpose of entering the schedule in Specification 4.6.H.1, it shall be as-sumed that the facility has been on a six-month inspection interval.
4 Once each operating cycle a representative sample of 10 hydraulic snubbers or approximately 10% of the hydraulie snubbers, which-ever is -less, shall be functionally testad for operability including veri-fication of proper piston movement, lock-up and bleed.
For each unit and subse-quant unit found inoperable, an additional 10% or ten (10) hydraulic snubbers shall be so tested until no more failures are found or all units per category tested have been casted.
Snubbers of rated capacity greater than 50,000 lbs.
need not be functionally casted!
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Amendment No. 67 I
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DAEC-1 r
3.6.A & 4.6,A BASES:
I Thermal and Pressurization Limitations The thenal limitations for the reactor vessel meet the
(
requirements of 10CFR50, Appendix G.
The allowable rate of heatup and cocidewn for the reactor vessel 8
contained fluid is 100 F per hour averaged ever a peried of one hour. Tnis rate has been chosen based en past ex;:erience with opersdng ;cwer plants. The asscciatad time pericd for heatup and cocidown cycles when the 1000F per hour rata is limiting provides for efficient, but safe, plant operation.
Specific analyses were made based en a heating and c: cling
.i rata.of~1000F/heur appifed continuously over a tameerature 0
C range of' 100 F ts 546 F.
Calculatad stesses were withis ASME Boiler and Pressurt Vessel Code Section III stress intansity and ' fatigue limits even at the flange area where maxicaan stress occurs.
Chicago Bridge and Iron Company performed detailed stress
~ analysis as shown in FSAR Appendix X, " Field Fabricated Reactor Vessel". This analysis includes more severe thennal cxditions than those which would be. enc:untered during nonnal i
heating and cooling operations.
r The permissible flange to adjacent shell tamcerature differ-0 ential of 145 F is the maximum calculated for 1000F hour heating and cooling rate applied centinuously over a.1000F to Amendment No. 67 3.6-12
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8 550 F range. The. differential,is due to the sluggish tamperature
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response of the flange metal and its value decreases for any lower heating rata or the same rate applied over a narrower range.
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The coolant in the bottan of the vessel is at a lower tamperature than that in the upper regions of the vessel when there is no
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recirculation _ flow. This colder water is forcad up when recirculatien l
pumps a m started. This will not result in stresses which excaed l
ASME Boiler and Pressure Vessal Code, Secti:n III limits when the J
tamperature differential is not greater than 145CF.
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The react:r c:clant system is a primary barMer against the j
release of fission products to the envir:ns. In ordar ts t
provide assurance that tnis barrier is maintafiied at a htgh degree i
v l) of integrity', restrictions have been placad en the cpee? ting conditions 1
ta which it can be sdjected.
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The nil-ductility transition (NDT) tamperature is defined as t
the tamperature below which ferritic stael creaks in a brittle I
rather than ductile manner. hadiation exposure from fast neutrons ( >l mov) above about 10I7 nyt may, shift the NDT tamperature of the vessel base metal above the initiai value.
Extansive tests have established the r.agnitude of changes as a function of the integrated neutron exposure.
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Neutron flux wires and samples of vessel matarial are in-i stalled in the reactor vessel adjacent to ne vessal wail at the core nidplane level. Tne wirus and samples will be Amendment-No. 67 3,3 13 l
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removed and tasted according to 10CFR50 Appendix H.
Results of.these analyses will be used to adjust Figure 3.6-1 as appropriate.
As described in partgraph 4.2.5 of tr.e Safety Analysis report, detailed stress analy es have been made on the reactor vessel for both staady stata and transient conditions with respect to matariat fatigue. The results of these transients are c:m-pared to allowable stress limits. Requiring the c:clant tamo-0 erature in an idle recirculation locp ts be within 50 F of the operating Tcop tamperature before a recirculaticn pumo is
. startad assures that the changes in coolant tamcerature at the react:r vessel no::les and bott:m head region are acceptable.
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s-DAEC-1 6.0 ADMINISTRATIVE CONTROLS
-6.1 MANAGEMENT - AUTHORITY AND RESPONSIBILITY
' 6.1.1 The Chief Engineer has. primary responsibility for the safe operation of the DAEC, and reports to the Assistant Vice President-Nuclear Generation.
6.1.2 The overall responsibility for the fire protection program for CAEC is assigned to the Assistant Vice President - Nuclear Generation.
The OAEC Chief. Engineer is responsible for directing the operating plant fire protection crogram.
- 6.1.3 The Quality Control Superviscr is responsible for implementation of the Quality Assurance program at the CAEC and reports to the Manager - Quality Assurance.
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Duane Arnold Energy Center Iowa Electric Light and Power Company Technical Specifications
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DAEC-1 6.4 RETRAINING AND REPLACE. MENT TRAINING 6.4.1-A training program shall be established to maintain the overall pro-ficiency of. the operating organization. This program shall consist of both retraining and replacement training elements and shall meet
' or exceed the minimum provisions outlined in ANSI N18.1-1971.-
6Property "ANSI code" (as page type) with input value "ANSI N18.1-1971.-</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..4.2 A training program'for the fire brigade shall be maintained under the direction of the Chief Engineer and shall meet or exceed the require--
rnents'of Section 27 of the NFPA Code, except for fire brigade training sessions which shall be held at least quarterly.
