Text

Annual Radioactive Effluent Release Report, January 1, 2002 Through December 31, 2002
	ML030920668
Person / Time
Site:	Cook
Issue date:	03/27/2003
From:	Joseph E Pollock; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:3691-02
	Download: ML030920668 (235)
	v • d • e

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER March 27, 2003 AEP:NRC:3691-02 Docket Nos: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY 1, 2002 THROUGH DECEMBER 31, 2002 In accordance with Technical Specification (TS) 6.9.1.7, Indiana Michigan Power Company hereby submits the Annual Radioactive Effluent Release Report for Donald C. Cook Nuclear Plant (CNP). This report covers the period January 1, 2002 through December 31, 2002.

The calculations in this report were performed in accordance with the CNP Offsite Dose Calculation Manual (ODCM). Revisions 16 and 17 of the ODCM were issued during the reporting period. A copy of each revision is included in Appendix 3.0 to this report to fulfill the requirements of TS 6.14.1.c.

There are no new commitments in this submittal. Should you have any questions, please contact Mr. Brian A. McIntyre, Manager of Regulatory Affairs, at (269) 697-5806.

Sincerely, Joseph E. Pollock Site Vice President DB/rdw Attachmenth

U.S. Nuclear Regulatory Commission AEP:NRC:3691-02 Page 2 c: H. K. Chemoff, NRC Washington, DC, w/o attachment K. D. Curry, Ft. Wayne AEP, w/o attachment J. E. Dyer, NRC Region III J. T. King, MPSC, w/o attachment MDEQ - DW & RPD, w/o attachment D. Minnaar, MDEQ NRC Resident Inspector J. F. Stang, Jr., NRC Washington, DC

U.S. Nuclear Regulatory Commission AEP.NRC:3691-02 Page 3 be: A. C. Bakken III, w/o attachment J. P. Carlson, w/o attachment M. J. Finissi, w/o attachment D. W. Foster J. B. Giessner D. W. Jenkins, w/o attachment J. A. Kobyra, w/o attachment B. A. McIntyre, w/o attachment J. E. Newmiller, w/o attachment J. E. Pollock, w/o attachment D. J. Poupard, w/o attachment M. K. Scarpello, w/o attachment T. K. Woods, w/o attachment

ATTACHMENT TO AEP:NRC:3691-02 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY 1, 2002 THROUGH DECEMBER 31, 2002

Annual Radioactive Effluent Release Report January 1, 2002 through December 31, 2002 3AMERICANO M ELECTRIC POWER AeP.Ameldes EneWj aIitners

Annual Radioactive Effluent Release Report Donald C. Cook Nuclear Plant Units 1 and 2 January 1, 2002, through December 31, 2002

i TABLE OF CONTENTS Page i

Table of Contents Introduction 1 I.

2 II. Radioactive Releases and Radiological Impact on Man Liquid Releases 2 Gaseous Releases 2 3

Solid Waste Disposition 3

III. Meteorological 3

IV. Offsite Dose Calculation Manual (ODCM) Changes 3

V. Total Dose 4

VI. Radiation Monitors Inoperable Greater Than 30 Days 4

VII. Conclusion LIST OF APPENDICES Appendix Title 2002 Effluent and Waste Disposal Annual Report -

Al. 1 Supplemental Information Al .2 Summary of Maximum Individual Doses: First Quarter, Second Quarter, Third Quarter and Fourth Quarter 2002 A2 .1 Hours at Each Wind Speed and Direction: First Quarter, 2002 A2.2 Hours at Each Wind Speed and Direction: Second Quarter, 2002 A2 .3 Hours at Each Wind Speed and Direction: Third Quarter, 2002 A2.4 Hours at Each Wind Speed and Direction: Fourth Quarter, 2002 3.0 Offsite Dose Calculation Manual (ODCM) Changes 1

I. - INTRODUCTION This report discusses the radioactive discharges from Unit 1 and Unit 2 of the Donald C. Cook Nuclear Plant during 2002. This is in accordance with the requirements of Cook Nuclear Plant Technical Specification 6.9.1.7.

The table below summarizes the pertinent statistics concerning the Plant's operation during the period from January 1 to December 31, 2002. The data in this table and the descriptive information on plant operation are based upon the respective Unit's Monthly Operating Reports, Performance Indicators and Control Room Logs for 2002.

Parameter I Unit 1 Unit 2 Gross Electrical Energy Generation (MTWH) 7,741,000 7,688,010 Unit Service Factor (%) 88.9 83.8 Unit Capacity Factor - MDC Net (%) 88.4 82.8 Unit 1 entered the reporting period in Mode 1 at 100% Rated Thermal Power (RTP). Small power adjustments were made to facilitate main turbine valve testing throughout the year. The unit reduced power on 1/8/02 to -80% RTP to plug main condenser tubes. The unit returned to 100% RTP on 1/9/02. On 4/25/02 the Turbine Generator was taken off-line to repair a damaged Breaker K1 disconnect. The reactor remained at 8% RTP. The unit returned to 100% RTP on 4/28/02. On 5/4/02 the unit was shutdown and the scheduled UlC18 refueling outage commenced. The unit attained criticality on 6/8/02 and reached 87% RTP on 6/14/02 at which time the unit was shutdown due to a feedwater pump trip attributed to circulating water debris. The forced outage was extended to perform maintenance on steam generator stop valves and the control rod system. The unit was brought critical on 6/17/02 and returned to 100% RTP on 6/20/02. On 11/10/02 reactor power was reduced to 30% to allow containment entry to add oil to the #14 Reactor Coolant Pump, returned to 100% RTP on 11/12/02.

