ML19294A793
| ML19294A793 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/15/1979 |
| From: | Vandenburgh D Maine Yankee |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| WMY-79-19, NUDOCS 7903220181 | |
| Download: ML19294A793 (75) | |
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Proposed Change No. 68 MAME -
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ENGINEERING OFFICE WESTBORO, MASSACHUS E TTS 01581 617-3GG-9011 2
PC 68-1 B.3.2.1 WMY 79-19 March 15, 1979 United States Nuclear Regalatory Commission Washington, D. C.
20555 Attention: Office of Nuclear Reactor Regulation Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors Referencer: (a) License No. DPR-36 (Docket No. 50-309)
(b) Letter from B. K. Grimes, Assistant Director for Engineering and Projects, USNRC, to All Power Reactor Licensees, dated July 11, 1978 (c) Letter from B. K. Grimes, Assistant Director for Engineering and Projects, USNRC, to All Pressurized Water Reactor Licensees, dated November 15, 1978
Dear Sir:
Subject:
Maine Yankee Radiological Effluent Technical Specifications Pursuant to Section 50.59 of the Commission's Rules and Regulations, Maine Yankee Atomic Power Company hereby proposes the following modification to Appendix A of the Operating License.
Proposed Change: Reference is made to the Operating License DPR-36 and the Technical Specifications contained in Appendix A issued to the Maine Yankee Atomic Power Company for the Maine Yankee Atomic Power Station.
We propose to make the following changes:
- 1) Add to Section on " Definitions," new definitions for " Channel Calibration /
Channel Adjustment," " Source Check," and "Offsite Dose Calculation Manual," in accordance with the use of these terms in the new specifica-tions outlined below.
- 2) Replaca Section 3.16, " Release of Liquid Radioactive Effluents" with new Section 3.16.
The contents of this section dealing with radioactive liquid effluents has been revised to reflect the guidance put forth by the NRC in the draf t model Radiological Effluent Technical Specifications for PWR's (NUREG 0472). The proposed updated specification 3.16 addresses liquid ef fluent concentrations, resulting offsite doses from radioactive liquid effluents, the availability and use of liquid radwaste treatment equipment, and the sampling and analysis schedule for liquid waste streams.
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United States Nuclear Regulatory Conmission March 15, 1979 Attention: Office of Nuclear Reactor Regulation Page 2
- 3) Replace Section 3.17, " Release of Gaseous Radioactive Waste" with new Section 3.17.
The contents of this section dealing with radioactive gaseous effluents has been revised based on guidance put forth by the NRC in the draft model Radiological Effluent Technical Specifications for PUR's (NUREG 0472). The proposed updated specification 3.17 puts forth limits for gaseous effluent dose rates based on 10 CFR Part 20 considerations, limits on resulting offcite doses from radioactive gaseous effluents reflecting 10 CFR Fart 50, Appendix I conciderations, the availability and use of gaseous radwaste treatment equipment, the quantity of radioactivity which can be contained in a waste gas storage tank, and defines the sampling and analysis schedule for gaseous waste streams.
- 4) Insert new Section 3.25, " Radioactive Effluent Monitoring Systems."
The contents of this section concerning radioactive effluent monitoring and surveillance requirements has been included to reflect the guidance given by the NRC in the draf t model Radiological Effluent Technical Specifications for PWR's (NUREG 0472). The new specification deals with liquid and gaseous effluent monitoring instrumentation which per-forms a surveillance, protective, or controlling function on the release of radioactive effluents.
5)
Insert new Section 3.26 " Total Offsite Dose."
The contents of this specification reflect the guidance put forth in the draf t model Radio-logical Ef fluent Technical Specifications for PWR's.
It insures that the necessary dose evaluation will be done to demonstrate compliance with 40 CFR 190 with respect to doses to the general public resulting from plant operation.
- 6) Replace Section 4.8, " Operational Environmental Monitoring," with new Section 4.8, " Radiological Environmental Surveillance Program." The revision of this section reflects the guidance given in draft NURFG 0472.
The revised specification 4.8 addresses radiological environmental monitoring, annual land use census, and intercomparison program require-ments which deal with quality control of laboratory analyses performed as part of the Environmental Radiation Monitoring Program.
- 7) Insert new Section 4.13, " Surveillance of Radioactive Effluent Monitoring Systems." The contents of this section deal with the required type and frequency of surveillance of the radioactive effluent (liquid and gaseous) monitoring systems. This proposed specification is based on guidance given by NRC in NUREG 0472 and is included to insure the effluent monitor-ing systems will be maintained operable to perform their design functions.
- 8) As part of the Administrative Controls, add to Section 5.5.9, " Audits,"
the requirement to perform audits on the Offsite Dose Calculation Manual, and radiological Environmental Monitoring Program.
- 9) As part of the Administrative Controls, add to Section 5.8, " Procedures,"
the requirement to include procedures for Radiological Environmental Monitoring Offsite Dose Calculations, and effluent monitoring instrumen-tation setpoint determination in accordance with the new specifications which cover these items.
United States Nuclear Regulatory Commission March 15, 1979 Attention: Office of Nuclear Reactor Regulation Page 3
- 10) As part of the Administrative Controls, add to Section 5.9.1, " Reporting Requ?rements," a new Section titled " Annual Radiological Environmental Report," and a new Section titled " Semiannual Radioactive Effluent Release Report," in accordance with guidance put forth in the Effluent Standard Technical Specification.
- 11) As part of the Administrative Controls, add to Section 5.9.1, under
" Thirty Day Written Reports," the requirement to report the uncontrolled release of radioactive materials above given quantities and the measured levels of radioactivity in environmental sampling media in excess of given limits.
12)
In Section 5.9.2, of the Administrative Controls, add requirements for special reports under specifications 4.8.A, 4.8.B, 4.8.C, 3.16.B, 3.16.C, 3.17.B, 3.17.C and 3.17.D.
Revised Technical Specification pages are provided with this Intter.
It should be noted that pages 5.5-5, 5.8-1 and section 5.9 of the proposed changes to the Administrative Controls section are revisions to similarly numbered parts previously submitted on July 17, 1978, as part of proposed change No. 66, and.is presently awaiting Commission review and approval.
Reason for Change: The proposed changes are in direct response to the USNRC's request (References b and c) that Maine Yankee Atomic Power Company amend Maine Yankee's Operating License.
Basis for Change: The proposed technical specifications address issues put forth by the USNRC in their draft model Radiological Effluent Technical Specifications (References b and c) and are intended to implement the follow-ing Federal Regulations:
10 CFR Part 50, Section 50.36a, Section 50.34a(a),
Section 50.34a, 10 CFR Part 20, 10 CFR Part 50, Appendix A, General Design Criteria 60 and 64, and 40 CFR Part 190.
Since the solid waste disposal system at Maine Yankee complies with the regulations set forth in 10 CFR 50.34a and Appendix A Criterion 60, we believe that further commitments to your requests in Reference (c) cannot be made at this time. We would be interested in discussing this topic further with your staff, after receipt and review of the generic value-impact and specific value-impact assessment for Maine Yankee. As you are aware, NRC Chairman Joseph Hendrie committed to provide value-impacts in his July 21, 1978, response to Executive Order 12044. This response has been documented in Federal Register Vol. 43, No. 150, dated August 3, 1978, pages 34358 and 34359 which states in part:
"The Policy of the Nuclear Regulatory Commission is that value-impact analysis be conducted for any proposed regulatory actions that might impose a significant burden on the public... [and] where there are alternative means of realizing equivalent benefits... cost should be a prime consideration."
Certainly the resultant modifications of the waste disposal systems of many nuclear power plants to provide for the intended conversion of radioactive
United States Nuclear Regulatory Commission March 15, 1979 Attention: Office of Nuclear Reactor Regulation Page 4 wastes from liquid systems to a homcgeneous monolithic, immobilized free standing solid under the process centrol requirements of Reference (c) will impose a significant economic impact "... on the nuclear industry and hence on electric consumers." Lastly, we are interested in the documentation upon which your staff concluded that these proposed actions will provide substantial additional protection of the public health and safety as stip-ulated under 10 CFR 50.109.
We have included with this letter a draft of thee "Offsite Dose Calculation Manual (CDCM)" f or implementing the dose requirements of the proposed Technical Specifications. Work on this manual is continuing and a final version will be forwarded to you for your review within sixty dayc. Along with the ODCM, we have included one-line diagrams indicating the effluent flow paths for both liquid and gaseous radwaste, the location of all effluent monitors, and those equipment items which are included in the liquid and gaseous radwaste treatment system.
Safety Considerations: The changes proposed were requested by the USNRC and are not considered to constitute an unreviewed safety question. This change has been reviewed by the Nuclear Safety Audit and Review Committee.
Fee Determination: The major portion of the proposed change is an extension of the 10 CFR Part 50, Appendix I design study submitted to the USNRC on June 2, 1976, and constitutes completion of the requirements of Appendix I for the submittal of technical specifications. We conclude that this amend-ment is exempt from any fees defined in 10 CFR Part 170.12(c) since fees were not applicable when the requirements put forth by Appendix I to 10 CFR Part 50 became effecti e, and since submittal of this information has been delayed pending guidance from the USNRC which was issued in July, 1978 (Reference b).
Schedule of Change: These changes will be incorporated into Maine Yankee's Technical Specifications 90 days after approval by the Commission, but in no case b " ore November 30, 1979.
We trust this information is acceptable to you; however, should you have any questions, please contact us.
Very truly yours, MAINE YANKEE ATOMIC POWER COMPAh7 AW
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D. E. Vandenburgh Vice President Enclosures COMMONWEALTH OF MASSACHUSETTS)
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COUNTY OF WORCESTER
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March 15, 1979 Then personally appeared before me, D. E. Vandenburgh, who being duly
United States Nuclear Regulatory Commission March 15, 1979 Attention: Office of Nuclear Reactor Regulation Page 5 sworn did state that he is a Vice President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Maine Yankee Atomic Power Company, and that the statements therein are true to the best of his knowledge and belief.
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Robert H. Groce Notary Public
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.f My Commission Expires September 14, 1984 a
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MAINE YANKEE Gaseous Release Rathways Containment Air Radiation Monitor Alarm function of on line purge isolation Flow Indicator l
-e Containment Purge l
Stack Radiation Monitor l
Alarm & on line purge isolation Fuel Pool Building I
.j Stock Continuous Sampler Primary Auxiliary Building b
I RCA Storage Area Radiation Monitor - Alarm Fur;ction yt P
Condenser Air Ejectors 11 P
P P
C 11 C
a 11 C
11 Radiation Monitor - Alarm & Isolation Function Waste Gas System H
Blowdown Vent h-C 11 P
F Flow Indication d
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~ T Decay Compre-Surge Primarv Drums ssors Tank Vent IIcaders.
ent j
tack l
Ov u _ (60 dav buildup) _ _ _ __ __ __ __
C 11 P
Spray Pump Area P = Particulate Prefilter 11 = llEPA filter C = Charcoal filter
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MAINE YANKl:E Liquid Release Paths Pri:r.ary System llydror.enated Drains
- Vent to Waste Gas Surge Tank Primary Boron Filter W.iste Drain Tank Filter Degasifiers Demineralizer
- I' Evaporato Tank i
Waste Solidification d
Filter Y
\\
Eeric Acid Storage Tank Primary Systems Aerated Waste Aerated Drains Drain Liquide Test Tanks Evaporator Tanks Filter Filter w
3-Waste Solidification
/
V
'g Radiator Monitor - Ala rm Function Filter c.
