ML17053A619
| ML17053A619 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/12/1979 |
| From: | NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML17053A618 | List: |
| References | |
| NUDOCS 7904200219 | |
| Download: ML17053A619 (30) | |
Text
WASTE SOLIDIFICATION TECHNICAL SPECIFICATIONS ADMINISTRA:.IVECONTROL TECHNICAL SPECIFICATIONS V 904200 3lg
SECTION DESCRIPTION PAGE 6.11
- 6. 12 6.13 6.14=
6.-15 6.16 Radiation Protection Program Respiratory Protection Program High Radiation Area Fire Protection Inspection Process Control Program Offsite Dose Calculation Manual 260 260 263 268 269 270
LIMITING CONDITION FOR OPERATION
. SURVEILLANCE RE(UIREMENT 3.6.13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Applies to the operating status of the liquid, gaseous and solid eff1uent treatment systems.-
~0b ective:
To assure operability of the liquid, gaseous and solid effluent treatment system.
(a)
Li<iuid The liquid radwaste treatment system shall be OPERABLE.
The system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the dose due to liquid effluent releases to unrestricted areas (see Figure 3.6. 12-1) would exceed
- 0. 18 mrem to the total body or 0.6 mrem to any organ 'in any calendar quarter.
With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a
Special Report which includes the following information:
4.6.13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS A~li bill Applies to the surveillance requirements for the liquid, gaseous and solid effluen't treatment systems
~0b'ective:
To verify operability of the liquid, gaseous and solid effluent treatment system.
~Sufi (a) ti(iuid Doses due to liquid releases to unrestricted areas shall be calculated monthly when releases are made.
The liquid radwaste system shall b(, demonstrated OPERABLE quarterly un'less the liquid radwaste system has been-utilized to process radioactive liquid effluents during the previous quarter. ~
Identification of equipment or subsystems not OPERABLE and the reason for nonoperability.
Action(s) taken to restore the nonoperable equip-ment to OPERABLE s ta tus.
Summary description of action(s) taken to prevent a recurrence.
241ccc
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (c)
Solid
-A radwaste solidification system shall be OPERABLE and used to provide for the SOLIDIFICATION of wet solid wastes, for the SOLIDIFICATION and packaging of other radioactive wastes, and to ensure the meeting of the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.
With the requirements of 10 CFR Part 20, 10 CFR Part 71, and the PROCESS CONTROL PROGRAM of Specification 6.15, not satisfied, suspend shipments of defective containers of solid radioactive wastes from the site.
With the solid radwaste system not OPERABLE for more than 31 days, when required to meet 10 CFR Part 20 and 10 CFR Part 71, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:
Identification of equipment of subsystems not OPERABLE and 'the reasons for inoperability.
4.6. 13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (c)
Solid The solid radwaste system shall be demonstrated operable at least once per quarter by performance of functional tests of the equipment and components of
. a solid radwaste system or by operating a
solid radwaste system at least once in the previous quarter in accordance with the process control program or demonstrate the capability for solidification for verification of the existence of a valid contract for SOLIDIFICATION to be performed in accordance with the PROGRESS CONTROL PROGRAM.
The PROCESS CONTROL PROGRAM of Specification
- 6. 15 shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each. type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, radioactive chemical wastes and filter media).
The test specimens shall be processed in accordance
. with procedures of the PROCESS CONTROL PROGRAM.
I 24ldddl I
LIMITINGCONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6.13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (cont'd)
'c)
Solid (cont'd)
Action(s) taken to restore the inoperable equipment to OPERABLE status.
A description of alternative used for SOLIDIFICATION and packaging of wastes.
Summary description of action(s) taken to prevent a recurrence.
.6. 13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (cont'd)
(c)
Solid (cont'd)
If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspende
'ntil such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION.
SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFI-CATION'arameters determined by the PROCESS CONTROL PROGRAM.
If the initial test specimen from.a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive. batch ~
'the same type of wet waste until. 3
~ consecutive initial test specimens
, demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required to assure SOLIDIFICATION of subsequent batches of waste.
