ML20205P255
| ML20205P255 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/27/1987 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Central Iowa Power Cooperative, Corn Belt Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20205P258 | List: |
| References | |
| DPR-49-A-141 NUDOCS 8704030218 | |
| Download: ML20205P255 (8) | |
Text
~.
D UNITED STATES
~ *
- 8 NUCLEAR REGULATORY COMMISSION n
WASHINGTON, D. C. 20555 Y
'%..... S IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARN0LD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.141 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Iowa Electric Light and Power Company, et al, dated August 29, 1986 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 without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's. regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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-49 is hereby amended to read as follows:
EPn!M058886 PDR
~
g..
-,-v-
. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.141, are hereby incorporated.
in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A
Daniel R. Muller, Director BWR Project Directorate 32 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 27, 1987 i
r
ATTACHMENT TO LICENSE AMENDMENT NO.141 FACILITY OPERATING LICENSE N0. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Pages 8
3.3-6 3.3-8 3.3-12 3.3-18 3.3-19 i
a a
h l
4 i
i
., ~., _
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C.
Scram Insertion Times C.
Scram Insertion Times 1.
The average scram insertion 1.
After each refueling outage all time, based on the de-OPERABLE rods shall be scram l
energization of the scram time tested from the fully pilot. valve at time zero, of withdrawn position to the l
all OPERABLE control rods in drop-out of the reed switch at the reactor power operation the rod position required by condition shall be no greater Specification 3.3.C.
The than:
nuclear system pressure shall be above 950 psig (with Average Scram saturation temperature) and the Rod Insertion requirements of Specification Position Times (Sec) 3.3.B.3.a met. This testing shall be completed prior to 46 0.35 exceeding 40%. power.
Below 30%
38 0.937 power, only rods in those 26 1.86 sequences (A12 and A g or_B12 3
06 3.41 and B g) which are fully 3
withdrawn in the region from' 2.
The average scram insertion
-100% rod density to 50% rod times for the three fastest density shall be scram time control rods of all groups of tested.
During all scram time four control rods in a 2 x 2 testing below 30% power, the array shall be no greater Rod Worth Minimizer shall be than:
OPERABLE or a second licensed operator shall verify that the Average Scram operator at the reactor console Rod Insertion is following the control rod Position Times (Sec) program.
46 0.37 38 1.01 26 1.97 06 3.62 3.
Maximtsn scram insertion time to rod position 04 of any OPERABLE control rod should not exceed 7.00 seconds.
Amendment No. 114, 117, 141 3.3-6 i
DAEC-1 3.3 and 4.3 BASES l
A.
Reactivity Limitation l
1.
The requirements for the control rod drive system have been identified by evaluating the need for reactivity centrol via control rod movement over the full spectrum of plant conditions and events. As discussed in Subsection 4.6.1 of the Updated FSAR, the control rod system design is intended to provide sufficient control of core reactivity that the core could be made subcritical 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-free condition and will show that the reactor is subcritical by at least R + 0.38% a k/k with the analytically determined j
strongest control rod fully withdrawn.
Amendment No. Jif, 141 3.3-8
,_g
. - - ~
DAEC-1 l
B.
Control Rod Withdrawal l
1.
Control. rod drop accidents as discussed in the Updated FSAR can lead to significant core damage.
If coupling integrity is maintained, the possibility of a rod drop 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 (RWM).
l 2.
The control rod housing support restricts the outward 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, will not contribute to any damage to the primary coolant system.
The design basis is given in Subsection 4.6.1 of the Updated FSAR and the safety evaluation is given in Subsection 4.6.2 of the Updated FSAR.
This support is i
not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing. Additionally, the support is not required j
if all control rods are fully inserted and if an adequate shutdown l
margin with one control rod withdrawn has been demonstrated, i
3.3-12 l
Amendment No. fff, 141 s
DAEC-1 During the use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.
It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either-when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other personnel qualified to perform this function may be designated by the Plant Superintendent, Nuclear.
l C.
Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR-from becoming less than the safety limit.
After initial fuel loading and subsequent refuelings when operating above 950 psig, all control rods shall be scram tested within the constraints imposed by the Technical Specifications and before the 40%
power level is reached.
The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
{
I Amendment No. J20,141 3.3-18 i
i I
l
DAEC-1 D.
Reactivity Anomalies l
During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.
The magnitude of this excess reactivity may be : inferred from the critical rod configuration.
As fuel burnup progresses, anomalous behavior in the excess reactivity may be dete' ted by comparison of the critical rod pattern at selected base states to c
the predicted rod inventory at that state.
Power operating base conditions provide the most sensitive and directly interpretable data relative to core f
reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.
Requiring a rea'ctivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% AK/K.
l Deviations in core reactivity greater than 1% AK/K are not expected and require l
thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.
E.
Recirculation Pumps
]
APRM and/or LPRM oscillations in excess of those specified in section 3.3.E could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be taken.
These specifications are based upon the guidance of GE SIL #380, Rev. 1, 2/10/84.
Amendment No. JJ9, 120, 141
- _. _