ML20084J238
| ML20084J238 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/13/1984 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20084J234 | List: |
| References | |
| NUDOCS 8405090045 | |
| Download: ML20084J238 (117) | |
Text
{{#Wiki_filter:ENCLOSURE I TO SCANR48498 REVISIONS TO ANO-1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS i i I i i l l
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TABLE OF CONTENTS SECTION TITLE PAGE ( 1. DEFINITIONS 1.1 RATED POWER 1
- 1. 2 1
- 1. 3 REACTOR OPERATING CONDITIONS 1 OPERABLE
- 1. 4 2 1.5 PROTECTION INSTRUMENTATION LOGIC 2
- 1. 6 INSTRUMENTATION SURVEILLANCE 3 POWER DISTRIBUTION 1.7 REACTOR BUILDING 4
1.8 5
- 1. 9 FIRE SUPPRESSION WATER SYSTEM 5 STAGGERED TEST BASIS 1.10 5 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS) DEFINITIONS
- 2. Sa 2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 7 SAFETY LIMITS, REACTOR CORE 2.2 SAFETY LIMITS, REACTOR SYSTEM PRESSURE 7
2.3 10 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION
- 3. 11 3.1 LIMITING CONDITIONS FOR OPERATION 16 REACTOR COOLANT SYSTEM 3.1.1 Operational Components 16 3.1.2 Pressurization, Heatup and Cooldown Limitations 16 3.1.3 Minimum Conditions for Criticality 18 3.1.4 Reactor Coolant System Activity 21 3.1.5 Chemistry 23 3.1.6 Leakage 25
( 3.1.7 3.1.8 3.1.9 Moderator Temoerature Coefficient of Reactivity Low Power Physics Testing Restrictions 27 30 31 Control Rod Ooeration 3.2 32 3.3 MAKEUP AND CHEMICAL ADDITION SYSTEMS 34 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING, 3.4 AND REACTOR BUILDING SPRAY SYSTEMS 36 3.5 STEAM AND POWER CONVERSION SYSTEM 40 INSTRUMENTATION SYSTEMS 3.5.1 Operational Safety Instrumentation 42 3.5.2 Control Rod Group and Power Distribution Limits 42 3.5.3 46 Safety Features Actuation System Setpoints 3.5.4 Incore Instrumentation 49 3.5.5 Fire Detection Instrumentation 51 3.5.6 Radioactive Liquid Effluent Instrumentation 53d 3.5.7 53f 3.6 Radioactive Gaseous Effluent Instrumentation 531 REACTOR BUILDING 3.7 AUXILIARY ELECTRICAL SYSTEMS 54 3.8 56 FUEL LOADING AND REFUELING 3.9 58 CONTROL ROOM EMERGENCY AIR CONDITIONING AND ISOLATION SYSTEM 3.10 60 SECONDARY SYSTEM ACTIVITY 3.11 66 EMERGENCY COOLING POND 3.12 66a 3.13 MISCELLANEOUS RADI0 ACTIVE MATERIALS SOURCES 66b PENETRATION ROOM VENTILATION SYSTEM 66c L
SECTION TITLE
.,, g 3.14 HYDROGEN PURGE SYSTEM 3.15 66e
[ 3.16 FUEL HANDLING AREA VEllTILATION SYSTEM 66g ' SHOCK SUPPRESSORS (SNUBBERS) 3.17 66i FIRE SUPPRESSION WATER SYSTEM 3.18 66m 3.19 FIRE SUPPRESSION SPRINKLER SYSTEMS 66n CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS 3.20 FIRE HOSE STATIONS 660 3.21 66p PENETRATION FIRE BARRIERS 3.22 66q 3.23 REACTOR BUILDING PURGE FILTRATION SYSTEM 66r REACTOR BUILDING PURGE VALVES 66t 3.24 3.25 EXPLOSIVE GAS MIXTURE (RESERVED) 66u RADIOACTIVE EFFLUENTS 3.25.1 Radioactive Liquid Effluents 66v 3.25.1.1 Concentration 66v 3.25.1.2 Dose 66v 3.25.1.3 Waste Treatment 66w 3.25.1.4 Liquid Holdup Tanks 66x 3.25.2 66y 3.25.2.1 Radioactive Gaseous Effluents 66z Dose Rate 3.25.2.2 Dose - Noble Gases 66z 3.25.2.3 Dose - Iodine-131, Tritium, and Radionuclides 66aa 3.25.2.4 in Particulate Form 66bb Gaseous Radwaste Treatment 3.25.2.5 Gas Storage Tanks 66cc 3.25.3 Total Dose 66dd
- 4. SURVEILLANCE STANDARDS 66ee
( 4.1 OPERATIONAL SAFETY ITEMS 67 ( 4.2 67 4.3 REACTOR COOLANT SYSTEM SURVEILLANCE 76 4.4 TESTING FOLLOWING OPENING OF SYSTEM 78 REACTOR BUILDING 4.4.1 Reactor Building Leakage Tests 79 4.4.2 Structural Integrity 79
- 4. 5 85 EMERGENCY CORE COOLING SYSTEM AND REACTOR 4.5.1 BUILDING COOLING SYSTEM PERIODIC TESTING 92 Emergency Core Coolino Systems 4.5.2 Reactor Building Cooling Systems 92 4.6 95 AUXILIARY ELECTRICAL SYSTEM TESTS 100 4.7 REACTOR CONTROL R0D SYSTEM TESTS 4.7.1 Control Rod Drive System Functional Tests 102 4.7.2 Control Rod Program Verification 102 4.8 104 EMERGENCY FEEDWATER PUMP TESTING 105 4.9 REACTIVITY ANOMALIES 4.10 106 CONTROL ROOM EMERGENCY AIR CONDITIONING AND ISOLATION SYSTEM SURVEILLANCE 107 4.11 4.12 PENETRATION ROOM VENTILATION SYSTEM SURVEILLANCE 109 4.13 HYDROGEN PURGE SYSTEM SURVEILLANCE 109b EMERGENCY COOLING POND 4.14 110a 4.15 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE ~110b AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT 110c
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SECTION TITLE PAGE 4.16 SHOCK SUPPRESSORS (SNUBBERS) 110e 4.16.1 Hydraulic Shock Suppressors 110e C 4.17 FUEL HANDLING AREA VENTILATION SYSTEM SURVEILLANCE 110h 4.18 STEAM GENERATOR TUBING SURVEILLANCE 110j 4.19 FIRE DETECTION INSTRUMENTATION 110p a 4.20 FIRE SUPPRESSION WATER SYSTEM 110q 4.21 SPRINKLER SYSTEMS 110t 4.22 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS 110u 4.23 FIRE HOSE STATIONS 110v 4.24 PENETRATION FIRE BARRIERS 110w 4.25 REACTOR BUILDING PURGE FILTRATION SYSTEM 110x 4.26 REACTOR BUILDING PURGE VALVES 110z 4.27 DECAY HEAT REMOVAL 110aa 4.28 EXPLOSIVE GAS MIXTURE (RESERVED) 110bb 4.29 RADI0 ACTIVE EFFLUENTS 110cc 4.29.1 Radioactive Liquid Effluents 110cc 4.29.1.1 Concentration 110cc 4.29.1.2 Liquid Holdup Tanks 110gg 4.29.1.3 Liquid Radioactive Effluent Instrumentation 110hh 4.29.2 Radioactive Gaseous Effluents 110jj 4.29.2.1 Dose Rate 110jj 4.29.2.2 Gas Storage Tanks 110mm 4.29.2.3 Radioactive Gaseous Effluent Monitoring Instrumentation 110nn 4.29.3 Dose Calculations for Radioactive Effluents 110rr 4.30 RADIOLOGICAL ENVIRONMENTAL MONITORING 110ss 4.30.1 Radiological Environmental Monitoring Program { Description 110ss 4.30.2 Land Use Census 110zz 4.30.3 Interlaboratory Comparison Procram 110bbb
- 5. DESIGN FEATURES 111 5.1 SITE 111 5.2 REACTOR BUILDING 112 5.3 REACTOR 114 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 116
- 6. ADMINISTRATIVE CONTROLS 117 6.1 RESPONSIBILITY 117 6.2 ORGANIZATION 117 6.3 FACILITY STAFF QUALIFICATIONS 117 6.4 TRAINING 117 6.5 REVIEW AND AUDIT 117 6.6 REPORTABLE OCCURRENCE ACTION 126
- 6. 7 SAFETY LIMIT VIOLATION 126 6.8 PROCEDURES 127 6.9 RECORD RETENTION 128 6.10 RADIATION PROTECTION PROGRAM 129 6.11 HIGH RADIATION AREA 129 6.12 REPORTING REQUIREMENTS 140 6.13 ENVIRONMENTAL QUALIFICATION 147 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 148 l
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1.10 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS), DEFINITIONS 1.10.1 Dose Equivalent I-131 f ( The Dose Equivalent I-131 shall be the concentration of I-131 (microcurie / gram) 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. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." 1.10.2 Source Check A Source Check shall be the qualitative ar,sessment of channel response when the channel sensor is exposed to a radioactive source. 1.10.3 Offsite Dose Calculation Manual (ODCM) The Offsite Oose Calculation Manual shall contain the methodology . and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. 1.10.4 Liquid Radwaste Treatment System A Liquid Radwaste Treatment System is a system designed and used i for holdup, filtration, and/or demineralization of radioactive ! { liquid effluents prior to their release to the environment. { 1.10.5 Gaseous Radwaste Treatment System A Gaseous Radwaste Treatment System is any system designed and installed to reduce radioactive gaseous effluents by collecting gases from radioactive systems and providing for decay or holdup for the purpose of reducing the total radioactivity prior to re-lease to the environment. i 1.10.6 Ventilation Exhaust Treatment System 9 A Ventilation Exhaust Treatment System is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Fea-ture (ESF) atmospheric cleanup systems are not considered to be Ventilation Exhaust Treatment Systems. 1.10.7 Purge - Purging Purge or Purging is the controlled process of discharging air or (- gas from a confinement to reduce the airborne radioactivity con-centration in such a manner that replacement air or gas is required to purify the confinement. Sa
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1.10.8 Member (s) of the Public Member (s) of the Public shall include all persons who are not occupationally associated with the plant. This category does not C include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does in-clude persons who use portions of the site for recreational, occupational or other purposes not associated with the plant. 1.10.9 Exclusion Area The exclusion area is that area surrounding ANO within a minimum radius of .65 miles of the reactor buildings and controlled to the extent necessary by the licensee for purposes of protection of individuals from exposure to radiation and radioactive ma-terials. 1.10.10 Unrestricted Area An unrestricted area shall be any area beyond the exclusion area boundary. t Sb
3.5.6 Radioactive Liquid Effluent Instrumentation Applicability: During releases via this pathway. ( Objective: To provide instrumentation for radioactive liquid releases. Specification: 3.5.6.1 The radioactive liquid effluent monitoring instrumentation shown in Table 3.5.6-1 shall be operable with their alarm / trip setpoints set to ensure that the limits of specification 3.25.1.1 are not exceeded. 3.5.6.2 With alarm / trip setpoints less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, until the setpoint is changed to an acceptably conservative value. 3.5.6.3 With less than the minimum number of channels operable, take the action shown in Table 3.5.6-1. Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Semiannual Radio-active Effluent Release Report why the inoperability was not corrected. 3.5.6.4 Specifications 3.0.3, 3.0.4, and 6.12.3 are not applicable. Bases: The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with the methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. L 53f
p O O Table 3.5.6-1 Radioactive Liquid Monitoring Instrumentation i Minimum Operable Instrument Channels Applicability Action
- 1. Liquid radwaste effluent monitor (automatic termination) 1 During releases via this pathway A
- 2. Liquid radwaste effluent flow monitor 1 During releases via this pathway B i
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Table 3.5.6-1 (Cont.inued) , , , Jable Notation ( Action A. Descrintion With the number of channels operable less than required, effluent releases may be resumed provided that prior to initiating a release: 1. At least two independent samples of the tank's contents are analyzed in accordance with Specification 4.29.1.1; 2. At least two technically qualified members of the facility staf f independently verify that the computer input data is correct and; 3. At least 2 members of the facility staff independently verify the discharge valve lineup. Otherwise, suspend release of radioactive effluents via this pathway. B. With the number of channels operable less than required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump curves may be used to estimate flow. 53h
3.5.7 Radioactive Gaseous Effluent Instrumentation .,, Applicability: As shown in Table 3.5.7-1. h Objective: To provide instrumentation for radioactive gaseous releases. Soecification: 3.5.7.1 The radioactive gaseous effluent monitoring instrumentation shown in Table 3.5.7-1 shall be operable with their alarm / trip setpoints set to ensure that the limits of Specification 3.25.2.1 are not exceeded. 3.5.7.2 With a channel alarm / trip setpoint less conservative than required, declare the channel inoperable. 3.5.7.3 With less than the minimum number of channels operable, take the action shown in Table 3.5.7-1. Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected. 3.5.7.4 Specifications 3.0.3, 3.0.4, and 6.12.3 are not applicable. Bases: The radioactive gaseous effluent instrumentation is provided to monitor ( and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.105. L 531 _ - _ _ _ _ _ _ _ _ _ _ _ _ - - - - _ .)
b Table 3.5.7-1 Radioactive Gas Effluent Monitoring Instrumentation Instrument Operable Applicability Parameter Action
- 1. Waste Gas Holdup System Noble gas activity monitor 1 During releases via Radioactivity A (provides alarm and automatic this pathway (DRVTP) termination'of release) ,'
Effluent flow monitor 1 DRVTP System flow B l
- 2. Auxiliary Building Ventilation Sy.ctem a) Noble gas activity 1 DRVTP Radioactivity C monitor .