Bloudown
- ~
Con d en s e r -+"--
Filter Filter
^11
- Radlation Monitor Alarm & Isolation U
Service Water Radiation O^
-Turbine Blde. Sump Monitor - Alarm Function 1
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To Seal Pit %
- Radioactive liquid waste effluent treatment
i.
3 MAINE YANKEE PROPOSED RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS
TECHNICAL SPECIFICATIONS Table of Contents Definitions Page 1.1 Fuel Storage 1.1-1 1.2 Site Description 1.2-1 1.3 Reactor 1.3 1 1.4 Containment 1.4-1 2.0 Safety Limits and Maximum Safety Settings 2.1 Limiting Safety Settings Reactor Protection System 2.1-1 2.2 Safety Limits - Reactor Core 2.2-1 2.3 Safety Limits - Reactor Coolant System Pressure 2.3-1 3.0 Limiting Conditions for Operation 3.1 Incore Instrumentation 3.1-1 3.2 Reactor Coolant System Activity 3.2-1
- 3. 3 Reactor Coolant System Operational Components 3.3-1 3.4 Combined lieat-up, Cooldown and Press-Temp. Limits 3.4-1 3.5 Chemical and Volume Control System 3.5-1 3.6 tore Cooling and Containment Spray Systems 3.6-1 3.7 Boron and Sodium Hydroxide Available for Containment Spray System 3.7-1
- 3. 8 Reactor Core Energy Removal 3.8-1 3.9 Operational Safety Instrumentation and Control Systems 3.9-1 3.10 CEA Group and Power Distribution Limits 3.10-1 3.11 Containment 3.11-1 3.'12 Station Service Power 3.12-1 3.13 Refueling Operations 3.13-1 3.14 Primary Systen Leakage 3.14-1 3.15 Reactivity Anomalies 3.15-1 3.16
_ Radioactive Liquid Kaste Release 3.16-1 3.17 Radioactive Gaseous Waste Release 3.17-1 3.18 Chemistry 3.18-1 3.19 Safety Injection Valving 3.19-1 3.20 Shock Suppressors 3.20-1 3.21 Steam Generators 3.21-1 3.22 (Reserved) 3.22-1 3.23 Fire Protection Systems 3.23 1 3.24 Secondary Coolant Activity 3.24-1 3.25 Radioactive Effluent IIonitoring Systems 3.25-1 3.26 Total Offsite Dose 3.26-1 4.0 Surveillance Requirements 4.0-1 4.1 Inst rumentation and Centro 1 4.1-1 4.2 Equipment and Sampling Tests 4.2-1 4.3 Reactor Coolant System Leak Tests 4.3-1 4.4 Containment Testing 4.4-1 Amendment No.
TECH?i1LAL SPFCIFICATIONS Table of Contents.
Definitiens Page 4.5 Energency Pctter System Periodic Testing 4.5-1 4.6 Periodic Testing 4.6-1 4.7 Reactor Coolant System Surveillance Teating 4.7-1 4.8 Radiological Environmental Surveillance Program 4.8-1 l
4.9 Shock Suppressor (Snubber) Surveillanr.a 4.9-1 4.10 Steam Generator Tube Surveillance 4.10-1 4.11 (reserved) 4.12 (reserved) 4.13 Radioactive Effluent Monitoring Systens 4.13-1 5.0 Administrative Centrols 5-1 5.1 Responsibility 5-1 5.2 Organization 5-1 5.3 Facility Staff Qualificaticas 5-5 5.4 Training 5-5 5.5 Revicu and Audit 5-6 5.6 Reportable Occurrence Action 5-12 5.7 Safety Linit violation 5-13 5.8 Procedures 5-14 5.9 Reporting Recuirctents 5-15 5.10 Record Retention 5-23 5.11 Radiation Protection Progran 5-24 5.12 High Radiation Area 5-24 Ar.enduent So.
a ENGINEERED SAFECUARDS SYSTEMS (Continued)
Subsystem (Continued)
Safeguards manual and automatic initiation.
Degree of Redundancy The dif ference between the number of operable channels and the number of channels which when tripped will cause an automatic system trip.
INSTRUMENTATION SURVEILLANCE Channel Check A qualitative determination of acceptable operability by observation of channel behavior during normal plant ope ra t ion.
This determination shall, where feasible, include comparison of the channel with other independent channels measuring the same variable.
Channel Functional Test Injection of a simulated signal into the channel to verify that it is operable, including any alarm and/or trip initiating action.
C_hannel Calibration / Channel Adjustment Adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including equipment action, alarm, interlocks or trip and shall include the channel functional test.
The channel calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
MISCELLANEOUS DEFINITIONS Operable A system or component is operable if it is capable of fulfilling its safeguard and operating functions.
Operating A system or component is operating if it is performing its safeguard or operating functions.
Amendment No.
4
i MISCELIRJEOUS DEFINITIONS (Continued)
Control Rods All full-length shutdown and regulating control elements assemblies (CEA).
Partial-Length Control Element Assemblies Control element assemblies (CEA) that contain neutron absorbing material only in the lower quarter of their length.
Containment Integrity Containment integrity is defined to exist when all of the following are true:
a.
All non-automatic containment isolation valves and blind flanges are closed.
b.
The equipment hatch is properly c losed and sealed.
c.
At least one hatch in the personnel air lock is properly closed and sealed.
d.
All automatic containment isolation valves are operable or are locked closed.
e.
The uncontrolled containment leakage satisfies Specification 4.4 Section I.B.3.
Reportable Occurrence A reportable occurrence is defined in Section 5.9.1 of these specifications.
Radio Isotope Release Limits The Maine Yankee radio isotope release limits are as defined in Technical Spec-ification 3.16, paragraph A, item 1, for liquid releases and Technical Specifi-cation 3.17, paragraph A, item 1 for gaseous releases.
FREQUENCY NOTATION The f requency notation specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 0.1.
Amendment No.
F i r e_S u_pp r e_s s_l o n__Wa t e r_ Sy s t e m A fire suppression water system shall consist of: A water source (s); gravity tank (s) or pump (s); and distribution piping with associated sectionalizing control or isolation valves.
Such valves shall include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, host standpipe or spray system riser.
OFFSITE DOSE CALCULATION MANUAL (ODCM)
A manual containing the methodology and parameters to be used in the calcul-ation of offsite doses due to radioactive gaseous and liquid ef fluents.
Amendment No.
6
TABLE 0.1 FREQUENCY NOTATION Notation Frequency S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W
At least once per 7 days M
At least once per 31 days Q
At least once per 92 days if the plant is in the cold shutdown condition SA At least once per 184 days A
At least once per year R
At least once per 18 months P
Prior to each reactor startup PR Completed prior to each release l
N.A.
Not applicable Amendment No.
0.1
3.16 RELEASE OF LIQUID RADIOACTIVE EFFLUENTS Applicability:
Applies to the controlled release of all liquid waste discharged f rom the plant which may contain radioactive materials.
Obj e ct ive :
To establish conditions for the release of liquid waste containing radioactive materials and to assure that all such releases are within the concentration limits specified in 10 CFR Part 20, and also assure that the releases of radioactive materials in liquid wastes (above background) to unrestricted areas are kept "as low as is reasonably achievable" in accordance with 10 CFR 50, Appendix I.
Specification:
A.
Liquid Effluents: Concentration 1.
The concentration of radioactive material in liquid effluents released from the site to unrestricted areas (see Figure 3.16-1) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than noble gases and 2 x 10-4 pC1/ml total activity concentration for all dissolved or entrained noble gases.
2.
With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits, immediately decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits and provide notification to the Commission pursuant to Specification 5.9.1.13.
B.
Liquid Effluents: Dose 1.
The dose commitment to an individual from radioactive materials in liquid effluents released to unrestricted areas (see Figure 3 16-1) shall be limited:
a.
During any calendar quarter to 5.1.5 mrem to the total body and/or to 5,5 mrem to any organ, and b.
During any calendar year to 3.3 mrem to the total body and/or to 5,10 mrem to any organ.
2.
With the calculated dose commitment from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and Ame nd me n t No.
3.16-1
submit to the Commission within 30 days, pursuant to Specification 5.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid ef fluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose or dose commitment to an individual from such releases during this period is within 3 mrem to the total body and 10 mrem to any organ.
C.
Liquid Waste Treatment 1.
The Boron Recovery Evaporator (for hydrogenated liquid wastes), or the Waste Disposal Evaporator (for aerated liquid wastes) shall be routinely used to reduce the radioactive materials in the liquid radwaste effluent stream prior to its discharge, to unrestricted areas; or 2.
With the appropriate evaporator unavailable for use due to maintenance, testing, or incompatibility of waste stream with equipment, the release of hydrogenated or aerated liquid waste to unrestricted areas may continue without further action provided that the resultant cumulative doses in unrestricted areas does not exceed 50% of the limits of Specification 3.16.B.l.a.
3.
With hydrogenated or aerated liquid waste streams being discharged without treatment, and with the resultant doses exceeding 50% of the limits of Specification 3.16.B.l.a. prepare and submit to the Commission within 30 days pursuant to 5.9.2, a Special Report which includes the following information:
a.
Identification of evaporator not operable and the reason for inoperability.
b.
Action (s) taken to restore the inoperable evaporator to operable status.
~
Summary description of action (s) taken to c.
prevent a recurrence.
D.
Liquid Effluents:
Sampling and Analysis Liquid radioactive waste sampling and activity analysis shall be performed in accordance with Table 3.16-1.
Amendment No.
3.16-2
Records shall be maintained and reports of the sampling and analysis results shall be submitted in accordance with Section 5.9 of these Specifications.
Basis:
A.
Liquid Effluents: Concentration 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 icvels of radioactive materials in bodies of water outside the site will not result in exposures within (1) the Section II.A design objectives of Appendix I,
10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 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.
B.
Liquid Effluents: Dose This specification is provided to implement the requirements of Sections II.A, 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.A of Appendix 1.
The specification provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid ef fluents will be kept "as low as is reasonably achievable".
In addition, since the facility is located on a salt water estuary, the release of radioactive waste in liquids will not result in radionuclide concentrations in finished drinking water which would be in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conf ormance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially unde re s t ima t ed.
Amendment No.
3.16-3
C.
Liquid Waste Treatment The maintenance and use of the Boron Recovery Evaporator and Waste Disposal Evaporator to perform their design function ensures that this system will be available to treat liquid effluents prior to release to the environment. The requirement that these systems be used routinely provides assurance that the releases of radioactive materials in liquid effluents will be kep t "as low as is reasonably achievable", while at the same time, allows flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20.
This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives of Appendix I to 10 CFR Part 50.
The action level given in Specification 3.16.C.2 and 3.16.C.3 (i.e., 50% of the limits of Specification 3.16.B.l.a) was chosen as a suitable fraction of the quarterly dose limit for liquid effluent given in 10CFR50, Appendix I, and is equivalent to the annual dose objectives of Appendix 1 for liquid effluents.
D.
Sampliny and Analysis The sampling and analysis program outlined in Table 3.16-1 provides reasonable assurance that the release limits given in Specification 3.16.A.1 are met at the discharge point to the unrestricted area at all times.
The results of the analysis program also provides the necessary isotopic release inf ormation which allows dose assessments to be performed such that compliance with Specification 3.16.B can be demonstrated.
Amendment No.
3.16-4
TABLE 3.16-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Sampling Minimum Type of Activity of Detection Liquid Release Type Frequency Analysis Analysis (LLD)
Frequency (uCi/ml)a PR PR A.