On 12/22/02 power was reduced to approximately 52% to again allow oil to be added oil to the #14 Reactor Coolant Pump, returned to 100% RTP on 12/23/02. On 12/23/02 reactor power was reduced to 58% to allow repair of a weld leak on a West Main Feed Pump discharge pressure instrument piping, returned to 100% RTP on 12/26/02. The unit exited the reporting period at 100% RTP.

Unit 2 entered the reporting period in Mode 1 at 100% RTP. Small power adjustments were made to facilitate main turbine valve testing throughout the year. On 1/19/02 the unit was shutdown and the scheduled U2C13 refueling outage commenced. The unit attained criticality on 2/25/02 and increased power to 28% RTP on 2/28/02. On 2/28/02, during the power increase the unit reduced power to 2.5% RTP in order to adjust steam generator stop valve 2-MRV-220, 100% RTP was attained on 3/5/02. On 4/5/02 power was reduced to 40% RTP to replace three 2AB 250V DC Station Battery cells that exhibited signs of cracking. The unit returned to 100% RTP on 4/6/02. On 5/12/02 the unit experienced a turbine/reactor trip from 100% RTP due to a failure of a power supply in the channel 1 control group. The power supply was replaced, the unit attained criticality on 5/15/02 and returned to 100% RTP on 5/17/02. The unit was shutdown on 5/25/02 due to a turbine stop valve body steam leak. The unit was taken critical on 6/1/02 and returned to 100% RTP on 6/3/02. On 7/22/02 the unit automatically tripped from 100% RTP due to low-1

low main condenser vacuum while flushing the 'C' main turbine condenser waterbox. The unit was taken critical on 7/23/02 and achieved 94% RTP on 7/27/02 at which time the unit was shutdown following the #23 Circulating Water pump discharge valve failing closed. The unit was taken critical on 8/3/02 and returned to 100% RTP on 8/5/02. The unit exited the reporting period at 100%

RTP.

II. RADIOACTIVE RELEASES AND RADIOLOGICAL IMPACT ON MAN Since a number of release points are common to both units, the release data from both units are combined to form this two-unit, Annual Radioactive Effluent Release Report. Appendix A1.1 through A2.4 of this report presents the information n accordance with section 6.9.1.7 of Appendix A to the Facility Onerating Licenses, as specified in the Technical Specification, Regulatory Guide 1.21 and 10 CFR Part 50, Appendix I.

The "MIDAS System" is a computer code that calculates doses due to radionuclides that were released from the Donald C. Cook Nuclear Plant.

All liquid and gaseous releases were well within Offsite Dose Calculation Manual limits and Federal Limits.

There were no abnormal liquid or gaseous releases.

Liquid Releases During 2002 there were 70 liquid batch releases. During the first quarter there were 23 liquid batch releases. During the second quarter there were 26. During the third quarter there were 10.

During the fourth quarter there were 11.

Estimated doses (in millirem) to maximally exposed individuals via the liquid release pathways are given in appendix 1.2 of this report.

Gaseous Releases During the first quarter of 2002 there were seven batch releases from Waste Gas Decay Tanks (GDT), one from containment purge and 83 Containment Pressure Reliefs (CPR). During the second quarter there were six batch releases from GDT, one from containment purge and 120 CPR. During the third quarter there were no GDT or purge batch releases and 169 CPR. During the fourth quarter there were no GDT or purge batch releases and 149 CPR. CPR continue to be listed as batch releases in accordance with NRC inspections 50-315/89016 (DRSS) and 50-316/8917 (DRSS). There were a total of 13 GDT, two unit purges and 521 CPR during 2002.

In calculating the dose consequences for continuous and batch gaseous releases during 2002, the meteorological data measured at the time of the release were used.

The estimated doses (in millirem) to maximally exposed individuals via the gaseous release pathways are given in appendix 1.2 of this renort.

Solid Waste Disposition There were 16 shipments of radioactive waste made during 2002.

This included shipments made from the site and the various radioactive waste processors to the ultimate disposal site.

III. METEOROLOGICAL Appendices A2.1, A2.2, A2.3, and A2.4- of this report contain the cumulative joint frequency distribution tables of wind speed and wind direction, corresponding to the various atmospheric stability classes for the first, second, third and fourth Quarters of 2002.

Hourly meteorological data is available for review and/or inspection upon request.

IV. OFFSITE DOSE CALCULATION MANUAL (ODCM) CHANGES The Offsite Dose Calculation Manual, PMP-6010-OSD-001, was changed during the report period. The reasons for the changes and the Plant Operations Review Committee approval are documented on the procedure Review and Approval Tracking Form. These changes did not reduce the accuracy or reliability of dose calculations or setpoint determinations. Appendix 3.0 contains the revised ODCM with changes indicated by marginal bars.

V. TOTAL DOSE Section 3.2.5 of the ODCM requires that the dose or dose commitment to a real individual from all uranium fuel cycle sources in Berrien County be limited to no more than 25 millirem to the total body or any organ (except the thyroid, which is limited to no more than 75 millirem) over a period of 12 consecutive months to show conformance with the requirements of 40 CFR Part 190. The maximum cumulative dose to an individual from liquid and gaseous effluents during 2002 was well within the ODCM limits. Measurements using thermoluminescent dosimeters at 11 offsite stations indicate that the dose due to direct radiation is negligible compared to preoperational doses and current background levels. This is fully evaluated in the Annual Radiological Environmental Operating Report for 2002.

