ML20147B491
| ML20147B491 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/23/1988 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20147B449 | List: |
| References | |
| NUDOCS 8803020108 | |
| Download: ML20147B491 (10) | |
Text
r-p o'es:
- /
y g UNITED STATES NUCLEAR REGULATORY COMMISSION
- t WASHINGTON. D. C. 20666
% ,. . . . . ,o NEBRASKA PUBLIC POWER DISTRICT ,
1 DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.117 License No. DPR-46
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Nebraska Public Power District (the licensee) dated December 22, 1987, 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 (1) that the activities authorized by this amendment can be conducted withcut 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 I public; and i 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. t 1
l 8803020100 800223 PDR ADOCK 05000290 P PDR
- 2. Accordingly, the license is amended by changes to the Technical Specifi-cations 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 Specification The Technical Specifications contained in Appendix A, as revised through Amendment No. , 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 COMISSION
. L
/ Jose alvo,Diheter Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 23, 1988
n ATTACHMENT TO LICENSE AMEN 0 MENT NO.117 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Pages 95 96 99 100 101 101a 104 I
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.. *'LIMITXNG CONDITXON-FOR OPERAT20N SURVE2LLANCE REQUIREMENT
- c. During each refueling outage ob -
serve-that any drive which has been_ uncoupled from and'subse-quently recoupled to its control rod does not go to the overtravel-
- 2. The_ control-rod drive housing position.
support systes shall be in place .
during reactor power operation 2. The control rod drive housing or when the reactor coolant support systen shall be-inspected system is pressurized above at- after reassembly and the results sospheric pressure with fuel in of the inspection recorded.
the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is set.
3.a. Whenever the reactor is in the startup or run mode below 20%
rated power and control rod 3a. Prior to entering the Group movement is within the Group Notch mode during startup and Notch mode after 50% of the prior to attaining 20% rated control rods have been power during rod insertion while withdrawn, the rod sequence shutting down, the capability of control system shall be operable. the Rod Sequence Control System If the system is decernined to to properly fulfill its function be inoperable during power de- shall be verified by the checkst
. scent in accordance with checks in Specification 4.3.3.3, power (1) Durina'Startup and Shutdown l may be maintaiped above 20% rated -
power until repairs are made. Group Notch Portion - Test the six comparator circuits. Go through
- b. During the shutdown procedure each comparator inhibit, initiate rod movement is restricted teste verify error, and reset.
to the RSCS sequence following Af ter comparator checks initiate the testing performed between test and observe completion of 35% and 20% power level and the cycle indicated by illumination of automatic reinstatement of the test complete light.
RSCS restraints at the preset power level until 50% control b. Prior to the start of control rod rod density is reached, withdrawal towards criticality and Alignment of rod groups shall be prior to attaining 20% rated power accomplished prior to the during rod insertion while shutting reinstatement of the Rod Sequence down, the capability of the Rod Worth Control Systes restraints. Minimiser (RWM) to properly fulfill its function shall be verified by
- c. Whenever the reactor is in the the following checks.
startup or run modes below 20%
rated power the Rod Worth Minimiser shall be operable or a second licensed operator or other qualified employee shall verify that the operator at the reactor console is following the control rod progras.
- d. Deleted.
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. LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT 3.3.B.3 (cont'd) ' 3.B.3.b (cont'd) k e. If Specifications 3.3.B.3a 1) The correctness of the Banked L through d cannot be =et, the Positten Withdrawal Sequence r reactor shall not be started, input to the RWM computer shall
. or if the reactor is in the be verified.
I run or startup modes at less
'_ than 20% rated power, it shall 2) The RWM computer on line diag-g be brought to a shutdown nostic test shall be sucess-condition immediately. -
fully performed.
- f. The sequence restraints imposed 3) Proper annunciation of the se-I on the control rods may be re- lection error of at least one
_ moved by the use of the individua] out-of-sequence control rod in E rod position bypass switches for each fully inserted group shall T scram testing only those rods be verified.
- which are fully withdrawn in the 100% to 50% rod density range. 4) The rod block function of the RWM shall be verified by with-y drawing the first rod as an out-
. of-sequence control rod no more ir than to the block point.
E c. When required, the presence of a second licensed operator or other qualified employee to verify the following of the E correct rod program shall be verified.
L T- 4. Prior to control rod withdrawal
- 4. Control rods shall not be with- for startup, verify that at drawn for startup unless at least liast two source range channels two source range channels have an c. ave an observed count rate of observed count rate equal to or at least three counts per second.
greater than three counts per second. 5. .W hen a limiting control rod pattern exists an instrument
- 5. During operation with limiting functional test of the RBM shall control rod patterns, as deter- be performed prior to withdrawal mined by the designated quali- of the designated rod (s).
fied personnel, either:
- a. Both RBM channels shall be
. operable: or
- b. Control rod withdrawal shall be blocked: or
- c. The operating power level shall be limited so that the MCPR will remain above the safety limit assuming a single error that results in complete with-drawal of any single operable ccntrol rod.
