ML20195C837
| ML20195C837 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/19/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Nebraska Public Power District (NPPD) |
| Shared Package | |
| ML20195C845 | List: |
| References | |
| DPR-46-A-099, TAC 56601 NUDOCS 8605300634 | |
| Download: ML20195C837 (12) | |
Text
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4, UNITED.ITATES 8
NUCLEAR REGULATORY COMMISSION o
h3' WASHINGTON, D. C. 20555 p$
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l NERRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 99 License No. DPR-46 1.
The Nuclear Regulatory Comission (the Ccmission) has found that:
A.
The application for amendment by Nebraska Public Power District dated December 20, 1984, as supplemented February 22, 1985, 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; B.
The facility will operate in conformity with the application, 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 Comission'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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the licensee 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. DPR-46 is hereby amended to read as follows:
r6053oo634 e60519 DR ADOCK 0500 8
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- w. (2) Technical Specification The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 99, are hereby incorporated in the license. The licensee shall operate the facility-in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 1
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Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Charges to the Technical Specifications Date of Issuance: May 19, 1986 I
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ATTACHMENT TO LICENSE AMENDMENT NO. 99 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET N0. 50-298 Replace the following pages of the Apper. dix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
_ Pac.e_s_
4 41 59 76 81 219 220 237
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8.
Simulated Automatic Actuation - Simulated automatic actuation means applying
,s a simulated signal to the sensor to actuate the circuit in question.
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9.
Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.
Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.
J.
Limiting Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility.
When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.
Limiting Conditions for Operation (LCO) shall be applicable during the operational conditions specified for each specification.
Adherence to the requirements of the LCO within the specified time interval shall constitute compliance with the specification. In the event the LCO is restored prior to expiration of the specified time interval, completion of the LCO action is not required.
In the event an LCO cannot be satisfied because of circumstances in excess of those addressed in the specification, the facility shall be placed in HOT SHUTDOCN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless
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i corrective measures are completed that permit operation un '.er the LCO for the
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specified time interval as measured from initial discover 3 Exception to these requirements shall be stated in the individual specifications.
Entry into an operational condition shall not be made unless the conditions of the LCO are met without reliance on the actions specified in the LCO unless otherwise excepted. This provision shall not prevent passage through operational conditions required to comply with the specified actions of an LCO.
When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:
(1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s),
component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in at least HOT SHUTDOWN within 6 hcurs, and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
This specification is not applicable in the cold condition or the refueling mode.
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l Amendment No. 26, 99,
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y K.
Limiting Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a
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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:
4 1.
Refuel Mode - The reactor is in the refuel mode when the mode switch is in the REFUEL 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 825 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.
3.
Shutdown Mode - The reactor is in the shutdown mode when the mode switch is in the SHUTDOWN position.
1 4.
Startun/ Hot Standby Mode - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are bypassed, 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 - Operability - Operating
- l 1.
Operable - Operability - A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).
Implicit in this definition shall be the assumption i
that all necessary attendant instrumentation, controls, normal and emergency electrical power sources (except as specified in Sections 1.0.J and 3.9),
l cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform 4
its function (s) are also capable of performing their related support function (s),
f 2.
Operating - Operating canna a system, subsystem, train, component, or device 1
is performing its intended function in its raquired manner.
N.
Deleted.
O.
Operating Cycle - Interval between the end of one refueling outage and the end of j
the nevr subsequent refueling outage.
l P.
Primary containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the fo11cwing conditions are satisfied:
I 1.
All manual containment isolation valves on lines connected to the reactor coolant system or containment, and which are not required 1
to be open during accident conditions, are closed.
2.
At least one door in each airlock is closed and sealed.
~4-Amendment No.
JHI 93, f
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 BASES (Cont.d) 4.1 BASES (Cont.d) ence paragraph VII.5.7 FSAR). Thus full scale flow signal will be sent l the IRM System is not required in to half of the APRM's resulting in the "Rua" mode. The APRM's cover a rod block condition. Thus, if the l only the power range. The IRM's calibration were performed during and APR!t's provide adequate coverage operation, flux shaping would not in the startup and intermediate range, be possible.
