ML20003E356
| ML20003E356 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/25/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003E353 | List: |
| References | |
| NUDOCS 8104030076 | |
| Download: ML20003E356 (12) | |
Text
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NUCLE AR REGULATORY COMMISSION
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W ASHINGTON, D. C. 205S6 j
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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment b. 57 License No. DPR-40 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for anendment by the Omaha Public Power District (the licensee) dated May 19, 1980, 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 app 11-cation, 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 Comission's regulations; D.
The issuance of this_ amendment will rot 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.
8t104030o%
. 2.
Accordingly, Facility Operating License No. DPR-40 is hereby amended as follows:
(1) Amend paragraph 3.B. to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 57, are hereby incorporated in the license. The ifcensee shall operate the facility in accordance with the Technical Specifications.
(2) Delete paragraphs 3.C., 3.E., 3.F., and 3.G.
(3) Renumber existing paragraph 3.D. to read 3.C.
(4) Delete paragraphs 4.A., 4.B., 4.C., 4.D., 4.E., and 4.F.
(5) Renumber existing paragraph 5 to read 4.
3.
This license amendment is effective as of the date of its issuance.
FOP, THE NUCLEAR REGULATORY COMMISSION
~
"A h
l i
Robert A. Clark, Chief l
Operating Reactors Branch #3 Division of Licensing Attachments:
- 1) Revised page 4 to OPR-40
- 2) Changes to the Technical l
l-Specifications Date of Issuance: March 25, 1981
ATTACHMENT TO LICENSE AMENDMENT NO. 57 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 1.
Replace the following pages of the Operating License and the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert a) Operating License 4
4 4a 5
b) Appendix "A" Technical Specifications 2-57c 2-57c 2-57d 2-57d 2-57e 2-57e 2-57f 5-20 5-20 5-21 (added) 6-2 6-2 2.
Remove the following blank pages from the Appendix "A" lechnical Specifications.
Remove i
Figure 1-4 Figure 1-5 Figure 1-6 Figure 1-7 2-7a 2-55A 2-55B 2-55C 2-55D 2-55E k
n
. _ _ _. - = - - _ - _ _ _ _ -
. A.
Maximum Power Level l
Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not to exceed 1500 megawatts thermal (rated power).
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.-
Security Plan The licensee shall maintain in effect and fully implement all provisions of the Commission-approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of proprietary documents, collectively titled,
" Fort Calhoun Station Unit No. 1 Site Security Plan," dated April 7, 1978, with Revision No. 1 dated July 31, 1978.
4.
This amended license is effective as of the date of issuance and shall expire at midnight on June 7, 2008.
FOR THE ATOMI ENERGY COMMISSION Original signed by A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing
Enclosures:
Appendices A and B - Technical Specifications fDate of Issuance: Aug 9,1973
r TECHNICAL SPECIFICATIONS J
TABLE OF CONTENTS Page DEFINITIONS................................................
1 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 OPERATION............................
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 Maximum 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 e.....
2-15 2.1.7 Pressurizer Operability.........................
2-16a 2.2 Ct.emical and Volume Control System......................
2-17 2.3 Emergency Core Cooling byscem........................... 2-20 2.4.
Cantainment Cooling.....................................
2-24 2.5 Steam and Feedwater Systems.............................
2-28 2.6 Containment System......................................
2-30 2.7 Electrical Systems......................................
2-32 2.8 Refueling Operations..........,.........................
2-37 2.9 Radioactive Materials Release.....................
2-40 2.10 Reactor Core......................................
2-48 i
l 2.10.1 Minimum Conditions for Criticality..............
2-48
(
2.10.2 Reactivity Control System and Core Physics l
l Parameter Limits..............................
2-50 l
2.10.3 In-Core Instrumentation.........................
2-54 i
2.10.4 Power Distribution Limits.......................
2-56 1
l i
l L
2.11 Containment Building and Fuel Storage Building Crane....
2-58 l
2.12 Control Room Systets....................................
2-59 l
2.13 Nuclear Detec tor Cr oling Sys tem.........................
2-60 2.14 Engineered Safety : eatures System Initiation Instrumentation ?.ettings..............................
2-61 2.15 Instrumentation and Control. Systems.....................