The annual dose to the maximum individual will be estimated by first, summing the quarterly total body air dose, the quarterly skin air dose, the quarterly critical organ dose from iodines and particulates, the quarterly total body dose from liquid effluents, the quarterly critical organ dose from liquid effluents, and the direct radiation monitoring program. These quarterly values will be summed and compared to the annual limit. The table that follows here represents the above verbal description:

3

Dose 1St Quarter] 2 d Quarter 3 rd Quarter I4" Quarter Total Body or any 2.30_-022 6.52E-02 4.87E-02 1.74E-02 organ, except thyroid (Air) I_ _ _ _ _ _

Total Body (Air) 6.837-04 1.05X-02 1.445-02 1.02E-03 Skin (Air) 1.28E-03 2.20E-02 2.62E-02 1.965E-03 Total Body (liquid) j9.74E-03 !6.16E-03 1.61E-03 3.70E-03 Maximum organ (liquid) 9.87E-03 J1.18E-02 1.73E-03 3.76E-03 Direct Radiation 0. 00E+00 O.OOE+00 0.005-00 O.OOE-00 Total 14.46:-02 1.17E-01 9.26--02 2.78E-02 Cumulative Total Dose (Total Body or any other organ) mrem 2.82E-01 Annual Dose Limit (mrem) 2.50E+01 Percent of Limit 1.13E-00 For individuals that are within the site boundary, the occupancy time is sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary VI. RADIATION MONITORS INOPERABLE GREATER THAN 30 DAYS There were no radiation monitors inoperable for greater than 30 davs while there was a release via that pathway.

VII. CONCLUSION Based on the information presented in this report, it is concluded that the Donald C. Cook Nuclear Plant Units 1 and 2 performed their intended design function with no demonstrable adverse affect on the health and safety of the general public.

4

2002 Effluent and Waste Disposal Annual Report SUPPLEMENTAL INFORMATION Facility: Donald C. Cook Plant Licensee: Indiana Michigan Power Company I REGULATORY LIMITS 1.1 Noble Gases The air dose in unrestricted areas due to noble gases released in gaseous effluents shall be limited to the following:

1.1.1 During any calendar quarter, to

5 mrad for gamma radiation and

10 mrad for beta radiation.

1.1.2 During any calendar year, to

10 mrad for gamma radiation and

20 mrad for beta radiation.

1.2 Iodines - Particulates The dose to a member of the public from radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than eight days in gaseous effluents released to unrestricted areas shall be limited to the following:

1.2.1 During any calendar quarter to

7.5 mrem to any organ.

1.2.2 During any calendar year to

15 mrem to any organ.

1.3 Liquid Effluents The dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas shall be limited:

1.3.1 During any calendar quarter to

1.5 mrem to the total body and to

5 mrem to any organ.

1.3.2 During any calendar year to

3 mrem to the total body and to

10 mrem to any organ.

Al. 12-1

2002 Effluent and Waste Disposal Annual Report 1.4 Total Dose The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is limited to < 75 mrem) over a period of 12 consecutive months.

2 MAXIMUM PERMISSIBLE CONCENTRATIONS 2 1 Gaseous Effluents The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to the following:

2.1.1 For noble gases: < 500 mrem/yr to the total body and

< 3000 mrem/yr to the skin.

2.1.2 For all radiolodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than eight days: < 1500 mrem/yr to any organ.

The above limits are provided to insure that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area to annual average concentrations exceeding the limits in 10 CFR Part 20, Appendix B, Table 2, Column 1 2 2 Liquid Effluents The concentration of radioactive material released at any time from the site to unrestricted areas shall be limited to the concentrations specified in 10 CFR Part 20, Appendix 3, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 p.Ci/ml total activity.

Al 1-2

2002 Effluent and Waste Disposal Annual Report 3 AVERAGE ENERGY The average energy (E) of the radionuclide mixture in releases of fission and activation gases as defined in Regulatory Guide 1.21, Appendix B, Section A.3 is not applicable because the limits used forgaseous releases are based on calculated dose to members of the public.

4 MEASUREMENTS and APPROXIMATIONS of TOTAL RADIOACTIVITY 4.1 Fission and Activation Gases Sampled and analyzed on a 4096 channel analyzer and HpGe detector. Tritium analysis is performed using liquid scintillation counter.

4.2 Iodines Sampled on iodine adsorbing media and analyzed on a 4096 channel analyzer and HpGe detector.

4.3 Particulates Sampled on a glass filter and analyzed on a 4096 channel analyzer and HpGe detector. Sr-89 and Sr-90 analyses performed by offsite vendor.

4.4 Liquid Effluents Sampled and analyzed on a 4096 channel analyzer and HpGe detector. Tritium analysis is performed using liquid scintillation counter. Fe-SS, Sr-89 and Sr-90 analyses performed by offsite vendor.

