ML20116K621
| ML20116K621 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/08/1996 |
| From: | Kelly G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20116K625 | List: |
| References | |
| NUDOCS 9608150140 | |
| Download: ML20116K621 (8) | |
Text
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UNITED STATES yo j
j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. - a'*1
%...../
IES UTILITIES INC.
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 puANE ARNOLD ENERGY CENTER AMENDMEhT TO FACILITY OPERATING LlrrmE Amendment No. 217 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by IES Utilities Inc., et al.,
dated January 18, 1996, 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. OPR-49 is hereby amended to read as follows:
9608150140 96080s PDR ADOCK 05000331 P
' (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 217, 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 prior to startup from RF014.
.FOR THE NUCLEAR REGULATORY COMISSION
,/
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GlennB. Kelly,ProjedtManager Project Directorate 111-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: August 8, 1996 4
e
ATTACMENT TO LICENSE AMENDMENT NO. 217 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed p. ages. The revised areas are indicated by vertical lines.
Remove Insert 1.1-16 1.1-16 3.2-3 3.2-3 i
3.2-4 3.2-4 3.2-9 3.2-9 3.2-43 3.2-43 l
t
DAEC-1 the IRM channel closest to the withdrawn rod is by-passed. The results of.
this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the Safety Limit.
Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
4 B.
Scram and Isolation on Reactor Low Water Level 4
The setpoint for the low level scram is above the bottom of the separator i
skirt. This level has been used in transient analyses dealing with coolant 4
inventory decrease. Analyses show that scram and isolation of all process lines (except main steam and reactor water cleanup) at this level 4
adequately protects the fuel and the pressure barrier, because MCPR is greater than the Safety Limit in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 21 inches below the normal operating range and is thus adequate to avoid spurious scrams.
C.
Scram - Turbine Ston Valve Closure The turbine stop-valve closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram setting at 10 percent of valve closure, the resultant increase in surface heat flux l
Amendment No. !?,35,1:0,217 1.1-16
F Table 3.2-A 3g ISOLATION ACTUATION INSTRUMENTATION 3*
MINIMUM VALVE 5
OPERABLE GROUPS APPLICABLE CHANNELS ISOLATED OPERATING PER TR BY SYSTEM {P 3
SIGNAL ACTION TRIP FUNCTION TRIP LEVEL SETTING MODE c;
Common Isolation Sianals l
1,2,3 2
2 20 0
Reacter Water Level-tow 2 170 Inches 1 2,3, and
- 2 3(*)
26 T'
1,2,3 2
4 23 Reactor Water Level - Low-Low-Low 2 18.5 Inches 1,2,3 2
1 21 1,2,3 4(")
7 20 Drywell Pressure - High 5 2.0 psig 1,2,3 2
2 20 1,2,3 2
3(*)
26 m
1,2,3 2
4 23
- m d,
1,2,3 1")
9 23 Main Steam Line Isolation Main Steam Line Pressure - Low 2 850 psig 1
2 1
22 Main Steam Line Flow - High s 140% of Rated 1,2,3 2/line 1
20 Steam Flow Condenser Backpressure - High s 20 In. Hg 1,2**,3**
2 1
21 Main Steam Line Tunnel s 200*F 1,2,3 4(*)
1 21 Temperature - High Turbine Building Temperature - High 5 200*F 1,2,3 4
1 21 Main Steam Line Radiation - High' s 3 x Normal Rated 1,2,3 2
1 "')
21 Power U
Background )
Table 3.2-A (Continued) g k
ISOLATION ACTUATION INSTRUMENTATION 2
a MINIMUM VALVE OPERABLE GROUPS z
P APPLICABLE CHANNELS ISOLATED OPERATING PER TRIP BY 8
TRIP FUNCTION TRIP LEVEL SETTING MODE SYSTEM'*)
SIGNAL ACTION 3
L Refuel Floor Exhaust Duct -
s 9 mr/hr 1,2,3 and
- 1 3(*)
26 j?
High Radiation h!