2-65 2.16 River Level............................................. 2-71 2.17 Miscellaneous Radioactive Material Sources.............. 2-72 2-73 2.18 Shock Suppressors (Snubbers) 2.19 Fire Protection System..................................
2-89
. Amendment No. 32, 38, 52,-54, 57 i
TABLE OF CONTENTS (Cont'd)
?_agg 5.9 Repor t ing Re quire men t s................................. 5-10 5.9.1 Routf Reports................................. 5-10 5.9.2 Repc
'.2,l e Oc cur renc e s..........................
5-12 5.9.3 Special Reports.................................
5-15 5.9.4 Unique Reporting Requirements...................
5-15 5.10 Reco'
.etention......................................
5-18 5.11 Radiation Protection Program...........................
5-19 5.12 Environmental Qualifications...........................
5-20 5.13 Seconda ry 'Wa t e r Chemis t ry..............................
5-20 5.14 Systems Integrity......................................
5-21 5.15 Iodine Monitoring......................................
5-21 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS.....................
6-1 6.1 Limits on Reactor Coolant Pump Operation...............
6-1 6.2 Use of a Spent Fuel Shipping Cask......................
6-2 6.3 Auxiliary Feedwater' Automatic Initiation Setpoint 6-3 6.4 Operation with Less Than 75% of Incore Detector Strings Operable.....................................
6-4 F
_lii
[ Amendment No. 32,[54,f3,$4,[55,L57
.m m.
-m --
.a -. m
n.
2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)
(5) DNBR Margin During Power Operation Above 15% of Rated Power (a) The following DNB related parameters shall be maintained within the limits shown:
(i)
Cold leg Temperature 1 545"F *
(ii)
Pressurizer Pressure
> 2075 psia *
(iii) Reacv - Coolant Flow
[195,700gpm**
(iv)
Axial S.: ape Index, Y i Figure 2-7 g
(b) With any of the above parameters exceeding the limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Basis Linear Heat Rate The limitation on linear heat rate ensures that in W event of a LOCA, the peak temperature of the fuel cladding will not excee' 2200*F.
Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System, or the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.
The Excore Detector Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors and verifying that the axial shape index is maintained within the allowable
' limits of Figure 2-6 as adjusted by Specification 2.10.4.(1).(c) for the allowed linear heat rate of Figure 2-5, RC Pump configuration, and F T of Figure 2-9.
In conjunction with the use of the ecore monitoring system IXd in establishing the axial shape index limits, the follow;.x ass.=pticas are made:
(1) the CEA insertion limits'of Specification 2.10.2.(6) and lorg term insertion limits of Specification 2.10.2.(7) are satisfied. *?) the flux peaking augmentation factors are as shown in Figure 2-8, (3) the azimuthal power tilt restrictions of Specification 2.10.4.(4) are satisfied, and (4) the total planar radial peaking factor does not exceed-the limits of Specification 2.10.4.(3).
- Limit not appTicable during either a thermal power ramp in excess of 5% of rated thermal power per minute or a thermal power step of greater than 10%
of rated thermal power.
- This number is an actual limit (not including uncert inties).
All other values in this listing are indicated values and include an allowance for measurement uncertainty (e.g., 545 F, indicated, allows for an actual T f 547 F).
c
. FORT CALHOUN-2-57c Amendment No. 32, #3, 57
C 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) i 2.10.4 Power Distribution Limits (Continued)
The Incore Detector Monitoring System provides a direct measure of the peaking factors and the alarms which have been established for the indiviFial incore detector segments ensure that the peak linear heat rates will be continuously maintained within the allowable limits of Figure 2-5.
The setpoints for these alarms include allowances, set in the conservative dirtctions, for the factors listed in 2.10.4.(1).
I Total Planar and Integrated Radial Peaking Factors (F and FR ) and
- Y AzimuthalPowerTilt(Tg T
The limitations of F and T are provided to ensure that the assumptions xy q
used in the analysis.for establishing the Linear Heat Rate and Local Power Density.- High LCO's and LSSS setpoints remain valid during operation at the T
various allowable CEA group insertion limits..The limitations on F and T areprovidedtoensurethattheassumptionsusedintheanalysiseskablishi8g the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain validpurin9operationatthevariousallowableCEAgroupinsertionlimits.