Based on experience at other generating stations, drift The requirement to have the scram of instruments, such as those in functions indicated in Table 3.1.1 the Flow Biasing Network, is not operable in the Refuel mode assures significant, that shifting to the Refuel mode during reactor power operation dces Group (C) devices are active only not diminish the protection provided during a given portion of the by the reactor protection system.
operational cycle.
For example, the IRM is active during startup Turbine stop valve scram occurs at and inactive during full-power 10% of valve closure. Below 233 psig operation. Thus, the only test turbine first stage pressure (30% of that is meaningful is the one rated), the scram signal due to tur-performed just prior to shutdown bine stop valve closure is bypassed or startup; i.e.,
the tests that because the flux and pressure scrams are performed just prior to use are adequate to protect the reactor, of the instrument.
Turbine control valves fast closure Calibration frequency of the instru-initiates a scram based on pressure ment channel is divided into two switches senuing Electro-Hydraulic groups. These are as follows:
Control (LilC) system oil pressure.
The switches are located on the 1.
Passive type indicating devices Control Valve Emergency Trip oil that can be compared with like header, and detects the loss of units on a continuous basis.
oil to hold the valves open.
2.
Vacuum tube or semi-conductor This sc ram signal is also bypasned devices and detectors that when turbine first stage pressure drif t or lose sensitivity.
is less than 233 psig.
Experience with passive type instru-The requirements that the IRM's be in-menta in generating stations and sub-serted in the core when the APRM's rem stations indicates that the specified 2.5 indicated on the scale in the calibrations are adequate.
For those Startup and Refuel modes assures that devices which employ amplifiers, etc.,
drift specifications call for drift to be less that 0.4%/ month; i.e.,
in the period of a month a maximum drift of 0.4% could occur, thus providing for adequate margin.
Amendment No. 73, 99,
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COOPER NUCLEAR STATION 2f TABLE 3.2.H (Page 7)
<n AUTOMATIC DEPRI.SSURIZATION SYSTDI (AUS) CIRCtflTRY REQUIREMENTS E
2a Minimum Nermber Action Required When
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Instrument of Operable Components Component Operability Instrument I.D. No.
Setting Limit Per Trip System (1)
Is Not Assured g.
Reactor Low Water NBI-LIS-83, A & B
> +12.5" Indicated I
B
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Level Level h5 NBI-LIS-72, A.B.C & D > -145.5" Indicated 2
'A Level ADS Timer MS-TDR-K5, A & B
__ 120 sec.
I B
Low-Low Set NBI-PS-51, A,B,C & D SI-A Open Low Valve 2
B 1015210 psig (Increasing)
SI-B Close Low Valve 0,
875210 psig (Decreasing) 7 51-C Open High Valve 1025t10 psig-(Increasing)
St-D Close High Valve 875210 psig (Decreasing) c
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9 COOPER liUCLEAR STATION TABLE 4.2.B (Page 7)
ADS SYSTEM TEST & CAtlBl!ATION FisFyt!ENCIES a:s Instrument Item item I.D. No.
Functtonal Test Freq.
Calibration Freq.
Check g
_ Instruments e
1.
ADS Inhibit Switch MS-SW-S3A & B Once/ Month (1)
N.A.
None l
2.
Reactor Low Water Level NBI-LIS-83, A & B Unce/ Month (1)
Once/3 Months Once/ Day NBI-LIS-72, A,B,C, & D once/ Month (1)
Once/3 Months once/ Day 3.
ADS Timer MS-TDR-K5 A & B Once/ Month (1)
Once/Oper. Cycle None 4.
Low-I.ow Set NBI-PS-51, A,B,C, & D Once /!!on t h (1)
Once/Oper. Cycle hone Logic (4)(6) 1.
ADS Control Power Monitor Once/6 Months N.A.
2.
