ML20100K373
| ML20100K373 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/03/1985 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20100K375 | List: |
| References | |
| NUDOCS 8504150290 | |
| Download: ML20100K373 (42) | |
Text
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o UNITED STATES g
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NUCLEAR REGULATORY COMMISSION
,y 4p WASHINGTON, D. C. 20555
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OMAHA PUBLIC POWER DISTRICT DOCKET N0. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 86 License No. DPR-40 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Omaha Public Power District (the licensee) dated March 21, 1978 and supplements dated March 30, ments of the Atomic Energy Act of 1954, as amended (the Act) quire-1979 and October 18, 1984 complies with the standards and re
, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the previsions of the Act, and the rules and regulations of the Comission; 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 not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
8504150290 850403 ADOCK 0500 5
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. 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. 86, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective within six months of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION I
,lt(
James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
. Changes to the Technical Specifications Date of Issuance: April 3, 1985 n.
=
ATTACHMENT TO LICENSE AMENDMENT N0. 86 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 the area of change.
Remove Pages Insert Pages Remove Pages Insert Pages i
i 3-64 3-64 11 ii 3-65 3-65 iii iii 3-66 3-66 4
4 3-67 3-67 7
7 3-68 3-68 8
8 3-69 3-69 2-40 2-40 3-70 3-70 2-41 2-41 3-71 3-71 2-42 2-42 3-71a 2-43 2-43 3-72 3-72 2-44 2-44 3-73 3-73 2-45 2-45 3-74 3-74 2-46 2-46 3-75 3-75 2-47 2-47 5-8 5-8 2-47a 5-8a 2-62 2-62' 5-15 5-15 2-64 5-16 5-16 2-64 2-64a 2-64a 5-17 5-17 3-13 3-13 5-19 5-19
~3-19 3-19
f TECHNICAL SPECIFICATIONS 1
TABLE OF CONTENTS i
Page.
i 4
DEFINITIONS.........................................................
1 1.0 ' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS..............
1-1 l
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-O' 2.0.1 General Requirements..............................
2 2.1 Reactor Cool ant Sys tem...................................
2-1 2.1.1 Operable Components...............................
2-1 2.1.2 Heatup and Cool down 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 Concentrati o ns.................................. 2-13 2.1.6 Pressurizer.and Steam System Safety Valves........
2-15 2.1.7 Pres s uri ze r Ope rabil i ty.. :........................
2-16a 2.1.8 Reactor Coolant System Vents......................
2-16b 2.2 Chemical and Vol ume Control System.......................
2-17 2.3 Emergency Core Cooling System............................
2-20 2.4 Containment Cooling..........e............................. 2-24 2.5 Steam and.Feedwater Systems..............................
2-28 2.6 Containment System.......................................
2-30 2.7 El ectri cal Sys tems.......................................
2-32 2.8 Refuel i ng Operations....................................
2-37 2.9 Radioactive Effluentss........,...........................
2-40 l
2.9.1 Li quid and Gaseous Effl uents...................... 2-40 2.9.2 Solid Radioactive Waste...........................
2-47a 2.10 Reactor Core.............................................
2-48 2.10.1. Minimum Condi tions for Criticality................
2-48 2.10.2 Reactivity Control Systems and Core Physics Parameter Limits................................
2-50 2.10. 3 In-Core Instrumentation...........................
2-54 2.10.4 Power Distribution Limits.........................
2-56 2.11 Containment Building and Fuel ' Storage Building Crane.....
2-58 I
i Amendment No. 32,38,52,54,57,67, 89, SJ, 86 O
TABLE OF CONTENTS (Continued) 4 Page 2.12 Co ntrol Room Sys tems.....................................
2-5 9 l
2.13 Nuclear Detector Cooling System..........................
2-60 2.14 Engineered Safety Features System Initiation Instrumentation Settings...............................
2-61 2.15 Instrumentation and Control Systems......................
2-65 2.16 R i v e r L e v el.............................................. 2 -71 i
2.17 Miscellaneous Radioactive Material Sources'...............
2-72 l
2.18 Shock Suppressors (Snubbers).............................
2-73 2.19 Fire Protection System...................................
2-89
. 2.20 Steam Generator Cool ant Radioactivi ty....................
2-96 2.21 Post-Accident Monitoring Instrumentation.................
2-97 4
3.0 SURVEILLANCE REQUIREMENTS..'....................................
3-1 3.1 Instrumentation and Contro1..............................
3-1 l
3.2 Equi pment and Sampling Tests.............................
3-17 3.3 Reactor Coolant System, Steam Generator Tubes, and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing
. S u rve 11 1 a nce...........................................
3 - 21 3.4 Reactor Coolant System Integrity Testing................. 3-36 3.5 Containment Test.........................................
3-37
- 3.6 Safety Injection and Containment Cooling Systems
'~..
Tests..................................................
3-54 3.7 Emergency Power System Periodic Tests....................
3-58 3.8 Main Steam Isolation Va1ves..............................
3-61 3.9 Auxil i ary Feedwater System............................... 3-62 3.10 Reactor Core Parameters..................................
3-63 3.11 Radiological Environmental Monitoring Programs...........
3-64 3.12 Radiological Waste Sampling a'nd Monitoring...............
3-69 3.12.1 iiquid and Gaseous Effluents.....................
3-69 3.12.2 Solid Radioactive Waste..........................
3-71a 3.13. Radioactive Material Sources Surveillance................
3-76 3.14 Shock Suppressors (Snubbers).............................
3-77 3.15 Fire Protection System...................................
3-80 4.0 DESIGN FEATURES................................................
4-1 4.1 Site.....................................................
4-1 4.2 Containment Design Features..............................
4-1 4.2.1 Contai nment Structure............................ 4-1 4.2.2 Penetrations.....................................
4-1 4.2.3 Containment Structure Cooling Systems............
4-2 I
ii Amendment No. 35,43, #6, 54, 69, 86, 86 9
O TABLE OF CONTENTS (Continued Page 4.3 Nucl ear Steam Supply Sys tem (NSSS)........................
4-3 4.3.1 Reactor Cool ant System.............................
4-3 4.3.2 Reactor Core and Contro1...........................
4-3 4.3.3 Emergency Core Cooling.............................
4-3 4.4 Fuel Storage..............................................
4-4 4.4.1 New Fu el S to ra g e...................................
4-4 4.4.2 Spent Fuel Storage.................................
4-4 5.0 ADMINISTRATIVE CONTR0LS........................................
5-1 5.1 Responsibility............................................
5-1 5.2 Organization..............................................
5-1 5.3 Facil i ty Staff Quali fications.............................
5-la 5.4 Training..................................................
5-3 5.5 Review and Audit..........................................
5-3 5.5.1 Plant Review Committee (PRC).......................
5-3 5.5.2 Safety Audit and Review Committee (SARC)...........
5-5 5.5.3 Fi re Protecti on Inspection.........................
5-8a l
5.6 Reportable Occurrence Action.................... J.........
5-9 5.7 Safety Limit Violation....................................
5-9 5.8 P ro c ed u re s................................................
5-9
- 5. 9 - Reporti ng Requi rements....................................
5-10 5.9.1 Routi ne Reports....................................
5-10 5.9.2 Reportabl e Occurrences.'............................
5-12 5.9.3 Special Reports..................................... 5-15 5.9.4 Un.1que Reporti ng Requi rements......................