24lddd 2
LIMITINGCONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6. 13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (cont')
4.6. 13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS (cont'd)
~Re orts The semiannual Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report period:
container
- volume, total curie quantity (determined measurement or estimate),
principal gaoma radionuclides (determined by measurement or estimate).
type of waste (e.g.,
LSA, Type A, Type B, Large quantity), 'and solidification agent (e.g.,
semen urea formaldehyde).
24lddd3
BASES FOR 3.6.13 AND 4.6.13 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment.
The requirements that the appropriate portions of this system can be used when specified provides assurance that the releases of radioactive materials in liquid effluents 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 design objective Section 11.D of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.
The OPERABILITY of the gaseous radwaste treatment system ensures that the systems.will be available for use whenever gaseous effluents require treatment prior to release to the environment.
The requirement that the appropriate portions of these systems be used when specified provides reasonable'ssurance that. the releases of radioactive materials in gaseous effluents 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 design objective Section IID of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
The OPERABILITY of the solid waste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite.
This specification implements the requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.
The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.
24leee
Continued The semi-annual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the format of Appendix B thereof.
The semi-annual Radioactive Effluent Release Reports shall include a summary of the meteoro-logical conditions concurrent with the-release of gaseous effluents during each quarter as outlined in Regulatory Guide 1.21, with data surmarized on a quarterly basis following the format of Appendix B thereof.
The semi-annual Radioactive Effluent Release Reports shall include an assessment of the radiation doses from radioactive effluents to individuals due to their activities inside the unrestricted area boundary (Figure 3.6. 12-1) during the report period.
All assumptions used in making these assessments (e.g., specific activity, exposure time and location) shall be included in these reports.
The semi-annual Radioactive Effluent Release Reports shall include the following information for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents:
a)
A descr'iption of the event and equipment involved.
b)
Cause(s) for the unplanned release.
c)
Actions taken to prevent recurrence.
d)
Consequences of the unplanned release.
The semi-annual Radioactive Effluent Release Reports shall include an assessment of the radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21, Appendix B, Section El, 2, 3.
In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.
The meteorological conditions concurrent with the releases of effluents shall be used for determining the gaseous pathway doses.
The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual.
The semiannual radioactive effluent reports shall include any changes to the Process Control Program or to the Offsite Dose Calculation Manual made during the reporting period, as provided in Specifications
- 6. 15 and 6.16.
257d
Process Control Pro ram The Process Control Program shall be a manual containing the equipment operating procedures, process parameters, set points, drawings and controls, and the laboratory procedures detailing the program of sampling,
- analysis, and evaluation within which solidification of radioactive wastes from liquid systems is assured, consistent with Specifications 3.6. 13c and 4.6. 13c.
a.
Changes to the Process Control Program shall be made by either of the following methods:
(1)
Licensee initiated changes:
Shall be submitted to.the Commission by inclusion in the semiannual-Radioactive Effluent Release Report for the period in which the change(s) was made and shall contain:
sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; a determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and documentation of the fact that the change has been reviewed and found acceptable by the Site Operations Review Committee.
Shall become effective upon review and acceptance by the Site Operations Review Committee, unless otherwise acted upon by the Commission through written notification to the licensee.
(2)
Commission initiated changes:
Shall be determined by the Site Operations Review Coomittee to be applicable to the facility after consideration of facility design; The >>~pensee shall provide the Commission with written notification of its determ>nation of applicability including any necessary revisions to reflect facility design; Shall be reviewed by the Safety Review and Audit Board.
269
Offsite Dose Calculation Manual The Offsite Dose Calculation Manual shall describe the methodology and parameters to be used in the calcu-lation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm/trip setpoints consistent with the applicable Limiting Conditions for Operation contained in these Technical Specifications.
Any changes to the Offsite Dose Calculation Manual shall be made by either of the following methods: t (1)
Licensee initiated changes:
Shall be submitted to the Commission by inclusion in the semiannual Radioactive Effluent Release Report the period in which the change(s) was made and shall contain:
sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and documentation of the fact that the change has been reviewed and found acceptable by the Site Operations Review Committee.
Shall become effective upon review and acceptance by the Site Operations Review Committee, unless otherwise acted upon by the Commission through written notification to that licensee.