ADS Actuation once/6 !!onths N.A.
3.
Low-Low Set Logic Once/6 Honths N.A.
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t-NOTES FOR TABLES 4.2.A THROUCH 4.2.F 1.
Initially once.every month until exposure (M as defined on Figure 4.1.1) is 2.0 X 10 ; thereafter, according to Figure 4.1.l(after NRC approval). The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of CNS.
2.
Functional tests shall be performed before cach startup with a required frequency not to exceed once per week.
3.
This instrumentation is excepted from the functional test definition. The functional test will consist of applying simulated inputs. Local alarm lights representing upscale and downscale trips will be verified but no rod block will be produced at this time.
The inoperative trip will be initiated to produce a rod block (SRM and IRM inoperative also bypassed with the mode switch in RUN). The functions that cannot be verified to produce a rod block directly will be verified during the operating cycle.
4.
Simulated automatic actuation shall be performed once cach operating cycle, Where possible, all logic system functional tests will be performed using j
the test jacks.
5.
Reactor low water level, high drywell pressure and high radiation main steam line tunnel are not included on Table 4.2.A since they are tested on Table 4.1.2.
6.
The logic system functional tests shall include an actuation of time delay relays and timers necessary for proper functioning of the trip systems.
7.
These units are tested as part of the Core Spray System tests.
8.
The flow bias comparator will be tested by putting one flow unit in " Test" (producing a rod block) and adjusting the test input to obtain comparator l
rod block. The flow bias upscale will be verified by observing a local upscale trip light during operation and verifying that it will produce a rod blo'ck during the operating cycle.
9.
Perf ormed during operating cycle. Portions of the Icgic is checked more frequently during functional tests of the functions that procuce a rod bleck.
10.
The detector will be inserted during each operating cycle and the proper i
amount of travel into the core verified, i
Amendment !!o. 75, 99, -
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6.0 ADMINISTRATIVE CONTROLS o
e 6.1 ORGANIZATION
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6.1.1 Responsibility b
The Division Manager of Nuclear Operations shall have the over-all fulltime K
onsite responsibility for the safe operation of the Cooper Nuclear Station.
During periods when the Division Manager of Nuclear Operations is unavailable, he may delegate his responsibility to one of the managers in the Nuclear Operations Division.
e 6.1.2 Offsite i
The portion of the Nebraska Public Power District management which
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relates to the operation of this station is shown in Figure 6.1.1.
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6.1.3 Plant Staff - Shift Complement The organization for conduct of operation of the station is shown in j
Fig. 6.1.2.
The shift complement at the station shall at all times 1
meet the following requirements. Note: Higher grade licensed operators d
may take the place of lower grade licensed or unlicensed operators.
3 1
]
A.
A licensed senior reactor operator (SRO) shall be present at the station at all times when there is any fuel in the reactor.
B.
A licensed reactor operator shall be in the control room at all times when there is any fuel in the reactor.
C.
Two licensed reactor operators shall be in the control room during all startup, shutdown and other periods involving signif-icant planned control rod manipulations. A licensed SRO shall either be in the Control Room or inimediately available to the Control Room during such periods.
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D.
'A licensed senior reactor operator (SRO) with no other concur-1 rent duties shall be directly in charge of any refueling opera-i j
tion, or alteration of the reactor core.
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A licensed reactor operator (RO) with no other concurrent i
duties shall be directly in charge of operations involving the handling of irradiated fuel other than refueling or reactor core alteration operations.
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E.
An individual who has been trained and qualified in health physics techniques shall be on site at all times that fuel is s
on site.
F.
Minimum crew size during reactor operation shall consist of d
four licensed reactor operators (two of whom shall be licensed l
l SRO) and three unlicensed operators. Minimum crew size during l
reactor cold shutdown conditions shall consist of two licensed j
reactor operators (one of whom shall be licensed SRO) and one unlicensed operator.
In the event that any member of a minimum shift crew is absent i
or incapacitated due to illness or injury a qualified replace-
{
ment shall be designated to report on-site within two hours.