5-15 5.10 Re c o rd s Re t e n ti on.........................................
5-18 5.11 Radiati on Protection Program..............................
5-19 5.12 Envi ronmental Qualification...............................
5-20 5.13 Seconda ry Water Chemistry.................................
5-20 5.14 Sys tems I nte g ri ty.........................................
5-21 5.15 Iodine Monitoring.........................................
5-21 5.16 Sampling and Analysis of Plant Effluents..................
5-21
-C.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.........................
6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpoint.........
6-1 S.4 Operation With Less Than 75% of Incore Detector Strings 0perable........................................
6-1 tii Amendment.No. 32, 34, AE, EA, EE, 57,73,"ES,86 6
a
L DEFINITIONS PROTECTIVE SYSTEMS (Continued)
Engineered Safety Feature Logic (2)
The system which utilizes relay contact outputs from individual instrument channels to provide a dual channel signal to independently initiate the actuation of the engineered safety feature equipment. Two logic subsystems, termed A and B, are provided; each subsystem is composed of four channels wired to provide independent safety feature initiation signals on a 2-out-of-4 basis.
Degree of Redundancy The difference between the number of operable channels and the number of channels which when tripped will cause an automatic system trip.
INSTRUMENTATION SURVEILLANCE Channel Check A qualitative determination of acceptable operabi.11ty by observation of channel behaviour during normal plant operation. 'iThis determination shall where feasible, include comparison of the channel with other independent channels measuring the same variable.
Channel Functional Test Injection of a simulated signal into the' channel to verify that it is operable, including. any alarm and/or trip initiating action.
Channel Calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including equipment action, alarms, interlocks or trip, and shall be deemed to include the channel functional test.
Source Check Verification of channel response when the channel sensor is exposed to a radioactive source.
4 Amendment No. 86 O
t
e p g n m ons Azimuthal Power Tilt - Tn Azimuthal Power Tilt shall be the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that axial half (upper or lower) of the core divided by the average power of all quadrants in that axial half (upper or lower) of the core.
Unrodded Planar Radial Peaking Factor - Fxy The Unrodded Planar Radial Peaking Factor is the maximum ratio of the peak to average power density of the ' individual fuel rods in any of the unrodded horizontal planes, excluding azimuthal tilt, T.q Unrodded Integrated Radial Peaking Factor - FR The Unrodded Integrated Radial Peaking Factor is the ratio of the peak pin
' power to the average pin power in an unrodded core, excluding azimuthal tilt, T.
q Fire Suppression Water System The fire suppression water system consists of-fire pumps and distribution piping with associated sectionalizing control or isolation valves.
Such valves include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprink1br, hose standpipe or spray system riser.
A manual or set of operating procedures detailing the program of sampling, analysis, and evaluation.
Dose Equivalent I-131 That concentration of I-131 (pCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.
In Ot' r words, 7
Amendment No. 32,38,67,86 4
9
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' DEFINITIONS Dose Equivalent I-131 (uCi/gm) = pCi/gm of I-131
+ 0.0361 x pCi/gm of I-132
+ 0.270 x uCi/gm of I-133
+ 0.0169 x uCi/gm of I-134
+ 0.0838 x pCi/gm of I-135
'E - Average Disintegration Energy E is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 minutes making up at least 95% of the total non-iodine radioactivity in the coolant.
Offsite Dose Calculation Manual (0DCM)
A manual containing the methodology and parameters to be used in the:
- 1) calculation of doses in the unrestricted area 'due to radioactive liquid
~
and gasecus effluents, 2) calculation of liquid and gaseous effluent monitoring instrumentation setpoints, and 3) specific details pertinent to the radiological environmental monitoring program.
Purge-Purging A means for the removal and replacement of gases within the contat, ment building.
Venting A means for the reduction of pressure greater than atmospheric within the containment structure.
References (1) USAR, Section 7'.2 (2) USAR, Section 7.3 8
Amendment No. 67,86 9
-.-..---,e
2.0 LIMITING CONDITIONS FOR OPERATIONS 2.9 Radioactive Effluents 2.,9.1 Liquid and Gaseous Effluents Applicability
. Applies to' the controlled release of radioactive materials 'in liquid and gaseous effluents from the facility. The provisions of Technical Speci-fication 2.0.1 for Limiting Condition for Operation are not applicable.
Objective i
To define the limits and conditions for the controlled release of radio-active materials in liquid and gaseous effluents to the environs to ensure that these releases are as low as is reasonably achievable in conformance with 10 CFR Part 50.34a and 50.36a, and to ensure that these releases result in conc'entrations of radioactive materials in liquid and gaseous effluents released to unrestricted areas are within the limits specified in 10 CFR Part 20.
To ensure that the releases of radioactive materials above background to unrestricted areas are as low as is reasonably achievable, the following design objectives apply.
A.
Liquid Effluents (1) The d:se or dose commitment to a member of the,public during any calendar year should not exceed 3, millirems to the total body.
(2) The dose or dose commitment to a member of the'public during any calendar year should not exceed 10[ millirems to any organ.
B.
Gaseous Effluents i
2 (1). The calculated annual air dose due to gamma radiation at any location which could be occupied by individuals in unrestricted areas should not exceed 10 millirads; (2) The calculated annual air dose due to beta radiation at any location which could be occupied by individuals in unrestricted areas should not exceed 20 millirads; and (3) The calculated annual total quantity of iodine-131, tritium, and all radioactive material in particulate form with half-lives greater than 8 days should not result in an annual dose or dose commitment to any organ of an individual in an unrestricted area from all pathways of exp'osure in excess of 15 millirems.
1 2-40 Amendment No. 86 I
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2.0 LIMITING CONDITIONS FOR OPERATION 2.9 Radioactive Effluents (Continued) 2.9.1 Liquid and Gaseous Effluents (Continued)
(1) Specifications for Liquid Waste Effluents a.
(i):
The release rate of radioactive material in liquid effluents shall De controlled such that the instan-taneous concentrations for radionuclides, other than dissolved or entrained noble gases, do not exceed the values specified in 10 CFR Part 20, Appehdix B, for unrestricted areas.
For dissolved or entrained noble gases, the concentration shall be limited to 2.0 t - 04 mci /ml total activity.
(ii)
With the conc.entration of radioactive material released to unrestricted areas exceeding the above limits, appropriate corrective actions shall be taken immediately to restore concentrations within the above limits.
b.
The cumulative dose contributions from radioactive materials in liquid effluents released to unrestricted areas shall.be deter-mined, in accordance with the ODCM, on' a quarterly basis.
If the dose contributions, due to the cumulative release of liquid effluents averaged over a calendar quarter, exceed one-half of the design objectives, the following ~ course of actions shall be taken:
(i)
Make an investigation to! identify the causes for such releases.
(ii)
Define and initiate a program of action to reduce such releases to the desig.n levels.
(iii) Submit a special report, pursuant to Specification 5.9.3, fwithin 30 days from the end of the quarter during which release occurred, identifying the causes and describing the proposed program of action to reduce such release to the design levels.
c.
The equipment or subsystem (s) of the liquid radwaste treatment system as identified in the ODCM shall be operated prior to the discharge of radioactive materials in liquid wastes.