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b) Iodine sampler 1 DRVIP D c) Particulate sampler 1 DRVTP D d) Effluent flow monitor 1 DRVTP System flow B l e) Sampler flow monitor 1 DRVIP Sample flow B
- 3. Spent Fuel Pool Area When the system is Ventilation-System in operation a) Noble gas activity 1 Radioactivity C monitor '
i b) Iodine sampler 1 D i i......
f O ' Table 3.5.7-1 (Continued) Radioactive Gaseous Ef fluent Monitoring Instrumentation Instrument Operable Applicability Parameter Action c) Particulate sampler 1 D. d) Effluent flow-monitor 1 System flow B e) Sampler flow monitor 1 Sample flow B
- 4. Reactor Building Purge and When the system is Ventilation System in the operation a) floble gas activity' Radioactivity
?
monitor 1 - m C, E N b) Iodine sampler 1 D l 1 c) Particulate sampler 1 0 l d) Effluent flow i monitor 1 System flow l B e) Sampler flow l monitor 1 Sample flow B i - I
l 1 Table 3.5.7-1 (Continued) . Table Notation ( Action Description A. With the number of channels operabl. less than required, the contents of the tank may be rel to the environ-ment provided that prior to initiating t e release:
- 1. At least two independent samples of the tank's contents are analyzed, and
- 2. At least two technically qualified members of the facility staff independently verify the computer input data, and
- 3. At least 2 members of the facility staff inde-pendently verify the correct discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway. B. With the number of channels operable less than required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours. C. With the number of channels operable less than required, effluent releases via this pathway may continue provided ( grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. D.
. With the number of channels operable less than required, i
ef fluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary t sampling equipment as required in Table 4.29-3. E. When purging the reactor building, immediately suspend purging if less than the required number of monitoring channels are operable. Purging may be resumed provided that prior to initiating the purge:
- 1. At least two independent samples of the reactor building atmosphere are analyzed, and
- 2. At least two technically qualified members of the j
facility staff independently verify the computer input data. L 531
3.25 RADI0 ACTIVE EFFLUEllTS 3.25.1 _ Radioactive Liouid Effluents .( 3.25.1.1 Concentration Applicability: At all times Objective: To ensure that the limits of 10 CFR 20 are met. Specifications: 3.25.1.1 A. The concentration of radioactive material released to the discharge canal shall be limited to the concentration specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides noble gases. other than dissolved or entrained For dissolved or entrained noble gases, the total concentration released shall be limited to 2 x 10-4 pCi/ml. B. With the concentration of radioactive material released exceeding the above limits, immediately initiate action to restore concentration to within limits and provide notification to the Commission within 24 hours. In lieu of any other report, prepare and submit a special report within 30 days pursuant to specification 6.12.5. C. Specifications 3.0.3, 3.0.4 and 6.12.3 are not applicable. Bases: This specificatinn is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the c.oncentration levels specified in 10 CFR Part 20, Appendix 8, Table II. This limitation provices additional ase.urante that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than the Section II A dt. sign objectives of Appendix I,10 CFR Part 50, to a member of the public.
'Ihe concentration limit for noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its I1PC in air (submersion) was converteu to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
L 66v
Radioactive Liquid Effluents 3.25.1.2 Dose Applicability: At all times Objective: To ensure that the of 10 CFR 50 Appendix I Section IV A are met. Specifications: 3.25.1.2 A. The dose commitment to a member of the public from radio-active material in liquid effluents released from ANO-1 to the discharge canal shall be:
- 1) During any calendar quarter less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and
- 2) During any calendar year less than or equal to 3 mrem to the total body and less than or equal to 10 mrem to any organ.
- 8. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a special report to the Commission within 30 days, pursuant to specification 6.12.5.
C. The provisions of specifications 3.0.3, 3.0.4, and 6.12.3 are not applicable. Bases: Specification 3.25.1.2 provides assurance that releases of liquid efflu-ents will result in concentrations far below the limits of 10CFR20. The specification provides the required operating flexibility and at the same time assures that the release of radioactive material in liquid effluents will be kept "as low as reasonably achievable". I 66w
Radioactive Liquid Effluents 3.25.1.3 Waste Treatment ( Applicability: At all times Objective: To assure that the amount of radioactive material in liquid effluents will be "as low as reasonably achievable." Specifications: 3.25.1.3 A. The appropriate parts of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid waste prior to their discharge when it is pro-jected that the cumulative dose during a calendar quarter due to liquid effluent releases would exceed 0.18 mrem to the total body or 0.625 mrem to any organ. B. The provisions of this specification do not apply to the laundry tanks due to their incompatability with the rad-waste system. C. With racios.tive liquid waste being disc.harged without treatment and in excess of the above limits, in lieu of any other report, prepare and submit a special report to the Commission within 30 days per specification 6.12.5. D. The provisions of Specifications 3.0.3, 3.0.4 and 6.12.3 are not applicable. Bases: The requirement.s that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable 10 fraction CFR Part 50, for of the guide liquid set forth in Section II A of Appendix I, effluents. The values of 0.18 mrem and 0.625 mrem are approximately 25% of the yearly design objectives on a quarterly basis. Section The II. yearly design objectives are given in 10 CFR 50 Appendix I, r L 66x
Radioactive Liquid Effluents 3.25.1.4 Liquid Holdup Tanks ( Applicability: At all times. Objective: To ensure that the limits of 10 CFR 20 are not exceeded. Specifications: 3.25.1.4 A. The quantity of radioactive material contained in each unprotected
- outside temporary radioactive liquid storage tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.
B. With the quantity of radioactive material exceeding the above limit, immediately suspend all additions of radioactive material to the affected tank and within 48 hours reduce the tank contents to within the limit. C. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. Bases: This specification is provided to ensure that in the event of an uncontrolled release of the contents of the tank
- the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in the unrestricted area.
- Tanks included in this specification are those outdoor temporary tanks that 1) are not surrounded by liners, dikes, or walls capable of holding the tank contents, and 2) do not have overflows and surrounding f
area drains connected to the liquid radwaste treatment system. L 66y
O 3.25.2 Radioactive Gaseous Effluents 3.25.2.1 Dose Rate Applicability: At all times Objective: To ensure that the dose rate in unrestricted areas from gaseous effluents will be within the limits of 10 CFR 20. Specifications: 3.25.2.1 A. The dose' rate in unrestricted areas (see Figure 5.1-1)'due to radioactive materials released in gaseous effluents from the site shall be:
- 1) For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.
- 2) For iodine-131, for tritium and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.
During periods of reactor building purging the dose rate may be averaged over a one hour interval. B. With the dose rate (s) exceeding the above limits, without delay restore the release rate to within the above limit (s) C. Specifications 3.0.3, 3.0.4 and 6.12.3 are not applicable. Bases: This specification is provided to ensure that, at any time, the dose rate due to gaseous effluents from all units on the site will be within the i limits of 10 CFR 20 for unrestricted areas. This specification applies to the release of gaseous effluents from all reactors at the site. I t 66z
I Radioactive Gaseous Effluents . 3.25.2.2 Dose - Noble Gases r~ ( Applicability: At all times Objective: To ensure that the design objective doses of 10 CFR 50 Appendix I Section IV A are not exceeded. Specifications: 3.25.2.2. A. The dose due to noble gases released in gaseous effluents from ANO-1 to unrestricted areas (see Figure 5.1-1) shall be: 1) During any calendar quarter, less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
- 2) During any calendar year, less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta raciation.
B. With the calculated dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a special report to the Commission within 30 days, pursuant to Specification 6.12.5. (, C. The provisions of Specifications 3.0.3, 3.0.4, and 6.12.3 are not applicable. Bases: Specification 3'.75.2.2 implements the design guides specified in 10 CFR 50 Appendix I Section II and the limiting condition for operation as set forth in Section IV A of Appendix I. i The specifications provide the required operating flexibility and at the same time implement the guides set forth in Section IV A Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". These limits provide reasonable assurance that radioactive material dis-charged in gaseous effluents will not result in the exposure of an indi-vidual in an unrestricted area, to annual average concentrations exceed-ing the limits specified in Appendix B, Table II of 10 CFR Part 20 [10 CFR Part 20.106(b)]. For individuals who may at times be within the ex-clasion area boundary, the occupancy of the individual will be suffi-ciently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary. g L 4 66aa
Radioactive Gaseous Effluents 3.25.2.3 Dose - Iodine-131, Tritium, and Radionuclides in Particu-late Form Applicability: At all times Objective: To ensure that the dose limits of 10 CFR 50 Appendix I Section IV A are met. Specifications: 3.25.2.3 A. The dose to a member of the public from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from ANO-1 to unrestricted areas (see Figure 5.1-1) shall be: " 1) During any calendar quarter, less than or equal to 7.5 mrems -to any organ, and
- 2) During any calendar year, less than or equal to 15 mrems to any organ.
B. With the calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half-lives greater thar. 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a special report to the Com-mission within 30 days, pursuant to Specification 6.12.5. C. The provisions of Specifications 3.0.3, 3.0.4, and 6.12.3 are not applicable. Bases: Specification 3.25.2.3 implements the design guides set forth in 10 CFR 50 Appendix I, Section II C and the limiting conditions for operation as set forth in Appendix I, Section IV A. The specifications provide the required operating flexibility and at the same time implement the guides set forth in Section IV A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable". 4 L 66bb
Radioactive Gaseous Effluents 3.25.2.4 Gaseous Radwaste Treatment (_ Applicability: At all times Objective: To assure that the amount of radioactive material in gaseous effluents is "as low as reasonably achievable." Specifications: 3.25.2.4 A. Ventilation exhaust treatment systems shall be used to reduce radioactive materials in gaseous waste prior to discharge when the projected doses due to gaseous ef-fluent releases from ANO-1 to unrestricted areas (see Figure 5.1-1) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation over a calendar quarter; or when the projected doses due to iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days would exceed 1.0 mrem to any organ over a calendar quarter. B. When degasifying the reactor coolant system, the gaseous radwaste treatment system shall be utilized to process the degassing effluent to reduce the concentration of ra-dioactive materials prior to discharge when the projected doses due to gaseous effluent releases from ANO-1 to un-restricted areas (see Figure 5.1-1) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation over a calendar quarter. C. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other re-
, port, prepare and submit to the Commission within 30 days
'a special report, per Specification 6.12.5.
D. The provisions of Specification 3.0.3, 3.0.4 and 6.12.3 are not applicable. Bases: The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radio-active materials in gaseous effluents will be kept "as low as reasonably achievable." The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II B and II C of Appendix I,10 CFR Part 50, for gaseous effluents. The values 0.625 mrad, 1.25 mrad, and 1.0 mrem are approximately 25% of the yearly design objectives on a quarterly basis. The yearly design objectives are given in Specifications 3.25.2.2 and 3.25.2.3. L 66cc
Radioactive Gaseous Effluents 3.25.2.5 Gas Storage Tanks ( Applicability: At all times Objective: To restrict holdup tank.the amount of activity in a radioactive gas Specification 21 . 3.25.2.5 A. The quantity of radioactivity contained in each gas storage tank shall be limited to 300,000 curies noble gases (Xe-133 equivalent). B. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and limit.within 48 hours reduce the tank contents to within the C. The provisions of Specification 3.0.3 and 3.0.4 are not applicable. Bases: The value of 300,000 curies is a suitable fraction of the quantity of radioactive material which if released over a 2 hour period, would result in a total body exposure to a member of the public at the exclusion area ( boundary of 500 mrem. ETSB 11-5 in NUREG-0800, This Julyis1981. consistent with Branch Technical Position L 66dd
RADI0 ACTIVE EFFLUENTS 3.25.3 Total Dose Applicability: At all times Objective: To ensure that the limits of 40 CFR 190 are not exceeded. Specifications: 3.25.3.1 The calculated doses from the release of racioactive materials a in liquid or gaseous effluents shall not exceed twice the limits of Specification 3.25.1.2, 3.25.2.2, or 3.25.2.3. 3.'25.3.2 With the calculated doses exceeding the above limits, prepare and submit a Special Report pursuant to 10CFR Part 20.405C. 3.25.3.3 If the limits of 40CFR190 have been exceeded, obtain a vari-ance from the Commission to permit further releases in excess of 40CFR190 limits. 'A variance is granted until staff action on the request is complete. 3.25.3.4 The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. Bases: This specification is provided to meet the dose limitations of 40 CFR 190 that have now been incorporated into 10 CFR Part 20. The specifi-( cation requires the preparation arid submittal of a Special Report when-ever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public.will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Re-port will describe a course of action that should result in the limita-tion of the annual dose to a member of the public to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contribu-tions from other nuclear fuel cycle facilities within a radius of 8 km must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already keen corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in Specifications 3.25.1 and 3.25.2. An individual is not considered to be a member of the public during any period in which he/she is engaged in carrying out any opera-tion that is part of the nuclear fuel cycle. L 66ee
4.29 RADI0 ACTIVE EFFLUENTS 4.29.1 Radioactive Liquid Effluents 4.29.1.1 Concentration Applicability: At all times Objective: To ensure that the limits of Specification 3.25.1.1 are met. Specifications: 4.29.1.1 -A. Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analyses program of Table 4.29-1. B. The results of the radioactivity analyses shall be used in accordance with the ODCM to assure that the concentra-tions at point of release are maintained within the limits of Specification 3.25.1.1. Bases: This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to
- unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than the Section II A design objectives of Appendix I,10 CFR Part 50, to an individual. The concentration limit for noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
1 a 9 f 110cc
TABLE 4.29-1 RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSES PROGRAM C l -l Sampling i Minimum l Type of I Lower Limit l Liquid Release l Frequency l Analyses l Activity l l Type l of Detection l l l l Frequency l Analyses I (LLD) l l l I l l l (uCi/ml) (,) l I I I I l P l P l yisotopic(*) l A. Batch Waste leach Batch l Each Batchl I 5 x 10-7 (b) ~ l l Release (d) l l l l l l l l l l I-131 1 1 x 10-6 l l P l l l l l l l l l Dissolved and l 1 x 10-5 l l 10ne Batch /M1 M l Entrained l l l l l Gases l l j- l(Gamma Emitters)l l l l l P l 1 l M l H-3 l 1 x 10 3 l l l lEachBatchl'Compg) l l l l l site l Gross Aloha i 1 x 10-7 l l P l Sr-89, Sr-90 l l l 5 x 10 6 l leach Batch l Q l l 1 l Comp l l l Fe-55 l 1 x 10-6 l l l l site [c) l l l l l l I i ! 3
~
t L - 110dd
TABLE 4.29-1 (Continued) , TABLE NOTATION
-( a.