(2)
Commission initiated changes:
Shall be determined by the Site Operations Review Committee to be applicable to the facility after consideration of facility design.
The licensee shall provide the Commission with.written notification of their determination of applicability including any necessary revisions to reflect facility design.
Shall be reviewed by the Safety Review and Audit Board.
270
OFFSITE DOSE CALCULATION MANUAL
NINE MILE POINT NUCLEAR STATION UNIT 1 OFF-SITE DOSE CALCULATIONALMANUAL LICENSE NO.
DPR-63 DOCKET NO. 50-220 APPROVALS SIGNATURES Radiochemistry and Radiation Protection Supervisor E.
10.
Leach DATE AND INITIALS REVISION 0 REVISION l REVISION 2 I'tation Superi,ntendent NMPNS T. J.
Perkins J
J General Superintendent Nuclear Generation Chairman of S.O.R.C.
T.
E.
Lempges gliolpp NIAGARA MOHAWK P011ER CORPORATION
TABLE OF CONTENTS DOSE CALCULATIONALhlANUAL SECTION PAGE A.
Liquid Effluent monitor Alarm Setpoints B.
Service Wager Monitor Alarm Set Point Determination C.
Gaseous Effluent Monitor Setpoint t
D.
Liquid Effluent Concentration and Dose Calculations E.
Gaseous Effluent Dose Rate and Dose Calculations 1-2 3-4 4-6 6-8 Table 1
Average Energy per Disintegration Figure 1
Liquid and Solid'Radwaste System Figure 2
. On-site Environmental Station and TLD Locations 10 12
" ~
-t(
~
A.
LI VID PASTE EFFLUENT MONITOR ALARM SE OINT DETERMINATION (Reference T.S. Section 3.6.11.a, 4.6.ll.a, 3.6.12.a.l):
The alarm setting on the liquid radwaste effluent monitor is based on dilution water flow, waste tank discharge rate, the isotopic composition of the liquid to be discharged, the back-ground count rate of the monitor, and the efficiency of the monitor.
Due to the variability of these parameters,'an adjustable alarm setpoint is provided and will be set just prior to the release of each batch of liquid waste.'he following methodology (specified by Procedures Nl-PSP-4 and Nl-OP-28)
~is u:ed for determining setpoint of the High and High High alarms:
1.
A sample will be taken from the tank to be discharged and a
gamma isotopic analysis performed.
2.
From the tank analysis and the MPC values (10CFR20 Table II, Column 2) a Required Dilution Factor will be calculated, using the relation:
RDF
=
E (Ci/MPCi) 1(here:
Ci = concentration of nuclide i, pCi/ml MPCi = Maximum Permissible Concentration for nuclide i, pCi/ml 3.
The maximum discharge rate permitted is determined by Circulating and Service l(ater pump configuration, and Required Dilution Factor.
For a given dilution flow, discharge flow is calculated (by nomograph) according to the equation:
discharge flow = dilution flow RDF x F i%ere:
F
= Conservative safety factor (1.1 to 2) used to account for non-gamma emitters.
4.
The liquid effluent monitor alarm (High High) setpoint, which ensures that the limits of Specification 3.6.12.a.l are not exceeded, is calculated as follows:
Setpoint
= MPC x RDF x Fl x F2 i%ere:
Set Point
= Dial setting on alarm potentiometer MPC RDF
= Maximum permissible concentration in discharge canal for isotopic mixture being discharged, or a lesser concentration.
= Isotopic mixture Required Dilution Factor as defined above.
A.
LIQUID RAD1UASTE EFFLUENT MOiVITOR SET POINT DETERMINATION (Cont.)
4.
Fi = Monitor calibration factor,
~c s
pCi/ml FZ
= Potentiometer setting factor, relating turns setting to alarm set point in cps.
The set points may be tabulated to maintain a conservatively low MPC in the discharge card, or calculated for a specific discharge.
B.
SERVICE lUATER MONITOR ALARM SET POINT DETERMINATION 1.
The monitor high high alarm point maximum setting permitted is calculated in accordance with:
Alarmcpm = MPCcana]
x Canal Flo<<
x Fl cpm/pCi/ml Service 1Uater Flow 2.