If the radioactive liquid wastes were discharged without treatment by one or more of the pieces of equipment or subsystem (s) identified in the ODCM and it appears that one-half of the annual objective will be exceeded during the calendar quarter, a special report, pursuant to Specification 5.9.3, shall be prepared and submitted to the Commission within 30 days.
This report shall include the following information:
(i)
Identification of equipment or subsystems not operable and reason for inoperability.
2-41 Amendment No. 23,86
~
2.0 LIMITING ~ CONDITIONS FOR OPERATIONS 2.9 Radioactive Effluents (Continued) 2.9.1 Liquid and Gaseous Effluents (Continued)
(11)
Action (s) taken to restore the inoperable equipment to status.
(iii) Summary description of action (s) taken to prevent a recurrence.
d.
During release of radioactive liquid waste excluding releases from the steam generators, the following conditions shall be met:
(i)
At least one circulating water pump shall be in operation to provide a dilution flow of approximately 120,000 gpm in the discharge tunnel.
(ii)
The overboard header effluent radiation monitor shall be set in accordance with the ODCM to alarm and automatically close the discharge valve prior to exceeding the limits specified in 2.9.1(1)a.(i) above.
(iii) The gross liquid waste activity and flow rate shall be continuously monitored and recorded during the release.
If the effluent radiation monitor'is inoperable, effluent releases may continue provided that prior to initiating a release:
1.
At least two independent samples are analyzed in accordance with Specification 3.12.1(1).
2.
At least two qualified individuals independently verify the release rate calculations.
.If the flow rate indicator is inoperable, effluent
' releases may continue provided the flow rate is deter-mined at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual release.
If the radioactivity cannot be recorded automatically, effluent releases may continue provided the gross radio-activity level is recorded manually at least onceiper'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual release.
e.
Whenever steam generator liquid is being released to the dis-charge tunnel 1) the steam generator blowdown radiation monitors shall be set to alarm and automatically close the blowdown isola-tion valves prior to exceeding the limits specified in 2.9.1(1)a(1) above, and 2) the gross activity for each blowdown line shall be monitored and recorded by the blowdown radiation monitors.
If one of the two radiation monitors is inoperable, the activity for both blowdown lines shall be monitored by the 2-42 Amendment No. 86
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i 2.0 LIMITING CONDITIONS FOR OPERATIONS 2.9 Radioactive Effluents (Continued) 2.9.1 Liquid and Gaseous Effluents (Continued) operable radiation monitor.
If both radiation monitors are inoperable, steam generator liquid release may continue provided appropriate grab samples are analyzed for principal gamma emitters at a sensitivity of 5.0E-07 pCi/ml and recorded at least daily when the specific activity of the sample is less than or equal to 0.01 uCi/ gram dose equivalent I-131 and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 uCi/ gram dose equivalent I-131.
If the radioactivity cannot be recorded automatically, effluent releases may continue provided the gross radioactivity level is recorded manually at least once per four hours during actual release.
(2)
Specifications for Gaseous Waste Effluents a.
(i)
The release rate of radioactive materials in gaseous effluents shall be controlled such that the instan-taneous concentrations of radionuclides do not exceed the values specified in 10 CFR Part 20, Appendix B, for unrestricted areas.
(ii)
With the concentration of radioactive material released to unrestricted areas exceeding the above limits, appro-priate corrective actions shall be taken immediately to restore concentration within the above limits.
b.
The cumulative dose contributions to each of the 16 cardinal sectors, from radioactive naterials in gaseous effluents shall be determined, in accordance-with the ODCM, on a quarterly basis.
If the dose contributions, due to'the cumulative release of gaseous. effluents averaged over a calendar quarter exceed one-half of the ' design objectives, the following course of actions shall be taken:
(i)
Make an investigation to identify the cause for such release rates.
(ii)
Define and initiate a program of action to reduce such releases to design levels.
(iii) Submit a special report, pursuant to' Specification 5.9.3, within 30 days from the end of the quarter during which release occurred, identifying the causes and describing the proposed program of action to reduce dose contribu-tions.
c.
The equipment or subsystem (s) of the gaseous radwaste treatment system as identified in the ODCM shall be operated prior to the discharge of radioactive materials in gaseous wastes.
If the radioactive gaseous wastes were discharged without treatment by 2-43 Amendment No. 86 A
~
2.0 LIMITING CONDITIONS FOR OPERATIONS 2.9 Radioactive Effluents (Continued) 2.9.1 Liouid and Gaseous Effluents (Continued) one or more of the equipment or subsystem (s) identified in the ODCM, a special report, pursuant to Specification 5.9.3, shall be prepared and submitted to.the Comission within 30 days.
This report shall include the following information:.
(i)
Identification of equipment or subsystem (s) not operable and reason for inoperability.
(ii)
Action (s) taken to restore the inoperable equipment to operable status.
(iii) Summary description of action (s) taken to prevent a recurrence.
d.
The hydrogen and oxygen monitors shall be monitoring the in-service gas decay tank during the transfer of waste gases to the gas decay tank and the concentration of hydrogen and oxygen shall be limited to below' flammability, concentrations. Whenever the monitors are inoperatie, transfer of waste gases to a gas decay tank may continue p ovided grab samples are taken from the gas decay tank and analysed:
(1) every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during degassing operations, and (2) daily during other operations.
wr e.
(i)
The stack monitors for gaseous, particulate and iodine activities may be inoperable provided that 1) releases from a. gas decay tank, containment pressure relief line, and the containment purge line are secured,'and 2) when-ever the ventilation stack gas or particulate monitor is inoperable, appropriate grab samples will be taken and analyzed once per eight (8) hours.
(ii) i)uring power operation, the condenser air ejector dis-charge shall be monitored for gross radioactivity.
If this monitor is inoperable, grab samples shall be taken and analyzed daily for principal ganna emitters.
f.
During release of gaseous radioactive wastes from th.e gaseous waste discharge header or during containment venting to the ventilation stack, the following conditions shall be met:
(i)
The gas, iodine, and particulate monitors shall be monitoring the vent stack.
(ii)
At least one exhaust fan shall be in operation.
(iii) The effluent control radiation monitors shall be set in accordance with the ODCM to alarm and automatically terminate the releases specified in 2.9.1(2)a(prior to exceeding the limits 1)above.
2-44 Amendment No. J2, 86
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2.0 LIMITING CONDITIONS F0'R OPERATIONS 2.9 Radioactive Effluents (Continued)
)-
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2.9.-1 Liquid and Gaseous Effluents (Continued)
(iv)
The activity shall be monitored and recorded. The i
i flow rate shall be monitored and recorded, cr deter-mined by calculation.
(v)
'During the release of gaseous wastes from the contain-ment purge line, a conteinment gas monitor and a particulate monitor shall monitor the containment, in i
additiontoconformingwith(i)through(iv)above.
Basis Releases of radioactivity in liquid wastes within the design objective i
levels provide reasonable assurance that the resulting annual exposure from liquid effluents will not exceed the limits specified in Appendix I to 10 CFR Part 50. These specifications provide reasonable assurance that the resulting exposure will not exceed 3 mrem to total body or 10 mrem to any organ. At the same time, these specifications permit the flexibility of operation, compatible with considerations of health and safety,- to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in n
releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20.