Batch Waste Re-Each Batch Each Batch Principal Gamma 5 x 10-7 b e
Emitters 8 lease Tanks I-131 1 x 10-6 Dissolved and 1 x 10-5 Entrained Gases PR H-3 1 x 10-5 Each Batch M
C Composite Gross alpha 1 x 10 P-32 1 x 10-6 PR Sr-89, Sr-90 5 x 10~
Each Batch Q
Composite Fe-55 1 x 10-6 c
B.
Plant Continuous Principal Gamma
-7b Releases f Emitters 8 5 x 10 W
W Grab Sample I-131 1 x 10-6 Dissolved and Entrained Gases 1 x 10-5 W
-5 Grab sample M
H-3 1 x 10 d
Composite
-7 Gross alpha 1 x 10
-6 1 x 10 P
Sr-89, Sr-90 5 x 10-8 d
Composit
-6 Fe-55 1 x 10 Amendment No.
3.16-5
TABLE 3.16-1 (Continued)
TABLE NOTATION a.
The lower limit of detection (LLD) is defined in Table Notation a.
of Tc51e 4.8-2 of Specification 4.8.
b.
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 ci r cums t ance s, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (i.e., 5 x 10-7/I, where I is the photon abundance expressed as a decimal f raction), 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, Appendix B, Table II, Column 2.
c.
A composite sample is one in which the quantity of liquid sampled is 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.
To be representative of the quantities ami concentrations of radioactive ma t e rials in liquid effluents, grab samples shall be collected and composited in proportion to the rate of flow of the effluent stream.
Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
c.
A batch release is the discharge of liquid wastes of a discrete volume.
f.
A continuous release is the discharge of liquid wastes of a non-discrete volume ; e.g., f rom a volume of system that has an input flow during the continuous release.
g.
The principal gamat emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144.
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 f or the analysee should not be reported as being present at the LLD level.
When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the semiannual Radioactive Effluent Release Report.
Amendment No.
3.16-6
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_m LIQUID EFFLUENT RESTRICTED AREA BOUNDARY FOR MEETING REQUIREMENTS OF MAINE YANKEE 10CFR20 APP. B, TABLE II, COL. 2.
FIG l'It E 3.16-1
3.17 _ RELEASE OF GASEOUS RADIOACTIVE WASTE Applicability:
Applies to the controlled releases of all gaseous waste discharged f rom the plant which may contain radioactive materials.
Objective:
To establish conditions in which gaseous waste containing radioactive materials may be released and to assure that all such releases are within the dose limits specified in 10 CFR Part 20, and also assure that the releases of radioactive materials in gaseous waste (above background) to unrestricted areas are kept "as low as is reasonably achievable" in accordance with 10 CFR 50, Appendix 1.
Specification A.
Gaseous Effluents: Dose Rate 1.
The instantaneous dose rate in unrestricted areas (see Figere 3.17-1) due to radioactive materials released in gaseous ef fluents f rom the site snall be limited to the following:
a.
The dose rate limit for noble gases shall be
$500 mrem /yr to the total body, and 13000 mrem /yr to the skin b.
The dose rate for all radioiodines and radioactive uaterials in particulate form and radionuclides other than noble gases with half lives greater than 8 days shall be 11500 mrem /yr to any organ.
2.
With the dose rates exceeding tne above limits, immediately decrease the release rate to comply with the limit and provide nctification to the Commission pursuant to Specification 5.9 1.13.
B.
Caseous Ef fluents: Dose from Noble Gases 1.
The air dose in unrestricted areas (see Figure 3.17-1) due to noble gases released in gaseous effluents shall be limited to the following:
During any calendar quarter to 55 mrad f or a.
gamma radiation and 110 mrad f or beta radiation; b.
During any calendar year,110 mrad for gamma radiation and 120 mrad for beta radiation.
Amendment No.
3.17-1
2.
With the calculated air dose from radioactive noble gaseous in gaseous ef fluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 5.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose during this period is within 10 mrad f or gamma radiation and 20 mrad for beta radiation.
C.
Gaseous Ef fluents:
Dose from Radiciodines, Radioactive Ma t e ria! in Particulate Form and Radionuclides Other
_Than Noble Gases 1.
The dose to an individual from radiolodines, radioactive materials in particulate form and radionuclides with half-lives greater than 8 days other than noble gases in gaseous ef fluents released to unrestricted areas (see Figure 3.17-1) shall be limited to the following:
a.
During any calendar quarter 57.5 mrem, b.
During any calendar year 115 mrem 2.
With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous ef fluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 5.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases of radioiodines, radioactive materials in particulate form, and radionuclides with half lives greater than 8 days other than noble gases in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose or dose commitment to an individual f rom such releases during the period is within 15 mrem to any organ.
D.
Caseous Waste Treatment 1.
The Gaseous Decay Holdup tank (s) and appropriate subsystem of the ventilation filter system shall be routinely used to reduce radioactive materials in gaseous waste prior to their discharge to unrestricted areas.
Amendment No.
3.17-2
2.
With the Gaseous Decay Holdup Tank (s) or appropriate subsystems of the ventilation filter system unavailable for use due to maintenance, testing, or incompatibility of waste stream with equipment, the release of gaseous vaste to the plant stack and unrestricted areas may continue without further action provided that the resultant cumulative doses in unrestricted areas does not exceed 50% of the limits of Specification 3.17.B.l.a or 3.17.C.1.a.
3.
With gaseous waste streams being discharged without appropriate filtration or holdup, and with the resultant doses exceeding 50% of the limits of Specification 3.17.B.l.a or 3.17.C.l.a. prepare and submit to the Commission within 30 days pursuant to Specification 5.9.2, a Special Report which includes the following information:
a.
Identification of equipment or subsystems not operable and the reason for inoperability.
b.
Action (s) taken to restore the inoperable equipment to operable status.
c.
Summary description of action (s) taken to prevent a recurrence.
E.
Gaseous Ef fluents: Cas Storage Tanks 1.
The quantity of radioactivity contained in a waste gas storage tank shall be limited to <9.0E+04 curies noble gases (considered as Xe-133).
2.
With the quantity of radioactive material in a waste gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> either reduce the tank contents to within the limit or provide notification to the Commission pursuant to Specification 5.9.1.13.
The written report shall include a description of activities planned and/or taken to reduce the tank contents to within the above limit.
F.
Gaseous Ef fluents: Sampling and Analysis Gaseous radioactive waste sampling and activity analysis shall be perf ormed in accordance with Table 3.17-1.
Records shall be maintained and reports of sampling and analysis results shall be submitted in accordance with Section 5.9 of these Specifications.
Amendment N
- 3.17-3
Basis:
A.
C a s e mi s Effluents: Dose Rate This specification is provided to ensure that the dose rate at anytime at the restricted area boundary from gaseous effluents will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II.
These limits provide reasonable assurance that radioactive material discharged in gaseous ef fluents will not result in the exposure of an individual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix B, Tablo II of 10 CFR Part 20 (10 CFR Part 20.106(b)).
For individuals who may at times be within the restricted area, the occupancy time will be suf ficiently low to compensate for any increase in the atmospheric diffusion factor above that at the restricted area boundary.
The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the restricted area boundary to 5500 mrem / year to the total body or to 13000 mrem / year to the skin.
These release rate limits also restrict, at all times, the corresponding thyroid dose rate above bachground to an infant via the milk-infant pathway to $1500 mrem / year for the nearest real milk animal to the plant.
B.
Gaseous Ef fluents: Dose from Noble Gases This specification is provided to implement the rquirements 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 specification provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I assure that the releases of radioactive material in gaseous ef fluents will be kept "as low as is reasonably achievable." Conformance with the guides of Appendix 1 is shown in the ODCM by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The ODC4 equations for determining the air doses at the restricted area boundary are based upon the historical average atmospheric conditions.
Amendment No.
3.17-4
C.
Dose, Radiciodines, Radioactive Material in Particulate Form and Radionuclides Other Than Noble Gases 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 f or Operation are the guides set forth in Section II.C of Appendix I.
The specification provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive materials in gaseous ef fluents will be kept "as low as is reasonably achievable." The ODCM calculational methods implement the requirements in Section III.A of Appendix I that conformance with the geides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.
These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area.
The pathways which are examined in the development of these calculations are: 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 ma n, and 4) deposition on the ground with subsequent exposure of man.
D.
Gaseous Waste Treatment The requirement that the Gaseous Decay Holdup Tank (s) and ventilation filter system be used routinely provides reasonable assurance that the releases of radioactive materials in gaseous ef fluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives of Appendix I to 10 CFR Part 50.
The action levels governing the use of appropriate portions of the gaseous radwaste Amendment No.
3.17-5
treatment system 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 for the resultant dose impact stated in Specification 3 17.B.l.a and 3.17.C.l.a correspond to one-half of the quarterly design dose objective values of Appendix I,Section II.B and II.C of 10 CFR Part 50.
E. Caseous Effluants: Gas Storage Tank Restricting the quantity of radioactivity (considered Xe-133) contained in a waste gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body e xpos u re to an individual at the nearest exclusion area boundary be below 10 CFR Part 100.
The total body gamma dose is calculated using an exclusion boundary dispersion coef ficient (x/Q) value of 5.93 x 10-4
-3 sec m and a Xe-133 total body gamma dose conversion factor of 2.94 x 10+2 F. Gaseous Effluents: Sampling and Analysis The sampling and analysis program outlined in Table 3.17-1 provides reasonable assurance that the release limits given in Specification 3.17.A are met at the restricted area boundary. The results of the sample and analysis program also provides the necessary isotopic release inf ormation which allows dose assessments to be performed such that compliance with Specifications 3.17.B and 3.17.C can be demonstrated.
Amendment No.
3.17-6
TABLE 3.17-1 P
RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM R
M E
z Sampling Minimum Lower Limit of Caseous Release Type Frequency Analysis Type of Detection (LLD)
Frequency Activity Analysis (uCi/ml)a A.
Waste Gas Storage PR PR Tank Each Tank Each Tank Principal Gamma Emitters 1x 10-4b Grab Sample B.
Containment Purge PR PR C
f 1 x 10-4b Each Putge
Each Purge Principal Gamma Emitters Grab Sample H-3 1 x 10-6
{
f C.
Plant Vent Stack Mc gc Principal Gamma Emitters 1 x 10-4b u
Grab 1
Sample H-3 1 x 10-6 gd I-131 1 x 10-12 e
Continuous Charcoal Sample I-133 1 x 10-10 f
b d
Principal Gamma Emitters 1 x 10_yg Continuouse Particulate (I-131, Others)
Sample Continuous M
Gross alpha 1 x 10-11 Composite Particulate Sample Continuous Q
Sr-89, Sr-90 1 x 10-11 Composite Particulate Sample
TABLE 3.17-1 (Continued)
TABLE NOTATION a.
The lower limit of detection (LLD) is defined in Table Notation a.
of Table 4.8-2 of Specification 4.8.
b.
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 circumstances, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (i.e., 1 x 10-4/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% of the MPC value specified in 10 CFR 20, Appendix B, Table II, Column 1.
Analyses shall also be performed following shutdown, startup, or similar c.
operational occurrence which could alter the mixture of radionuclides if continuous monitoring channels show a significant increase.
d.
Analyses shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following each shutdown, startup or similar operational occurrence which could lead to significant increases in radioiodine releases, if the reactor coolant (see Specification 4.2, Table 4.2-1) I-131 activity level is at least a factor of 10 greater than the previously measured 1-131 coolant concentration taken before the transient.