The Lower Limit of Detection (LLD) is the smallest concentrat of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radio chemical separation): t LLD = 4.66 S b E V 2.22 Y exp(-Aat) where LLD is the lower limit of detection as defined above (as pCi per unit mass or volume) s b isthestandarddeviationofthebackgroundcountingrate or of the counting rate of a blank sample (in counts per minute). , E is the counting efficiency (as counts per transformation) V is the sample size (in units of mass or volume) 2.22 is the number of transformations per minute per picocurie ( Y is the fractional radiochemical yield (when applicable) A is the radioactive decay constant for the particular rad,i o,nucl i de ' At is the elapsed time between sample collection (or end of the sample collection period) and time of counting Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the LLD is an a Priori (before the fact) limit representing the capability of a measurement system and not an a Posteriori (af ter the fact) limit for a particular measurement. i a t L { 110ee
TABLE 4.29-1 (Continued) , ( _ TABLE NOTATION ' b. For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations. Under these circumstances, it will be more appropriate to calculate the concentration of such radionuclides using observed ratios with those radionuclides which are measurable, c. A composite to proportional sample is one in which the quantity of liquid sampled is the quantity the method of sampling employeo ic ults in a specimen which i representative of the liquids released. d. A volume.batch release is the discharge of liquid wastes of a discrete Prior to sampling, each batch shall be isolated and mixed to ensure representative sampling. e. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: CoS8, C060, Zn65, Mo99, Cs134, Cs137, Cel41, and Ce144.Mn54, Fe59, does not mean that only these nuclides are to be detected and re-This list ported. with the above nuclides, shall also be identified and r Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circum-stances result in LLD's higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report. D Daily P Prior to Release M Monthly Q Quart'erly R Every 18 months { L V 110ff i
Radioactive Liquid Effluents 4.29.1.2 -Liquid Holdup Tanks '{ Apolicability: At all times. Objective: To ensure that the limits of 10 CFR 20 are not exceeded. Specifications: 4.29.1.2 The quantity of radioactive material contained in an out-side temporary radioactive liquid storage tank shall be determined to be within the liinit of Specification 3.25.1.4 by analyzing a representative sample of tne contents of the tank at least once per 7 days when radioactive materials are being added to the tank. Bases: This specification is provided to ensure that in the event of an uncontrolled release of the contents of the tank the resulting concentrations would-be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest the potable unrestricted water supply and the nearest surface water supply in area. ( b L 110gg t
~
Radioactive Liquid Effluents 4.29.1.3 Liquid Radioactive Effluent Instrumentation Applicability: Applies to the instrumentation in the liquid radwaste , system that is used to limit the amount of radioactivity released to the environs. i Objective: To provide surveillance specifications for the instruments required in Specification 3.5.6. Specifications: 4.29.1.3 Each radioactive liquid effluent monitoring instrumen-tation channel shall be demonstrated operable by performance of the channel check, source check, channel calibration, and channel test at the frequencies shown in Table 4.29-2. Bases: To ensure that the instrumentation for the liquid radwaste system is operable. The channel test demonstrates that automatic isolation of this i pathway and control room alarm annunciation occur if the instrument
- indicates measured levels above the trip setpoint. The channel test l also demonstrates that alarm annunciation occurs if any of the following conditions exist
- 1. Power to the detector is lost.
- 2. The instrument indicates a downscale failure.
- 3. Instrument controls are not set in the operate mode.
The initial channel calibration is performed using one or more of the
! reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance accivities with NBS. These standards permit calibrating the system over its intended range of energy and measurement range. For subsequent channel calibration, sources that have been related i to the initial calibration are used.
4 s-f 4 e h 110hh
p O O Table 4.29-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirement Channel Source Channel Instrument Channel Check Check Calibration Test
- Liquid radwaste i effluent line
! Radiation monitor D* P** R j (automatic termination) Q Flow monitor D* NA R NA Notation DDuring releases via this pathway
[ i 2 D A check check source. source is not required if the background activity is greater than the activity of the I D Daily . P Prior to release
- H Monthly Q Quarterly R Every 18 months i
l i i : t E
RADI0 ACTIVE EFFLUENTS 4.29.2 Radioactive Gaseous Effluents [. ( 4.29.2.1 Dose Rate Applicability: At all times Objective: To ensure that the dose rate, at any time, in unrestricted areas from gaseous effluents will be within tha dose limits of 10 CFR 20.
- Specifications:
4.29.2.1 A. The dose rate, due to noble gases in gaseous effluents shall be determined in accordance with the ODCM to be within the limits of-Specification 3.25.2.1. B. The dose rate in unrestricted areas, due to iodine-131
-tritium, and all radionuclides in particulate form with half-lives greater than 8 days released in gaseous efflu-ents, shall be-determined in accordance with the ODCM to be within the required limits by using the results of the sampling and analyses program, specified in Table 4.29-3.
Bases: l' ( This specification provides for sampling and analyses to ensure that ( Specification 3.25.2.1 is met. L 110jj
f
- O O TABLE 4.29-3 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSES PROGRAM l l l Minimum l Lower Limit of l Gaseous Release Type Sampling Analyses l l l
l l l Type of I Detection, l l Frequency Frequency I Activity Analyses l l l l l(uCi/ml)'QLD) l l l l 1 P l l P l l A. Waste Gas Each Tank l l Storage Tank
! l Each Tank l Principal Gamma Esaitters(b) I 1 x 10 '8 (9) l l l Grab Sample l l l l l l l l I B. Reactor Bldg. l P l P 1 I 1
l Purge i Each Purge i Each Purge ! l l i Principal Gamma Emitters (b) I 1 x 10-4 (9) I l Grab Sample i i li-3 l l I l 1 x 10-6 l l I l l C. Unit Vents l M (c) (d) l M l Principal Gamma Imitters(b) i 1 x 10 '8 (9) = l l Grab Sample Il-3 l o I (Auxiliary Bldg. ) l 1 l l l 1 x 10-6 l l r I (Spent Fuel Pool l l l l W (f) l l l Area Ventilation) l Continuous (") I CIIarcoal I-131 l l l Sample l l 1 x 10-12 l l l l (Rx Bldg. Ventilation) l l l l l l l l l l W (t) l I l l Continuous (") I l l Particulate i Principal Gamma Emitters (b) I 1 x 10-22 l l l Sample 1 (I-131, Others) l l l 1 l 1 I l I 1 I it i i I l l Continuous (") 1 Particulate I Gross Alpha l 1 x 10-11 l l l l Sample I l l l i i l 1 I l l l Q l I I l l Continuous (*) I Composite l Sr-89, Sr-90 l 1 x 10-22 y l I l l l Particulatel I i l l Sample i I l l l i I I l i l l floble Gas l floble Gases Gross l 1 x 10% l l l Continuous (*) 1 Monitor i Beta or Gamma I (Xe-133 equiv.) I l I I I I I
~
TABLE 4.29-3 (Continued) _ TABLE NOTATION ( a. See definition in Table 4.29-1, Table Notation. b. The principal gamma emitters for which the LLD specification will apply are exclusively 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, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semiannual effluent report, c. Tritium grab samples -shall be taken from the Reactor Building venti-lation exhaust at least once per 24 hours when the refueling canal is flooded. d. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel area, whenever spent fuel is in the spent fuel pool. e. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation 3.25.2.2, made in accordance with Specification 3.25.2.1, and 3.25.2.3. f. Samples shall be changed at least once per 7 days at.1 analyses shall be completed within 48 hours af ter changing (or after removal from 1 the sampler). g. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these cir-cumstances, the LLD may be increased inversely proportional to the magnitude of the gamma yeld (i.e., (IE-4/I), where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10*f, of the MPC value specifled in 10 CFR 20, Ap-pendix B, Table II, Column 1. D Daily P Prior to Release W Weekly M Monthly Q Quarterly R Every 18 months 11011
' Radioactive Gaseous Effluents 4.29.2.2 Gas Storage Tanks
-( Applicability: At all times Objective: To ensure meeting the requirements of Specification 3.25.2.5. Specifications: 4.29.2.2 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the limits of Specification 3.25.2.5 at least once per 24 hours when radioactive materials are being added to the tank and the reactor coolant activity exceeds the limits of Speci-fication 3.1.4.1.b. Bases: 1 This specification is provided so that the requirements of Specification 3.25.2.5 are met. L 110mm
Radioactive Gaseous Effluents 4.29.2.3 Radioactive Gaseous Effluent Monitoring Instrumentation Applicability: Applies to the instrumentation in the gaseous radwaste system that is used to limit the amount of activity released to the environs. Objective: To provide surveillance specifications for the instruments listed in Specification 3.5.7. Specifications: 4.29.2.3 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated operable by performance of the channel cneck, source check, channel calibration, and channel test at the frequencies shown in Table 4.29-4. Bases: To ensure that the instrumentation for the gaseous radwaste system is operable. The channel test demonstrates that control room alarm annunciation occurs if any of the following conditions exist:
- 1. The instrument indicates measured levels above the alarm /
trip setpoint.
- 2. Power to the detector is lost. -
- 3. The instrument indicates a downscale failure.
- 4. Instrument controls are not set in the operate mode.
For the waste gas holdup system noble gas activity monitor, the channel test also demonstrates that automatic isolation of the release pathway occurs if the instrument indicates above the trip setpoint. The initial channel calibration is performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards permit calibrating the system over its intended range of energy and measurement range. For subsequent channel calibration, sources that have been related to the initial calibration are used. 110nn
{ O O Table 4.29-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Channel Channel Source ** Channel Functional Instrument Check Check Calibration Test
- 1. Waste Gas Holdup System
- a. Noble Gas Activity monitor (provides automatic termination of release D* P R Q
- b. Effluent Flow Monitor D* N/A R N/A
- 2. Auxiliary Building Ventila-tion System
.-- a. Noble Gas Activity y Monitor D* M R o Q
- b. Ef fluent Flow Monitor D* N/A R N/A
- c. Sampler Flow Monitor D* N/A R N/A
- d. Iodine Sampler Cartridge l W*(1) N/A N/A N/A l
- e. Particulate Sampler !
Filter W*(l) N/A N/A N/A l
+
f Table 4.29-4 (Continued) Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Channel Channel Source ** Channel Functional Instrument Check Check Calibration Test
- 3. Spent Fuel Pool Area Ventilation System
- a. Noble Gas Activity Monitor D*
! M R Q I
- b. Effluent Flow Monitor D* N/A R N/A
- c. Sampler Flow Monitor D*
j N/A R N/A i d. Iodine Sampler Filter W*(1) N/A
, N/A N/A
- ~;; e. Particulate Sampler lzg Filter W*(1) 1 N/A N/A N/A
- 4. Reactor Building Purge I System
- a. Noble Gas Activity Monitor D* M R P
- b. Effluent Flow Monitor D*
t N/A R N/A
- c. Sampler Flow Monitor D* N/A R N/A
- d. Iodine Sampler Filter '
W*(1) N/A N/A N/A i e. Particulate Sampler Filter W*(1) '. N/A N/A N/A . i 1 l 1
Table 4.29-4 (Continued) .., Table Notation ( *During releases via this pathway.
**A check source is not required if the background activity is greater than the activity of the check source.
P Prior to release W Weekly D Daily M Monthly Q Quarterly R Once per 18 months
~.