The design objectives have been developed based on operating ~ experience, calculational procedures based on models and data set forth in Regulatory Guide 1.109, and the evaluation of Fort Calhoun facility in accordance
{
with Appendix I of 10 CFR Part 50 dose design objectives. The design objectives take,into account a corrbination of variables including fuel failures, primary system leakage, primary-to-secondary system leakage i
and the performance of various radioactive waste treatment systems.
t 1
Specification 2.9.1(1)a requires the licensee to limit the concentration of radioactive materials in liquid effluents released from the site to 4
I levels specified in 10 CFR Part 20, Appendix B, fer unrestricted areas.
This specification provides assurance that no member of the general public will be exposed at any time to liquid containing radioactive materials in i-excess of limits considered permissible under the Commission's Regulations.
Specification 2.9.l(l)b establishes the frequency of dose calculations in
~accordance with the ODCM.
This specification also establishes the report-ing requirements in accordance with Section IV.A of Appendix I'to 10 CFR Part 50, in addition to the requirements of Section 5.9 of these Technical
)
Specifications.
Specification 2.9.l(1)c requires the operation of the equipment or sub-i system (s) of the radioactive liquid waste system, as identified in the ODCM, j.
to reduce the release cf radioactive materials in liquid effluents to as l
low as reasonably achievable, consistent with the requirements of 10 CFR Part 50.36a, and General Design Criterion 60 of Appendix A to 10 CFR Part 50.
i Nonnal use of the equipment or subsystem (s) in the radioactive liquid waste i
system provides reasonable assurance that the quantity released will not i
exceed the-design objectives.
1 2-45 Amendment No. JZ,86 i
=.
l i
2.0 LIMITING CONDITIONS FOR OPERATIONS 1
2.9 Radioactive Effluents (Continued)
E 2.9.1 Liquid and Baseous Effluents (Continued)
Basis (Continued) j Specification 2.9.1(1)d, consistent with the requirements of General Design j
Criteria 60 and 64 of 10 CFR Part 50, Appendix A, requires operation of suitable equipment to dilute, control, and monitor the releases of radio-active materials in liquid wastes, other than steam generator liquid, from the overboard hedder during any period when releases are taking place.
P j
Specification 2.9.1(1)e requires the monitoring of the steam generator j
liquid when releases are being discharged to the environment.
Inoperabil-ity of one radiation _ monitor will not affect the monitoring capabilities j
as the.other radiation monitor would serve the intended purpose.
If both j
radiation monitors are found inoperable and if steam generator liquid is being released to the environment, the specified sampling frequency pro-i i
vides assurance that no major activity is released during a limited period of time when repairs are being made.
The release of radioactive materials in gaseous waste effluents.to unre-stricted areas will not result in concentrations that exceed limits i
specified in 10 CFR Part 20 at any time and should be as low as is reason-j ably achievable in accordance with the requirements of 10 CFR-Parts 50.34a j
and 50.36a. These specifications provide reasonable assurance that the resulting annual air dose due to gamma radiation will not exceed 10 mrad and that the resulting annual air dose to beta radiation will not exceed j
20 mrad from the gaseous waste effluents.from the plant.
These specifica-tions also provide reasonable assurance that no individual. in an unrestricted area will receive an annual dose to the total body greater than 5 mrem or an annual dose to the skin greater than 15 mrem from these gaseous effluents, and that the annual dose to any organ,of an individual from radiciodines and radioactive material in particulate form will not exceed 15 mrem.
i At the same timel these specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided with a dependable source of power even under unusual operating conditions which.may temporarily result in releases higher than such numerical guides'for design objectives but still within levels that assure that the average population exposure is equivalent to small fractions of doses from natural bar.kground radiation.
Specification 2.9.1(2)a '.equires the licensee to limit the concentration of radioactive materials in gaseous effluents from the station to levels speci-fied in 10 CFR Part 20, Appendix B, for unrestricted areas. This specifica-tien provides assurance that no member of the general public will be exposed at any time to gases containing radioactive materials in excess of limits specified in the Connission's regulations.
Specification 2.9.1(2)h establishes the frequency of dose calculations in accordance with the ODCM. This specification also establishes the reporting 2-46 Amendment No. Lt. 46
-w
,nmewe--,.-.wLn-n,.,. - ~,,.
_w,,n,--------.,,--~-w,--,,~---w-----
=
2.0 LIMITING CONDITIONS FOR OPERATIONS 2.9 Radioactive Effluents (Continued) 2.9.1 Liquid and Gaseous Effluents (Continued)
Basis (Continued) requirements in accordance with Section IV.A of Appendix I to 10 CFR Part 50, in addition to the requirements of Section 5.9 of these Tech-nical Specifications.
Specification 2.9.1(2)c requires the operation of equipment or subsystem (s) of the radioactive gaseous waste system, as identified in the ODCM, to reduce the release of radioactive materials in gaseous effluents to as low as reasonably achievable, consistent with the requirements of 10 CFR Part 50.36a, and General Design Criterion 60 of Appendix A to 10 CFR
^
Part 50.
Normal use of the equipment or subsystem (s) in the radioact e gaseous waste system provides reasonable assurance that the quantity released will not exceed the design objectives.
Specification.2.9.1(2)d ensures that the concentration of potentially explosive gas mixtures entrained in the gas decay tank (s) will be main-tained below the flammability limits of hydrogen and oxygen.
Maintaining the concentration of hydrogen and oxygen below their flammability limits i
with a measurement program provides assurance that the releases of radio-active materials will be controlled in conformance with the requirements of General Design Criteri.on 60 of Appendix A to 10 CFR Feet 50.
Specification 2.9.1(2)e provides assurance that releases from gas decay tank, containment pressure relief line, and containment. purge line are not made whenever the ventilation stack monitors are inoperable. This specification also assures that the gross radioactivity, during power operation, is monitored from the condenser air ejector discharge.
Specification 2.9el(2)f requires operation of suitable equipment to dilute, control, and monitor in order to provide assurance that radioactive materials released in the gaseous effluents are properly controlled and monitored in accordance with the requirements of General Design Criteria 60 and 64 of 10 CFR Part 50, Appendix A.
2-47 Amendment No. 86
~
2.0 LIMITING CONDITIONS FOR OPERATIONS 2.9 Radioactive Effluents (Continued) 2.9.2 Solid Radioactive Waste Applicability This specification applies to the processing and packaging of solid and compacted radwaste.
Objective To ensure conformance with 10 CFR Part 20 and 10 CFR Part 71 prior to shipment of solidified radwaste from the facility. The provisions of Technical Specification 2.0.1 for Limiting Conditions for Operation are not applicable.
Specification
()
The equipment or subsystem (s) of the solid radwaste system, as identified in the Process Control Program (PCP), shall be operated to provide for the solidification of vet solid wastes and the compaction of compressible wastes. Waste solidification will be verified by requirements specified in the PCP.
If solidified radwaste fails to meet the above !' objective" regulations or the acceptance criteria of the PCP, no offsi.te shipments shall be made of the non-conforming materials.-
Basis The solid radwaste, system is generally op'erated on a batch basis, and is available.to perform abnornal or emergency functions. The proper opera-tion of the solid radwaste system ensures that the pertinent requirements of 10 CFR Part 20 and 10 CFR Part 71 will be implemented. This specifica-tion also complies with the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The operating procedures, process parameters and the acceptance criteria, included in the Process Control Program, will provide compliance with these requi.re-ments.