Daily sampling shall continue until either (1) the charcoal sampling shows that the radioiodine release rate has remained at, or decreased to the release rate range prior to the power transient or (2) the daily radioiodine release rates have stabilized within a factor of 3 of each other for at least 3 consecutive days. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLL's may be increased by a factor of 10.
e.
The ratio of the scmple flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.17.A, 3 17.B and 3.17.C.
f.
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 f or 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.
Amendment No.
3.17-8
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- l'1 eci, m RESTRICTED AREA AND MAINE YANKEE SITE BOUNDARY FIG Uit E 3.17-1 9
3.25 RADIOACTIVE EFFLUENT MONITORING SYSTEMS Applicability:
Applies at all times to radioactive effluent monitoring systems which perform a surveillance, protective, or controlling function on the release of rsdioactive ef fluents from the plant.
Obiective:
To assure the operability of the radioactive effluent monitoring systeas to perform their design functions.
Specification:
A.
Radioactive Liquid Effluent Instrumentation 1.
The radioactive liquid ef fluent monitoring instrumentation channels shown in Table 3.25-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.16.A.1 are not exceeded during periods of release through the pathway monitored.
2.
With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative tha1 a value which will ensure that the limits of 3.16.A.1 are met, suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.
B.
Radioactive Gaseous Effluent Instrumentation 1.
The radioactive gaseous process and ef fluent monitoring instrumentation channels shown in Table 3.25-2 shall be OPERABLE vith their alarm / trip setpoints set to ensure that the limits of Specification 3.17.A.1 are not exceeded.
2.
With a radioactive gaseous process of ef fluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits of 3.17.A.1 are met, declare the channel inoperable.
Basis:
A.
Radioact ive Liquid Ef fluent Instrumentation The radioact ive 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 are to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
Amendaent No.
3.25-1
e B.
Radioactive Caseous Ef fluent Instrumentation The radioactive gaseous ef fluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous ef fluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments are to ensure that the alarm / trip will occur prior to exceeding tne limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requiraments of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
Amendment No.
3.25-2
TABLE 3. 2 5-1 k
RADIOACTIVE LIQUID, /LUENT MONITORING INSTRUMENTATION 2c MINIMUM CllANNELS 5
INSTRUMENT OPERABLE APPLICABILITY ACTION 1.
. Gross Radioactivity Monitors Providing Automatic Termination of Release a.
Liquid Radwaste Effluent Line (1) 1 2.
Gross Radioactivity Monitors Not Providing Automatic Termination of Release a.
Service Water System Effluent Line (1) 3 Y
Ul b.
Steam Generator Blowdown Line (1) 2 3.
Flow Rate Measurement Devices a.
Liquid Radwaste Effluent Line (1) 5 b.
Discharge Canal N.A.
N.A.
c.
Steam Generator Blowdown Effluent (1) 4 Line
- During release via this pathway.
- Pump curves are utilized to estimate flow.
TABLE 3.25-1 (Continued)
TABLE NOTATION ACTION 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed f or up to 14 days, provided that prior to initiating a release:
1.
At least two independent samples are analyzed in accordance with Specification 3.16, Table 3.16-1.
2.
At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valving; o the rwis e, suspend release of radioactive effluents via this pathway.
ACTION 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided grab samples are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-7 udi/ gram; 1.
At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> wnen the specific activity of the secondary coolant is >0.01 uCi/ gram DOSE EQUIVALENT l-131.
2.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is <0.01 uCi/ gram DOSE EQUIVALENT I-131.
ACTION 3 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10- uCi/nl.
ACTION 4 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, and the secondary system has detectable radioactivity effluent releases via this pathway may continue for up to 14 days provided the flow rate is estimated at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during actual releases.
ACTION 5 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases via this pathway may continue for up to 14 days provided the flow rate is estimated at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during actual release.
Amendment No.
3 25-4
TABLE 3.25-2 g
o E.
RADI0 ACTIVE GASEGUS EFFLUENT MONITORING INSTRUMENTATION 8
S MINIMUM z
CHANNELS
?
INSTRUMENT OPERABLE APPLICABILITY ACTION 1.
Waste Gas Holdup System (a) a.
Noble Gas Activity Monitor (1) 6 b.
Effluent System Flow Rate (1) 7 2.
Containment Purge Monitoring System a.
Noble Gas Activity Monitor (1) 8 b.
Effluent System Flow Rate (1) 7 Measuring Device F
w T
w (a) Monitor provides alann and automatic isolation function.
- During releases via this pathway.
TABLE 3.25-2 (continued)
I RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION B"
MINIFUM CllANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION 3.
Plant Stack (Vent Header System) a.
Noble Gas Activity Monitor (1) 8 b.
Iodine Sampler Cartridge (1) 9 c.
Particulate Sampler Filter (1) 9 d.
Effluent System Flow Rate (1) 7 u>
'o Measuring Device r
Y e.
Sampler Flow Rate !!easuring (1) 7 Device
- During releases via this pathway.
TABLE 3.25-2 (Continued)
TABLE NOTATION ACTION 6 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided that prior to initiating the release:
independent samples of the tank's contents are 1.
At least two analyzed, and 2.
At least two technically qualified members of the Facility Staf f independently verify the release rate calculations and discharge valve lineup; O the rwis e, suspend release of radioactive effluents via this pathway.
ACTION 7 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided the flow rate is estimated at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
ACTION 8 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 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 9 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment for periods on the order of seven (7) days and analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the end of sample collection.
Amendment No.
3.25-7
3.26 TOTAL OFFSITE DOSE Applicability:
Applies to all dose contributing sources of radioactive materials (liquid, gaseous, and direct radiation) released from the plant to unrestricted areas.
Objective:
To establish reporting requirements for total dose assessments to any real individual in unrestricted areas due to all sources of radiation exposure from the plant.
Specification:
A.
Dose 1.
The dose commitment to a real individual from all station sources is limited to 5,25 mrem to the total body or any organ (except the thyroid, which is limited to 5,75 mrem) over a period of 12 consecutive months.
2.
With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3 16.B.1, 3.17.B.1, or 3.17.C.1, prepa re and submit to the Commission with 90 days, pursuant to Specification 5.9.2 a Special Report and limit the subsequent releases such that the dose or dose commitment to a real individual from all sources is limited to 5,25 mrem to the total body or any organ (except thyroid, which is limited to 575 mrem) over 12 consecutive months. This Special Report shall include an analysis which demonstrates that radiation exposures to all real individual f rom all sources (including all ef fluent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. O the rwis e, obtain a variance f rom the Commission to permit releases which exceeds the 40 CFR Part 190 Standard.
Basis:
A.
Dose This specification is provided to meet the reporting requirements of 40 CFR 190.
It is assumed that the sum of all reasonably postulable contributions f rom sources other than the immediate site will be small compared to this standard and can be ignored.
Amendment No.
3.26 -1
4.8 RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE PROGRAM Applicability:
This section applies to radiological environmental surveillance and land use census.
Obje ct ive :
To verify that plant operations have no significant radiological effect on the environment and that continued operation will not result in radiological effects detrimental to the environment. The program also shall verify that any measurable concentrations of radioactive materials related to plant operations are not significantly higher than expected based on ef fluent measurements and modeling of the environmental exposure pathways.
So e c i f icat ion :
A.
Radiological Environmental Monitoring 1.
The radiological environmental monitoring program shall be conducted as specified in Table 4.8-1 with lower limits of detection (LLD's) as specified in Table 4.8-2.
2.
With the radiological environmental monitoring program not being conducted as specified in Table 4.8-1, prepare and submit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
3.
With the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table 4.8-1 exceeding the limits of Table 4.8-3 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from receipt of the Laboratory Analyses, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 4.8-3 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Repcrt.
4.
Any change or permanent modification of any of the sample collection locations which have previously been sampled as part of the routine environmental radiological monitoring program, shall be identified and included in the Annual Radiological Operating Report.
Amendment No.
4.8-1
B.
Land Use Census 1.
An annual land use census shall be conducted to identify the location of the nearest cow and residence and the nearest garden
- goat, of the nearest greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles.
2.
With a land use census identifying a location (s) which yields a calculated dose commitment at least 15 percent greater than the values currently being calculated in Specifications 4.14.B.1 and 4.15.C.1, prepare and submit to the Commission within 30 days, pursuant to Specification 5.9.2, a Special Report which identifies the new location (s).
3.
With a land use census identifying a location (s) which yields a calculated dose commitment (via the same exposure pathway) at least 15 percent greater than at a location from which samples are currently being obtained in accordance with Specification 4.8.A.1, prepare and submit to the Commission within 30 days, pursuant to Specification 5.9.2 a Special Report which identifies the new location.
If permission from the owner to collect samples can be obtained and sufficient sample volume is available, then this new location shall be added to the radiological environmental monitoring program within 30 days.
The sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted at this time.
C.
Intercomparison Program 1.
As part of the Environmental Radiological Monitoring Program, analyses shall be performed on radioactive materials supplied as part of an Intercomparison Program which has been approved by NRC; or 2.
With analyses not being performed as re quired above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
- Broad leaf vegetation sampling may be performed at the point with the highest predicted D/Q in lieu of the garden census.
Amend me nt No.
4.8-2
Basis:
A.
Radiological Environmental Monitoring The radiological environmental monitoring 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 ef fluent 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 ef fluent measurements and modeling of the environmental exposure pathways. Program changes may be initiated based on operational experience.
A two-zone sample collection network has been established for environmental surveillance.
Samples are collected in Zone I at locations in the vicinity of the plant where concentrations of plant effluents may be detectable. These samples are compared to samples which have been collected simultaneously at locations in Zone II where the concentration of plant ef fluents is expected to be negligible. The Zone II samples provide a running background which will make it possible to distinguish significant radioactivity introduced into the environment by the operation of the plant from that introduced by weapons testing or other sources.
Marine biological monitoring stations have been established at river and bay sampling points upstream and downstream f rom the plant discharge structure and in the immediate vicinity of the discharge structure.
The function of monitoring the marine environment is to identify and determine the magnitude of any radionuclide reconcentration in the marine food chain.
Special attention is given to gamma spectrum analysis of media in order to identify and reference nuclides present in plant effluents.
The detection capabilities required by Table 4.8-2 are state-of-the-art for routine environmental measurements in industrial laboratories.
Amendment No.
4.8-3
e B.
Land Use Census This specification is provided to ensure that changes in the use of 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 than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored.
In lieu of the garden census, broad leaf vegetation samples from the site boundary in the direction sector with the highest D/Q may be substituted. The use of the maximum offsite D/Q value predicted for gaseous effluents from the plant stack will generate the maximum possible calculated dose and thus no real garden located at any other point could have a greater calculated dose commi tmen ts.
The addition of new sampling locations to Specification 4.8.A.1 based on the land use census is limited to those locations which yield a dose commitment at least fifteen percent greater thaa the calculated dose commitment at any location c:arrently being sampled.
This eliminates the unnecessary changing of the environmental radiation monitoring program for new locations which, within the accuracy of the calculation, contributes essentially the same to the dose or dose commitment as the location already sampled. The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than fifteen percent, would not be expected to result in a significant increase in the ability to detect plant effluent related nuclides.
C.
Intercompa rison Program The requirement for participation in an Intercomparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are perf ormed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
Amendment No.
4.8-4
TABLE 4.8-1 RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE PROGRAM ( (
g 5c.
{
Exposure Pathway Number of Sampling and Type and Frequency 2) and/or Sample Sample Locations Collection Frequency of Analysis n
z 1.
AIRBORNE a.