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Amendment No. 67 6.4-1
f' DAEC-1 6.5 REVIEW AND AUDIT 6.5.1 Operations Comittee 6.5.1.1 Function The Operations Comittee shall function to advise the Chief Engineer.
on all matters related to nuclear safety.
6.5.1.2 Composition The Operations Comittee shall be composed of the Assistant Chief Engineers and 5upervisors frem the following departments: Operations,
Maintenance, Reactor and Plant Engineering, Radiation Protection, Quality Control, and Technical Engineering.
The Assistant Chief Engineer - Operations shall act as the Chairman.
One or more of the members shall be designated as Vice Chairman.
6.5.1.3 Alternates All alternate members shall be appointed in writin by the Chief Engineer to serve on a permanent basis; hcwever, ro more than three alternates shalf participate as voting members in Operations Comittee activities at any one time.
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DAEC-1 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 Any reportable occurrence shall be reported immediately to the Chief Engineer and to the Assistant Vice President - Nuclear Generation, and promptly reviewed by the Operations Committee.
6.6.2 The Operations Committee shall prepare a scparate report of each reportable occurrence..This report shall include an evaluation of the cause of the occurrence..a record of the corrective action taken, and also recommendation for appropriate action to prevent or reduce the probability of a recurrence.
6.6.3-Copies of all such reports shall be submitted to the Safety Commf ttee for review and to the Assistant Vice President - Nuclear Generation for review and approval of any recommendations.
1 Amendment No. 67-6.6-1
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-6.7 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED 6.7.1 If a safety limit is exceeded, the reactor shall be shut down and reactor operation shall only be resumed when authorized by the NRC.
- 6. 7. 2 An imediate report shall be made to the Assistant Vice President -
l Nuclear Generation and the Safety Comittee. The Assistant Vice l
President - Nuclear Generation shall prcmptly report the circumstances to the NRC as specified in Subsection 6.12, Plant Reporting Requirements.
6.7.3 A complete analysis of the circumstances leading up to and resulting from the situation together with reccmendations to prevent a re-j currence shall be prepared by the Cperations Comittee. This report shall be submitted to the Assistant Vice President - Nuclear Generation
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and the Safety Comittee. Appropriate analyses or reports will be submitted to the NRC by the Assistant Vice President - Nuclear Generation as specified in Subsection 6.12, Plant Reporting Requirements.
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I Amendment No. 67 6.7-1 w.
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.DAEC-1 SPECIFICATICN' SURVEILLANCE REQUIRE 3GNT
'6.9.2' Source Leakage Test-T.9.2 Source Leakage Test Radioactive sources shall A.1 Tests for leakage and/or be leak' tested for con-contamination shall be
- tamination.. The leakage performed by the licensee stest'shall be capable of or by other persons spe-
. detecting.the presence-cifically authorized by of 0.005 microcurie of the Commission or an radioactive material on agreement State, as follows:
the' test sample.
If the test reveals the presence 1 1.
Each sealed source, except of 0.005 microcurie or startup sources subject to more of removable con-core flux,. containing tamination, it shall radioactive material,. other immediately be withdrawn.
than Hydrogen 3, w3 th a from use,. decentaminated, half-life greater than and repaired, or be dis-thirty days and in any posed of in.accordance form other than gas shall be with Ccamission regula-tested for leakage and/or tions.
contamination.at intervals not to exceed six months.
.Those: quantities of by-product material thac 2.
The periodic leak test exceed the quantities required dces not apply to listed in 10 CIR 30.71 sealed sources that are l
Schedule B are.to be leak stored and not being used.
tested in accordance with The sources excepted f cm the schedule shown in this test shall he tested Surveillance Requirements.
for leakage prior to any All other sources (includ-use or transfar to another ing alpha emitters) user unless they have been i
containing greater than leak tested within six O.1 microcurie are also months prior to the date_ of to be leak tested in use or transfer.-
In the i
accordance with the absence of a certificate Surveillance Requirements.
from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
3.
Startup. sources shall be leak tested prior to and follow-i ing any repair or maintenance i
and before being subjected to cora flux.
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i SPECIFICATION SURVEILIANCE REQUIRDENT
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Reporting Reqttirements Results of the leak tests performed on sources shall be included in the Annual' Operating Report if the tests reveal the presence of 0.005 microcurie or more of removable contamination.
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DAEC-1 6.9.2 BASES-(
d Ingestion or inhalation of source material may give rise to total body or organ' irradiation.
This specification assures that leakage from radioactive material sources does not exceed allowable limits.
In the unlikely event that those quantities of radioactive by-product materials of interest to this specification which are exempt from leakage testing are ingested or inhaled, they represent less than one mavhum permissible body burden for total body irradiatien.
The limits for all other sources (including alpha emitters) are based upon 10 CFR 70.39 (c) limits for plutonium.
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