ML20093G436
| ML20093G436 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/12/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20093G440 | List: |
| References | |
| NUDOCS 8407240151 | |
| Download: ML20093G436 (7) | |
Text
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UNITED STATES o
t NUCLEAR REGULATORY COMMISSION h.
,j WASHINGTON, D. C. 20555 O
Y g
OMAHA PUBLIC POWER DISTRICT DOCKET N0. 50-285 FORT CALHOUN STATION, UNIT NO. I h-AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 81 License No. DPR-40 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.. The application for amendment by the Omaha Public Power District (the licensee) dated March 9,1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and t'.e Commission's tules 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 endansaring 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.
e 8407240151 840712 PDR ADOCK 05000286 P
PDR O
l 2.
Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.
DPR-40 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 81, 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 the date of its issuance.
FOR THE NUCLEAR REGULA ORY COMMISSION
,/
/
James R. Miller, Chief Operating Rea~ctors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of issuance:
July 12, 1984 9
J
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ATTACHMENT TO LICENSE AMENDMENT NO. 81 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below.
The revised pages are identified by amendment number and contain vertical lines indicating
'C the area of change.
Remove Pages Insert Pages i
1 1
2-97 2-98 (Table 2-9) 3-13 (Table 3-3) 3-13 (Table 3-3) f 4
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1 1
l TECHNICAL SPECIFICATIONS l
TABLE OF CONTENTS' Page DEFINITIONS......................................................... I
- 1. 0.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM......................
1-1 1.1 Safety Limits - Reactor Core..............................
1-1 1.2 Safety Limit, Reactor Coolant System Pressure............
1-4 1.3 Limiting Safety System Settings, Reactor Protective System................................................. 1-6 2.0 LIMITING CONDITIONS FOR 0PERATION'..~.'..........................
2-0 2.0.1 General Requirements.............................. 2-0 2.1 Reactor Coolant System....................................
2-1 2.1.1 Operable Components............................... 2-1 2.1.2 Heatup and Cooldown Rate.......................... 2-3 2.1.3 Reactor Coolant Radioactivity..................... 2-8 2.1. 4 Reactor Coolant System Leakage Limits............. 2-11 2.1.5 Maximum Reactor Coolant Oxygen and Halogens Concentrations...................................
2-13 2.1.6 Pressurizer and Steam System Safety Valves........ 2-15 2.1.7 Pressurizer Operabili ty............................ 2-16a 2.1.8 Reactor Coolant System Vents...................... 2-16b c.__
2.2 Chemical and Vol ume Control Sys tem....................... 2-17 2.3 Emergency Core Cooling Sys tem............................. 2-20 j
2.4 Containment Cooling...................................... 2-24 2.5 Steam and Feedwater Systems.............................. 2-28 2.6 Containment System....................................... 2-30 2.7 El ec tri ca l Sys tems....................................... 2-32 i
2.8 Refueling Operations..................................... 2-37 2.9 Radi oacti ve Materials Rel ease............................ 2-40 2.10 Reactor Core............................................. 2-48 l
2.10.1 Minimum ~ Conditions for Criticali ty................ 2-48 qi 2.10.2 Reactivity Control System and Core Physi.cs P a rame te r L i mi t s................................
2-50 l
. 2.10. 3 In-Core Ins trumentation........................... 2-54 2.10.4 Power Distribution Limits......................... 2-56 2.11 Containment Building and Fuel Storage Building Crane...... 2-58 2.12 Control Room Sys tens..................................... 2-59 2.13 Nuclear Detector Cooling System.......................... 2-60 2.14 Engineered Safety Features System Initiation Instrumentation Setti ngs............................... 2-61 2.15 Ins trumentation and Control Systems...................... 2-65 2.16 River Level.............................................. 2-71 2.17 Miscellaneous Radioactive Material Sources............... 2-72
' 2.18 Shock Suppressors (Snubbers )............................. 2-73 d
2.19 Fi re Protection Sys tem................................... 2-89 2.20 Steam Generator Coolant Radioactivity.................... 2-96 2.21 Post-Accident Monitoring Instrumentation................. 2-97 l-
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. Amendment No. 32, 38, 52, 54, 57 L 67, 80, ;81
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2.0 LIMITING CONDITIONS FOR OPERATION 2.21 Post-Accident Monitoring Instrumentation Applicability Applies to post-accident monitoring instrumentation not included as part of the Reactor Protective System or Engineered Safety Features.