Radiciodine and 5
Continuous operation of Radioiodine canister.
Particulates sampler with sample Analyze at least once collection as required per week for I-131.
by dust loading but at least once per week.
Particulate sampler.
Analyze for gross beta radioactivity > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following filter change.
Perform gamma isotopic analysis on each sample when gross beta activity d
3 is > 0.5 pC1/m.
Perform m
1 gamma isotopic analysis on composite (by location) sample at least once per quarter.
2.
DIRECT RADIATION 8
Once per month.
Gamma dose.
At least once per month.
TABLE 4.8-1 (continued)
(1)(3)
RADIOLOGICAL ENVIRONFENTAL SURVEILLANCE PROGRAM P
5 g
Exposure Pathway Number of Sampling and Type and Frequency (2) and/or Sample Sample Locations Collection Frequency of Analysis o
z
,o 3.
WATERBORNE a.
Surface 2
Composite
- sample Gamma isotopic analysis and collected over a gross beta of each com-period of $31 days posite sample.
Tritium analysis of composite sample at least once per quarter.
b.
Ground 2
Once per quarter.
Gamma isotopic and tritium analyses of each sample.
c.
Sediment from Shoreline 1
Once per 6 months.
Gamma isotopic analysis of each sample.
Y 4.
INGESTION a.
Milk 4
Once per two weeks when Gamma isotopic and I-131 milk animals are on analysis of each sample.
pasture;**
Once per month at other times.
Composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Grazing season extends from June 1 to November 1.
TABLE 4.8-1 (continued)
RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE PROGRAM (
I e
Exposure Pathway Number of Sampling and a
Type and Frequency ( )
f and/or Sample Sample Locations Collection Frequency of Analysis n
p b.
Fish and 2
One sample in season.
Gamma isotopic analysis Invertebrates One sample of each of on edible portions.
at least two of the following species:
1.
Alewife 2.
Winter flounder 3.
Tomcod 4.
Mussel 5.
Lobster 6.
Crab c.
Vegetation and 3
At time of harvest.
Gamma isotopic analysis Food Products One sample of any of on edible portion.
e the following classes 2
of food products:
0 1.
Fruit 2.
Tuberous vegetable 3.
Above ground vege-table 1
At time of harvest.
I-131 analysis.
One sample of any broad leaf vegetation.
(1)
Specific sample locations for all media are specified in the Offsite Dose Calculation Manual and reported in the Annual Radiological Environmental Operating Report.
(2) See Table 4.8-2 for maximum values for the Lower Limits of detection.
(3)
Deviations are permitted from the required sampling schedule if specimen are unobtainable 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.
TABLE 4.8-2 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)#
i a,
e f
Airborne Particulate Water or Gas Fish Milk Food Products Sediment 3
Analysis (pCi/l)
(pCi/m )
(pCi/kg, wet)
(pCi/1)
(pci/kg, wet)
(pCi/kg, dry) b gross beta 4
1 x 10~
3 2000 g
54 15 130 Mn 59 30 260 pe 58,60 15 130 Co T
65 30 260 Zn cc 95 15 Zr 131 7 x 10-2 1
60c 7
134,137 15, 18 1 x 10-130 15 80 150 Cs d
d 140 15 15 Ba
TABLE 4.8-2 (Continued)
TABLE NOTATION a-The LDD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability and that only a 5% probability exists of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4.66 Sb 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) sh is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute)
E is the counting ef ficiency (as counts per transformation)
V is the sample size (in units of mass or volume) 2.22 is the number of transformation per minute per picocurie Y is the f ractional radiochemical yield (when applicable)
A is the radioactive decay constant for the particular rad ionuclide At is the elapsed time between sample collection (or end of the sample collection period) and time of counting.
The value of sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.
In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples).
Ame nd me n t No.
4.8-9
TABLE 4.8-2 (Continued)
TABLE NOTATION Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable.
In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.
b - LLD for surface water.
c - LLD for leafy vegetables.
d - The Ba-140 LLD and concentration can be determined by the analysis of its short-lived daughter product La-140 subsequent to an 8 day period following collection.
The calculation shall be predicted on the normal ingrowth equations for a parent-daughter situation and the assumption that any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6 percent of its original value). The ingrowth equations will assume that the supported La-140 activity at the time of collection is zero.
e - If the measured concentration minus the 5 sigma counting statistics is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.
Amendment No.
4.8-10
TABLE 4.8-3 g
REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES i
2er x
Water Airborne Particulate Fish Milk Vegetables
?
Analysis (pCi/l) or Gases (pCi/m )
(pci/Kg, wet)
(pCi/1)
(pCi/Kg, wet) 3 0
11 - 3 3 x 10 Mn-54 1x 10 3 x 10 Fe-59 4 x 10 1x 10
3 co-58 1 x 10 3 x 10' 2
Co-60 3 x 10 1 x 10 2
Zn-65 3 x 10 2 x 10 n
2 m
Zr-95 4 x 10 b
~
I-131 0.9 3
1 x 10 Cs-134 30 10 1 x 103 3
60 1 x 10 3
Cs-137 50 20 2 x 10 70 2 x 103 2
2 Ba-140 2 x 10 3 x 10
4.13 SURVEILLANCE OF RADI0 ACTIVE EFFLUENT MONITORING SYSTEMS Applicability:
Applies to the required surveillance of the radioactive effluent (liquid and gaseous) monitoring systems.
Objective:
To specify the type and frequency of surveillance to be applied to the radioactive effluent monitoring systems.
Specification:
A.
Radioactive Liquid Ef fluent Instrumentation 1.
Each radioactive liquid ef fluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHECK, SOURCE CHECK, INSTRUMENT CALIBRATION, and INSTRUMENT FUNCTIONAL TEST operations at the frequencies shown in Table 4.13-1.
2.
Setpoints and setpoint calculetions shall be traceable to ensure that the limits of Specification 3.16.A.1 are met.
B.
Radioactive Gaseous Ef fluents Instrumentation 1.
Each radioactive gaseous process or ef fluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHECK, SOURCE CHECK, INSTRUMENT CALIB RATION, INSTRUMENT FUNCTIONAL TEST operations at the frequencies shown in Table 4.13-2.
2.
Setpoints and setpoint calculations shall be traceable to ensure that the limits of Specification 3.17.C.1 a re met.
Basis:
A.
Radioactive Liquid Ef fluent 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. The alarm / trip setpoints for these instruments are calculated to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
Amendment No.
4.13-1
TABLE 4.13-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUFENTATION SURVEILLANCE REQUIRESENTS 2
CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST l.
Gross Beta or Gamma Radioactivity Monitors Providing Alarm and Auto-matic Isolation a.
Liquid Radwaste Ef fluent Line D*
PR R(3)
Q(1) 2.
Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation (0) a.
Service Water System Effluent Line D*
M R(3)
Q(2) b.
Steam Generator Blowdown I
Line D*
PR R(3)
Q(1)
C da 3.
Flow Rate Measurement Devices a.
Liquid Radwaste Effluent Line D(4)
N.A.
R Q
b.
Steam Generator Blowdown Line D(4)
N.A.
R Q
c.
Discharge Canal N.A.
N.A.
N.A.
- During releases via this pathway.
- Pump curves are utilized to estimate flow.
TABLE 4.13-2 f
RADIOACTIVE CASE 0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E.
!c CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL f
INSTRUMENT CHECK CHECK CALIBRATION TEST 1.
Waste Gas Holdup System a.
Noble Gas Activity Monitor PR*
PR R(3)
Q(1) b.
System Effluent Flow Rate Measuring Device PR*
N/A R
Q 2.
Containment Purge Vent System a.
Containment Noble Gas Activity Monitor D*
M R(3)
Q(2) b.
System Effluent Flow Rate D*
N/A R
Q L
Measuring Device Sampler Flow Rate Measuring D*
N/A R
Q c.
Device 3.
Plant Stack (Vent Header System) a.
Noble Gas Activity Monitor D*
M R(3)
Q(2) b.
Iodine Sampler D*
N.A.
N.A.
N.A.
c.
Particulate Sampler D*
N.A.
N.A.
N.A.
d.
System Effluent Flow Rate Measurement Device D*
N.A.
R Q
e.
Sampler Flow Rate Measurement Device D*
N.A.
R Q
- During releases via this pathway.
TABLES 4.13-1 and 4.13-2 TABLE NOTATION (1)
The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
1.
Ins t rumen t indicates measurec. levels above the alarm / trip setpoint.
2.
Ins t rume n t indicates a downscale failure.
3.
Ins t rume nt controls not set in operate mode.
(2)
The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions, exist:
1.
Ins t rume nt indicates measured levels above the alarm / trip setpoint.
2.
Instrument indicates a downscale failure.
3.
Ins t rume n t controls not set in operate mode.
(3)
The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation shall be performed using one or more of the ref erence 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 shall pe rmit calibrating the system over its intended range of energy and rate capabilities.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used, at the refueling interval.
(4)
CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.
Amendmen t No.
4.13-5
B.
Radioactive Gaseous Ef fluent 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.
The alarm / trip setpoints for these instruments are calculated to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
Amendmen t No.
4.13-2
h.
Reports and meeting minutes of the Plant Operation Review Committee.
1.
Perf orm special reviews and investigations and render reports thereon as requested by the Assistant Vice President of Engineering and Operations or to his delegated alternate.
9.
AUDITS Audit s of facility activities shall be performed under the cognizance of the NS AR Committee. These audits shall encompass:
a.
The conformance of f acility operation to provisions contained within the Technical Specifications and applicable; license conditions at least once per 12 months.
b.
The performance, training and qualification of those members of the f acility staf f who have a direct relationsaip to operation, maintenance or technical aspects of the plant, at least once per 12 months.
c.
The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or 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, a t least once per 24 months.
The Facility Emergency Plan and implementing procedures at least e.
once per 24 months.
f.
The Facility Security Plan and implementing procedures at least once per 24 months.
g.
The Facility Fire Protection Program and implementary procedures at least once per 24 months.
h.
An independent fire protection and loss prevention inspection and audit shall be perf ormed annually utilizing either qualified of f site 1.icensee personnel or an outside fire protection firm.
1.
An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.
j.
T..e OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.
k.
The radiological environmental monitoring program and the results thereof at lcast once per 12 months.
1.
Any other area of facilit; operation considered appropriate by the NSAR Committee or the Vice President.
10.
AUTHORITY The NSAR Committee shall report to and advise the Vice President on those areas of responsibility specified in Sections 5-B-8 and 5-B-9.
Amendment No.
5.5-5
5.8 PROC EDURES 5.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
The applicable procedures recommended in Appendix "A" of Regulatory a.
Guide 1. 33, Novembe r, 1972.
b.
Refueling operations.
c.
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
Emergency Plan implementation.
e.
f.
Fire Protection Program implementation.
g.
Offsite Dose Calculationals.
h.
Radiological Environmental Monitoring.
1.
Ef fluent monitoring instrumentation setpoint determinations.
5.8.2 Each procedure of 5.8.1 above, and changes thereto, shall be reviewed by the PORC and approved by the (Plant Manager) prior to implementation and reviewed periodically as set forth in administrative procedures.
5.8.3 Temporary changes to procedures of 5.8.1 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 management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
The change is documented, reviewed by the PORC and approved by c.
the Plant Manager within 14 days of implementation.
Amendment No.
5.8-1
5.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 5.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regalations, the following reports shall be submitted to the Director of the Regional Of fice of Inspection and Enforcement unlass otherwise noted.