Al. 1-3

2002 Effluent and Waste Disposal Annual Retort 5 BATCH RELEASES 5.1 Liquid 5.1.1 Number of batch releases:

23 releases in the ls- quarter, 2002 26 releases in the 2ad quarter, 2002 10 releases in the 3rd quarter, 2002 11 releases in the 4-C quarter, 2002 5.1.2 Total time period for batch releases:

10490 minutes 5.1.3 Maximum time for a batch release:

205 minutes 5.1 4 Average time period for batch release:

150 minutes 5.1.5 Minimum time period for a batch release:

51 minutes 5.1.6 Average stream flow during periods of release of effluent into a flowing stream:

7.37E+5 gpm circulating water A-_ 1-4

2002 Effluent and Waste Disposal Annual Report 5.2 Gaseous 5.2.1 Number of batch releases:

91 releases in the 1st quarter, 2002 127 releases in the 2 nd quarter, 2002 169 releases in the 3 rd quarter, 2002 149 releases in the 4 th quarter, 2002 5.2.2 Total time period for batch releases:

18733 minutes 5.2.3 Maximum time for a batch release:

276 minutes 5.2.4 Average time period for batch release:

34.9 minutes 5.2.5 Minimum time period for a batch release:

9.0 minutes Al. 1-5

2002 Effluent and Wasce Disposal Atnual Report 6 ABNORMAL RELEASES 6.1 Liquid 6.1.1 Number of Releases:

1 S' Quarter 2. Quarter 3 rd Quarter 4th Quarter 0 0 0 0 6.1.2 Total activity released (Ci):

13t Quarter 2nd Quarter 3 rd Quarter 4th Quarter 0 0 0 0 6.2 Gaseous 6.2.1 Number of Releases:

1St Quarter 2 "a Quarter 3 Quarter 4 'h Quarter 0 0 0 0 6.2.2 Total activity released (Ci):

1 5t Quarter 2 nd Quarter 3r?1 Quarter 4 th Quarter 0 0 0 0

~.._1-5

2002 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT GASEOUS EFFLUENTS-GROUND LEVEL RELEASES CONTINUOUS MODE lNuclides releasedl Unit I1st Quarterl 2nd Quarterl 3rd Quarterl 4th Quarterl l1. FISSION GASES I I I I I I l H3 I Ci I 4.65E+01 I 4.25E-01 I 2.84E+01 I 2.39E+01 I l XE133 I Ci I -------- I 6.67E+01 I __- - I ______

-- I ITotal for Period I Ci I 4.65E+01 I 1.09E+02 I 2.84E+01 I 2.39E+01 I lNuclides releasedl Unit I1st Quarterj 2nd Quarterl 3rd Quarterl 4th Quarterl

12. IODINES I I I I I I l I131 I Ci I 1.60E-06 I 1.05E-03 I 1.73E-06 I -------- I I *I132 I Ci I 1.01E-04 I 5.71E-05 I -------- I -------- I lTotal for Period I Ci I 1.03E-04 I 1.11E-03 I 1.73E-06 I -------- I l3. PARTICULATES l I I I I I l C058 l Ci I _-____ 1.37E-11 I -------- - I _-______I l C060 l Ci I -------- I 1.35E-08 I -------- I 4.77E-06 I l CS134 I Ci I -------- l 1.93E-08 I -------- I 1.01E-06 I l CS137 I Ci I -------- l 1.07E-06 l -------- I 1.08E-04 ITotal for Period I Ci I -------- I 1.10E-OG I -------- I 1.14E-04

DENOTES SUPPLEMENTAL ISOTOPES A1.1-7

2002 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT GASEOUS EFFLUENTS-GRCOUJND LEVEL RELEASES BATCH MODE INuclides releasedj Unit I1st QuarzerJ 2nd Quarter] 3rd Quarter] 4th Quarter]

l1. FISSION GASES l I I I l H3 I Cl l 63E-01 1 l 1.10--01 j 3.46E-01 I 1 72E-Q1 I l AR41 Ci l 5 44E-01 j 2 02--00 l 3.84E5i-0 I 6.10E-C1 I I KR85 I Ci I 2.73E-02 l 2 35E-01 l 2 08E+00 I 4.98E-01 I l X-131M I Cl I 4.48E-02 I 2.76E-02 I -------- I -------- I I XE133M I Ci J 1 87E-02 l 2.665-02 l -------- I -------- I l XE133 I Ci I 3.36E+00 l 6 025+00 l 8.23E+00 2.31E500 I I XE135 I Ci I 6.77E-02 j 1.50E-0l I 2.525-01 l 3 32E-02 I lTotal for Period I Ci I 4.23E+00 I 8.65E+00 j 1.47E-01 I 3.62E+00 I

12. IODINES I I I I I I l T131 I Ca j -------- I 4.38E-07 --------

- I -------- I l I133 I Cl I -------- I 1.12E-07 -------- - I -------- I lTotal for Period I Cl I -------- I 5.50E-07 J ------- I -------- I l3 PARTICULATES I I I I I I I CS137 I Ci I ________ I 7.41_-07 -------- - I -------- I lTotal for Period I Ci j -------- I 7.41E-07 I -------- I ----- ___ I 1:-