==
3.3 and 4.3 BASES A. Reactivity Limitation
- 1. The requirements for the control rod drive system have been identified by evaluating the need for reactivity control via control rod movement over the full spectrum of plant conditions and events. As discussed in subsection III.4 of the Updated Safety Analysis Report (USAR) the l-control rod system design is intended to provide sufficient' control of core reactivity that the core could be made suberitical with the strongest rod fully withdrawn. This reactivity characteristic has been a basic assumption in the analysis of plant performance. Compliance with this requirement can be demonstrated conveniently only at the time of initial fuel loading or refueling. Therefore, the demonstration must be such that it will apply to the entire subsequent fuel cycle. The demonstration shall be performed with the reactor core in the cold, xenon-fres condition and will show that the reactor is suberitical by at least R + 0.38% Ak/k with the analytically determined strongest control rod fully withdrawn.
The value of "R", in units of %Ak/k, is the amount by which the core reactivitty, in the most reactive condition at any time in the subse-quent operating cycle, is calculated to be greater than at the time of the demonstration. "R", therefore, is the difference between the .
calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life tore reactivity. The value of "R" must be positive or zero and must be determined for each fuel cycle. ,
The demonstration is performed with a control rod which is calculated to be the strongest rod. In determining this "analytically strongest" rod, it is assumed that every fuel assembly of the some type has identical material troperties. In the actual core, however, the control cell
- material n roperties vary within allowed manufacturing tolerances, and the strongest rod is determined by a combination of the control cell geometry asd local k=. Therefore, an additional margin is included in the shutdows margin test to account for the fact that the rod used for the demonsetstion (the "analytically strongest") is not necessarily the strongest rod in the core. Studies have been made which cempara experimental c.-iticals with calculated criticals. These studies have shown that actual criticals can be predicted within a given tolerance band. For gado11nia cores the additional margin required due to control cell material manufacturing tolerances and calculational uncertainties has experimer : ally been determined to be 0.38% Ak/k. When this additional margin is deaonstrated, it assures that the reactivity control requirement is met.
- 2. Reactivity margin - inoperable control rods.
SpecificatAwn 3.3. A.2 requires that a rod be taken out of service if it
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% 3.3 and 4.3 BASES (cont'd.)
cannot be moved with drive pressure. If the rod is fully inserted and then disarmed electrically, it is in a safe position of maximum con.
tribution to shutdown reactivity. If it is disarmed electrically in a non. fully inserted position, that position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.l.
This assures that the core can be shutdown at all times with the remaining control rods assumin8 the strongest operable control red does not insert. An allowable pattern for control rods valved out of service, which shall meet this Specification, vill be determined and made available to the operator. .
In order to perform shutdown margin and control red drive scram time tests subsequent to any fuel loading operation as required by the Technical Specifications, the relaxation of the following Rod Sequence Control System restraints is required: (a) The sequence restraints imposed on the control rods may be removed by the use of the individual rod position bypass switches for scram testing only those rods which are fully withdrawn in the 100t to 50% rod density range. (b) Verify that subsequent to the use of the rod position bypass switches rod movement in the 50% rod density to preset power level range is restricted to the single notch mode.
If damage within the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be ruled out, then a generic problem effecting a number of drives cannot be r41ed out. Circumferential cracks resulting from stress assisted intergranular corrosion have occured in the collet housing of drives at several BVRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods. Limiting the period cf operation with a potentially severed collet housins and requiring increased surveillance after detecting one stuck rod vill assure that the reactor vill not be operated with a large number of rods with failed collet housings.
- 5. Control Rod
- 1. Control rod drop accidents as discussed in the USAR can lead to l
. significant core damage. If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.
Absence of such response to drive movement could indicate an uncoupled condition. Rod position indication is required for proper function of the rod sequence control system and the rod worth minimizer (RVM).
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- 2. The control rod housing support restricts the outvard movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, vill not contribute to any damage to the primary coolant system. The design basis is given in subsection III.8.2 of the USAR and the safety evaluation is l giver, in subsection VIII.8.4. This support is not required if the reactor coolant system is at atmospheric pressure since there vould then be no driving force to rapidly eject a drive housing. Additionally, the support is not required if all control rods are fully inserted and if an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain subcritical even in the event of complete ejection of the strongest control rod.
- 3. The Rod Vorth Minimizer (RVM) and the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control rods to prospecified sequences.