NA Not applicable (1) Verify presence of cartridge or filter only. i s. D' i i a d 110qq
RADI0 ACTIVE EFFLUENTS
- 4. f.9. 3 Dose Calculations for Radioactive Effluents Applicability: At all times Objective:
To ensure that the requirements of 10CFR50, Appendix I, Section IIIA are met. Specifications: 4.29.3 Cumulative dose contributions and dose projections for liquid effluents and for gaseous effluents shall be de-termined in accordance with the Offsite Dose Calculation Manual at least once per 31 days. Bases: These calculations provide the dose values to be compared to the limits of Specifications 3.25.1.2, 3c25.1.3, 3.25.2.2, 3.25.2.3, 3.25.2.4 and 3.25.3. i L 110rr i I
4.30 RADIOLOG1 CAL ENVIRONMENTAL MONITORING ,, 4.30.1 Radiological Environmental Monitoring pronram Description ( Applicability: Applies at all times. Objective: To provide information on the radiological effects of station operation on the environment. Specifications: 4.30.1.1 The-radiological environmental monitoring samples shall be collected pursuant to Table 4.30-1 and shall be analyzed pursuant to the requirements of Tables 4.30-1 and 4.30-2. The sample locations shall be shown in Table 4-1 in the ODCM. 4.30.1.2 a. With the radiological environmental monitoring program not being conducted as specified in Table 4.30-1, prepare and submit to the Commission in the Annual Radiological Environmental Report a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. (Deviations are permitted from the required sampling schedule if specimens ' are not obtainable due to hazardous conditions, seasonal unavailability, or to malfunction of sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period).
- b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at one or more of the locations specified in Table 4.30-1 exceeding the limits of Table 4.30-3 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days from the end of the affected quarter, a report which includes an evaluation i of any release conditions, environmental factors or other aspects which caused the limits of Table 4.30-3 to be exceeded, and defines the actions taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the ;
calendar year limits of Specifications 3.25.1.2 and 3.25.2.2. When more than one of the radionuclides ' in Table 4.30-3 are detected in the sampling medium, this report shall be submitted if: , Concentration (1) reporting level (1) + Concentration reporting level(2)(2) + **** #- 1.0 When radionuclides other than those in Table 4.30-3 I are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a member of the public is equal to or L 110ss
greater than the calendar year limits of ,Specifica-tions 3.25.1.2 and 3.25.2.2. This report is not re-quired if the measured level of radioactivity was ( not the result of plant effluents, however, in such an event, the condition _shall be reported and de-scribed in the Annual Radiological Environmental Re-port. c. With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 4.29-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the causes of the unavailability of samples and identify the new location (s) for obtaining re-placement samples in the next Semiannual Radioactive Effluent Release Report and also include in the re-port a revised table for the ODCM reflecting the new location (s), d. 4 The provisions of Specifications 3.0.3, 3.0.4, and 6.12.3 are not applicable. 4.30.1.3 The results of analyses performed on the radiological environmental monitoring samples shall be summarized in the Annual Radiological Environmental Report. Bases: The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluents monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program of commercial operation.will be effective for at least the first three years Following this period, program changes may be initiated based on operational experience. The detection capabilities required by Table 4.30-2 are state of-the art for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141. L 110tt
4 p O O TABLE 4.30-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Sampling and and/or Sample Sample Locations
- Type of Frequency Collection Frequency of Analyses
- 1. AIRBORNE 3
- a. Radiciodine 5 Locations Continuous operation of Radiciodine canister, and Particulates sampler with sample Analyze at least once collection as required per 7 days for I-131.
by dust loading but at least once per 7 days. Particulate sampler. Analyze for gross beta radioactivity > 24 hours following filter change. Perform gamma isotopic % analysis on each sample @ when gross beta activity C is > 10 times the mean of control sample. Perfort gamma isotopic analysis on composite (by location) sample at least once every 92 days.
- 2. DIRECT RADIATION 40 Locations At least once per 92 2 dosimeter per Gamma dose. At least days. once per 92 days.
location
" Sample locations are shown in the Offsite Dose Calculation Manual (0DCM).
[ O D TABLE 4.30-1 (Continued) Exposure Pathway Number of and/or Sample Sampling and Type of Frequency Sample Locations
- Collection Frequency of Analyses
- 3. WATERBORNE
- a. Surface 2 Locations Composite ** sample Gamma isotopic analysis collected over a period of each sample by 5 31 days. location. Tritium analysis of composite sample at least once every 92 days.
- b. Ground 2 Locations At least once per 92 Gamma isotopic and days. , tritium analyses of each sample.
- c. Drinking 1 Location Monthly grab sample I-131 analysis of each sample; and Gross beta and gamma isotopic analyses of each sample. Tritium analysis of composite sample at least once every 92 days.
- d. Sediment from 2 Locations At least once per 184 Shoreline Gamma isotopic analysis days of each sample .
- Sample locations are shown in the ODCM.
** Composite samples shall be collected by an aliquot at intervals not exceeding 24 hours.
f- D TABLE 4.30-1 (Continued) Exposure Pathway Number of and/or Sample Sampling and Type of Frequency Sample Locations
- Collection Frequency of Analyses
- 4. INGESTION
- a. Milk 4 Locations At least once per 31 days Gamma isotopic and when animals are on pasture. I-131 analyses of each sample.
- b. Fish 2 Locations One sample in season, or at Gamma isotopic analysis least once per 184 days if on edible portions. '
not seasonal. One sample of each of the following species: p;
- 1. Catfish '
E
- 2. Crappie or Bass
- c. Food Products ** 3 Locations At time of harvest. One Gamma isotopic analysis sample of each of the on edible portions, following classes of food products: 8
- 1. Fruits
- 2. Flowering Vegetable
- 3. Tubular Vegetable 1 Location At time of harvest. One I-131 analysis.
sample of broad leaf vegetetion.
^ Sample locations are shown in the ODCM. **If these food products are available.
O
l p O D l TABLE 4.30-2 MAXIMUM VALUES OF THE LOWER LIMITS OF DETECTION (LLD(a)) Analyses Airborne Particulate Water or Gas Fish (pCi/1) (pCi/m2 ) Milk Food Products Sediment (pCi/kg, wet) (pCi/l) (pCi/kg, wet) (pCi/kg, dry) gross beta 4(b) 1 x 10'2 3g (1000(b)) 54gn 15 130 59 pe 30 260 , 58,60 15 Co 130 65 30 Zn 260 '. 9b 10 Zr-Nb 131 g 1*(b) 7 x 10'2 1 60(c) 134,137 Cs 15(10(b)),18 1 x 10 -2 130,150 15,18 60,80 150,180 140 15 Ba-La 15
*For Monthly grab samples l
(a) LLD (b) See for definition drinking ofwater LLD in table notation of Table 4.29-1. (c) LLD for leafy-vegetables. ! 8 i i e i L (
M f O O i
. TABLE 4.30-3 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Water Airborne Particulate Fish Analyses (pCi/1) Milk Food Products '
or Gases (pCi/m3) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) H-3 3 x 104(a) Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 i Co-58 1 x 103 3 x 104 j Co-60 3 x 102 - 1 x 104 ! Zn-65 3 x 102 2 x 104 ! Zr-Nb-95 4 x 102(b) w{ ! I-131 2 0.9 3 1 x 102 } Cs-134 30 10 1 x 103 60 1 x 103 CS-137 50 20 2 x 103 70 2 x 103 l Ba-La-140 2 x 102(b) 3 x 102(b) l (a) For drinking water samples (b) Total for parent and daughter ) I i s : l { i
-- _ . _ . ~ . . _ . _. - -
Radiological Environmental Monitoring 4.30.2 Land Use Census ( Applicability: Applies at all times. Objectives: This specification will identify changes in use of the unrestricted areas. Specifications: 4.30.2.1 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles from the ANO-1 reactor building.
4.30.2.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1, by door-to-doorauthorities. agricultural survey, aerial survey, or by consulting local 4.30.2.3. a. With a land use census identifying a location (s) which yields a calculated dose commitment due to I-131, tritium, and radionuclides in particulate form greater than the values currently being calculated in Unit I Specification 4.29.3 and Unit 2 Specification 4.11.2.3 submit location description in the Semiannual Radioactive Effluent Release Report per Specification 6.12.2.6. b. With a land use census identifying a location (s) which yields a calculated dose commitment (via the same exposure pathway) greater than at a location from which samples are currently being obtained in accordance with Specification l 4.30.1.1. The new location shall be added to the radio-logical environmental monitoring program within 30 days, if possible. The sampling location having the lowest calculated dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October
- 31) of the year in which this land use census was con-ducted.
4.30.2.4 ( The results of the land use census shall be included in the Annual Radiological Environmental Report. 4.30.2.5 The provisions of Specifications 3.0.3, 3.0.4 and 6.12.3 are not applicable.
- Broad Leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest 0/Q in lieu
- of the garden census.
l 110zz l
Bases: l This specification is provided to ensure that changes in the use of C unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater inan 500 square feet provides assurance that significant exposure pathway via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determire this minimum garden size, the following assumptions were used,1) that 20% of the garden was used for growing broad leaf vegetation yield of 2 kg/(i.e., similar square to lettuce and cabbage), and 2) a vegetation meter. : l i i e ( ! I i l 110aaa
._ _. _ _ . - - -- _. __ _-- _ __ _ _ = . _ - _ _ _ _
Radiological Environmental Monitoring 4.30.3 Interlaboratory Comparison Prooram Applicability: Applies to the off-site radiochemistry laboratory Objective: To provide independent checks on the accuracy of the measurements of radioactive material in environmental samples. Specifications: i. 4.30.3.1 Analyses shall be performed on radioactive materials ' supplied as part of Interlaboratory Comparison Program which has been approved by NRC. 4.30.3.2 With analyses not being performed as required above, report ' the corrective actions taken to prevent a recurrence to the
- Commission in the Annual Radiological Environmental Report.
4.30.3.3 The results of analhses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Report 4 pursuant to Specification 6.12.2.5. 4.30.3.4 The provisions of Specifications 3.0.3, 3.0.4 and 6.12.3 are not applicable. Bases: i 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 enviromental sample matrices.are performed as part of a quality assurance program for environmental monitoring-in order to demonstrate that the results are reasonably valid. l f l l l t i 3 l i-l 110bbb i
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- g. Review of facility operations to detect potential nuclear safety hazards,
- h. Performance of special reviews, investigations and reports thereon as requested by the General Manager.
- 1. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the General Manager.
- j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the General Manager.
- k. Review of all changes to the Offsite Dose Calculation Manual and the Process Control Program.
AUTHORITY 6.5.1.7.1 The Plant Safety Committee shall:
- a. Recommend to the General Manager written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
- b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above consitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the Assistant Vice-President, Nuclear Operations and the Safety Review Committee of disagreement between the PSC and the General Manager: however, the General i
Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above. RECORDS l f 6.5.1.8 The Plant Safety Committee shall maintain written minutes of.each PSC meeting that, at a minimum, docu-ment the results of all PSC activities performed under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the General Manager and Chairman of the Safety Review Committee. 6.5.2 Safety Review Committee (SRC) FUNCTION 6.5.2.1 The Safety Review Committee shall function to provide l independent review and audit of designated activities in the areas of:
- a. nuclear power plant operations
- b. nuclear engineering
- c. chemistry and radiochemistry l
122
REVIEW 6.5.2.'7 The SRC shall review;
- a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
- b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- c. Proposed tests or experiments which involve an 3
unreviewed safety question as defined in Section 50.59, 10 CFR.
- d. Proposed changes in Technical Specifications or licenses.
- e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
- f. Significant operating abnormalities or deviation from normal and expected performance of unit equipment that affect nuclear safety.
- g. Events requiring 24-hour notification to the Commission.
- h. All recognized indications of an unanticipated deficiency in some aspects of design or operation of structures, systems, or components that could affect nuclear safety.
- i. Reports and meeting minutes of the Plant Safety Committee.
- j. Changes to the ODCM and PCP.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the SRC. These audits shall encompass:
- a. The conformance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per year.
- b. Th'e performance and retraining of all members of the plant management and operations staff, and the performance, training, and qualifications of new members of the entire plant staff at least once per year.
- c. The results of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per six months.
124
- d. The Facility Emergency Plan and implementing procedures at least once per 12 months. ;
- e. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
- f. The Facility Security Plan and implementing procedures at least once per 12 months.
- g. Any other area of facility operation considered appropriate by the SRC or the Senior Vice-President, Energy Supply (SRVP,ES).
- h. An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified off-site licensee personnel or an outside fire protection firm.
- i. The radiological environmental monitoring program and the results thereof at least once per 12 months.
- j. The Offsite Dose Calculation Manual and Process Control Program and implementing procedures at least once per 24 months.
- k. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of 10 CFR 50 Appendix "B", at least once per 24 months.
AUTHORITY 6.5.2.9 The SRC shall report to and advise the Senior Vice President, Energy Supply (SRVP,ES) on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8. RECORDS 6.5.2.10 Records of SRC activites shall be prepared, approved and 4 distributed as indicated below:
- a. Minutes of each SRC meeting shall be prepared, approved and forwarded to the Senior Vice-President, Energy Supply (SRVP,ES) within 14 days following each meeting.
- b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Senior Vice-President, Energy Supply (SRVP,ES) within 14 days following completion of the review.
- c. Audit reports encompassed by Section 6.5.2.8 above shall be forwarded to the Senior Vice-President, Energy Supply (SRVP,ES) and to the management positions responsible for the areas audited within 30 days after completion of the audit.
125
- a. The facility shall be placed in at least hot shutdown within one hour.
- b. The Nuclear Regulatory Commission shall be notified and a report submitted pursuant to the requirements of 10 CFR 50.36 and Specification 6.12.3.1.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November, 1972.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
- e. Emergency Plan implementation.
- f. Fire Protection Program implementation.
- g. New and spent fuel storage
- h. Offsite Dose Calculation Manual and Process Control Program implementation at the site.
, 6.8.2 Each procedure of 6.8.1 above, and changes theret3, shall be reviewed by the PSC and approved by the General Manager prior to implementation and reviewed periodically as set forth in administrative procedures. 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
- a. The intent of the original procedure is not altered.