The monitor High High alarm may be set at any level below this calculated maximum value.
3.
The monitor High alarm set point may be set between 2x and Sx background.
4.
MPC used is that associated with the isotopic mixture of the most probable contaminant, the Reactor Closed Loop Cooling system, unless significant activity is found in the Service lUater, in which case, the Service 1Uater isotopic analysis will be used.
S.
Upon receipt, of the Hi alarm, an investigation will commence to determine and isolate the input source to the Service lUater System.
6.
A grab sample should be taken immediately, and as often as required to account for activity discharged by isotopic analysis of the samples.
7.
Upon receipt of the Hi alarm, the Hi Hi alarm set point should be re-calculated and set in accordance with the mixture actually present in the service water.
C.
GASEOUS EFFLUENT i~lONITOR SETVOINTS (Technical Specifications Section 3.6.1l.b, 4.6.ll.b)
A gaseous sample from the stack shall be taken and analyzed for isotopic composition monthly.
From this analysis the total release
- rate, in uCi/sec, will be calculated, using the following equation:
Stack Release Rate
= Stack Flow Rate x Stack Concentration uCi/dec
= cm3/sec x EiuCii/cm3 2.
For each stack monitor, the monitor reading, net cpm or net cps, will be related to Release Rate by a stack Calibration Factor:
= pCi/sec net cpm
'or CF
= uCi/sec net cps The appropriate factors for monitors 07, 08, 11 and 12 are posted at the stack monitors, and -reflect the latest repre-sentative isotopic analysis.
3.
The limits on noble gas releases imposed by Specification 3.6.12.b.l shall not be exceeded prior to receiving an alarm on the stack monitor.
4.
The stack monitor maximum instantaneous release rate alarm shall be set to alarm at or below a release rate Qs given by the equation:
Ei Qis (1.6 EiY
+ 0.3 Eig)
~
1 Gamma and beta E are based on stack gaseous release gamma isotopic analysis and Table 1.
S.
Normally the high alarm will be set at approximately 25'f this calculated value, and the High High alarm to approximately 75~a of this value.
The set points will be reset quarterly, or if an isotopic analysis results in a calculated release rate limit which varies by 25~a from that calculated the last time it was set.
C.
GASEOUS EFFLUENT bIONITOR SETPOINTS (Cont.)
6.
A gaseous sample from the recombiner discharge shall be taken and analyzed for isotopic composition at least monthly.
From this analysis the release rate, in pCi/sec at the re-combiner discharge, will be calculated:
Recombiner Release Rate' Off Gas Flow Rate x Off Gas Concentration pCi/sec
= cm3/sec x Ei uCii/cm3 7.
For each off gas monitor, the reading in mr/hr will be related to the release rate at the Recombiner Discharge by an off gas calibration constant, posted at the monitor:
K = pCi/sec mr/hr 8.
The design Off Gas Treatment System holdup time (flow rate adjusted) will be applied to each Xe and Kr nuclide detected to determine the portion of the Stack Release attributable to Off Gas.
9.
The Off Gas monitor (at the recombiner discharge) shall be set to initiate automatic closure of the waste gas discharge valve prior to exceeding a stack release rate Qs given by the equation:
Ei Q.s (1.0 Ei~
+03E8) <1 D.
LIQUID EFFLUENT CONCENTRATION AND DOSE CALCULATIONS (Technical Specifications 3.6. 12.a. 1, 3.6. 12. a. 2, 4.6. 12, a. 1, 4.6. 12.a.2)
Each liquid waste tank to be discharged is GeLi analyzed, a Required Dilution Factor assigned, and a maximum discharge flow determined as discussed in Sections Al, 2 and 3.
This procedure ensures meeting the concentration requirements of Tech Spec 3.6 '2.a.l and surveillance requirements of Speci-fication '4.6.12.a.l.
2.
Prior to each radioactive liquid discharge, an estimate of Curies to be discharged is made (based on the GeLi analysis, and tank volume) and added to previous discharge estimates for the calendar quarter.
If these estimates total 1.25 Ci for the current calendar quarter, an evaluation of operability*
of the liquid radwaste treatment system must be made.