2-47a Amendment No. 86
l 2.0 LIMITING CONDITIONS FOR OPERATIONS 2.14 Enoineered Safety Features System Initiation Instrumentation l
Settings (Coltinued)
(3) Containme.it High Radiation (Air Monitoring)
(Con.tinued)
The setpoints for the isolation function will be calculated in accordance with the ODCM.
Each channel is supplied from a separate. instrument A.C. bus
- and each auxiliary relay requires power to operate.
On failure of a single A.C. supply, the A and B matrices will assume a one-out-of-two logic.
(4) Low Steam Generator Pressure
' signal is provided upon sensing a low pressure in a steam A
generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of 500 psia includes a +22 psi uncertainty and was the setting used in the safety analysis.(3)
As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam. generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steamline break.
The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.
(5) SIRW Tank Low Level Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal'. The switchover point of 16 inches above tank bottom is set to prevent the pumps from running dry during the 10 seconds required to stroke the valves and to hold in -
reserve approximately 28,000 gallons of at least 1700 ppm (4) borated water.
The FSAR loss of coolant accident analysis assumed the recirculation started when the minimum usable volume of 283,000 gallons had been pumped from the tank.
(6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to one or two steam generators upon sensing a low water level in the steam generator (s) if the l
2-62 Amendment No. 5, 32, A3, 65, 86
i S
t 8
TABLE 2-1 z
P k-Engineered Safety Features System Initiation Instrument Setting Limits b
Functional Unit Channel Setting Limit 1 5 psig Safety Injection 1.
liigh Containment Pressure a.
Containment Spray (3) b.
c.
Containment Isolation
- d.. Containment Air Cooler DBA Mode II)
Safety injection
- 1600 psia 2.
Pressurizer Low / Low Pressure a.
Containment Spray (3) b.
c.
Containment Isolation d.
Containment Air Cooler DBA Mode nL 3.
Containment liigh Radiation Containment Ventilation Isolation In accordance with the Offsite Dose Calculational Manual 4.
Low Steam Generator Pressure a.
Steam Line Isolation
> 500 psia (2) b.
Aur.iliary Fee'dwater Actuation E466.7 psia 5.
SIRW Low level Switches Recirculation Actuation 16 inches +0, -2 in, above tank bottom 6.
4.16 KV Emergency Bus low a.
Loss of Voltage (2995.2 & 104) volts 20
- Trip Voltage 5.9(4y
.8 1
seconds b.
Degraded Voltage
> 3825.52 volts i) Bus IA3 Side T4.8 i.5) seconds
> Trip
g.
3 n
TABLE 2-1 (Continued) x P
Engineered Safety Features System Initiation Instrument Setting Limits b
Functional Unit Channel Setting Limit eo 6.
(Continued) b.
(Continued) ii) Bus IA4 Side
> 3724'.08 volts
\\~I"i P T4.8 i.5) seconds 7.
Low Steam Generator Water Level Auxiliary Feedwater Actuation
> 28.2% of wide range tap span 8.
High Steam Generatcr Delta a
Pressure Auxiliary Feedwater Actuation
< 119.7 psid u
~
5 a.
(1) 11ay be bypassed below 1700 psia and is automatically reinstated above 1700 psia.
(2)Maybebypassedbelow550psiaandisautomaticallyreinstatedabove550 psia.
((3) Simultaneous high containment pressure and pressurizer low / low pressure
- 4) Applicable for bus voltage < 2995.2 - 20.8 volts only.
(For voltage > (2995.2 - 20.8) volts, time j
delay shall be > 5.9 seconds.)
TABLE 3-3 MINIMUM FREQUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS Surveillance
~
Channel Description Function Fr#quency : -
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.
1
- c. Calibrate R
c.
Physically measured CEDM position used to verif.y 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-tion, out-of-sequence, and overlap monitoring systems.
- c. Calibrate R
c.
Calibrate secondary CEA position indication w
system and CEA interlock alarms.
3.
Area, Process, and
- a. Check D
a.
Normal readings observed and internal test Post-Accident signals used 'to' verify instrument operation.
Radiation Monitors Except Effluent
- b. Test M
b.
Detector exposed to remote operated radiation Radiation Monitors (l) check source or test signal.
R
~
- c. Calibrate R
c.
RM-063L, M, and H and RM-064 - One time factory I
calibration is acceptable provided linearity a
solid sources are used to check the integrity of the detectors.
RM-091A and B - In situ calibra-2 P
tion by electronic signal substitution is accept-E able for all range decades above 10 R/hr.
In situ calibration for at least one decade below
$t 10 R/hr shall be by means of calibrated radiation source. All other monitors - Exposure to known radiation source.
1 I
(1) The surveillance requircments for effluent radiation monitors are described under Specification 3.12.1.
Effluent radiation monitors are:
RM-054A, RM-054B, RM-055, RM-055A, RM-060, RM-061, and RM-062.
RM-050 and RM-051 are considered effluent radiation monitors when monitoring the ventilation stack.
i
m I
l i
~
[~
TABLE 3-4 (Continued)
MINIMUM FREQUENCIES FOR SAMPLING TEST Type of Measurement Sample and.
,is and Analysis Frequent 1.
Reactor Coolant (Continued)
(c)
Cold Shutdown (1) Chloride 1 per 3 days (d) Refueling (1) Chloride I per 3 days Operation (2)BoronConcentration 1 per 3 days 2.
SIRW Tank Boron Concentration 1 per 31 days s
3.
Concentrated Boric Boron Concentration 1 per 31 days l
Acid Tanks
~
4.
SI Tanks E,oron Concentration 1 per 31 days l~ ~
5.
Spe: t Fuel Pool Boron Concentration 1 per 31 days
~ l~
(1) Until the radioactivity of the reactor coolant is restored to < 1 uCi/gm DOSE EQUIVALENT I-131.
(2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
I Amendment No. 28. O' 86 3-19 s
9
~
3.0 SUP.VEILLANCE REQUIREMENTS 3.11 Radiological Environmental Monitoring Program Applicability Applies to radiological monitoring of plant environs.
Objective To' establish a radiological monitoring program adequate to measure changes in the levels of environmental radioactivity due to plant effluents.
Specifications (1) The radiological environmental monitoring program shall be conducted according to Table 3.9.
Additional details of the radiological environmental monitoring program Are in the ODCM.
No changes shall be made to the ODCM which might reduce the effectiveness of the program. Analytical results of this program and deviations from the sampling schedule shall be reported to the Comission pursuant to Specification 5.9.4.b.
(2)
If the level of radioactivity in an environmental sampling medium exceeds the reporting level specified in the ODCM, a non-routine report shall be prepared and submitted to the Comission pursuant to Specification 5.9.4.b.2.
(3) A land use survey shall be conducted once per 24 months between the dates of June 1 and October 1.
This survey shall identify the loca-tion of the nearest milk animal and the nearest residence in each of the 16 cardinal sectors within a distance of five miles. The results of the land use survey shall be submitted to the Commission pursuant to Specification 5.9.4.b.
The survey shall be conducted under the following conditions:
a.
Within a one-mile radius from the plant site, enumeration by door-to-door or equivalent counting technique.
b.
Within a five-mile radius, enumeration by using referenced information from county agricultural agents or other reliable sources.
If it is learned from this survey that milk animals are present at a location which yields a calculated thyroid dose greater than from previously sampled animals, the new location shall be added to the monitoring program. The sampling location having the lowest calcu-lated dose may then be dropped from the monitoring program at the end of the grazing season during which the survey was conducted.