- b. The change is approved by two members of the plant staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
- c. The change is documented, reviewed by the PSC and approved by the General Manager within 14 days of implementation.
127
l
- h. Records of in-service inspections performed pursuant to these Technical Specifications.
- i. Records of Quality Assurance activities required by Section 17 of the Quality Assurance Manual for Operations.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the PSC and the SRC.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.
- m. Records of the analyses required by the Radiological l Environmentai Monitoring Program.
l 6.10 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.11 HIGH RADIATION AREA 6.11.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area (as defined in 20.202(b)(3) of 10 CFR 20) in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and shall be controlled by requiring the issuance of a radiation work permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a pre-set integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been knowledgeable of them,
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device., This individual shall be responsible for providing positive control over the activities 'within the area shall perform periodic radiation surveillance at the frequency specified in the radiation work permit.
6.11.2 The requirements of 6.11.1 above, shall also apply to each high radiation area in which the intensity.of radiation is greater than 1000 l mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and access to these areas shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Health Physics Superintendent.
-129- (Next page is 140)
The dose assignments to various duty functions may be, estimates based on pocket dosimeter, TLD, or film badge measurements. 'Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole specificbody major dose received work from external sources shall be assigned to functions. 6.12.2.3 Monthly Operating Report Routine reports of operating statistics which include: (1) Average Daily Unit Power Level (2) Operating Data Report (3) Unit Shutdowns and Power Reductions (4) Narrative Summary of Operating Experience shall be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, 0:C. 20555, with a copy to the appropriate Regional Office by the fifteenth of each month following the calendar month covered by the report. 6.12.2.4 Annual Report All challenges to the pressurizer electromatic relief valve (ERV) and pressurizer safety valves shall be reported annually. 6.12.2.5 Annual Radiological Environmental Report
- C (a) Routine radiological environmental reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
'(b) The annual radiological environment.al reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate) and previous environmental surveillance reports and an assessment of the observed impact of the plant operation on the environment. The report shall also include the results of the land use census required by Speci fiction 4.30.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall proside an analysis of the problem and a planned courst of action to alleviate the problem.
+A single submittal may be made for ANO-1 and ANO-2. The submittal should combine those sections that are common to both units at the station.
L
. 141
l l
' j The annual radiological environmental reports shall include , summarized and tabulated results of all radiological en-vironmental samples taken during the report period. In the event that some results are not available for inclusion with the report, 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. 1 i
The report shall also include the following: a summary description of the radiological environmental monitoring j program including sampling methods for each sample type, size 1 and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; the result of Land Use Census required by the Specification 4.30.2, and the results of licensee participation in the Interlaboratory Comparison Program required by Specification 4.30.3 6.12.2.6 Semiannual Radioactive Effluent Release Report ** (a) Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. (b) The radioactive effluent release reports shall include e. summary of the quantities of radioactive liquid and gaseous effluents and solid waste release from the unit. The data will be summarized on a quarterly basis following the format of Regulatory Guide 1.21, Rev. O Appendix A. (c) The radioactive effluent release report shall include the following information for all unplanned releases to unrestricted areas of radioactive material in gaseous and liquid effluents;
- 1. A description of the event and equipment involved.
l ! 2. Cause(s) for the unplanned release.
- 3. Actions taken to prevent recurrence.
l 4." Consequences of the unplanned release. i (d) This report shall contain a description of any changes l to the ODCM and PCP made during the period of the report.
**A single submittal inay be made for ANO-1 and ANO-2.
t l l 141a a
6.12.3 Reportable Occurrences Reportable occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC as required below.
-( Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report original reportshall be completed and reference shall be made to the date.
6.12.3.1 Prompt Notification With Written Followup The types of events listed below shall be reported as expeditiously as possible, but within 24-hours, by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the appropriate Regional Office, or his designate no late'r than the first working day following the event, with a written followup report within two weeks. A copy of the confirmation and a written followup report shall also be sent to the Director, Office of Management and Program Analysis, USNRC. The < written report shall include, as a minimum, a completed copy of a licensee event report form. information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material the event. to provide complete explanation of the circumstances surrounding (a) Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the Technical Specifications or failure to complete the required protective function. NOTE: Instrument drift discovered as a result of testing need not be reported under this item but may be reportable under items (e), (f), or 6.12.3.2(a). l 4 C 142
(d) Abnormal degradation of systems other than those specified in item 6.12.3.1(c) above designed to contain radioactive material resulting from the fission process.
. NOTE:
Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this item. (e) An unplanned offsite release during any one hour period of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:
- 1. A description of the event and equipment involved.
- 2. Cause(s) for the unplanned release.
- 3. Actions taken to prevent recurrence.
- 4. Consequences of the unplanned release.
_.s i r l e l l l l l l r h
- ' 145 I
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6.12.5 Special Reports Special reports shall be submitted to the Director of the Office of { Inspection and Enforcement Regional Office within the time period specified for each report. activities identified in t.he applicable reference specification.These rep a. i Radioactive Liquid Effluents, Specification 3.25.1. b. Radioactive Gaseous Effluents, Specifications 3.25.2. This report shall include:
- 1) a description of the occurrence,
- 2) the cause(s) for exceeding the limit (s),
' 3) corrective action taken to mitigate the consequences of the occurrence,
- 4) action taken to prevent recurrence, and
- 5) a summary of the consequences of the occurrence.
i (:- k i I j -- i 146a
ENCLOSURE 2 TO OCAN048408 REVISIONS TO AN0-2 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS l
INDEX DEFINITIONS ( SECTION PAGE SOURCE CHECK............................................. 1-6 0FFSITE DOSE CALCULATION MANUAL (0DCM)................... 1-6 LIQUID RADWASTE TREATMENT SYSTEM......................... 1-6 GASEOUS RADWASTE TREATMENT SYSTEM........................ 1-6a VENTILATION EXHAUST TREATMENT SYSTEM..................... 1-6a MEMBER (S) 0F THE PUBLIC.................................. 1-6a PURGE - PURGING.............. ........................... 1-6a EXCLUSION AREA..................................... ..... 1-6a UNRESTRICTED AREA........................................ 1-6a ( i l I l ARKANSAS - UNIT 2 Ia l l
INDEM LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS ( SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE.................................. 3/4 2-1 3/4.2.2 RADIAL PEAKING FACT 0RS............................ 3/4 2-4 3/4.2.3 AZIMUTHAL POWER TILT.............................. 3/4 2-5 3/4.2.4 DNBR MARGIN....................................... 3/4 2-7 3/4.2.5 RCS FLOW RATE..................................... 3/4 2-11 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE.............. 3/4 2-12 3/4.2.7 AXIAL SHAPE INDEX................................. 3/4 2-13 3/4.2.8 PRESSURIZER PRESSURE.............................. 3/4 2-14 3/4.3 INSTRUMENTATION i 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION................ 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION................................. 3/4 3-10 3/4.3.3 MONITORING INSTRUMENTATION , Radiation Monitoring Instrumentation.............. 3/4 3-24 Incore Detectors.................................. 3/4 3-28 Seismic Instrumentation........................... 3/4 3-30 i i Meteorological Instrumentation..... .............. 3/4 3-33 l Remote Shutdown Instrumentation................... 3/4 3-36 Post-Accident Instrumentation..................... 3/4 3-39 Chlorine Detection Systems........................ 3/4 3-42 Fire Detection Instrumentation.................... 3/4 3-43 Radioactive Gaseous Effluent Monitoring i Instrumentation................................. 3/4 3-44a ' Radioactive Liquid Effluent Monitoring Instrumentation................................. 3/4 3-44j i ( 3/4.3.4 TURBINE OVERSPEED PROTECTION...................... 3/4 3-45 l ARKANSAS - UNIT 2 IV l
- - . _ . _ - - - ~ _ -_---_.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.................................. 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS................................. 3/4 11-7 3/4.11.3 TO TA L D O S E - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4
. . . 11-15 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING (REFERENCE TO ANO-1 PROGRAM)................................ 3/4 11-16 l
l r i I l ARKANSAS - UNIT 2 VIIIa t
INDEX GASES SECTION PAGE 3/4.0 APPLICABILITY.......................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.................................. B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.................................. B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES........................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE.................................. B 3/4 2-1 3/4.2.2 RADIAL PEAKING FART 0RS............... ............ B 3/4 2-2 3/4.2.3 AZIMUTHAL POWER TILT.............................. B 3/4 2-2 3/4.2.4 DNBR MARGIN....................................... B 3/4 2-3 3/4.2.5 RCS FLOW RATE..................................... B 3/4 2-4 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE.............. B 3/4 2-4 3/4.2.7 AXIAL SHAPE INDEX................................. B 3/4 2-4 3/4.2.8 PRESSURIZER PRESSURE.............................. B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION........................ B 3/4 3-1 ! 3/4.3.2 ENGINEERED SAFETt FEATURE INSTRUMENTATION......... B 3/4 3-1 l 3/4.3.3 MONITORING INSTRUMENTATION........................ B 3/4 3-2 l 3/4.3.4 TURBINE OVERSPEED PROTECTION...................... B 3/4 3-5 t I l l l l l t i ARKANSAS - UNIT 2 IX
_INDEX , EASES ( _SECTION PAGE 3/4.9.5 COMMUNICATIONS.................................... B 3/4 9-2 3/4.9.6. REFUELING MACHINE OPERABILITY..................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING........ B 3/4 9-2 3/4.9.8 COOLANT CIRCULATION............................... B 3/4 9-2 3/4.9.9 andSTORAGE 3/4.9.10 WATER LEVEL - REACTOR VESSEL AND POOL WATER LEVEL.......................... B 3/4 9-3 3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM............. B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN i MARGIN................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.......................................... B 3/4 10-1 3/4.10.3 REACTOR COOLANT L00PS............................. B 3/4 10-1 3/4.10.4 CENTER CEA MISALIGNMENT........................... B 3/4 10-1 ( 3/4.10.5 MINIMUM TEMPERATURE FOR CRITICALITY...... ........ B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.................................. B 3/4 10-2 3/4.11.2 r GASE0US EFFLUENTS.......... ...................... B 3/4 10-3 3.4.11.3 TOTAL 00SE........................................ B 3/4 10-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING (REFERENCE TO ANO-1 PR0 GRAM)............................... B 3/4 10-7 L ARKANSAS - UNIT 2 XII
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE OCCURRENCE ACTI0N............................. 6-12 S.7 SAFETY LIMIT VIOLATION................................... 6-13 6.8 PROCEDURES............................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES........ 6-14 6.9.2 SPECIAL REP 0RTS................................... 6-18a 6.9.3 SEMIANNUAL RADIOLOGICAL EFFLUENT RELEASE REPORT... 6-18b 6.10 RECORD RETENTION........................................ 6-19 6.11 RADIATION PROTECTION PR0 GRAM............................ 6-20 6.12 ENVIRONMENTAL QUALIFICATION............................. 6-20 6.13 HIGH RADIATION AREA..................................... 6-21 6.14 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM, LAND USE CENSUS, INTERLABORATORY COMPARISON PROGRAM......... 6-22 6.15 0FFSITE DOSE CALCULATION MANUAL (0DCM).................. 6-22 A ARKANSAS - UNIT 2 XV
DEFINfTIONE ENGINEERED SAFETY FEATURE RESPONSE TIME ( 1.24 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times applicable. where shall include diesel generator starting and sequence loading delays PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized by under the provisions of 10 CFR 50.59, or 3) otherwise approved the Commission. SOFTWARE 1.26 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation and procedures. F PLANAR RADIAL PEAKING FACTOR 1.27 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane ( average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt. SOURCE CHECK 1.28 A SOURC5 CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to the radioactive source. OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.29 An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. LIQUID RADWASTE TREATMENT SYSTEM 1.30 A LIQUID RADWASTE TREATMENT SYSTEM is a system designed and installed to reduce radioactive liquid effluents from the unit. This is accomplished by providing for holdup, filtration, and/or demineralization of radioactive liquid effluents prior to their release to the environment. L ARKANSAS - UNIT 2 1-6
DEFINITIONS GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASE0US RADWASTE TREATMENT SYSTEM is any system designed and installed _to reduce radioactive gaseous effluents from the plant by collecting offgases from radioactive systems and providing for decay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. VENTILATION EXHAUST TREATMENT SYSTEM 1.32 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Atmospheric cleanup systems that are Engineered Safety Feature (ESP) actuated are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS. MEMBER (S) 0F THE PUBLIC 1.33 MEMBER (S) 0F THE PUBLIC shall include all persons who are not oc-cupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of ( the site for recreational, occupational or other purposes not associated with the plant. PURGE-PURGING 1.34 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to reduce airborne radioactive concentrations in such a manner that replacement air or gas is required to purify the confinement. EXCLUSION AREA l 1.36 The EXCLUSION AREA is that area surrounding ANO within a minimum i radius of .65 miles of the reactor buildings and controlled to the extent necessary by the licensee for purposes of protection of indi-viduals from exposure to radiation and radioactive materials. UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area at or beyond the exclusion area boundary. L ARKANSAS - UNIT 2 1-6a
. .- __ _ _ _ _ - - . _ . _.. .=. _-._ .___ ._ - - . _ .