_S,TARTUP REPORT 5.9.1.1 A summa ry report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a d if ferent design or has been manufactured by a different fuel supplier, and (4) madifications that may have significantly altered the nuclear, thermal, or hydraulic perf ormance of tha plant.
5.9.1 2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictiona and specifications. Any corrective actions that were required to obtain satisf actory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
5.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticr'i(
whichever is earliest.
If the Startup Report does not cover all three events (i.e.,
ini *al criticality, completion of startup test program, and resumption or commencement of comme rcial powe r ope ration), supplementary reports shall be subnitted at least every three months until all three events have been completed.
ANNUAL REPORTS 5.9 1.4 Annual reports covering the activities of the unit as des:ribed below for the previous calendar year shall be submitted prior to March 1 of zach year.
The initial report shall be submitted prior to MarCa 1 of the year following initiil criticality.
5.'.l.5 Reports required on an annual basis shall include:
a.
A tabulation on an annual basis of the number of station. utility and ocher personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according
.o work and job functions,1*
e.g.,
reactor operations and surveillance, inse rvice inspection, routine maintenance, special maintenance (dr scribe maintenance),
waste processing and refueling. The dose assignment to 7arious duty f unctions may ba 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 body dose received from external sources shall be assigned to specific major work functions.
1*This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
Arendment No.
5.9-1
ANNUAL RADIOLOGICAL ENVIRONMENTAL REPORT 3 9.1.6 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May I of the year following initial criticality.
5.9 1.7 The annual radiological environmental operating reports shall include summa r ie s, i n t e rp re t a t ions, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational c on t rols (as appropriate), and previous environmental aurveillance reports and an assessment of the observed impacts of the plant operation on the e nvi r onme n t.
The reports shall also include the results of the land use censuses required by Specification 4.8.B.
If harmful ef fects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.
The annual radiological environmental operating reports shall include summarized and tabulated results of all radiological environmental 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.
The reports shall also include the following: a summary description of the radiological environmental monitoring program including a map of all sampling locations keyed to a table giving distances and directions from the reactor; the result of land use censuses required by the Specification 4.8.B; and the results of licensee participation in the Intercomparison Program required by Specification 4.8.C.
SEMIANNUAL RADIOACTIVE EFFLUENT RELEASL REPORT 5.9.1.8 Routine radioactive ef fluent release reports covering the operating of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. The period of the f irs t report shall begin with the data of initial criticality.
5.9.1 9 The radioactive ef fluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released f rom the unit with data summarized on a quarterly basis.
Amendment No.
5.9-2
The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter including doses from uncontrolled releases.
In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated. The meteorological conditione concurrent with the releases of effluents shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be perf ormed in accordance with the Of f site Dose Calculation Manual (ODCM).
Amendment No.
5 9-3
MONTHLY REACTOR OPERATING REPORT 5.9 1 10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Of fice of Management and Program Analysis, U.S. Nuclear Regulatory Commissior., Washington, D.C.
20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.
REPORTABLE OCCURRENCES 5.9.1 11 The REPORTABLE OCCURRENCES of Specification 5.9 1.12 and 5 9 1.13 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. ' Supplemental reports may be required to fully describe final resolution of occurrence.
In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 5.9.1.12 The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Of fice, or his designate no later than the f irs t working day following the event, with a written followup report within 14 days. The written followup report shall include, as a minimum, a completed copy of a licensee event report form.
Information prov!ded 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.
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.
b.
Operation of the unit or af fected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications.
Abnormal degradation discovered in fuel cladding, reactor coolant c.
pressure boundary, or prima ry containment.
Amendment No.
5.9-4
d.
Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1% k/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if suberitical, an unplanned reactivity insertion of more than 0.5% k/k; or occurrence of any unplanned criticality.
Failure or malf unction of one or more components which prevents e.
ar could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.
f.
Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
g.
Conditions arising f rom natural or man-made events that, as a direct result of the event, require unit shutdown, operation of safety systems, or other protective measures required by technical specifications.
h.
Errors discovered in the transient or accident analyses or in the cethods used for such analyses as described in the safety analysie report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less cons e rva t ive than assumed in the analyses.
1.
Performance of structures, systems, or components that requires remedial act ion or corrective measures to prevent operation in a manner less conservative than assumed in the c2cident analysis in the safety analysis report or technical specification bases; or discovery during unit life of conditions not specifically considered in the saf ety analysis repo rt or technical specifications that reqeire remedial action or corrective measures to prevent the existence or development of an unsafe condition.
Amend me n t No.
5.9-5
TiiIRTY DAY WRITTEN REPORTS 5.9.1 13 The types of events liste. belcw shall be the subject of written reports to the Director of th( Regional Of fice within thirty days of occurrence of the event. 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 waterial to provide complete explanation of the circumstances surrounding the event.
Reactor protection system or engineered safety feature instrument a.
settings which are found to be less conservative than those established by the technical specifications but which do not prevent the f ulf illment of the functional requirements of af fected systems.
b.
Conditions leading to operation in a degraded mode permitted by limiting condition for operation or plant shutdown required by a
a limiting condition for operation.
c.
Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduct ion of degree of redundancy provided in re actor protection systems or engineered safety feature systems.
d.
Abnormal degradation of systems other than those specified in 5.9.1.12.c abeve designed to contain radioactive material resulting f rom the fission process.
An uncontrolled offsite release of 1) more than 1 curie of e.
radioactive material in liquid ef fluents, 2) more than 150 curies of noble gas in gaseous ef fluents, or 3) more than 0.05 curies of radiciodine in gaseous ef fluents.
f.
Occurrence of radioactive material contained in gaseous holdup tanks in excess of that permitted by the limiting condition for operation established in the technical specifications.
Amendment No.
5.9-6
f.
Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 4.8-3 when averaged over any calendar quarter sampling period. When more than one of the radionuclides in Table 4 8-3 are detected in the sampling medium, this report shall be submitted if:
concentration (1) concentration (2) +...> 1.0
+
limit level (1) limit level (2)
When radionuclides other than those in Table 4.8-3 are detected and are the result of plant ef fluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.16.B.1, 3.17.B.1 and 3.17.C.1 This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
5.9.2 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.
a.
Reactivity anomalies, Specification 3.15.
b.
Radiological Environmental Monitoring, Specifications 4.8.A, 4.8.B and 4.8.C.
c.
Plans for restoration of 115 KV service, Specification 3.12.
d.
Containment Type A test failure, Specification 4.41C2.
e.
Integrated leakage rate test report, Specification 4.411I.
f.
Liquid Effluents, Specifications 3.16.B and 3.16.C.
g.
Gaseous Ef fluents, Specifications 3.17.B, 3.17.C and 3.17.D.
Amendment No.
5.9-7
O 4
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=I MAINE YANKEE ATOMIC POWER COMPANY OFFSITE DOSE CALCULATION MANUAL
DRAET q
[g hbE b TABLE OF CONTENTS jection Page e
1.0 INTRODUCTION
2.0 LIQUID RELEASE DOSE CALCULATIONS 2.1 Technical Specification 3.16 B, Dose to an Individual 2.1.1 Method I 2.1.1.1 Dose to the Total Body 2.1.1.2 Dose to the Critical Organ 2.1.1.3 Application of Method I 2.1.2 Method II 3.0 GASEOUS RELEASE DOSE CALCULATIONS 3.1 Technical Specification 3.17 A, Dose Rate Limits 3.1.1 Method I 3.1.1.1 Dose Rate Due to Noble Gases 3 1.1.1.1 Dose Rate to the Total Body 3.1.1.1.2 Dose Rate to the Skin 3.1.1.2 Dose Rate to the Critical Organ Due to Radioiodines and Particulates 3.1.1.3 Application of Method 1 3.1.2 Method II
4 DRAEI A
TABLE OF CONTENTS (continued)
S_ection Page 3.2 Technical Specifications 3.17 B, Dose to Air from Noble Cases 3.2.1 Method I 3.2.1.1 Air Dose Due to Gamma Radiation 3.2.1.2 Air Dose Due to Beta Radiation 3 2.1.3 Application of Method I 3.2.2 Method II 3.3 Technical Specification 3.17 C, Dose to an Individual 3.3.1 Method I 3.3.1.1 Dose to the Organ 3.3.1.2 Application of Method I 3.3.2 Method II 4.0 METEOROLOGY 5.0 ENVIRONMENTAL MONITORING APPENDIX I - Basis for the Dose Calculation Methods APPENDIX II - Basis for the Atmospheric Dilution Factors REFERENCES
DRAEI m
1.0 INTRODUCTION
The purpose of this manual is to provide methods to insure compliance with the dose requirements of the Technical Specifications.
Each method is based on a plant specific application of the models presented in Regulatory Guide 1.109.(1)
Methods are included to calculate the doses to individuals from both gaseous and liquid releases from the plant.
Under normal operations, experience has shown that the plant will be operated at a small fraction of the dose limits imposed by the Technical Specifications. For this reason the dose evaluations are presented at different levels of sophistication. The first method being the most conservative, but simplest to use; the second method requires a full analysis following the guidance presented in Regulatory Guide 1.109 The first is based on a critical organ, critical age group and as such it provides a conservative estimate of the doses required by the Technical Specifications.
If the Technical Specifications are met by application of the first method, no further analysis will be required.
If, however, it indicates that the Technical Specification limits are being approached, a more realistic estimate may be obtained by application of the second method. The second method will calculate the dose to seven organs of four age groups and is based on measured releases for each nuclide.
This method will be used to assess doses for the Semi-annual Radicactive Effluent Release Report. The basis for each of the dose calculation methods is described in Appendix I.
><e j\\
16 02 2.0 LIQUID RELEASE DOSE CALCULATIONS 2.1 Technical Specification 3.16 B, Dose to an Individual This section is to be used to insure compliance with the following Technical Specification:
LIMITING CONDITION FOR OPERATION The dose commitment to an individual from radioactive materials in liquid ef fluents released to unrestricted areas shall be limited:
a.
During any calendar quarter to 11.5 mrem to the total body and to 15 mrem to any organ, and b.
During any calendar year to 13 mrem to the total body and to 110 mrem to any organ.
The dose commitment to any individual from liquid releases is proportional to the quantity (curies) to which that individual is exposed.
The following equations shall be used to calculate the dose commitment resulting from a liquid release (in curies) from the Maine Yankee Station.
The specification requires a monthly evaluation, however the following equations can be applied for any duration of release.
l.\\
j/ hj-ON di 2.1.1 Method I 2.1.1.1 Dose to the Total Body The dose to the total body is:
Dtb(mrem) 0.0023 Q58Co + 0.05 Q60Co + 0.03 Q134Cs + 0.03 Q137Cs
=
where:
Q58Co = Cobalt-58 Release (C1)
Q60Co = cobalt-60 Release (C1)
Q134Cs - Cesium-134 Release (C1)
Q137Cs= Cesium-137 Release (C1) 2.1.1.2 Dose to the critical Organ The dose to the critical organ is:
organ (mrem) = 0.4 Q60Co + 0.13 Qi ny + 0.72 Q134Cs + 0.51 Q137Cs D
where:
- Cobalt-60 Release (C1)
Q60Co
= I dine-131 Release (C1)
Q1311 Q134Cs = Cesium-134 Release (C1)
Q137Cs - Cesium-137 Release (C1)
,9 c, 7]. !~
j ;, ;,
,r
{ j 'l.
- ]
3 221.1.3 Application of Method I Step 1.
Determine the number of curies of Cesium 134, Cesium 137, Iodine 131, Cobalt-58 and Cobalt-60 released during the period.
Step 2.