Also, any location (s) from which milk can no longer be obtained may be dropped and replaced if practicaole from the monitoring program and the Commission shall be notified pursuant to Specification 5.9.4.b.
3-64 Amendment No. 28, 75, 86
+-n
3.0 SURVEILLANCE REQUIREMENTS 3.11 Radiological Environmental Monitoring Program (Continued)
(4) Analyses shall be performed on radioactive materials as p)rt of an Interlaboratory Comparison Program that has been approved by the NRC. The results of these analyses shall be included in the Annual Radiological Environmental Operating Report.
Basis The radiological environmental monitoring program required by this
- specification provides measurements of radiation and of radioactive c.aterials in those exposure pathways und for those radionuclides which lead to the highest potential radiation exposures of individuals attributable to the operation'of Fort Calhoun Station.
The specification for land use survey s provided to ensure that changes in the use of unrestricted areas are i entified and that modifications to the monitoring program are made if equired by the results of this census. This census satisfies the requirements of Section IV.B'.3 of Appendix I to 10 CFR Part 50.
The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material. in environ-mental media are performed in order to demonstrate the validity of resul ts.
3-65 Amendinent No. 86 9
TABLE 3-9 Radiological Environmental Monitoring Program Exposure Pathway and/or Sample Collection Sitel Types of Analysis 2 Frequency
- 1. Direct
- a. Ten TLD indicator stat! ions Gamma Isotopic Quarterly Radiation One (1) control station, total of eleven (11).
- b. An inner-ring of sixteen Gama dose during Site Replaced Annually-(16) statioris, one -in each Area and General Emergen-
'metebrological sect 6r in the cies only.
general area of the site boundary and within 2.5 miles.3
- c. An outer-ring of sixteen (16)
Gamma dose during Site Replaced Annually i'
stations, one in each meteoro-Area and General Emergen-8 logical sector located outside cies only.
of the inner ring but no more distant than approximately 5 miles.J
- 2. Air Monitoring
- a. Indicator Stations
- 1. Filter for Gross Beta 4
- 1. Weekly p
,1. Three (3) Stations in the
- 2. Charcoal for I-131
- 2. Weekly
{
{ral area of the Site
- 3. Filter for Gamma Isotopic
- 3. Quarterly Com-s posite of g
- 2. City of Blair weekly filters g
3
- b. One (1) background station
- 3. Water
- 1. Monthly compos-s downstream drinking water ite for Gamma tiltake..
Isotopic Anals R
isis
- b. Missourt River downstream
- 2. Quarterly Com-'
near the mi.xi.ng zone.
posite for H-3 Analysis
- c. ijtssourt Rtyer upstream of plantintake-(background).
' TABLE 3-9(Continu:d)
Radiological Environmental Monitoring Program (Continued)
Exposure Pathway and/or Sample Collection Sitel Types of Analysis 2 Frequency
- 4. Milk 5
- a. Nearest family cow when avail-Gamma Isotopic and I-131
- 1. Semimonthly able, or one (1) Dairy farm grazing season within 8 kilometers.
(May to October)
- b. One (1) Dairy farm between 8 kilometers and 30 kilo-meters (background).
- 5. Fish
- a. Four fish samples within Gama Isotopic Once per season vicinity of plant discharge (May to October)
- b. One (1) background sample upstream of plant discharge.
O
- 6. Sediment One sample from downstream area Gama Isotopic Semi-annually on the Station side of the Missouri River.
I See Table 10 of the ODCM.
2 The lower limit of detection (LLD) for analysis is' defined in the ODCM in accordance with the wording of NUREG-0472, Rev. 2.
E 3
g Details of the Emergency TLD stations are contained in Emergency Preparedness Implementing Procedures.
4 k
When a gross beta count indicates radioactivity greater than 1E-12 pCi/ml or 1 pCi/md :a gama spectral g
analysis will be performed.
5 p
When milk samples are not available, a broad leaf Vegetation sample shall be collected monthly when available.
.E
\\
TABLE 3-10 DELETED 1
3.,68 Amendment No. 86 e
" 3.12 Radiological Waste Sampling and Monitoring 3.12.1 Liquid and Gaseous Effluents Applicability Applies to the sampling, monitoring; and testing used for liquid and gaseous effluents.
The specified frequencies may be ad,iusted to accommodate operation schedules except that variance should not exceed 1.25 times the specified interval.
Objective To ensure that radioactive liquid and gaseous releases from the 1 facility are' maintained as low as reasonably achievable and within the limits specified by Specification 2.9.l(1) and 2.9.l(2).
Specifications
~
(1) Liquid Effluents a'.
Radioactive liquid waste sampling and activity analyses shall be performed in accordance with Table 3-11.
The results of these analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is limited to the values in Specification 2.9.1(1)a.
b.
Prior to release of each batch of liquid effluent, the batch shall be mixed, sampled, and analyzed for principal. gamma emitters.
When operational or other limitations preclude-specific ganna-radionuclide analysis of each batch, gross radioactivity measurements shall be made to estimate the quantity and concentrations of radioactive materials released
~
in the. batch, and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for the principal gamma-emitting radionuclides.
c.
The overboard header radiation monitor shall ha a:
(i)
Source check prior to any release of radioact ve materials from the monitor or the hotel waste nks.
(ii)
Quarterly channel functional test.
(iii) Channel calibration at "R" frequency (every 18 months).
d.
The steam generator blowdown radiation monitors shall have:
(1)
Daily channel checks.
(ii)
Monthly source checks.
3-69 Amendment No. 28, 86
3.0 SURVEILLANCE REQUIREMENTS 3.12 Radiological Waste Sampling and Monitorina (Continued) 3.12.1 Liquid and Gaseous Effluents (Continued)
(iii)
Quarterly channel functional tests.
(iv)
Channel calibratio'n at "R" frequency (every 18 months),
e.
The steam generator blowdown effluent flow rate will be calibrated at "R" frequency (every 18 months) and visually.
determined operable daily.
f.'
Records shall be maintained of the radioactive concentrations and volume before dilution of each batch of liquid effluent released and of the average _ dilution flow and length of time over shich each discharge occurred. Analytical results shall be submitted to the Conunission in accordance with Section 5.9.4.a of these specifications.
(2) Gaseous Effluents a.
Radioactive gaseous waste sampling and activity analyses shall be performed in accordance with Table 3-12. The results of these analyses shall be used with the calculational methods in the ODCM to assure that the concentration of radioactive materials in unrestricted areas is limited to.the values in Specification 2.9.1(2)a.
b.
(1)
A ventilation stack radiation monitor.shall have a source check prior to any release of radioactive materials from a gas decay tank or the containment. A monthly source check will be perfonned during refueling outages if a purge or gas decay tank release is not done during that
" month.
(ii)
Each ventilation stack monitor shall have a quarterly channel functional test.
(iii)
Each ventilation stack monitor shall be calibrated at "R" frequency (every 18 months).
(iv)
The ventilation stack flow rate will be calibrated and functionally tested at "R" frequency (every 18 months).
The stack radiation monitor flow rate will be calibrated and functionally tested at "R" frequency (every 18 months).
Both will be determined operable by visual inspection daily.
c.
The condenser air ejector monitor shall have a:
(1)
Daily channel check.