i TABLE 1.2
-(
FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours. D At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. P Completed prior to each release. , N.A. Not applicable. 4 e s # I 1 l ARKANSAS - UNIT 2 1-8 l - i
INSTRUMENTATION ' RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION . LIMITING CONDITION FOR OPERATION 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. APPLICABILITY: During releases via this pathway. ACTION: a. With the following gaseous effluent monitoring in-strumentation channels alarm / trip setpoint less con-servative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel. 1. Waste Gas Holdup System Noble Gas Activity Monitor. (during periods of gaseous releases.) 2. ( Containment Purge and Ventilation System Noble Gas Activity Monitor. (during periods of containment building PURGE.) b. With less than the minimum number of monitoring instrumentation channels OPERABLE, take the action shown in Table 3.3-12. c. Return the instruments to OPERABLE status within 30 days or, in lieu of any other report, explain in the l next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected. d. ! The provisions of Specifications 3.0.3, 3.0.4, 4.0.4 l and 6.9.1.7 are not applicable. SURVEILLANCE REQUIREMENTS i l 4.3.3.9 ! Each radioactive gaseous effluent monitoring instrumenta- ! tion channel shall be demonstrated OPERABLE by performance i of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANilEL FUNCTIONAL TEST at the frequencies shown in l Table 4.3-12. l { L ARKANSAS - UNIT 2 3/4 3-44a l I
C D O TABLE 3.3-12 g; RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 9
=
i, MINIMUM
$1 CHANNELS
]' INSTRUMENT OPERABLE APPLICABILITY PARAMETER ACTION
(( 1. Waste Gas Holdup System
-4 i na a. Noble Gas Activity Monitor-(provides alarm and automatic
- 1 Radioactivity 25 termination of release)
- b. Effluent System Flow Monitor 1
- System Flow 26
- 2. Containment Purge and Ventilation System u, a. Noble Gas Activity Monitor
- 5 1 Radioactivity 27,29 u2 b. Iodine Sampler Cartridge Verify Presence of In *
@ 1 Cartridge 28
- c. Particulate Sampler Filter Vtrify Presence of 1
- Filter 28
- d. Effluent System Flow Monitor 1
- System Flow 26
- e. Sampler Flow Monitor 1
- Sampler Flow 26 f f
? .*
'\ %
,o N L' \ t
p A O' s,', . q . \- . .
\
TABLE 3.3-12 (Continued) [
- n RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION h,
MINIMUM g - Cl{ANNELS m INSTRUMENT i OPERABLE APPLICABILITY PARAMETER ACTION
- g. s 3. Spent Fuel Area Ventilation System yJ .
,y a. Noble Gas Activity
- t. -
Monitor
- 1 Radioactivity 27
- b. Iodine Sampler Cartridge Verify Presence of 1 Cartridge 28
, c. Particulate Sampler Verify Presence of Filter 1
- Filter 28
$ d. Effluent System Flow Monitor 1
- System Flow 26
- e. Sampler Flow Monitor
- 1 Sampler Flow 26
- 4. Auxiliary Building Area Ventilation System
- a. Noble Gas Activity Monitor 1
- Radioactivity 27
- b. Iodine Sampler Cartridge Verify Presence of 1 Cartridge 28
- c. Particulate Sampler Filter Verify Presence of 1
- Filter 28
- d. Effluent System Flow Monitor 1
- System Flow 26 .
- e. Sampler Flow Monitor 1
- Sampler Flow 26
;p 3 s
h O O
~#. . ,
e ! '">! '
+ 7 'i
,\
_, / ? 1 1 r x r' f
- 1 f,, s ;
; ;., - ; .^- o e
_ TABLE 3.3-12 (Continued) ,
'I -/J.s-
. g. .
. RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.
'$ Di MINIMUM 5 -
INSTRUMENT - CHANNELS-e OPERABLE APPLICABILITY PARAMETER ACTION g, S. Auxiliary Building Extension . - 3 Ventilation Sysism ,
" ~
- a. Noble Gas _ Activity.
.,Monit4 '
1 Radioactivity 27
- b. ' Iodine Sample Cartridge .
Verify Presence of
,1
- Cartridge
/
- 28
- c. Particulate Sampler -
Filter
. Verify Presence of 1 Fijter 28
- d. Ef fluent System Flow Monitor.
- 1 System Flow 26.
- e. Sampler Flow Monitor
- 1 Sampler Flow 26 t .
.e '
d r . a v
TABLE 3.3-12 (Continued) .,, TABLE NOTATION ( *During releasss via this pathway. ACTION 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released initiating to the environment provided that prior to the release:
- 1. At least two independent samples of the tank's con-tents are analyzed, and;
- 2. At least two technically qualified members of the Facility Staff independently verify the computer input data, and;
- 3. At least two technically qualified members of the facility staff independently verify the discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway. ACTION 26 With the number of char.nels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours. ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. ACTION 28 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are collected with auxiliary sampling equipment. Iodine sample cartridges and particulate sample filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours i after changing in accordance with Table 4.11-2. i ACTION 29 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving movement of fuel assemblies or CEAs within the pres-sure vessel. i I L ARKANSAS - UNIT 2 3/4 3-44e
-- -~ . . - -
.~-
TABLE 4.3-12 g RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS h CHANNEL 5 CHANNEL SOURCE CHANNEL FUNCTIONAL { INSTRUMENT CHECK CHECK CALIBRATION TEST
- 1. Waste Gas Holdup System m a. Gas Activity Monitor D* P** R Q (provides alarm and automatic termination of release)
- b. System Effluent Flow Monitor D* N/A R N/A
- 2. Containment Purge and Ventilation System w a. Gas Activity Monitor D* P** R M (1), P 4 b. Iodine Sampler Cartridge W*(2) N/A N/A N/A
- d. Particulate Sampler Fil+.er W*(2) N/A N/A N/A
- d. System Effluent Flow Monitor D* N/A R N/A
- e. Sampler Flow Monitor D* N/A R N/A
p O e TABLE 4.3-12 (Continued) g RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQ h
$ CHANNEL CHANNEL INSTRUMENT SOURCE CHANNEL FUNCTIONAL CHECK CHECK CALIBRATION TEST E 3. Spent Fuel Area Ventilation System m a. Gas Activity Monitor 'D* M** R Q
- b. Iodine Sampler Cartridge W*(2) N/A N/A N/A
- c. Particulate Sampler Filter W*(2) N/A N/A N/A
- d. System Effluent Flow Monitor D* N/A R N/A
- e. Sampler Flow Monitor D*
M N/A R- t1/A
^
- 4. Auxiliary Building Area Ventilation System Y
2 a. Gas Activity Monitor D* u:2 M** R Q
- b. Iodine Sampler Cartridge W^(2) N/A fl/A N/A
- c. Particulate Sampler Filter W^(2) ti/A 11/A ti/A
- d. System Ef fluent Flow Monitor D^ ti/A R 11/A
- e. Sampler Flow Monitor D* ti/A R t1/A 9
p A _O TABLE 4.3-12 (Continued) g RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMEN h
'" CHANNEL ' SOURCE CilANNEL CHANNEL INSTRUMENT FUNCTIONAL l CHECK CilECK CALIBRATION TEST
@ 5. Auxiliary Building Extension p Ventilation System m
- a. Gas Activity Monitor D* M** R Q
- b. Iodine Sampler Cartridge W*(2) N/A N/A N/A
- c. Particulate Sampler Filter W*(2) N/A N/A N/A
- d. System Effluent Flow Monitor D* N/A R, N/A
- e. Sampler Flow Monitor D* N/A R N/A Y
i
TABLE 4.3-12 (Continued) ., , TABLE NOTATION
*During releases via this pathway.
**A SOURCE CHECK is not required if the background activity is greater than the activity of the check source.
(1) During Containment Building ventilation operations. i' 4
. (2) Verify presence of cartridge or filter only.
s. 4 4 h f 1 i, L ARKANSAS - UNIT 2 3/4 3-44i 4 ' ~ v - e a r- een . e e <,. ,,~s- .,- ,.e-.r. ,,m-,mgw--,- , e-a--,-y, -----,,.m- _
,-,-m e, nns wey gn-yww<*4=~~rav,n-m-- ----w.w
INSTRUMENTATIO?i RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION h LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. APPLICABILITY: During releases via this pathway. ACTION: a.
~
With a radioactive liquid effluent monitoring instru-mentation channel alarm / trip setpoint less conser-vative than required by the above specification, immediately suspend the release of radioactive liquid ef fluents monitored by the affected channel, until the setpoint is changed to an acceptably conservative value. b. With less than the minimum number of monitoring instrumentation channels OPERABLE, take the action shown in Table 3.3-13. c. ( Return the instruments to OPERABLE status within 30 days or, in lieu of any other report, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected.
- d. The provisions of Specifications 3.0.3, 3.0.4, 4.0.4 and 6.9.1.7 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-13. L ARKANSAS - UNIT 2 3/4 3-44j
A D TABLE 3.3-13 3m
- o RADI0 ACTIVE LIQUID EFFLUENT. MONITORING INSTRUMENTATION MINIMUM
_ $1 - v) CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION (( 1. Gross Radioactivity Monitor (s)
- q (provides alarm and automatic ,
n, termination of release)
- a. Liquid Radwaste Effluent Line 1 During releases via this pathway 18
- 2. Flow' Monitor (s)
- a. Liquid Radwaste Effluent Line 1 During releases via this pathway 19 2
x v
=
+
TABLE 3.3-13 (Continued) , TABLE NOTATION { ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed provided that prior to initiating a release: 1. At least two independent samples are analyzed, and; 2. At least two technically qualified members of the Facility Staf f independently verify the release rate computer input data, and; 3. At least two technically qualified members of the facility staff independently verify the discharge valve lineup. Otherwise, suspend release of radioactive effluents via this pathway. -
- ACTION 19 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this cathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump curves may be used to estimate flow. ,,
l ( 4 J l ARKANSAS - UNIT 2 3/4 3-441 l t-
O O' TABLE 4.3-13 g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMEN h U CHANNEL SOURCE CHANNEL-CHANNEL FUNCTIONAL INSTRUMENT CllECK CHECK CALIBRATION TEST E 1. Gross Radioactivity Monitor (s) Z (provides alarm and automatic n3 isolation)
- a. Liquid Radwaste Effluents Line D* P** R Q
- 2. Flow Monitor (s)
- a. Liquid Radwaste Effluent Line D* N/A R N/A During releases via this pathway w
1 ** Athe SOURCE CilECK is not required if the background activity is greater than the activity of check source. e e
3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIOUID EFFLUENTS -( CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released from the site in liquid effluents to the discharge canal shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration released shall be limited to 2 x 10-4 uCi/ml. APPLICABILITY: At all times. ACTION: - a) With the concentration of radioactive material released exceeding the above limits, immediately initiate actions to restore concentrations to within the above limits. Provide notification to the Commission within 24 hours and in lieu of any other report, submit a Special Report - pursuant to Specification 6.9.2.g within 30 days. l b) The provisions of Specification 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analyses program of Table 4.11-1. 4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methods in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1 L ARKANSAS - UNIT 2 3/4 11-1
TABLE 4.11-1 . RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSES PROGRAM l l Sampling l Minimum Type of l Liquid Release l Frequency I Analyses l Activity i l Lower Limit l 1 Type 1 l Frequency I Analyses I of Detection i l l l (LLD) I l l 1 l l l (uCi/ml) (a) l I I i l l P l P l y isotopic (*)l I 1 A. Batch Waste leach Batch i Each Batchl I 5 x 10-7 (b) l l Release (d) i l l l l [ l l I-131 l l P 1 l 1 x 10-6 l i i l l l l i [ l Dissolved andl 1 x 10-5 I 10nc Batch /Ml M l Entrained i l i l l l Gases l I l l 1" l (Gamma l l l l l l l Emmitters) l l P l l l M l H-3 l 1 x 10 ' l l lEachBatchlCompgtel _ l l l l l Gross A'pha i 1 x 10-7 l l l P l l Sr-89, Sr-90 i S x 10 6 l l lEach Batch I Q l l l l l l l lCompogelFe-55 l 1 x 10-6 l l 1 1 l_ I I I ( I I I TABLE NOTATION a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probabil-ity of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): 4.66 s b llU
- E V 2.22 Y exp (-Mt)
Where: LLD is the lower limit of detection as defined above (as pico-curie per unit mass or volume). ss is the standard deviation of the background counting rate or oY the counting rate of a blank sample as appropriate (as counts per minute). E is the counting efficiency (as counts per transformation). L ARKANSAS - UNIT 2 3/4 11-2 4
TABLE 4.11-1 (Continued) ,,, V is the sample size (in units of mass or volume). { 2.22 is the number of transformations per minute per picocurie. Y is the fractional radiochemical yield (when applicable). A is the radioactive decay constant for the particular radio-nuclide, and at is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples). . Typical valu'es of E, V, Y, and at shall be used in the calculation. It should be recognized that the LLD is an a priori (before the fact) limit representing the capability of a measurement system and not an a posteriori (af ter the fact) limit for a particular measurement. b. For certain mixture's of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater l concentrations. Under these circumstances, it will be more appropriate to calculate the concentrati6n of such radionuclides using observed ratios with those radionuclides which are measurable. c. A composite sample is one in which the quantity of liquid sampled is l proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is ( representative of the liquids released. d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling, each batch shall be isolated and I mixed to assure representative sampling. ' e. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn54, Fe59, ; CoS8, Co60, Zn65, Mo99, Cs134, Cs137, Cel41, and Ce144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, I together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual i circumstances result in LLD's higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report. L ARKANSAS - UNIT 2 3/4 11-3
RADI0 ACTIVE EFFLUENTS i DOSE - LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose commitment to a MEMBER OF THE PUBLIC from radio-active materials in liquid effluents released from ANO-2 to the discharge canal shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 arem to any organ, and
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report submit a Special Report pursuant to Specification 6.9.2.g within 30 days.