2002 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES l Units 1st 2nd I 3rd I 4th lEst. I l l Quarter l Quarter Quarter l Quarter lTotal I I I I lError,%I lA.IFISSION AND I I I lACTIVATION GASES I I I I i li.lTotal Release I Ci I 4.07E+00l 7.52E+01l 1.44E+01l 3.45E+00l 12.9 l l2.[Average release luCi/secI 5.23E-01l 9.56E+00l 1.81E+00l 4.34E-01l I I irate for period I I I I I I I l3.lPercent of 1% Gammal 2.28E-021 3.60E-01l 4.94E-01l 3.36E-02l I I applicable limit I Beta I 1.02E-021 2.61E-011 1.62E-011 1.41E-021 I IB.IIODINES I I I I I I I Il.ITotal I-131 I Ci I 1.GOE-06l 1.05E-03I 1.73E-06I 0.OOE+00l 12.7 l l2.lAverage release luCi/secl 2.06E-07I 1.34E-041 2.18E-07I 0.OOE+00l I I Irate for period I I I I I I 13.lPercent of I % I 3.07E-01l 8.83E-011 6.49E-01 0.OOE+00l I I applicable limit I I I I I I IC.IPARTICULATES I I I I I I I ll.lParticulates withl Ci I 0.OOE+00l 1.84E-06l 0.OOE+00l 1.14E-04l 20.3 l l lhalf lives>8 daysl I I I I I I l2.lAverage release luCi/secl 0.OOE+00l 2.34E-07l 0.OOE+00l 1.43E-0Sl I I irate for period I I I I I I I 13.lPercent of I 0.OOE+00l 8.83E-01I O.OOE+00l 2.32E-Oll I applicable limit IlI I I I 14.lGross alpha I Ci 1<1.32E-061<2.06E-061<1.40E-061<5.52E-071 I radioactivity I I I I I I ID.lTritium I I I I I I I Il.lTotal Release I Ci I 4.67E+01l 4.26E+01l 2.87E+01l 2.41E+01l 10.8 l l2.lAverage release luCi/secl 6.OOE+00l 5.42E+00l 3.61E+00I 3.03E+00l I irate for period I I I I I I 13.lPercent of I % I 9.21E+01l 8.83E+01l S.69E+01l 7.87E+01l I applicable limit I I I I I Al.1-9

2002 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT LIQUID EFFLUENTS CONTINUOUS MODE lNuclides releasedl Unit I1st Quarterl 2nd Quarterl 3rd Quarterl 4_h Quarterl H3 IC I j 5.11E-02 l 6.03E-03 l 4 55E-03 I 3.19E-02 I 3ATCH MODE lNuclides releasedl Unit I1st Quarter! 2nd Quarter! 3rd Quarterl 4th Quarter[

l H3 I Ci I 2.18E-02 I 2.28E+02 I 6.90E+01 I 1 63E-02 l NA24 I Ci l 1.89E-04 l 1.71E-03 I _------- I ________

l CR51 I Ci I -------- I 2.38E-04 !- -- I ------- - I l MN54 I C l 1.54E-03 I 7.51E-04 I 1.43E-05 l 8 44E-06 I l FE55 I Ci I 4.59E-04 I 1.01E-03 I 8.05E-04 I _______ I l C058 I Ci I 2.59E-02 I 4.07E-02 I 2.82E-03 7.04E-04 I l C060 I Ci l 1.71E-03 l 4.93E-03 I 8.87E-04 l 9 83E-04 I l ZR95 l Ci I -------- I 9.53E-05 I -------- I 1.13E-05 I l NE95 I Ci I 1.33E-05 I 2.43E-04 I 6.20E-05 I 3.01E-05 I l AGlOM I Ci I -------- I 2.84E-04 I 2.03E-04 I 1.27E-04 I l C057 I Ci I 1 09E-05 I 2.00E-04 I 4.94E-06 --------

- I l SB124 I Ci I 1 40E-02 I 6.10E-02 I 2.62E-04 I 1.29E-04 I l SB125 I Ci I 5.27E-03 I 4.71E-03 I -------- I 1.35E-04 l TE132 I Ci I -------- l 2.76E-05 1 ____ - ______

-- -_ I l I131 I Ci I 1 32E-05 l 5.60E-03 1I -___ - I -------- I 1 I132 I Ci I -------- l 4 13E-05 I _____ I_ ------- I l CS137 I Ci l 4.07E-05 1 ---- _ - I -------- _ -- I l *A.aG08M I Ci -------- [

l 3.86E-05 1 _ __ I - I

SB12 Ci 3.82E-04 32 12E-03 -- - --------

SN-1,7M I Ci l 9 18E-06 5 61E-04 l 4.40E-06 --------

i *S3>26 l Ci 23 065E-0, ___ j ____ I_ ------- - I

XE-7-33 Ci -------- 3-------- --------

DENOTES SUDPLEMENTAL ISOTOPES A: 1-_ s

2002 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES CONTINUOUS Units 11st 2nd 3rd 1 4th lEst. I Quarter I Quarter I Quarter I Quarter ITotal I 1 1 1 I I I lError,%j IA.IFISSION AND I I I lACTIVATION I I I I I I I I lPRODUCTS I I I I I I Il.lTotal Release I Ci I 0.00E+00I 0.OOE+00l 0.ooE+00I 0.ooE+00I N/A I 12.lAverage diluted luCi/ml I 0.OOE+00I 0.OOE+00l 0.00E+001 0.OOE+00l I I concentration I I I I I I I during period I I I I I I 13.lPercent of I % I 0.OOE+001 0.OOE+001 0.00E+00l 0.OOE+00l I lapplicable limit I I I I I lB.-TRITIUM I I I I I I I l1.lTotal Release I Ci I 5.llE-021 6.03E-031 4.55E-031 3 19E-021 12.2 1 l2.lAverage diluted juCi/ml I 1.14E-10l 2.46E-lll l.55E-llI 3.76E-lll I I lconcentration I I I I I I during period I I I I I 13.-Percent of I %- l.14E-051 2.46E-061 1.55E-06l 3.76E-06l I I applicable limit I I I I I I lC.IDISSOLVED AND I I I I I lENTRAINED GASES I I I I ll.lTotal Release I Ci I 0.OOE+00l 0.OOE,001 0.00E+001 0.OOE+00l N/A I 12.lAverage diluted luCi/ml I 0.OOE+00l 0.00E+00l 0.OOE+00l 0.OOE+00l I 1concentration I I I I I I during period I I I I I 13.lPercent of 1 % 1 0.OOE+00I 0.OOE+00l 0.OOE+00l 0.OOE+00l I I applicable limit I I I I I I ID. Gross Alpha I Ci 1<5.41E-031<7.37E-031<7.31E-031<4.72E-031 N/A I I Radioactivity I I I I I I ITotal Release I I I I I IE.lVolume of Waste I Litersl 2.51E+07l 3.97E+06l 2.87E+061 1.96E+07l 2.00 1 ljReleased I I I I I I IF. Volume of I Litersl 4.50E+11l 2.45E+11l 2.93E+11l 8.49E+11l 3.48 l I IDilutlon Water I I I I I I I I used During I I I I I I I Period I I I I I I Al.l-li