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Amendment llo. 66, 82. EE, 99,
V 6.2 REVIEW AND AUDIT 6.2.1 The organization and duties of committees for the review and audit of station operation shall be as outlined below:
A.
Station Operations Review Committee (SORC) 1.
Membership:
a.
Chairman: Division Manager of Nuclear Operations b.
Technical Staff Manager c.
Operations Manager d.
Technical Manager e.
Operations Supervisor f.
Maintenance Supervisor g.
Instrument and Control Supervisor h.
Chemistry and Health Physics Supervisor l
1 1.
Plant Engineering Supervisor i
J.
Operations Engineer Supervisor l
k.
Computer Applications Supervisor 1.
Quality Assurance Manager - non-voting member.
Alternate members shall be appointed in writing by the Division 5
Manager of Nuclear Operations to serve on a temporary basis; however.
no more than two alternates shall serve on the Committee at any one time.
2.
Meeting Frequency: Monthly, and as required on call of the Chairman.
3.
Quorum: Division Manager of Nuclear Operations or his designated alternate plus four other members including alternates.
4.
Responsibilities:
Review all proposed normal, abnormal, maintenance and emergency a.
operating procedures specified'In 6.3.1, 6.3.2, 6.3.3, and 6.3.4 I
and proposed changes thereto: and any other proposed procedures or changes thereto determined by any member to effect nuclear safety.
.1 b.
Review all proposed tests and experiments and their results, which involve nuclear hazards not previously reviewed for conformance with technical specifications. Submit tests which may constitute an unreviewed safety question to the NPPD Safety Review and Audit Board for review.
Review proposed changes to Technical Specifications, c.
d.
Review proposed changes or modifications to station systems or equipment as discussed in the SAR or which involves an unre-viewed safety question as defined in 10CFR50.59(c). Submit changes to equipment or systems having safety significance i
to the NPPD Safety Review and Audit Board for review.
Review station operation to detect potential nuclear safety e.
l hazards.
P Amendment flo. 80, 82, 25, 99,
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NUC L E AR n.
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'E OPER AT IONS DIVISION 3
M AN AGER STATION OPERATIONS FIRE PROTECTION REVIEW COMMITTEE ENGINEER M (SO RC) g TECHN IC A L d
tm OUALITY ASSURANCE STAFF l
M ANAGER - CNS MANAGER
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I T R AI NING TECHNICAL OPERATIONS ADMINISTRATIVE j
MANAGER M AN AGER MANAGER MAN GER I
OPE R AT IO N S SENIOR l
l TRAINING r--
RAD / TECH SECURITY ADMINIS TR ATIVE M ATE RIAL SUPERVISOR I
ADVISOR SUPPORT I
SUPERVISOR SU PERVISO R SUPERVISOR 7- - --.3 l
l l
l ro COMPUTER PLANT OPERATIONS CHEM. S H.R APPLICATIONS ENGINEER ENGINEER SUPERV ISOR SUPERVISOR SU PERVISOR SUPERVISOR i
d I
I l
I/S ONE/ SHIFT MAINTENANCE I8C OPERATIONS 2/S TWO/ SHIFT SUPERVISOR 3/S THREE/ SHIFT SUP ERVISO R SUPER VISOR (S R O )
i RO-NRC REACTOR OPERATORS LICENSE SRO-NRC SENIOR REACTOR OPERATORS g
LIC EN SE y
SHIFT M FUNCTIONAL POSITION ONLY PHYSICALLY LOCATED N GE!ERAL OFFICE ELECTRICAL MAINT ENANCE MECHANICAL SUPERVISORS (SRO)
PLANNER /
l/S SUPERVISOR SCHEDULER SUPERVISOR l
CONTROL ROOM
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SUPERVISOR (SRO)
I 1/S Figure 6.l.2 l
NP.RD. Cooper Nuclear S Iation 2/S (UNIT OPER (RO) m Organization Chorf 3/s jTN1CEN U
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