- b. The provisions of specifications 3.0.3, 3.0.4 and 6.9.1.7 are not applicable. ,
SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days. I l l l l 6 i l l l ARKANSAS - UNIT 2 3/4 11-4 i
RADI0 ACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT { LIMITING CONDITION FOR OPERATION 3.11.1.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from ANO-2 to the discharge canal would exceed
.18 mrem to the total body or .625 mrem to any organ in any calendar quarter.
APPLICABILITY: At all times. ACTION: a. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.g within 30 days .
- b. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.7 are not applicable.
1 SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases shall be projected at least (. Once per 31 days in accordance with the ODCM. L ARKANSAS - UNIT 2 3/4 11-5 t-
RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS * ( LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each unprotected
- outside temporary radioactive liquid storage tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.
APPLICABILITY: At all times. ACTION: I a. With the quantity of radioactive material exceeding the above limit, immediately suspend all additions of radioactive material to the affected tank and within 48 hours reduce the tank contents to within the limit, b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each - ( unprotected outside temporary radioactive liquid. storage tank shall be determined to be within the above limit by analyzing a representative sample of the contents of the tank at least once per 7 days when radioactive materials [ are being added to the tank. Tanks included in this Specification are those outdoor temporary tanks that do not have 1) liners, dikes or I walls capable of holding the tank contents or 2) tank overflows and surrounding area drains connected to the ' LIQUID RADWASTE TREATMENT SYSTEM. l i i i
%v i
ARKANSAS - UNIT 2 3/4 11-6 4
l. RADI0 ACTIVE EFFLUENTS - 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to UNRESTRICTED AREAS (see i Figure 5.1-3) shall be limited to the following: i
- a. For noble gases: Less than or equal to the 500 mres/yr to the total body and less than or equal to 3000 mres/yr to the skin.
- b. For iodine-131, for tritium and for all radionuclides
- in particulate form with half lives greater than 8 days:
2 Less than or equal to 1500 mrem /yr to any organ. During periods of containment purging the dose rate may be averaged over a one hour interval. APPLICABILITY: At all times. l
- ACTION
l
- a. With the dose rate (s) exceeding the above limits, without delay restore the release rate to comply with the above j limit (s).
- b. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.7 4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents i shall be determined to be within the above limits in i accordance with the methods and procedures of the ODCM. I i 4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and j procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2. i i ARKANSAS - UNIT 2 3/4 11-7 l
.. . - - _ - _ . - - - . . - _ - - - _ _ . - . . - - - _ _ . ~ . . _ . . . - . - . _ - . - - - - _ . -
f O O-TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSES PROGRAM E 5 1 1 Minimum l 3l l Gaseous Release Type Sampling 1 Analyses l l Lower Limit of I l l l Type of I Detection l Frequency l c I l l Frequency l Activity Analyses I (uCi/ml) g LD) ll I I 1 O l l P I l
- l P l l A. Waste Gas Each Tank l Storage Tank l l Each Tank l Principal Gamma Emitters (b) ll 1 x 10-4 (9) l l l Grab Sample l l l l l l i I I I B. Reactor B1dg. I Each urge l Each Purge l Principal Gamma Emitters (U) l 1 x 10-4 (9) l Purge l Grab Sample l l 11-3 l 1 x 10-6 l
. I I I I I I C. Unit Vents l M (c) (d) l M I l Principal Gamma Emitters (b) I 1 x 10-4 (9) l w 1 l Grab Sampic l 3 1 l 11-3 l 1 x 10'G l (Auxiliary Bldg. Ext.)l l l l
- l (Spent Fuel Pool l l W (f) l l l
7' l Area Ventilation) Continuous (" Charcoal l l l l l l I-131 l 1 x 10-12 [ l l Sample i l (Rx Bldg. Ventilation)l l l
! l l l l (Radwaste Area Venti-l W (f) l l lation) Continuous I ") I l l l l Particulate l Principal Gamma EmittersID) i 1 x 10-11 l l l l Sample I (I-131, Others) l l l l l l l l l l l M i i I l l Continuous (") l Part iculate l Gross alpha l 1 x 10-11 l l l l Sample l l 1 l I I I I I I I I Q l i I l l Continuous (" l Compusite l Sr-89, Sr-90 1 1 x 10-11 ?
l l l l Particulate l l l l l l Sample i l l l l l l l 1 l -l l Noble Gas l Hoble Gases Gross l 1 x 10-6 l l l Continuous (") l Monitor l Beta or Gamma 1 (Xe-133 equiv.) l 1 l 1 I I I
TABLE 4.11-2 (Continued) ., TABLE NOTATION (_ a. The Lower LimitofofSpecification Detection (LLD) is defined in Table Notation a. of Table 4.11-1 3.11.1.1. b. The principal apply gamma emitters for which the LLO specification will are exclusivel f the followin; 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, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLO level for that nuclide. When unusual circumstances result in LLO's higher than required, the reasons shall be documented in the semiannual effluent report. c. Tritium grab samples shall be taken from the Reactor Building ventilation exhaust at least once per 24 hours when the refueling canal is flooded. d. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel area, whenever spent fuel is in the spent fuel pool. e. The ratio of the sample flow rate to the sampled stream flow rate ( shall be known for the time period covered by each dose or dose rate A calculation made in accordance with Specification 3.11.2.1, 3.11.2.2, and 3.11.2.3. f. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours af ter changing (or af ter removal from the sampler). g. For certain radionuclides with low gamma yield or low energies, or i for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLO. Under these circum- ' stances, the LLO may be increased inversely proportional to the magnitude of the gamma yield (i.e., 1 x E-4/I, where I is the photon i abundance expressed as a decimal fraction), but in no case shall ! the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10 CFR 20, i Appendix B, Table II, Column I. !L ARKANSAS - UNIT 2 3/4 11-9
RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The dose due to noble gases released in gaseous effluents, from ANO-2 to UNRESTRICTED AREAS (see Figure 5.1-3) shall be.
- a. During any calendar quarter, less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation and,
- b. During any calendar year, less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times. , ACTION
- a. With the calculated dose from radioactive' noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report, submit a Special Report pursuant ,
to Specification 6.9.2.g within 30 days.
- b. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.7 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations. Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days. ARKANSAS - UNIT 2 3/4 11-10
RADI0 ACTIVE EFFLUENTS DOSE - 10 DINE-131. TRITIUM. AND RADIONUCLIDES IN PARTICULATE FO ( LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, from. tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from ANO-2 to UNRESTRICTED AREAS (see Figure 5.1-1) shall be. a. During any calendar quarter, less than or equal to 7.5 mrems to any organ and, b. During any calendar year, less than or equal to 15 mrems to any organ. APPLICABILITY: At all times. ~ ACTION: a. With the calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with~ half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.g within 30 days.
- b. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.7 cre not applicable.
I t SURVEILLANCE REQUIREMENTS 4.11.2.3 Dose Calculations. Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the 00CM at least once per 31 days. I l L i ARKANSAS - UNIT 2 3/4 11-11 L
RADI0 ACTIVE EFFLUENTS GASEOUS RA0 WASTE TREATMENT ( LIMITING CONDITION FOR OPERATION 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEMS shall be to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent doses from ANO-2 to UNRESTRICTED AREAS, (see Figure 5.1-1), would exceed .625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter; or when the projected doses due to iodine-131, tritium, ar.d radio-nuclides in particulate form with half-lives greater than 8 days would exceed 1.0 mrem to any organ over a calendar quarter. APPLICABILITY: At all times. ACTION: a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other repcrt, submit30 within a days. Special Report pursuant to Specification 6.9.2.g b. The provisions of Specifications 3.0.3, 3.0.4 , and 6.9.1.7 are not applicable. SURVEILLANCE REOUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be pro-jected at least once per 31 days in accordance with the ODCM. L ARKANSAS - UNIT 2 3/4 11-12
RADI0 ACTIVE EFFLUEtlTS GASEOUS RADWASTE TREATMENT ( LIMITING CONDITION FOR OPERATION 3.11.2.5 When degasifying the reactor coolant system, the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radio-active material in gaseous waste prior to their discharge when the projected gaseous effluent doses from ANO-2 to UNRF5TRICTED AREAS, (see Figure 5.1-1) would exceed .625 mraa for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter. APPLICABILITY: At all times. ACTION: a. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.2.g within 30 days. b. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.7 are not applicable. SURVEILLANCE REQUIREMENTS ( 4.11.2.5.1 Doses due to gaseous releases from the site shall be pro-jected at least once per 31 days in accordance with the 00CM. e ARKANSAS - UNIT 2 3/4 11-13
1 RADI0 ACTIVE EFFLUENTS GAS STORAGE TANKS ( LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 300,000 curies noble gases (considered as Xe-133). APPLICABILITY: At all times. ACTION: a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours reduce the tank contents to within the limit. -
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas I storage tank shall be determined to be within the above C limit at least once per 24 hours when radioactive materials are being added to the tank and the reactor ; i coolant activity exceeds the limits of specification 3.4.8. ,i t ? i i l l ARKANSAS - UNIT 2 3/4 11-14
f RADI0 ACTIVE EFFLUENTS 3/4.11.3 TOTAL DOSE ( LIMITING CONDITION FOR OPERATION 3.11.3 The calculated doses from the release of radioactive materials in liquid or gaseous effluents shall not exceed twice the limits of Specification 3.11.1.2.a. 3.11.1.2b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b. APPLICABILITY: At all times. ACTION: a. With the calculated doses exceeding the above limits, prepare and submit a Special Report pursuant to 10CFR Part 20.405c. s. b. If the limits of 40CFR190 have been exceeded obtain a variance from the Commission to permit further releases in excess of 40CFR190 limits. A variance is granted until staff action on the request is complete. SURVEILLANCE REQUIREMENTS ( 4.11.4 Dose Calculations. Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCM. D i ARKANSAS - UNIT 2 3/4 11-15 l E
I 3/4.12 RADIOLOGTCAL ENVIRONMENTAL MONITORING ' 3.12 The Radiological Environmental Monitoring Program and Interlabor-atory Comparison Program required by the Arkansas Nuclear One - (- Unit 1 (Docket No. 50-313, License No. DPR-51) Technical Specifi-cations shall also apply to Arkansas Nuclear One - Unit 2. e 4 ( i L ARKANSAS - UNIT 2 3/4 11-16
INSTRUMENTATION BASES 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that suffi-cient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instru-mentation for Light-Water-Cooled Nuclear Plants to Assess Plant Condi-tions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations." 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection system ensures thAt sufficient l capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Ch1crine Release," February 1975. 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. ~ This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection instrumentation is in-i operable, except for detectors located in the containment during Modes 1 and 2, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. , 3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is.provided to monitor
~
and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. t ARKANSAS - UNIT 2 B 3/4 3-3
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s s For the radioactive gaseous effluent instrumentation surveillance
., requirements, the CHANNEL FUNCTIONAL TEST demonstrates that control room alarm annunciation occurs if any 'of the following conditiens exist:
1Im 3 The , instrument indicates measured levels above the alarm / trip
- ~ setpoint.
- 2. Power to the detector is lost.
_ 3 The instrument indicates a downscale failure. For the containment purge and the waste gas holdup system noble gas actlyity monitors, the CHANNEL FUNCTIONAL TEST also demonstrates that automatic isolation of. the release pathway occurs if the instrument f, indicates above the trip setpoint.
~
-The thitial CHANNEL CALIBRATION is performed using one or more of the s reference standards certified by the National Bureau of Standards or
, using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards permit calibrating the. system over its intended range of energy and measurement i range. For subsequent CHANNEL CALIBRATION, sources that have been related
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to the initial calibration are used. 9
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ARKANSAS - UNIT 2 8 3/4 3-4 P _ . . ,_ . _ _ - - _- _ - _- . _ __ - - - _ _ , . - - . - - - - - - . - - - - - - -. - - _ _ _ - -
, t-INSTRUMENTATION BASES 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided.to monitor and control', as applicable, the releases of radioactive materials in ~
liquid effluents during actual or potential releases of , liquid effluents. The alarm / trip setpoints for these instruments shall be' calculated in accordance with the procedures in the 00CM to ensure that the alare/ trip will occur prior to exceeding the limits of 10 CFR Part 20.
~
For the radioactive liquid effluent instrumentation surveillance requirements, the channel test demonstrates that automatic isolation ' of this pathway and control root alarm annunciation occur if the instrument indicates measured levels'above the trip setpoint. The channel test demonstrates that alana annunciation occurs if any of the following conditions exist:
- 1. Power to the detector is lost.
- 2. The instrument ~ indicates'a downscale failure.
- 3. Instrument controls are not set in the operate mode.
~
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The initial CHANNEL CALIBRATION is performed using one or more of the
- reference standards certified by the National Bureau of Standards or~using
, standards that have been obtained from suppliers that participate in measurement as'urance s activities with NBS. 1 These standards permit calibrating the system over its intended range of energy cnd measurement range. For subsequent CHANNEL. CALIBRATION, sources that have been related F 'to the initial calibration are used.