Reactor Building Exhaust D aft s 11 mr/hr 1,2,3 and
- I 3(*)
26 g
- High Radiation Offgas Vent Stack - High Note k Note m 1
3(*)
27 Radiation RHR System Shutdown Coolino
- [
Reactor Vessel Pressure - High s 135 psig 1,2,3 1
4 23 i
Reactor Water Cleanuo RWCU Differential Flow - High s 40 gpmd 1,2,3 1
5 23 RWCU Area Temperature - High s 130*F 1,2,3 1
5 23 N
RWCU Area Ventilation A 14*F )
1,2,3 1
5 23 Differential Temperature - High Standby Liquid Control System NA Note i 1/SBLC 5(
23 Initiation System RWCU Area Near TIP Room Ambient s Ill.5*F 1,2,3 1
5 23 Temperature - High Reactor Water Level - Low-Low 2 119.5 Inches 1,2,3 2
5 23
F I
g Table 4.2-A (Continued)
E l
g ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREMENTS o
g OPERATING l
CHANNEL NODES FOR WHICH
[a CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE l
TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED 5$
n S
RHR System Shutdown Coolina M
Reactor Vessel Pressure - High NA Q
Q 1,2,3 l
o i
Reactor Water Cleanuo RWCU Differential Flow - High D
Q Q
1,2,3 W
RWCU Area Temperature - High NA Q
A l', 2,3 RWCU Area Ventilation Differential Temperature NA Q
A 1,2,3 u,
k3
- High d3 Standby Liquid Control System Initiation NA R
NA Note b Reactor Water Level - Low-Low Once/ Shift Q Q
1,2,3 Reactor Core Isolation Coolina RCIC Steam Line Differential Pressure (Flow) -
NA Q
Q 1,2,3 High RCIC Steam Supply Pressure - Low NA Q
Q 1,2,3 RCIC Turbine Exhaust Diaphragm Pressure - High NA Q
R 1,2,3 RCIC Equipment Room Temperature - High D
Q A
1,2,3 RCIC Room Ventilation Differential Temperature D
Q A
1,2,3
- High RCIC Leak Detection Time Delay NA NA A
1,2,3 Suppression Pool Area Temperature - High D
Q A
1,2,3 Suppression Pool Area Ventilation Differential D
Q A
1,2,3 Temperature - High Nanual Initiation NA R
NA 1,2,3 RCIC System Initiation (N0-2404 Not Full Closed)
NA R
NA 1,2,3
3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. The objectives of the Specifications are:
1.
To ensure the effectiveness of the protective instrumentation when required including periods when portions of such systems are out of service for maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
2.
To prescribe the trip settings required to assure adequate performance.
Some of the settings on the instrumentation that initiate or control core and l
containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the 4
setting has a direct bearing on safety, are chosen at a level away from the j
i normal operating range to prevent inadvertent actuation of the safety system j
involved and exposure to abnormal situations.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
The trip level settings given for reactor water level repreunt the indicated water level.
The reactor water level trip settings are dr. fined or described i
in " inches" above the top of active fuel. The term top of active fuel, i
however, no longer has a precise physical meaning since the length of the fuel pellet columns has changed over time from that of the initial core load.
Since the basis of all safety analyses is the absolute level (inches above 4'
vessel zero) of the trip settings, the " top of active fuel" has been arbitrarily defined to be 344.5 inches above vessel zero. This definition is the same as that given by Figure 5.1-1 of the Updated Final Safety Evaluation Report (UFSAR) for the initial core and maintains the consistency between the various level definitions given in the UFSAR and the Technical Specifications.
The low water level instrumentation set to trip at 170" above the top of the active fuel closes all isolation valves except those in Groups 1,5,6,7, and 9.
l i
For valves which isolate at this level this trip setting is adequate to prevent uncovering the core in the case of a break in the largest line assuming a 60 second valve closing time. Required closing times are less than this.
t The low-low reactor water level instrumentation is set to trip when reactor 1
water level is 119.5" above top of the active fuel. This trip initiates the HPCI and RCIC, trips the recirculation pumps, and isolates RWCU. The low-low-l i
low reactor water level instrumentation is set to trip when the water level is 18.5" above the top of the active fuel.
This trip activates the remainder of the ECCS subsystems, closes Group 7 valves, closes Main Steam Line Isolation Valves, Main Steam Drain Valves, Recire Sample Valves (Group 1) and starts the 1
Amendment No. 109,128,193, 217 3.2-43