Perform the above multiplications to obtain the total body and organ doses for the period.
Step 3.
Record the total body and organ doses and maintain a cumulative dose f or the annual and quarterly periods.
2.1.2 Method II The dose calculated shall be in conformance with Regulatory Guide 1.109 and may be calculated by the computer program IDLE using site specific parameters applicable during the period of release.
. ~ ~
3.0 GASEOUS RELEASE DOSE CALCULATIONS 3.1 Technical Specification 3.17 A, Dose Rate Limits This section is to be used to insure compliance with the following Technical Specification:
LIMITING CONDITION FOR OPERATION The instantaneous dose rate in unrestricted areas due to radioactive materials released in gaseous effluents from the site shall be limited to the following:
a.
The dose rate limit for noble gases shall be 1500 mrem /yr to the total body, 53000 mrem /yr to the skin b.
The dose rate limit for all radiciodines and radioactive materials in particulate form and radionuclides other than noble gases with half lives greater than 8 days shall be 11500 mrem /yr to any organ.
3.1.1 Method I 3.1.1.1 Dose Rate Due to Noble Gases 3.1.1.1.1 Dose Rate to the Total Body The total body dose rate due to noble gases can be determined as h2AH
,\\ k A j follows:
.:d b ( rem) = 2.51 x 106 tb Qi DFBi yr 1
where:
h
= Release rate for each nuclide shown in Table 4.1.( C1) i see DFBi - Total body dose f actor for each nuclide shown in Table 4.1 3.1.1.1.2 Dose Rate to the Skin The skin dose rate due to noble gases can be determined as follows:
bskin ( yr ) " btb + 4.33 x 10 6
h DFS i
i i
where:
btb = Dose rate to the total body, as determined in 3.1.1.1.1.(mrem) yr h
= Release rate for each nuclide shown in Table 3.1.( C1) 1 see DFSi = Skin dose f actor for each nuclide shown in Table 3.1 3.1.1.2 Dose Rate to Critical Organ due to Radioiodines and Particulates The dose rate to the critical organ can be determined as follows:
b rem) = 5.92 x 108y organ ( y where:
Q1311 = Release rate f or Iodine-131 (Ci/sec).
,,,. q. d
~-
gg U'
3.1.1.3 Application of Method I Step 1.
Determine the release rate in Ci/sec for Iodine-131 and for each noble gas detected.
Step 2.
Find the dose rate to the total body by multiplying each noble gas release rate by the constant and by its total body dose f actor determined f rom Table 3.1 and summing the contributions f rom each nuclide.
Step 3.
Find the dose rate to the skin by multiplying each noble gas release rate by the constant and by its skin dose factor determined from Table 3.1 and summing the contributions f rom each nuclide.
Add, to this sum, the total body dose rate determined in Step 2 to obtain the total skin dose rate.
Step 4.
Find the dose ste to the critical organ by multiplying the Iodine-8 131 release rate by 5.93 x 10,
3.1.2 Method II The dose rates calculated will follow the guidance presented in Regulatory Guide 1.109.
The noble gas code rates may be calculated by the computer program IDLE and the dose rate due to radioiodines and particulates may be calculated by the computer program ATMODOS.
f-Table 3.1 Noble Cases and Their Dose Factors (To be used for Technical Specification 3.17 A)
Dose Factor, Total Body Dose Factor, Skin 3
3 Nuclide DFB (mrem-m )
DFS (mrem-m )
f t
pCi-yr pCi-yr Kr-83m 7.56E-08*
Kr-85m 1.17E-03 1.46E-03 Kr-85 1.61E-05 1.34E-03 Kr-87 5.92E-03 9.73E-03 Kr-88 1.47E-02 2.37E-03 Kr-89 1.66E-02 1.01E-02 Kr-90 1.56E-02 7.29E-03 Xe-131m 9.15E-05 4.76E-04 Xe-133m 2.51E-04 9.94E-04 Xe-133 2.94E-04 3 06E-04 Xe-135m 3.12E-03 7.11E-04 Xe-135 1.81E-03 1.86E-03 Xe-137 1.42E-03 1.22E-02 Xe-138 8.83E-03 4.13E-03 Ar-41 8.84E-03 2.69E-03 7.56E 7.56 x 10-8
'A
-3
~
s 7 u s
3.2 Technical Specification 3.17 B, Dose to Air Due to Noble Cases This section is to be used to insure compliance with the following Technical Specification:
LIMITING CONDITION FOR OPERATION The air dose in unrestricted areas due to noble gases released la gaseous ef fluents shall be limited to the following:
a.
During any calendar quarter to; 55 mrad for gamma radiation, and $10 mrad for beta radiation; b.
During any calendar year; to 110 mrad for gamma radiation, and $20 mrad for beta radiation.
3.2.1 Method I 3.2.1.1 Air Dose Due to Camma Radiation D(mrad) = 0.103 Qi(C1) DFi where:
= Number of curies of noble gas nuclide "1" released.
Qi DF
- Gamma dose factor to air for nuclide "i".
See Table 3.2.
i 3.2.1.2 Air Dose Due to Beta Radiation D(mrad) = 0.137 Qi DF i d
h_
where:
= Number of curies of noble gas nuclide "i" released.
Qi DF
- Beta dose factor to air for nuclide "i". See Table 3.2 i
3.2.1.3 Application of Method I Step 1.
Determine the number of curies released during the period for each noble gas detected.
Step 2.
Find the dose to air due to gamma radiation by multiplying each noble gas released by the constant and by its gamma dose factor determined from Table 3.2.
Sum the contributions from each nuclide to find the total air dose from gamma radiation.
Step 3.
Find the dose to air due to beta radiation by multiplying each noble gas release by the constant and by its beta air dose factor determined from Table 3.2.
Sum the contributions from each nuclide to find the total air dose from beta radiation.
3.2.2 Method II The dose calculated shall follow the guidance of Regulatory Guide 1.109 and may be calculated by the computer program AIRAD using the meteorological dispersion parameters applicable during the periods of release.
DRAFT Table 3.2 Noble Cases and Dose Factor to Air (to be used for Technical Specification 3.17 B)
Beta Air Dose Gamma Air Dose Factor Factor 3
3 Nuclide DF (mrad-m )
i(mrad-m )
i p01-yr pC1-yr Kr-83m 2.88E-04*
1.93E-05 Kr-85m 1.97E-03 1.23E-03 Kr-85 1.95E-03 1.72E-05 Kr-87 1.03E-02 6.17E-03 Kr-88 2.93E-03 1.52E-02 Kr-89 1.06E-02 1.73E-02 Kr-90 7.83E-03 1.63E-02 Xe-131m 1.llE-03 1.5bE-04 Xe-133m 1.48E-03 3.27E-04 Xe-133 1.05E-03 3.53E-04 Xe-135m 7.39E-04 3.36E-03 Xe-135 2.46E-03 1.92E-03 Xe-137 1.27E-02 1.51E-03 Xe-138 4.75E-03 9.21E-03 Ar-41 3.28E-03 9.30E-03 2.88E 2.88 x 10-4 tw q.;T m
7, 3.3 Technical Specification 3.17 C, Dose to an Individual This section is to be used to insure compliance with the following Technical Specification:
LIMITING CONDITION FOR OPERATION The dose commitment to an individual from radioiodines, radioactive materials in particulate form and radionuclides with half-lives greater than 8 days other than noble gases in gaseous effluents released to unrestricted areas shall be limited to the following:
a.
During any calendar quarter 57.5 mrem, b.
During any calendar year 115 mrem 3.3.1 Method I To insure that the dose limit to any organ is met, it is necessary to calculate the dose to the critical organ. The following equation is applicable:
3.3.1.1 Dose to the Organ organ (mrem) = 2.2 Q 0Co + 18.8 Q131I + 0.2 Q1331 + 6.1 Q137Cs D
6
'w r.
~ ~ - J iJ ' La u l
~
where:
Q60Co - Cobalt-60 Release (C1)
= I dine-131 Release (C1)
Q131I
= I dine-133 Release (C1)
Q133I Q137CS - Cesium-137 Release (C1) 3.3.1.2 Application of Method I Step 1.
Determine the number of curies of Iodine-131 released during the period.
Step 2.
Find the thyroid dose by performing the multiplication.
Step 3.
There will be 0.051 mrem / month dose to the critical f rom Carbon-14.
The correct fraction of this dose must be added to the dose determined in Step 2.
Step 4.
Record the thyroid dose and maintain a cumulative dose for the quarterly and annual periods.
Step 5.
If the calculated dose exceeds the Technical Specification, the dose may be calculated by Method II.
3.3.2 Method II The dose calculated shall be in conformance with Regulatory Guide 1.109 and may be calculated by the computer program ATMODOS using the meteorological parameters applicable during the periods of release.
4 DRAFT f
J
)
4.0 METEOROLOGY 4
Annual average dilution factors based on site meteorological data were computed for routine releases by the AEOLUS2 computer program. The values used are shown in Table 4.1.
a TABLE 4.1 s
Maine Yankee Dilution Factors Dose Rate Equations
- Dose Equations *
- 3 X/Q(sec/m )
4.33E-6 5.5E-7 2
D/Q(1/c )
1.97E-8 2.8E-9 3
[X/Q) (sec/m )
3.24E-6 9.4E-7 Noble gases only over Bailey's Cove, maximum concentration at 100 meters South-Maximum permanently occupiable ground level concentrations for all pathways 1220 meters North.
,b y)j='d,
etylh "M
5.0 ENVIRONMENTAL MONITOR"G The Radiological Environmental Monitoring stations are listed in Table 5.1.
The location of these stations with respect to the Maine Yankee facility are shown on topographic maps in Figure 5.1.
TABLE 6/
RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS
- EXPOSURE PATHWAY SAMPLE LOCATION DISTANCE FROM DIRECTION FROM AND/OR SAMPLE AND DESIGNATED CODE +
THE PLANT (kn)
THE PLANT 1.
AIRBORNE a.
Radioiodine and Particulate AP/CF-ll Montsweag Brook 2.7 NW AP/CF-13 Environmental Studies Laboratory 0.6 NE AP/CF-14 CMP Mason Station 4.8 NNE AP/CF-16 Westport Firchouse 1.8 S
AP/CF-29 Dresden Substation 19.8 N
2.
DIRECT RADIATION GM-11 Montsweag Brook 2.7 iG CM-12 Eaton Farm 0.5 W
GM-13 Environmental Studies Laboratory 0.6 NE CM-14 CMP Mason Station 4.8 NNE GM-15 Edgecomb Firehouse 5.6 ENE GM-16 Westport Firehouse 1.8 S
CM-17 Harrison's Trailer 6.4 SSW GM-29 Dresden Substation 19.8 N
Sample locations are shown on Figure
+ Station-lX's are indicator ctations and Station-2X's are control stations.
li. I V
TABLE (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS
- EXPOSURE PATHWAY SAMPLE LOCATION DISTANCE FROM DIRECTION FROM AND/OR SAMPLE AND DESIGNATED CODE THE PLANT (km)
THE PLANT 3.
WATERBORNE a.
Surface (Estuary)
WE-21 Plant Intake onsite WE-15 Plant Ou t f a l l ***
onsite b.
Groundwater WC-13 Environmental Studies Laboratory 0.6 NE WG-24 Morse Well 9.8 W
c.
Sediment from Shoreline SE-18 Foxbird Island Diffuser Discharge 0.6 SSE 4.
INGESTION a.
Milk TM-16 Baker Farm 7.2 W
TM-17 Kilkelly Farm 6.1 ENE TM-18 Potter Farm 7.8 NE TM-24 Knight Farm 9.8 WNW b.