-3)4.3.4 TURBINE OVERSPEED PROTECTION
_ This specification is provided to ensure ,that the turbine overspeed protection instrumentation and the turbine speed control valves are
- OPERA 8LE and will protect the turbine frou excessive overspeed. Pro-
. tection from turbine excessive overspeed is required since excessive
. overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.
e ARKANSAS - UNIT 2 B 3/4 3-5
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3/4.11 RADIOACTIVE EFFLUENTS BASES This specification applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system. 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents in unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20, 106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. 3/4.11.1.2 DOSE .( Provides assurance that releases of liquid effluents will result in con-centrations below the limits of 10CFR20. The specification provides the required operating flexibility and at the same time assures that the release of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology pro-vided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I." April 1977. L ARKANSAS - UNIT 2 B 3/4 10-2
RADI0 ACTIVE EFFLUEtlTS BASES 3/4.11.1.3 LIQUID RADWASTE TREATMENT The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials achievable". in liquid effluents will be kept "as low as is reasonably The specified limits governing the use of appropriate portions of the LIQUID RADWASTE TREATMENT SYSTEM were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. The values of .18 mrem and .625 mrem are approximately 25?! of the yearly design objectives on a quarterly basis. The yearly design objectives are given in 10CFR50, Appendix I, Section II. 3/4.11.1.4 LIQUID HOLDUP TANKS-Restricting the quantity of radioactive material contained in the speci-fled tanks provides assurance that in the event of an uncontrolled release of the contents of the tanks, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix 8, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UllRESTRICTED AREA. ( 4 ARKANSAS - UNIT 2 B 3/4 10-3
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time in UNRESTRICTED AREAS from gaseous effluents from all units on the site will be within the limits of 10 CFR Part 20.105(b). This specification applies to the release of gaseous effluents from all reactors at the site. 3/4.11.2.2 DOSE-N0BLE GASES This specification is provided to implement the requirements of Sections II.B. III. A, and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate path-ways is unlikely to be substantially underestimated. The dose calcula-tions established in the ODCM for calculating the doses due to the actual ( release rates of radioactive noble gases in gaseous effluents are consis-tent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estima-ting Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the site boundary are based upon the historical average atmospheric conditions. 1 ARKANSAS - UNIT 2 8 3/4 10-4 l
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RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.3 00SE-IODINE-131. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III. A, and IV. A of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials achievable".in gaseous effluents will be kept "as low as is reasonably The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The 00CM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision I, October 1977 and Regulatory Guide 1.111., " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical ( average atomospheric conditions. The release rate specifications for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man in the areas at or beyond the site boundary. The pathways that were examined in the development of these calculations were: .1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and
- 4) deposition on the ground with subsequent exposure of man.
3/4.11.2.4 and.5 GASEOUS RADWASTE TREATMENT ! The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is l reasonably achievable". This specification implements the requirements l l of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section 11.0 of Appendix l I to 10 CFR Part 50. The specified limits governing the use of i (s ARKANSAS - UNIT 2 8 3/4 10-5 i
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RADI0 ACTIVE EFFLUEllTS BASES 3/4.11.2.6 GAS STORAGE TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification to a quantity that is less than the quantity which provides assurance that in the event of an uncontrolled release of the contents of the tank, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest EXCLUSION AREA boundary will not exceed 0.5 rem in an event of 2 hours. Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event V an uncontrolled release of the tanks's contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest EXCLUSION AREA boundary will not exceed 0.5 rem. This is'July, NUREG-0800, consistent 1981. with Branch Technical Position ETS811-5 in 3/4.11.3 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 18525. that have now been incorporated into 10 CFR Part 20 by 46 FR The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors it is highly unlikely that the ( resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Partlevel. requirement 190 if the individual reactors remain within the reporting The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to.within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from ether nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the re-I quirements of 40 CFR Part 190, the Special Report with a request for a l variance (provided the release conditions resulting in violation of ! 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405c, is con- ! sidered to be a timely request and fulfills the requirements of 40 CFR l Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1 and 3.11.2. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in
- carrying out any operation that is part of the nuclear fuel cycle.
i i ARKANSAS - UNIT 2 8 3/4 10-6
3/6 12 RADIOLOGICAL ENVIRONMENTAL MONITORING , BASES 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING The bases for the Radiological Environmental Monitoring Program and Interlaboratory Comparison Program required by the Arkansas Nuclear One-Unit 1 (Docket No. 50-313, License No. DPR-51) Technical Specifications shall also apply to Arkansas Nuclear One - Unit 2. 4 4 1 L ARKANSAS - UNIT 2 B 3/4 10-7 _ _ - .. - ~- _. . . - - . - _ - . _ _ - - _ - - - - . _ - _ _ . . - ._. - - - _ .
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yyy FIGURE 5.1-3 ! MAXIMUM AREA BOUNDARY FOR RADIOACTIVE RELEASE CALCULATION (EXCLUSION AREAS) 1046 METER RADIUS FOR GASES AT END OF DISCHARGE CANAL FOR LIQUIDS (POINT A) l ARKANSAS - UNIT 2 5-3a
4 ADMINISTRATIVE CONTROLS
- f. Review of events requiring 24 hour written notification to the Commission.
- g. Review of facility operations to detect potential nuclear safety hazards.
- h. Performance of special reviews, investigations or analyses and reports thereon as requested by the ANO General Manager or the Safety Review Ccamittee.
- i. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the ANO General Manager and the Safety Review Committee.
- j. Review of the Emergency Plan and implementing procedures and .
shall submit recommended changes to the ANO General Manager and the Safety Review Committee.
- k. Review of proposed changes to the ODCM and PCP.
AUTHORITY 6.5.1.7 The Plant Safety Committee shall:
- a. Recommend in writing to the ANO General Manager approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
- b. Render determinations in writing with regard to whether or not l each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the Vice President, Nuclear Operations and the Safety Review Committee of disagreement between the PSC and the ANO General Manager; however, the ANO General Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The Plant Safety Committee shall maintain written minutes of each PSC meeting that, at a minimum, document the results of all PSC l activites performed under the responsibility and authority provisions of ! these technical specifications. Copies shall be provided to the ANO l General Manager and Chairman of the Safety Review Committee. l ARKANSAS - UNIT 2 6-7
ADMINISTRATIVE CONTROLS REVIEW 6.5.2.7 The SRC shall review:
- a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
- b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- d. Proposed changes to Technical Specifications or this Operating License.
- e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
- f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety.
- g. Events requiring 24 hour written notification to the Commission.
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.
- i. Reports and meeting minutes of the Plant Safety Committee.
I
- j. Proposed Changes to the ODCM and PCP.
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t l 1 l ARKANSAS - UNIT 2 6-10 l
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ADMINISTRATIVE CONTROLS AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the SRC. These audits shall encompass:
- a. The comformance of unit, operation to provisions contained within the Technical Specifications and 1pplicable license conditions at least once per 12 months.
- b. The performance, training and qualifications of the entire unit staff at least once per 12 months.
- c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems er method of operation that affect nuclear safety at least once per 6 months.
- d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B",
10 CFR 50, at least once per 24 months.
- e. The Facility Emergency Plan and implementing procedures at least once per 12 months.
- f. The Facility Security Plan and implementing procedures at least once per 12 months.
- g. Any other area of unit operation considered appropriate by the SRC or the Senior Vice President, Energy Supply (SRVP,ES).
- h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
- i. An independent fire protection and ' loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or l an outside fire protection firm.
t
- j. An inspection and audit of the fire protection and loss prevention program shall be performed by the qualified outside fire consultant at least once per 36 months.
- k. The OFFSITE DOSE CALCULATION MANUAL and Process Control Program and implementing procedures at least once per 24 months.
AUTHORITY , 6.5.2.9 The SRC shall report to and advise the Senior Vice President, Energy Supply (SRVP,ES) on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8. ARKANSAS - UNIT 2 6-11
ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. The unit shall be placed in at least HOT STANDBY within one hour.
- b. The Safety Limit violation shall be reported to the Commission, the Vice President, Nuclear Operations and to the SRC within 24 hours.
- c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PSC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the Commission, the SRC and the Vice President, Nuclear Operations within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below: f
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
I
- e. Emergency Plan implementation.
l f. Fire Protection Program implementation. l l
- g. Modification of Core Protection Calculator (CPC) addressable constants.
l NOTE: Modification to the CPC addressable constants based l I l on information obtained through the plant computer--CPC l l data link shall not be made without prior approval of the l l Plant Safety Committee. l
- h. New and spent fuel storage.
- i. OFFSITE DOSE CALCULATION MANUAL and Process Control Program implementation.
ARKANSAS - UNIT 2 6-13 I
A0filNISTRATIVE CONTROLS p 6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be ( reviewed by the PSC and approved by the General flanager prior to implementation and reviewed periodically as set forth in administrative procedures. ( s i i l I i 1 L ARKANSAS - UNIT 2 6-13a
ADMINISTRATIVE CONTROLS occurrence of the event. The written report shall include, as a minimum,
'C a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
a. Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems. b. Conditions leading to operation in the degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation. c. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems, d. Abnormal degradation of systems other than those specified in 6.9.1.8.c above designed to contain radioactive material resulting from the fission process.
- e. An unplanned offsite release of 1) more than 1 curie of
( radioactive material in liquid effluents, 2) more than 150 (. curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of rauioiodine in gaseous effluents. This report shall include the following information. -
- 1. Description of the occurrence.
2. Identify the cause(s) for exceeding the limit (s).
- 3.
Explain corrective action (s) taken to mitigate occurrence.
- 4. Define action (s) taken to prevent recurrence.
- 5. Summary of the consequence (s) of occurrence.
I I i I l l I D ARKANSAS - UNIT 2 6-18
ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the' Director of the Office of Inspection and Enforcement 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:
- a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
- b. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
- c. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
- d. Seismic event analysis, Specification 4.3.3.3.2.
- e. Inoperable Fire Detection Instrumentation, Specification 3.3.3.8.
- f. Inoperable Fire Suppression Systems, Specifications 3.7.10.1 and 3.7.10.2.
- g. Primary Coolant Specific Activity, Specification 3.4.8.
- h. Radioactive Effluents, Specifications 3.11.1.1, 3.11.1.2, 3.11.1.3, 3.11.2.1, 3.11.2.2, 3.11.2.3, and 3.11.2.4 and 3.11.2.5.
This report shall include the following:
- 1) Description of the occurrence.
- 2) Identify the cause(s) for exceeding the limit (s).
- 3) Explain corrective action (s) taken to mitigate occurrence.
- 4) Define action (s) taken to prevent recurrence.
- 5) Summary of consequence (s) of occurrence.
i l l j ARKANSAS - UNIT 2 6-18a
ADMINISTRATIVE CONTROLS 6.9.3 SEMIANNUAL RADIOLOGICAL EFFLUENT RELEASE REPORT 8 6.9.3.1 Routine radioactive effluent release reports covering the operating of the unit during the previous 6 months of operations shall be submitted within 60 days after January 1 and July 1 of each year.
- 6. S. 3. 2 The radioactive effluent release reports shall include a sum-mary of the quantities of radioactive liquid and gaseous effluents and solid waste release from the unit. The data will be summarized on a quar-terly basis following the format of Reg. Guide 1.21, Rev 0, Appendix A.
6.9.3.3 Any changes to the OFFSITE DOSE CALCULATION MANUAL and PCP shall be included in the semiannual report for the period in which the change (s) was made effective. 6.9.3.4 The radioactive effluent release reports shall include the following information for all unplanned releases to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents:
- 1. Description of the occurrence.
- 2. Identify the cause(s) for exceeding the limit (s).
- 3. Explain corrective actions taken to mitigate occurrence.
- 4. Define action (s) taken to prevent recurrence.
- 5. Summary of consequence (s) of occurrence.
3 A single submittal may be made for a multiple unit station. The-submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste system, the l submittal shall specify the releases of radioactive material from each - l unit. ARKANSAS - UNIT 2 6-18b l i - _ _ _. _ _ _ - _ _ _ _ _ . _ _ . _ _ _ _ . . _ _ _ _ _ . - _ . _ _ _ . _ _ . . _ .
ADMINISTRATIVE CONTROLS ( RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LAND USE CENSUS, INTERLABORATORY COMPARISON PROGRAM 6.14 The Radiological Environmental Monitoring ~ Program, Land Use Census, and Interlaboratory Comparison Program required by the Arkansas Nuclear One - Unit 1 (Docket No. 50-313, License No. DPR-51) Technical Specifications shall also apply to Arkansas Nuclear One - Unit 2. 6.15 10FFSITE DOSE CALCULATION MANUAL * (ODCM) FUNCTION 6.15.1 The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications. 6.15.2 Changes to the ODCM made by the Licensee shall:
- 1. Be reviewed and found acceptable by the PSC and SRC.
2.' Be submitted to the Commission by inclusion in the Semiannual Radiological effluent release report pursuant to Specification 6.9.3.3 for the period in which the change (s) was made ef-fective and shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided together with appropriate analyses or evaluations justifying the change (s);.
- b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations.
3. Shall become effective upon a date specified and agreed to by both the PSC and SRC following their review and acceptance of the change (s).
- This document is the same document as the ODCM required in the ANO-1 Technical Specifications.
L ARKANSAS - UNIT 2 6-22
ENCLOSURE 3 TO OCAN098310 REVISIONS TO ANO-1 & 2 0FFSITE DOSE CALCULATION MANUAL t L}}