(ii)
Monthly source check.
3-70 Amendment No. 86
I I
3.0 SURVE1L! ANCE REQUIREMENTS 3.12 3adiological Waste Sampling and Monitoring (Continued)
}
3.12.1 Liquid and Gaseous Effluents (Continued) i (iii) Quarterly channel functional test.
l (iv)
Channel calibration at "R" frequency (every 18 months).
d.
The hydrogen and oxygen monitoring system for the gas decay l
tanks shall have a:
(i)
Daily channel check.
I (ii)
Monthly cross comparison with a grab sample.
(iii) Quarterly channel calibration using gas mixtures with concentrations in the range of interest.
e.
Records shall be maintained and reports of the sampling and results of analyses shall be submitted to the Comission in accordance with Section 5.9.4.a of these specifications.
Basis The surveillance requirements given under Specification 3.12.1(2) provide assurance that radioactive gaseous effluents from the station are properly 4
controlled and monitored over the life of the station in conformance with the requirements of General Design Criteria 60 and 64 of 10 CFR Part 50, 3
Appendix A.
These surveillance requirements provide the data for the licensee and the Comission to evaluate the perfomance of the station relative'to radioactive gaseous wastes released to the environment. The existing minimum sensitivity of airborne effluent monitor RM-062 is SE-06 mC1/cc/100 cpm and this minimum sensitivity shall be maintained if the monitor is replaced.
Reports on the quantities of the radioactive
. _ _.J -
materials relea. sed in gaseous effluents shall be furnished to the Comis-sion on the basis of Section 5.9.4.a of these Technical Specifications.
On.the bas.ts,of such. reports and any additional information the Comission j
may oTsain from the licensee or others, the Comission may from time to'
~
time require the licensee to take such action as the Comission deems l
appropriate.
l The surveillance requirements given under Specification 3.12.1(1) provide assurance that liquid wastes are properly controlled and monitored in conformance with the requirements of General Design Criteria 60 and 64 of 10 CFR Part 50, Appendix A, during any planned relesse of radioactive materials in liquid effluents. These surveillance requirements provide j
the data for the licensee and the Comission to evaluate the station's performance relative to radioactive liquid wastes released to the environ-ment.
Reports on the quantities of radioactive materials released in liquid effluents shall be furnished to the Commission on the basis of Section 5.9.4.a of these Technical Specifications. On the basis of such l
reports and any additional information the Commission may obtain from the
)
licensee or others, the Commission may from time to time require the licensee to take such action as the Comission deems appropriate.
3-71 Amendment No. 86
~
O 3.0 SURVEILLANCE REQUIREMENTS 3.12 Radiological Waste Sampling and Monitoring (Continued) 3.12.1 Liquid and Gaseous Effluents (Continued)
II) 3.12.2 Solid Radioactive Waste II)
Applicability II)
Objective II)
Specification II)
Basis (1) The surveillance requirements for this se'ction will be incorporated under a supplement to this F.acility License Change after the Process Control Program
~
(PCP)hasbeenissued.
i 3-71a Amendment No. 86 4
TABLE 3-11 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS 1
A.
Monitor & Hotel Waste Tanks Releases Lower Limit of Type of.
Detection (LLD) aampling Frequency Activity Analysis (4)
(vCi/ml)
Principal Gamma Emitters (2)(5) 5.0 E-07
~
Each Batch I-131(2) 1.0 E-06 t
Dissolved.NobleGases(2) 1.0 E-05 Monthly'From One Batch (Gamma Emitters)
II) 1.0 E-05 H-3 Monthly Composite Gross a 1.0 E-07 Quarterly Composite (1)
SR-89, Sr-90 5.0 E-08 B.
Steam Generator Blowdown Lower Limit of Type of Detection (LLD)
Sampling Frequency Activity Analysis-(4)
(vCi/ml)
WeeklyComposite(I)
Principal Gamma Emitters (5) 5.0 E-07 I-131(6) 1.0 E-06 l
Weekly (3)(7)
Dose Equivalent I-131 1.0 E-06 (Gamma Emitters) i
>bnthly Dissolved Noble Gases 1.0 E-05 II) 1.0 E-05 H-3 Monthly Composite I
Gross a 1.0 E-07 II)
Quarterly Composite Sr-89, Sr-90 5.0 E-08 NOTES:
i (1) To be representative of the average quantities and concentrations of l
radioactive materials in liquid effluents, samples should be collected in proportion to the rate of flow of the effluent stream.
Prior to analyses, i
all samples taken for the composite should be mixed in order for the composite sample to be representative of the average effluent release.
i l
r I.
i 3-72 Amen'dment No.18,86
TABLE 3-11 RADI0ACTIVELIQUIDWASTESAMPLINGANDANALYSIS(Continued)
NOTES:
(2) Or gross radioactivity as described in Specification 3.12.1(1)b.
(3) When steam generator iodine activity exceeds 50 percent of limits in Specification 2.20, the sampling and analysis frequency shall be increased to a ~ minimum of five times per week. When the steam generator iodine activity exceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a minimum of once per day.
~(4) The lower limit of detection (LLD) is defined in the ODCM based on NUREG 0472, Rev. 3.
(5) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:
Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144.
~
(6) A weekly grab sample and analyses program including gamma isotopic identification will be initiated for the turbine building sump effluent when the steam generator blowdown water composite analysis indicates the
~
I-131 concentration is greater than 1.0 E-06 microcurie / milliliter.
(7) 1 per 7 days.
i I
3-73 Amendment No. 25,86 s
4 0
TABLE 3-12
~
RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS Lower. Limit of Sampling and Type of Activity Detection (LLD)
Gaseous Source Analysis Frequency Analysis (4)
(uti/ml)
A. Gas Decay Prior Principal Gamma (5) 1.0E-04(I)
Tank Releases to each release Emitters B. Containment Prior Principal Gama(5) 1.0E-04(1)
Purge Releases to each release -Emitters or Containment Prior Pressure Relief to each release H-3 1.0 E-06 Line Releases Monthly (3)
Tritium (H-3) 1.0 E-06 C. Condenser Air Monthly Principal Gamma (5) 1.0E-04(1)-
Ejector Releases Emitters
-(2)
Weekly)(Charcoal I-131 1.0 E-12 D. Continuous Sample Stack Releases Weekly (2)
Principal Gama(5)
.1.0 E-11 (Particulates)
Emitter.s I-131
& Particulates with half-lives greater than 8 days Monthly Composite Gross a 1.0 E-ll Quarterly Composite Sr-89, Sr-90 1.0 E-Il (Particulates) 4 NOTES:
(1)
For certain mixtures of gamma emitters, it may not be possible to measure radionuclides at levels near their sensitivity limits when other nuclides are present in the sample at much higher levels.
Under these circum-stances, it will be more appropriate to calculate the levels of such l
radionuclides using observed ratios with those radionuclides which are measurable.
1 i
l l
3-74 Amendment No. 86 9
- ~.-,--- -
TABLE 3-12 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS (Continued)
NOTES:
(2) To be representative of the average quantities and concentrations of radioactive materials in particulate form released in gaseous effluents, sample should be collected in proportion to the design flow rate of the effluent stream and the design flow rate will be used in estimating releases.
.(3) Required only when steam generator blowdown radioactivity for tritium (Table 3-11, Section B) exceeds 3.0E-03 microcurie / milliliter.