These sequences are established such that the drop of any in sequence control rod or control rod segment (i.e., one or more notches) would not cause the reactor to sustain a power excursion resulting in a peak fuel enthalpy in .
excess of 280 cal./gs. An enthalpy of 280 cal./gm. is well below the level at which rapid fuel dispersal could occur (i.e. , 425 cal./gm.). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Ref. Subsections III.6.6 and XIV 6.2 of the USAR and l Reference 1.
In performing the function described above, the RVM and RSCS are not required to impose any restrictions at core power levels in excess of 20% of rated.
Material in the cited references shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than 20s, regardless of the rod pattern. This is true for all normal and abnormal patterns including those which maximize the individual control rod worth, At power levels below 20% of rated, abnormal control rod patterns could produce rod worths high enough to be of concern relative to the 280 calories per gram rod drop limit. In this range the RVM and the RSCS constrain the control rod sequences and patterns to those which involve only acceptable rod worths.
The Rod Vorth Minimizer and the Rod Sequence Control System provide automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences. They serve as a backup to procedural control on control rod sequences, which limit the maximum reactivity worth of control rods. In the event that the Rod Vorth Minimiser is out of service, when required, a second licensed operator or other qualified technical plant employee whose qualifications have been reviewed by the NRC can manually fulfill the control rod pattern conformance functions of this system. In this case, the RSCS is backed up by independent procedural control to assure conformance.
The functions of the RVM and RSCS make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 204, these l
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s devices force adherence to acceptable rod patterns. Above 20% of rated power.
no constraint on rod pattern is required to assure that rod drop accident consequences are acceptable. Control rod pattern constraints above 20% of rated power are imposed by power distribution requicements as defined in Section 3.3.B.5 of these Technical Specifications. Power level for automatic cutout of. the RSCS function is sensed by first stage turbine pressure.
Because the instrument has an instrument error of 124 of full power, the nominal instrument setting is 22% of rated power. Power level for automatic cutout of the RWM function is sensed by feedwater and steam flow and is set nominally at 30% of rated power to be consistent with the RSCS setting.
Functional testing of the RWM prior to the start of control rod withdrawal at startup, and prior to attaining 20% rated thermal power during rod intertion while shutting down, will ensure reliable operation and minimize the probability of the rod drop accident.
The Reduced Notch Vorth Procedure for control rod withdrawal allows the Group Notch RSCS plants to take advantage of the Banked Position Withdrawal Sequence (BPVS) (Ref. 4). The BPVS has the advantage of having been proven statistically to have such low individual control rod worths that the possibility of a control rod drop accident (CRDA), which exceeds the 280 cal /gm peak fuel enthalpy limit, is precluded (Ref. 1).
The Reduced Notch Vorth Procedure is programmed into the RVM and is compatible with the hardwired Group Notch RSCS. In the pre checkerboard pattern (100% to 50% control rod density), the RVM will enforce the Reduced Notch Worth Procedure; while in the post checkerboard pattern (50% control rod density to RSCS low pove'r setpoint) the RSCS will enforce the rod pattern. Therefore, the RSCS is not required to be OPERABLE until the post checkerboard pattern is entered.
Af ter 50% of the control rods have been withdrawn (rod withdrawals may be simulated via use of the individual bypass switches), it is demonstrated that the Group Notch mode for the control rod drives is enforced. This demonstration is made by performing the hardware functional test sequence.
The hardware functional test sequence demonstrates that the Group Notch mode of the RSCS is operable prior to entering the Group Notch mode. The Group Notch restraints are automatically removed above 204 power.
During reactor shutdown, similar surveillance ,-hecks shall be made with regard to rod group availability as soon as automatic in?tiation of the RSCS occurs and subsequently at appropriate stages of the a ;,crol rod insertion.
4 The Source Range Monitor (SRM) sysrep performs no automatic safety system function; i.e., it has no scram function. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are functions of the initir, neutron flux. The requirements of at least 3 counts per second assures that,gny transient, should it occur, begins at or above the initial value of 10 t of rated power used in the analyses of transients cold conditions. Or.e operable SRM channel would be adequate to monitor the approach to critienlity using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.
101a.
s 3.3 and 4.3 BASES: (C:nt'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 volu=e (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. Licensing Topical Report GE-B'n'R Generic Reload Fuel Application, NEDE-24011-P-A, (most current approved submittal).
- 2. "Supplemental Reload Licensing Submittal for Cooper Nuclear Station Unit 1,"
(applicable reload document).
- 3. General Electric Service Information Letter No. 380, Revision 1, dated February 10, 1984
- 4. General Electric Service Information Letter No. 316. Reduced Notch k' orth Procedure, Novemb er , 19 79.
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