This specification is applicable while in modes 1, 2 and 3.
s+
Objective To assure that instrumentation necessary to monitor plant parameters during post-accident conditions is operable or that backup methods of analysis are available.
l Specifications Post-accident instrumentation shall be operable as provided in Table 2-9.
If the required instrumentation is not operable, then the appropriate action specified in Table 2-9 shall be taken.
Basis i
Post-accident monitoring instrumentation provides information, during and following an accident, which is considered helpful to the operator in determining the plant condition.
It is' desirable that this instru-mentation be operable at all tires during operation of the plant.
However, none of the post-accident monitors are required for safe shutdown of the plant nor are any control or safety actions initiated *
.by the monitors.
In general, the post-accident monitors provide wide range capabilities j
for parameters which are beyond the range of normal protective and j
control. instrumentation. They also provide remote sampling and j
analysis capability to reduce personnel, exposure under post-accident.
conditions.
Because the information necessary to assess the effect of
- an accident (i.e., core damage) can be obtained from other sources and by manual methods, it is not necessary that the post-accident monitors be operable at all times.
i l
t, 2-97.
Amendment No. 81-j%:
l l
l TABLE 2-9 Post-Accident Monitoring Instrumentation Operating Limits Minimum Operable Instrument Channels Action 1.
Containment Wide Range Radi5 tion Monitors (RM-091A & B) 2 (a) 2.
Wide Range Noble Gas Stack Monitor RM-063L (Noble Gas Portion Only) ~'
I a)
RM-063M (Noble Gas Portion Only) 1 a)
RM-063H (Noble Gas Portion Only)
I a) 3.
Main Steam Line Radiation Monitor (RM-064) 1 (a)
(.a) With the number of OPERABLE channels less than required by the minimum channels operable requirements, initiate the pre-planned alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and 1.
either restore the inoperable channel (s) to OPERABLE status
{
within 7 days of the event, or 2.
prepare and submit a special report to the Commission pursuant to specification 5.9.3 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedules for restoring the system to i
OPERABLE status.
1 2-98 Amendment No. 81
l TABLE 3-3 MINIMUM FREQUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS Surveillance Channel Descriptiori Function Frequency Surveillance Method
.1.
Primary CEA Position
- a. Check S
a.
Comparison of output data with secondary CEAPIS.
Indication System
- b. Test M
b.
Test of power dependent insertion limits, devia-tion, and sequence monitoring systems.
I
- c. Calibrate R
c.
Physically measured CEDM position used to verify system accuracy. Calibrate CEA position inter-locks.
2.
Secondary CEA Position
- a. Check S
a.
Comparison of output data with primary CEAPIS.
Indication System
- b. Test M
b.
Test of power dependent insertion limit, devia-
-i tion, out-of-sequence, and overlap monitoring 1
systems.
- c. Calibrate R
c.
Calibrate secondary CEA position indication w
.L system and CEA interlock alanns.
w 3.
Area, Process, and.
- a. Check D.
a.
Normal readings observed and internal test Post-Accident signals used 'to verify ihstrument operation.
Radiation Monitors
- b. Test M
b.
Detector exposed to remote operated radiation i
check source or test signal.
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- c. Calibrate R
c.
RM-063L, M, and H and RM-064 - One time factory
[
3 calibration is acceptable provided linearity i
g solid sources are used to check the integrity of the detectors.
RM-091A and B - In situ calibra-t tion by electronic signal substitution is accept-o g
able for all range decades above 10 R/hr.
In situ calibration for at least one decade below I
10 R/hr shall be by means of calibrated radiation t
source. All other monitors - Exposure to known radiation source.
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