(4) The Lower Limit of Detection (LLD) is defined in the ODCM based on NUREG 0472, Rev. 3.
(5) The princ.ipal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:
Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, Cs-134, Cs-137, Ce-141, Ce-144 for particulate emissions.
f 3-75 Amen &sent No. 86 9
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n 5.0 ADMINISTRATIVE CONTROLS i
5.5.2.8 e.
The Fort Calhoun Station Emergency Plan and implementing procedures at least once every twelve months.
f.
The Site Security Plan and implementing procedures at least once every twelve months.
g.
The Safeguards Contingency Plan and implementing procedures at least once every twelve months.
h.
The Radiological Effluent Program including the Radiological Environ-mental Monitoring Program and the results thereof, the Offsite Dose Calculation Manual and implementing procedures, and the Process Control Program for the solidification of radioactive wastes at least once per 2 years.
i.
fny other area of facility operation considered appropriate by the l
Safcty Audit and Review Committee or the Assistant General Manager -
Nuclear Production, Production Operations, Fuels, and Quality Assurance & Regulatory Affairs.
Author'ty i
5.5.2.9 The Safety Audit and Review Comittee shall report to and advise the Assistant General Manager - Nuclear Production. Production _ Operations,. -
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Fuels, and Quality Assurance & Regulatory Affairs on those areas of responsibility specified in Sections 5.5.2.7 and 5.5.2.8.
Records 5.5.2.10 Records of Safety Audit and Review Comittee activities shall be prepared,
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approved and distributed as indicated below:
Minutes of:each Safety Audit and Review Comittee meeting shall be a.
f prepared, approved and forwarded to the Assistant General Manager -
Nuclear Production, Production Operations, Fuels, and Quality Assur-ance & Regulatory Affairs within 14 days following each meeting.
b.
Reports of reviews encompassed by Section 5.5.2.7e. f, g, h,'and i l
above shall be prepared, approved and forwarded to the Assistant General Manager - Nuclear Production, Production Operations, Fuels, and Quality Assurance & Regulatory Affairs within 14 days following completion of the review.
c.
Audit reports encompassed b/ Section 5.5.2.8 above shall be forwarded to the Assistant General Manager - Nuclear Production, Production l
Operations, Fuels, and Quality Assurance & Regulatory Affairs and to i
the responsible management positions designated by the Safety Audit
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and Review Committee within 30 days after completion of the audit.
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5-8 Amendment No., p. 79,38,69,7fa,54,86 l
5.0 ADMINISTRATIVE CONTROLS 5.5.3 Fire Protection Inspection a.
An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site licensee personnel or an 'outside fire protection firm. The audit and inspection program responsibility shall rest with the Safety Audit and Review Committee.
b.
An' inspection and audit of the fire protection and loss prevention pr. gram by an outside qualified fire consultant shall be performed o
at intervals no greater than 3 years.
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5-8a Amendment No. 86
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. 5.9 3 Special Reports Special reports shall be submitted to the Regional Administrator of the appropriate NRC Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:
a.
In-service inspection report, reference 3.3.
b.
Tendon surveillance, reference 3.5.
c.
Containment structural tests, refa ence 3.5.
d.
Special maintenance reports.
e.
Containment leak rate tests, reference 3.5.
f.
Radioactive effluent releases, reference 2.9.
g.
Materials radiation surveillance specimens reports, reference.3.3.
h.
Fuel performance following each refueling outage.
i.
Fire protection equipment outage, reference 2.19.
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5.9.4 Unique Reporting Requirements a.
Radioactive Effluent Release Report A report covering the operation of the Fort Calhoun Station during the previous six months shall be submitted within 60 days after January 1 and July 1 of each yea'r.
The radioactive effluent release report shall include a sunmary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant as outlined in Regulatory Guide 1.21, Revision 1.
The radioactive effluent release report shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter as outlined in Regulatory Guide 1.21, Revision 1.
The radioactive effluent release report shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in l
Regulatory Guide 1.21, Revision 1.
In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shs11 be evaluated.
The meteorological conditions concurrent with tne 5-15 Amendment No. 9, 24, SS, A6, 86 6
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. 5.9.4 Unique Reporting Requirements (Continued) a.
Radioactive Effluent Release Report (Continued) releases of effluents shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be per-formed in accordance with the 'Offsite Dose Calculation Manual (ODCM).
The radioactive effluent release report shall include any changes (l) to the Process Control Program (PCP) or to the Offsite Dose Calculation Manual (0DCM) made during the reporting period.
A level of detail comensurate to the significance of the change will be provided.
b.
Radiological Environmental Operating Reports 1.
Annual Report An annual report containing the data taken in the radiological environmental monitoring program, in accordance with.the ODCM, for the previous calendar year of operation shall be submitted prior to May 1 of each year. The content of the report shall include:
(a) Sumarized and tabulated results of the. radiological environmental surveillance activities following the format of Regulatory Guide 4.8, Table 1.
In the event that some results are not available, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
(b) : Interpretations and statistical evaluation of the results, including an assessment of the observed impacts of the plant operation on the environment.
(c) The results of participation in the Interlaboratory Comparison Program.
(d) The results of land use survey required by Specifica-tion 3.11(3).
(1) These changes can be initiated either by the licensee (implementa-tion:
subject to review by the PRC) or by the Commission (imple-mentation:
subject to their applicability to the Fort Calhoun Station design, review by the PRC and followed by a review by the SARC).
5-16 Amendment No. 86
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5.9'4 Unique Reporting Requirements (Continued)
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b.
Radiological Environmental Operatino Reports (Continued) 2.
Non-Routine Report If a confirmed measured radionuclide concentration in an environmental sampling medium average over any calendar quarter sampling period exceeds the reporting level referenced in Table 3-9, Footnote 2, and if the radio-activity is attributable to plant operation, a written report shall be submitted to the Commission within 30 days from the end of the quarter.
The report shall include an evaluation of any release conditions, environmental factors, or other aspects
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necessary to explain the anomalous result.
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5-17 Amendment No. 9, JS, 86
5.0 ADMINISTRATIVE CONTROLS 5.10.2 The following records shall tne retained for the duration of the Facility Operating License:
a.
Records of drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Anal-ysis Report.
- b.. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
c.
Records of facility radiation and contamination surveys.
d.
Records of radiation exposure.for all individuals entering radia-tion control areas.
e.
Records of gaseous and liquid radioactive material released to the environs.
f.
Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.
g.
Records of training and qualification for current members of the plant staff.
h.
Records of in-service inspections performed pursuant to these Technical Specifications.
i.
Records of Quality Assurance activities required by the QA Manual, j.
Records of reviews performed for changes made to pr'ocedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k.
Records of meetings of the Plant Review Connittee and the Safety Audit and Review Committee.
1.
Records of Environmental Qualification which are covered under the provisions of Section 5.12 of these Technical Specifications.
Recordsof5heservicelivesofallhydraulicandmechanical m.
snubbers listed on Table 2-6(a) and (b) including the date at which the service life commences and associated installation and maintenance records.
n.
Records of analyses required by the Radiological Environmental Monitoring Program.
5.11 Radiation Protection Program
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Procedures for personnel radiation protection shall be prepared consis-tent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radia-tion exposure.
5-19 Ottif $4t6610/2&lSD, Amendment No. 59, 86 O
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