2002 EFFLUEN A.ND WASTE DISPOSAL AMNUAL REPORT LIQU:D EFFLUENTS-SUMMATION OF ALL RELEASES BATCH I Units [ 's. j 2nd 3rd 4tEst.

j Quarter l Quarter l Quarter Quarter lTctal I i I i I $rror,%I I IA.IFISSION AND [ i i I I I lACTWIATION I i j j !

l lPRODUCTS i i i Il.Total Release I Ci j 4.95--021 1 25E-011 5 06E-031 2 13E-031 11.9 I

12. Average diluted luCi/ml I 7.43E-091 l 06E-081 1 0l-E-091 3 58E-101 I I Iconcentration I I I I I I I I ldurzng period j I I I I I I 13.jPercent of j 6.28E-021 1 55E-0ll 1 04E-021 6 87E-03I I I applicable limit lI I I I I I lB ITRITIUM I I I I I I I l.lTotal Release I Ci l 2.18E+021 2.28E+021 6 90E+011 1.63E+021 10.1 I 12.jAverage diluted IuCi/ml I 3.30E-05l l.94E-05I l.38E-05I 2.74E-051 I I Iconcentration I I I I I I I I Iduring period I I I I I I I 3 jPercent of I k I 3.30E+001 1 94E+00 1.38E+00j 2.74-+001 I I applicable limit I I I I I I I jC.lDISSOLVED AND I I I I I I I I IENTRAINED GASES I I I I I l1 lTotal Release I Ci I 0.00E00! 3.92E-041 0.00E+001!.OOE+00l 28 6 I 12 lAverage diluted IuCi/ml I 0 OOE-00j 3.33E-llI 0.OOE+00l 0.00E+00l I I Iconcentration I I I I I I during period j I I I I I I3.IPercent of j I 0 OOE-001 l.67E-051 0.OOE+00l 0.00E+00Qa I applicable limit I I II I I I ID.IGross Alpha I Cl 1<9.23E-05I<1.71E-04Ic4.SOE-05I<2 37E-051 N/A I

! Radloactivity I I ! I i i I I ITotal Release I jI Ij I I IE.ivolume of waste I Latersl 1 24E-06! 1.45E-06j 5 63E-051 6.24E-05l 2 00 I IjReleased j I.IE1volume of I Lacersl 6 62E-09, 1 13Ef+10 S OCE-09! 5 95E-091 3 48 1 I IDilution Water I i j I uased Durin! i I Pericd I I

_I 1 -

2002 Effluent and Waste Disposal Annual Report Solid Waste and Irradiated Fuel Shipments Solid Waste Shipped Offsite for Burial or Disposal

1) Type of Waste Unit Estimated Estimated Total Error, %

amount a) Spent resins, filters, sludge, m3 6.30E-01 1.OOE+00 evaporator bottoms, etc. Curies 1.85E+01 3.75E+00 b) Dry compressible waste, m3 2.86E+01 1.OOE+OO contaminated equipment, etc Curies 7.97E+00 6 48E+00 3

c) Irradiated components, control Curies rods, etc.

d) Other m3 Curies _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

2) Estimate of Principle Radionuclide Composition a) Cs-I 34 2% Ni-634 Cs-I 37 2% Co58 69 %

Fe-55 2% H-3 18 %

Co-60 3% Mn-54 1%

b) H-3 5% Sb-1 25 2%

Cs-1 37 2% Ni-63 19 %

Co-60 36 % Fe-55 3 3)Solid Waste Disposition No. of Shipments Mode of Transportation Destination 8Truck Barnwell, SC 8L Truck Clive, UT

4) Type of Containers used for Shipment- Containers used are strong, tight metal boxes, drums and high Iintegrity containers.

11

15) Solidification Agent: There were no solidifications promdduring this repot eriod.

Al 1-13

2002 Effluent and Waste Disposal Annual Report Yearly Release i Rates GASES Fission and Activation Gases Total Release 9.71E+01 Curies Average Release Rate 3.08E-00 uCi/sec

% of Applicable Limits I 1.14E-02 %

3 2 24E-01 %NO lodines Total I-131 Release 1 05E-03 Curies Averagze Release Rate 3 34E-05 ptCiIsec

%Oof Applicable Limit 1 02E-'(00 %

Particulates Total Release I 16E-04 Curies Average Release Rate 3 67E-06 uCi/sec

° of Applicable Limit 1.04E+00 %

LIQUIDS Fission and Activation Products Total Release 1.82E-01 Curies Average Diluted Concentration 6.19E-9 uCi/ml j % of Applicable Limits Total Body 7 OOE-01 %

Organ 2 72E-01 %

A1.-1-14

The following distances were used in the calculation of the maximum individual doses:

Sector Direction Boundary (Meters) Nearest Residence (Meters)