If it is
- See Section D-5
LI UID EFFLUENT CONCENTRATION AND DOSE CALCULATIONS (Cont.)
2.
determined that waste is being discharge without, treatment and in excess of this quantity, a Special Report is to be submitted to the Commission, if confirming dose evaluation
'of actual releases indicate Specification 3.6.13.a has been exceeded (Since the composite analyses which allow complete dose calculations to be made are required quarterly, pre-mature dose calculations are not required.
See Section D-6 for dose calculation methods.).
3.
The total tank curies discharged each month is calculated frog tank concentration and volume data.
The dose speci-fications for liquid waste (T.S. -3.6.12.a.2) ~are considered to be satisfied if the total curies discharged is less than 10 curies per calendar quarter (excluding tuition and dis-solved gases).
4.
If the calendar quarter curies discharged approaches the value given in Section D-3, a dose calculation will be made to determine compliance with Specifications 3.6.12.a.2 in accordance with Section D - 6d.
New quarter'nd year maximum curie discharge limits may then be set on the basis of those calculations.
S.
Liquid Waste Treatment System Operability a)
The liquid waste treatment system is considered 'operable if no liquid waste discharges have occurred during the calendar quarter.
b)
The liquid waste treatment system 'is considered operable and the, liquid waste is considered discharged with treatment if each discharge (excluding laundry drain tanks) has been made from a Waste Sample Tank, which has been processed through a filter and demineralizer, or filter and demineralizer following a waste concentrator as described in submittal Nine Mile Point Nuclear Station Unit '1, Docket S0-220, Compliance with 10CFRSO Appendix I (1976).
(See Figure 1, updated to reflect system modifications).
c)
It is anticipated that through usage of one or two liquid waste concentrators NMP Unit 1 will maintain virtually zero discharge during normal operation.
Therefore projected dose calculations used as a tri-ggering mechanism for use of appropriate subsystems of the liquid waste processing system are unnecessary and inappropriate.
6.
Liquid Effluent Dose Calculation 51ethods a)
The dose calculations required shall be accomplished by ratio of curies released to the curies released which were used in calculating compliance with 10CFRSO Appendix I, (Reference 1).
It was determined, using, methods
D.
LBIUID EFFLUEiNT CONCENTNATIOiiI AND DOSE CALCULATIONS (Cont.)
6.
a) referenced therein, that 1.3 curies released resulted in an adult Whole Body dose of 0.089
- mrem, and an infant thyroid dose of 0.22 mrem.
These were the highest doses calculated, and were due principally to ingestion of fish (adult WB dose) and potable water (infant thyroid dose).
This method is intended to satisfy Technical Specifications 4.6.12.a.2 and 4.6.13.a.
b)
Whole Body Dose:
E Curies Released
= Whole Bod Dose (Adult) 1.3 curies 0.089 mrem Therefore, WB Dose
= Curies released x 0.068 c)
Maximum Organ Dose:
Curies Released
= Or an Dose (Infant Th roid) 1.3 curies 0.22 mrem Therefore, Organ Dose
= Curies Released x 0.17 d)
As an alternative, the methodology of Regulatory Guide 1.109 may be used with actual release data (isotopic analysis) to calculate whole body dose and organ dose.
The site-specific values of Reference 1 are to be used,,
as appropriate, to calculate dose due to ingestion of potable water (City of Oswego water supply) and inges-tion of fish (caught at the edge of the initial mixing zone).
E.
GASEOUS EFFLUENT DOSE RATE AND DOSE CALCULATIONS (Technical Specifications 3.6.12.b, 4.6.12.b) 1.
The dose rate limit for noble gases is considered to be met if the stack release rate Q (in Ci/sec) is restricted to a value obtained from the equation:
i Q is (1
6 Eiy + o 3 Eig)
~
~
E.
GASEOUS EFFLUENT DOSE RATE AND DOSE CALCULATIONS (Cont.).
2.
The dose rate limit for all radioiodines and for all radio-active materials in particulate form is considered to be met if the stack release rate Q (in Ci/sec) is restricted, for these materials, to a value obtained from the equation:
Q s 8.3'x 10 3.