Fish and Invertebrates FH-11 Long Ledge 1.1 S
FH-24 Sheepscot River 11.2 S
c.
Food Products TF-Il Hodgdon Farm 2.1 N
TF-12 Environmental Studies Laboratory 0.6 NE TF-23 Litchfield, Maine 32.2 NW TV-12**
Environmental Studies Laboratory 0.6 NE Sample locations are shown on Figure
- TV-12 Station is for leafy vegetables
- A dilution factor of 10 shall be applied to the radioactivites detected in the sample at this station.
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DRA e DRAgW f APPENDIX I A. Liquid Release Dose Calculations There are two methods for calculating the doses f rom liquid releases. The second method is calculation by the computer program IDLE which uses the models of Regulatory Guide 1.109. The only pathways available at Maine Yankee are the fish, invertebrate and shoreline pathways. Regulatory Guide 1.109 parameters were used with a plant discharge flow rate of 936 ft3/sec and mixing ratios of 0.1 (aquatic food) and .04 (shoreline). The doses were calculated for the 14 quarter period from Jauary 1975 to June 1978 to serve as a basis of comparison of the Method 1 and Method II models. The model for Method I was obtained by applying the Regulatory Guide 1.109 models to the Maine Yankee Station for an assumed release of on.e curie of each nuclide observed in the quarterly releases for the 14 quarter period. This resulted in a site specific dose conversion f actor providing the number of millirems per curie that would result for each nuclide. 1. Dose to the Total Body The total body dose is approximated in Method I by using the site specific dose conversion f actors for the four major contributions to the total body dose: D4 E ~ ,b C l-* Metho I J Dtb(mrem) = 0.0023 QS8Co + 0.05 Q60Co + 0.03 Q134Cs + 0.03 Q137Cs where: 0.0023 = mrem /C1 for Cobalt-58 f or child total body dose. 0.05 = mrem /Ci for Cobalt-60 for child total body dose. 0.03 = mrcm/Ci for Cesium-134 for adult total body dose. 0.03 = mrem /Ci for Cesium-137 f or adult total body dose. Method I has been applied to quarterly releat,es during January 1975 to June 1978 and the comparison to the IDLE calculation is shown in Figure A.l. 2. Dose to the Critical Organ A similar approach was used to find the dose to the critical organ. It was found that the following expression will provide a dose that is quite close to the critical organ dose for each release in the period January 1976 to January 1978: Method I Dorgan(mrem) = 0.4 Q60Co + 0.13 Q1311 + 0 72 Q134Cs + 0.51 Q13 7Cs where: 0.4 = mrem /Ci for Cobalt-60 for the adult gastro-intestinal 5 h y ,h j R w: , a'f,7*:ag f track dose. O.13 - mrem /ci for Iodine-131 for the adult thyroid dose. 0.72 = mrem /Ci for Cesium-134 for the teen liver dose. 0.51 = mrem /Ci for Cesium-137 f or skin dose. The organ dose calculated by Method I is not for a real organ, but for a combination of critical organs. Figure A.2 shows a comparison of the Method I " organ" dose calculation with the Regulatory Guide 1.109 models. B. Dose Rate to an Individual 1. Dose Rate in the Total Body The dose rate to the total body due to noble gases is calculated by the methods of Regulatory Guide 1.109 as follows: b re 4 (Cipc1 n) [X) (sec) tb ( yr ) = 1.11 S 3.17x10 p sec Q 3 m 3 x Q1 (C_.i.) DFBf (mem ) i yr pCi-yr where: S = Shielding Factor = 0.7 p [3] = Annual Average Dilution Factor = 3.24 x 10-6 (sec) Q 3 m btb ( rem) 7.98x10-2(pCi-vr-sec) 7 Q (CJ)3.15x10 (sec)DFB = 1 f yr Ci-sec-m3 yr yr /ag[;I jt~,,y hg'3 i e '1 ~y i A _ U iy y 6 3 = 2.51x10 (pCi-sec) qi(S )DFB (mrem-m ) 6tb ( rem) f yr Ci-m3 sec pC1 -yr 2. Dose Rate to the Skin The dose rate to the skin is calculated by the methods of Regulatory Guide 1.109 as follows: skin (m m) 3 total body + 1.1012[ j f } I DFS ( Ci y f undepleted i c where: undepleted = Annual Average Undepleted 3 = 4.33 x 10-6 sec Q 3 m h(Ci)DFS1 (*Ci-yr total body + 4.33x10 (pCi-sec) 6 bskin ( )" 3 ) yr Ci-m3 sec p 3. Dese Rate due to Iodines and Particulates The dose rate due to iodines and particulates is based on limiting the dose rate to the infant thyroid due to Iodine-131. bthyroid(mrem) - 18.8 t'Ci-released) 3.15x10 (sec) y ( C1) mrem 7 yr yr see bthyroid (mrem) = 5.92 x 10 (mrem-sec) y ( Ci) 8 yr Ci-yr see C. Dose to Air due to Noble Gases The dose to air due to noble gases has been calculated by the methods of Regulatory Guide 1.109. n
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1. Gamma Dose to Air w ~ J The gamma dose is calculated as follows: D(mrad) = 3.17 x 104 (pCi yr) []X) (sec) Ci sec Q 3 m )) ( Qi(C1) DFi (yr-pCi i where: [.X. ] = Annual Average Gamma Dilution Factor = 3.24 x 10-6 sec 3 m Q Q = Number of curies or noble gas "1" released i which leads to: D(mrad) = 0.103 Qi(C1) DFi 2. Beta Dose to Air The beta dose to air is calculated in the same manner as the gamma dose except the undepleted X/Q is used '4.33 x 10-6 (sec/m3) and the dose factor from beta radiation is applied, leading to: Dorad = 0.137 Qi(C1) DFi D. Dose to an Individual due to Radioiodines and Particulates There are two methods for calculating the dose due to radioiodines and particulates. The second method is calculation by the computer program ATMODOS which used the methods of Regulatory Guide 1.109. All pathways except the cow-milk pathway were considered. The milk pathway considered was the goat. Regulatory Guide 1.109 parameters h h I ~';" b f. *. N)kf,/ ' I' R ' Q; ? were used with milk and meat animals assumed to be on pasture 50% 'of the time and consuming 100% of their feed from pasture during that period. The doses were calculated for a 14 quarter period from January 1975 to June 1978 to serve as a basis of comparison for the Method I model. The model for Method I was obtained by applying the Regulatory Guide 1.109 models to the Maine Yankee Station for an assumed release of one curie per each nuclide observed in the quarterly release during the 14 quarter period. This resulted in a site specific dose conversion f actor providing the number of millirems per curie that would result for each nuclide. For the 14 quarter period, the critical organ, age I5 k and Mcombinations were found to be the skin for Co-60, the inf ant thyroid fo r I-131 and I-13 3 and the inf ant liver for Cs-13 7. Figure A.3 shows the comparison cf Model I and Regulatory Guide 1.109 models for the 14 month period. This leads to: Dorgan (mrem) = 2.2 Q60Co + 18.8 Q1311 + 0.2 Q133I + 6.1 Q137Cs where: 2.2 = mrem /Ci for Cobalt-60 for the skin dose. 18.8 = mrem /Ci for Iodine-131 for the inf ant thyroid dose. 0.2 = mrem /ci for Iodine-133 for the inf ant thyroid dose. 6.1 = mrem /Ci for Cesium-13 7 for the inf ant liver dose. ,, c 4 3f APPENDIX II Annual average dilution factors based on onsite meteorological data were computed for routine (long-term) releases by the Yankee Atomic Electric Company's (YAEC)AEOLUS(2) Computer Code. AEOLUS is based, in part, on the straight-line airflow model as discussed in Regulatory Guide 1.111(3) and includes the following basic features: -hourly meteorological data input (wind direction, wind speed, vertical temperature dif terence, and, optionally, direction fluctuation, air temperature, sea water surf ace temperature and solar radiation), straight-line air flow model with Gaussian dif fusion, plume centerline and sector-average models with single or split (vertica?. / horizontal, i.e., split-sigma) atmospheric stabilities, part-time ground level and part-time elevated releases (split-H model), seabreeze option for coastal sites (split-H, split-sigma, trapping and fumigation), multi-energy sector-averaged finite cloud diluticn f actors for gamma dose calculations (both normal and seahreeze cases), terrain features, plume rise (buoyant or momentum), depletion in transit (2 models), UA y O 'c.u ' recirculation correction factors (built-in options for flat terrains and river valleys, or user selected values), deposition rates (2 models), and dose statistical distributions for postulated accidental radioactive releases and exposure intervals (based on dose rate data per unit dilution f actor (A/Q) as input; thyroid, total body beta, total body gamma and skin doses. AEOLUS produces hourly and long-term average of non-depleted dilution f actors for evaluating ground level concentrations of noble gases, tritium, carbon 14 and non-element.al iodines, depleted dilution f actor for estimating ground level concentrations of elemental radiciodines and other particulates, effective gamma dilution f actors for evaluating gamma dose rates from a sector-averaged finite cloud (multiple-energy undepleted source), and deposition f actors for computing dry deposition of elemental radiciodines and other particulates. A more detailed description of the AEOLUS dif fusion model is provided in section 2 3.5 (long-term diffusion estimates) of the NEP 1&2 PSAR(4) and the AEOLUS computer code manual (2), DRMT Annual average non-depleted dilution factors, effective gamma dilution f actors and deposition (D/Q) rates for Maine Yankee were calculated using the following AEOLUS options; Sector-average model with temperature difference ( T) atmospheric stabilities, split-H model, buoyent plume rise, and no recirculation correction f actors. A/Q and D/Q values for the restricted area boundry critical sector are provided in Table 41. Hourly dilution f actors are computed for either ground level ot elevated releases using both the plume centerline model and the sector average model. Plume centerline values are for estimating short-term atmospheric dispersion (up to 8 hours) and sector average values are for dispersion during longer periods of time. In the plume centerline model, for ground level releases during neutral and stable atmospheric conditions and low wind speeds, atmospheric dispersion is corrected for either plume meander effects or for the additional dispersion of the effluent plume within the wake caused by buildings adjacent to the release point. During til other atmospheric stability and/or wind speed conditions, credit is taken only for building wake ef fects. In the sector-average model, dispersion is based on Regulatory Guide 1.111 and meander effects are ignored. Both building wake and plume meander are excluded f rom the equation for elevated releases. Atmospheric dilution factors are computed by the above models for each sequential hour of measured ceteorological data and for receptors positioned in the 16 cardinal compass directions around the plant. The hourly dilution f actors obtained as described and the corresponding direction
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,s h jD h/ l} iI [ 4Q (*, J in which the wind is blowing durin'g each hour are then stored in sector dependent arrays for sequential processing. This involves the averaging of selected hourly dilution values over successive, overlapping time intervals. For each selected interval size, the processing begins with the first hourly dilution value on record and then repeated for the same interval size starting with each subsequent hour of dispersion data. In the averaging process, the only values within a given interval that are considered in evaluating the mean dilution factor for the interval are those f or the specific wind direction being analyzed. Missing data are handled by imposing the condition that at least half of the entries within an averaging interval correspond to valid observations. Missing data points are not included in the averaging. The average dilution factors computed as described are subsequently classified, f or each 22-1/2 degree sector, into groups, and corresponding cumulative probability distributions are prepared. For each sector the dilution f actors at a number of percentile points are determined. These points define the percent of time a dilution value is equalled or exceeded. }}