A N 651 659 B NNE 617 660 C NE 789 943 D ENE 1497 1747 E E 1274 1716 F ESE 972 1643 G SE 629 1640 H SSE 594 1417 J S 594 1026 K SSW 629 942 A" . -15

Summary of Maximum Individual Doses First Quarter 2002 EFFLUENT APPLICABLE ESTIMATED AGE LOCATION % OF LIMIT ORGAN DOSE (mrem) GROUP DIST DIR (M) APPLICABLE (mrem)

. (Toward) LIMIT QTR Liquid Total Body 9.54E-03 Child Receptor I 6.36E-01 l.5E+O Liquid GI-Tract 9.87E-03 Child Receptor 1 1.97E-01 5.OE+0 Noble Gas Air Dose 1.14E-03 651 (N) 2.28E-02 5.OE+O (Ganrna-mrad)

Noble Gas Air dose 1.02E-03 651 (N) 1.02E-02 1.OE+l (Beta-mrad)

Iodines and Thyroid 2.30E-02 Child 659 (N) 3.07E-01 7.5E+O Particulates Al 2-1

Summary of Maximum Individual Doses Second Quarter 2002 EFFLUENT APPLICABLE ESTIMATED AGE LOCATION % OF LIMIT ORGAN DOSE (mrem) GROUP DIST DIR (M) APPLICABLE (mrem)

(Toward) LIMIT QTR Liquid Total Body 6 16E-03 Child Receptor 1 4.1 1E-O I 15E-0 Liquid Thyroid I 18E-02 Child Receptor I 2.36E-01 5 OE-tO Noble Gas Air Dose I SOE-02 651 (N) 3.60E-01 5 OE+O (Gamma-mrad)

Noble Gas Air dose 2 61 E-02 651 (N) 2.61 E-0 I 1 OE+1 (Beta-rnrad) lodines and Thyroid 6 62E-02 Child 659 (N) 8.83E-01 7 5E+O Particulates Al 2-2

Summary of Maximum Individual Doses Third Quarter 2002 EFFLUENT APPLICABLE ESTIMIATED -AGE LOCATION % OF LIMIT ORGAN DOSE (mrem) GROUP DIST DIR (M) APPLICABLE (mrem)

(Toward) LIMIT QTR Liquid Total Body 1.61E-03 Child Receptor I 1.07E-01 1.5E+O Liquid GI-Tract 1.73E-03 Child Receptor 1 3.46E-02 5.0E+0 Noble Gas Air Dose 2.47E-02 651 (N) 4.94E-0I 5.OE+O (Garnma-mrad)

Noble Gas Air dose 1.62E-02 651 (N) 1.62E-0 I l.OE+1 (Beta-mrad)

Iodines and Thyroid 4.87E-02 Child 659 (N) 6.49E-01 7.5E+O Particulates A1.2-3

Summary of Maximum Individual Doses Fourth Quarter 2002 EFFLUENT APPLICABLE ESTIMATED AGE LOCATION 0%O OF LIMIT ORGAN DOSE (mremn) GROUP DIST DIR (M) APPLICABLE (mnrem)

(Toward) LIMIT QTR Liquid Total Body 3 70E-03 Child Receptor 1 2 47E-01 I 5E-0 Liquid GI-Tract 3 76E-03 Child Receptor 1 7 52E-02 5 OE+O Noble Gas Air Dose I 6SE-03 651 (N) 3 36E-02 5 OE+O (Gamma-mrad)

Noble Gas Air dose I 41E-03 651 (N) 1 41E-02 I.OE+I1 (Beta-mrad) lodines and Total Body I 74E-02 Child 659 (N) 2 32E-01 7.5E+O Particulates Al 2-4

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02010101 - 02033124 STABILITY CLASS: A DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 0 13 8 0 0 21 0

NNE 0 4 1 0 0 5 0

NE 0 3 3 0 0 6 0

ENE 0 1 8 0 0 9 0

E 0 2 1 0 0 3 0

ESE 0 4 6 1 0 11 0

SE 1 7 8 0 0 16 0

SSE 0 4 13 5 0 _ 22 0

S 0 3 21 5 0 29 0

SSW 0 2 4 1 0 7 0

SW 0 9 17 2 0 28 0

WSW 0 21 32 7 2 62 0

W 0 7 15 1 0 23 0

WNW 0 19 10 0 0 29 0

NW 1 14 4 0 0 19 0

NNW 0 13 12 1 0 26 TOTA_____________2____12______ 163_______23__ 2_____0____316____

TOTAL 2 126 163 23 2 0 316 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2 .1-1

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02010101 - 02033124 STABILITY CLASS- 3 DT/DZ ELEVATION SPEED SP1OM DIRECTION DIR1OM LAPSE-DTG0M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-13 19-24 >24 TOTAL N 0 4 3 0 0 0 7 NNE 0 4 1 0 0 0 5 NE 0 3 1 0 0 0 4 ENE 0 2 0 0 0 0 2 E 1 4 2 0 0 0 7 ESE 2 1 1 0 0 0 4 SE 1 2 2 0 0 0 5 SSE 2 3 10 0 0 0 15 S 0 3 23 2 0 0 28 SSW 0 3 12 3 0 0 18 SW 0 2 10 0 0 0 12 WSW 0 2 5 0 1 0 8 W 0 8 8 1 0 0 17 WNW 0 8 8 0 0 0 16 NW 1 8 4 0 0 0 13 NNW 2 8 7 1 0 0 18 TOTAL 9 65 97 7 1 0 179 PERIODS OF CALM(HOURS) 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA- 5 12 1 2