The air dose limit for noble gases is considered to be met if the calendar quarter average stack release rate Q (in Ci/sec) is restricted to a value obtained from the equations:
I Zi (Q is Eiy) 1.0,x 10
~i (Q is ig 4.
The dose to an organ of an individual. from radioiodines and radioactive materials in particulate form with half-lives greater than 8 days is considered to be met if the calendar quarter average stack release rate Q (in Ci/sec) is restricted for these materials, to a value obtained from the equation:
6 7 x 10 6
5.
The indication of noble gas release rate is the stack monitor; readings are evaluated at intervals sufficient to ensure meeting the limits specified in paragraphs 1 and 3 of this section.
Other surveillance is performed as specified by Technical Specification Table 4.6.12-2, 6.
The monthly cumulative dose contribution to the 1'Role Body shall be calculated by ratio of the average release rate to the release rate corresponding to 500 mrem/year, for the appropriate isotopic analysis period, then by sum of the dose for each such period for the month.
(An analogeous calculation will be made for dose due to iodines and parti-culates).
interval 500 x d/365 Q max. inst.
Q max inst
= Qs, release
- rate, calculated from isotopic
- analysis, corresponding to 500 mrem/yr (See Section C-4).
days of the month the above release rate limit (and isotopic analysis) was applicable.
Dose interval
= Dose at the site boundary corresponding to the isotopic analysis period.
0
~
E.
GASEOUS EFFLUENT DOSE RATE AND DOSE CALCULATIONS (Cont.)
7.
If the release rates cited in Sections E-l, E-2, E-3 and E-4 are approached, dose calculations may be made using the methodology of Regulatory Guide 1.109 in order to determine new allowable release rates based on the isotopic analysis of material being discharged.
8.
The semi-annual Radioactive Effluent Release Report will include the calculations required by Technical Specifications 6.9.1e:
a)
> The assessment of radiation doses from radioactive effluents to individuals due to their activiti:es inside the unres-tricted area boundary shall be made using the readings obtained'rom On-Site TLD locations (see figure 2).
b)
The evaluation of any unplanned release will be made using methodology appropriate to the release pathway.
c)
Total body and skin 'doses to individuals exposed at the point of maximum off-site ground-level concentrations of radioactive materials in gaseous effluents will be eval-uated using the methodology of Regulatory Guide 1.109.
d)
Organ doses to individuals in unrestricted areas from radioactive iodine and radioactive material in particu-late form from all pathways of exposure will be evaluated using the methodology of Regulatory Guide 1.109.
REFERENCES 1.
Nine Mile Point Nuclear Station Unit 1, Docket 50-220, Compliance
. with 10CFR50 Appendix I, 1976.
2.
Regulatory Guide 1.109, Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I, Revision 1, October 1977.
3.
4.
Radiological Technical Specifications, Appendix A to Facility Operating License No.
DPR-63 for the Niagara Mohawk Power Corporation Nine Rile Point Nuclear Station Unit 1, Docket No. 50-220.
Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radio-activity in Solid Wastes and Releases of Radioactive Materials in Liquid and'aseous Effluents from Light-ifater-Cooled Nuclear Power Plants, Revision 1, June 1974.
0 TABLE i AVERAGE ENERGY PER DISINTEGRATION Isotope EY, mev/dis (Ref)
EB, mev/dis ( )
(Ref)
Kr-83m Kr-85 Kr-85m Kr-87 0.00248 0.0022
- 0. 159
- 0. 793 0.0371 0.250 0.253
- 1. 95
- 2. 22
- 2. 10 0.0201 0.0454 0.042 0.247 0.432 0.194 1.18 (2)
(2) 0.377
- l. 37
- 1. 01 0.-143
- 0. 135
- 0. 19 0.317 0.095 1.64
- 0. 611 (2)
(2)
(1)
ORNL-4923, Radioactive Atoms Su lement I, M. S. Martin, November 1973.
(2)
NED0-12037, "Summary of Gamma and Beta Emitters and Intensity Data";
M. E. Meek, R.
S. Gilbert, January 1970.
(The average B energy was
~
computed from the maximum energy using the ICRP II equation, not the 1/3 value assumption used in this reference).
(3)
.The average 8 energy includes conversion electrons.
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