If F
,F or T exceed their basic limitations, operation may continue undeFthe$dditio8alrestrictionsimposedbytheactionstatementssincethese additional restrictions provide adequate assurance that the assumptions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCO's and LSSS setpoints remain valid.
An azimuthal power tilt >0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of
.this unexpected tilt.
T, The value of T that must be used in the equation F T=Fxy(1 + T ) and FR F (1 + T ) is Ehe' measured tilt.
xy q
R q
T T
ihe surveillance requirements for verifying that F p
nd T are within p,
their limits provide assurance that the agtual va}5Xs,of F 9
and T do not exceed the assumed values.
Verifying F and F aftereIEhfuel90ading prior,toexceeding70%ofratedpowerpNvidesahditionalassurancethatthe core was properly loaded.
DNBR Margin During Power Operation Above 15% of Rated Power The selection of limiting safety system settings and reactor operating limits
'T such that:
1.
No specified acceptable fuel design limits will be exceeded as a result of the design basis anticipated operational occurrences, and 2.
- The consequences of the design basis postula cd accidents will be no more severe than the predicted acceptable consequences of the accident analysis in Section 14.
FORT CALHOUN.
=2-57d Amendment 32, 57
2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)
In order for these objectives to be met, the reactor must be operated consistent with the operating limits specified for margin to DNB.
The parameter limits given in-(5) and Figure 2-9 along with the parameter limits on quadrant tilt and control element assembly position (Figure 2-4) provide a high degree of assurance that the DNB overpower margin will be maintained during steady state operation.
The actions specified assure that the reactor is brought to a safe condition.
The reactor coolant pump differential pressure monitoring system that will be used to measure flow provides an accurate method of determining reactor coolant flow.
The procedure for determining individual pump and reactor vessel flow will be as follows:
1.
Obtain a pump casing AP, using the precision resistor and high accuracy digital voltmeter and converting to pressure.
2.
Obtain cold leg temperature and pressurizer pressure.
3.
Correct the reading to the curve specific gravity.
4.
Obtain pump flows from individual pump casing vs. flow curves.
5.
Add the individual pump flows to obtain the best estimate reactor vessel flow.
i FORT CALHOUN
.2-57e Amend 1ent No. 32, 57
5.0 ADMINISIRATIVE CONTROLS 5.12 Environmental Qualification 5.12.1 By no later than June 30, 1982 all safety related electrical equipment in the facility shall be qualified in accordance with the provisions of:
Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (00R Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," December 1979.
Copies of these documents are attached to Order for Modification of License DPR-40 da' d October 24, 1980.
5.12.2 By no later than December 1, 1980, complete and auditable records must be available and maintained at a central lccation which describe the environ-mental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guide-lines or NUREG-0588.
Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
- 5. 13 Secondary Water Chemistry A secondary water chemistry monitoring program to inhibit steam generator tube degradation shall be implemented.
This program shall be described in the station chenistry manual and shall include:
1.
Identificatir, of a sampling schedule for the critical parameters and control points for these parameters; 2.
Identification of the procedures used to measure the values of the critical parameters; 3.
Identification of process sampling points; 4.
Procedures for the recording and management of data; 5.
Procedures defining cerrective actions for off control point chemistry conditions; and 6.
A procedure identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective actions.
Ofdit ditid Ottindt lli 1980 FORT CALHOUN
'5-20 Amendment No. 57
6 5.0 ADMINISTRATIVE CONTR0!,5 5.14 Systems Integrity A program to rev. ace leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels shall be implemented.
This program still include the following:
1.
Provisions establishing preventive maintenance and periodic
.sual inspection requirements, and 2.
Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
5.15 Iodine Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions shall be implemented.
This program shall include the following:
1.
Training of personnel, 2.
Procedures for monitoring, and 3.
Provisions for maintenance of sampling and analysis equipment.
FORT CALHOUN-5-21 Amendment No. 57
m=
6.0 INTERIM SPECIAL TECHNICAL SPECIFICATION 6.2 'Use of Spent Fuel Shipping Cask i
(This Specification is Deleted - Page Intentionally left Blank) 4 t
i
- Amendment Nc. 5, 32, - 57 6-2-