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02010101 - 02033124 STABILITY CLASS: C DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT6OM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 1 9 3 2 0 0 15 NNE 1 4 1 0 0 0 6 NE 1 4 6 0 0 0 11 ENE 0 3 6 3 0 0 12 E 3 3 1 0 0 0 7 ESE 0 3 0 0 0 0 3 SE 2 4 0 0 0 0 6 SSE 0 4 2 0 0 0 6 S 1 3 8 4 0 0 16 SSW 0 4 6 0 0 0 10 SW 1 0 8 0 0 0 9 WSW 0 4 7 0 3 0 14 W 0 5 6 2 0 0 13 WNW 1 3 10 2 0 0 16 NW 2 3 7 0 0 0 12 NNW 0 10 10 0 0 0 20 TOTAL 13 66 81 13 3 0 176 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2. 1-3

SITE- AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD 02010101 - 02033124 STABILITY CLASS: D DT/DZ ELEVATION SPEED SPlOM DIRECTION DIR1OM LAPSE DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 3-12 13-18 19-24 >24 TOTAL N 2 32 33 6 73 NNE 7 20 0 0 6 0 33 NE o 15 0 0 3 0 24 ENE 3 22 7 0 0 0 32 E 7 23 0 0 3 0 33 ESE 2 14 0 0 3 0 19 SE 5 0 0 16 8 0 29 SSE 5 25 0 0 42 0 _ 72 S 5 34 -245 0 50 -1 - 1'4 SSW 3 39 0 95 13 0 1,0 SW 4 30 0 81 7 0 122 WSW 2 20 66 0 12 5 105 W 4 37 33 0 6 3 83 WNW 6 43 45 0 0 0 99 NW 6 27 0 31 0 0 64 NNW 10 33 29 0 1 0 73 TOTAL 77 435 540 69 9 0 1130 PERIODS OF CALM(HOURS)- 0 VARIABLE DIRECTION- 0 HOURS OF MISSING DATA. 5 A2 1-4

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02010101 - 02033124 STABILITY CLASS: E DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DTGOM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 2__ 0-0 - - -

0-- - -

0 N 2 0 0 0 0 2 0

NNE 6 2 0 0 0 8 0

NE 4 4 0 0 0 8 0

ENE 0 3 0 0 0 3 0

E 4 9 0 0 0 13 0

ESE 3 4 0 0 0 7 0

SE 4 18 2 0 0 24

-0 SSE 5 33 13 3 0 54

-0 S 4 49 15 2 0 70 0

SSW 4 17 8 0 0 29 0

SW 0 11 8 0 0 19 0

WSW 3 10 5 0 0 18 0

W 2 9 0 0 0 11 0

WNW 0 1 0 0 0 1 0

NW 1 1 0 0 0 2 0

NNW 2 2 0 0 0 4 TOT____________44______173_______51_____5____0____0____273____

TOTAL 44 173 51 5 0 0 273 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2.1-5

SITE. AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD. 02010101 - 02033124 STABILITY CLASS: F DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT5OM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL

- 0 N 0 0 0 0 0 0 0

NNE 1 0 0 0 0 1 0

NE 2 2 0 0 0 4 0

ENE 1 1 0 0 0 2 0

E 2 2 0 0 0 4 0

ESE 2 0 0 0 4 2 0 SE 6 3 0 0 0 9 0

SSE 8 6 0 0 0 '4 0

S 11 10 0 0 0 21 0

SSW 2 0 0 0 3 0

SW 1° 0 1 0 0 0 1 0

WSW 1 0 0 0 0 1 0

W 0 0 0 0 0 0 0

WNW 0 0 0 0 0 0 0

NW 0 0 0 0 0 0 0

NNW 2 0 0 0 0 2 TOTAL 37 29 0 0 0 0 66 PERIODS OF CALM(HOURS): 0 VARIA3LE DIRECTION 0 HOURS OF MISSING DATA 5 A2 1-6

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02010101 - 02033124 STABILITY CLASS: G DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 0 0 0 0 0 0 0 ENE 1 0 0 0 0 0 1 E 4 0 0 0 0 0 4 ESE 3 0 0 0 0 0 3 SE 4 0 0 0 0 0 4 SSE 1 0 0 0 0 0 1 S 1 0 0 0 0 0 1 SSW 0 1 0 0 0 0 1 SW 0 0 0 0 0 0 0 WSW 0 0 0 0 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 0 0 0 0 0 0 0 NNW 0 0 0 0 0 0 0 TOTAL 14 1 0 0 0 0 15 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2.1-7

SITE AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02010101 - 02033124 STABILITY CLASS ALL DT/DZ ELEVATION: SPEED SP1OM DIRECTION DIR1OM LAPSE DTGOM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-13 19-24 >24 TOTAL 0

N S 53 52 3 0 123 0

NNE 15 34 9 0 0 58 0

NE 13 31 13 0 0 57 0

ENE 5 32 21 3 0 61 0

E 21 43 7 0 0 71 0

ESE 12 28 10 1 0 51 0

SE 23 50 20 0 0 93 0

SSE 21 75 80 3 0 134 S 22 102 117 37 1 0 279 SSW 3 68 125 17 0 0 218 SW 5 53 124 9 0 0 191 WSW 6 57 115 19 11 0 208 W 6 66 62 10 3 0 147 WNW 79 73 2 0 0 161 NW 11 53 46 0 0 0 110 NNW 15 66 53 3 0 0 143 TOTAL 196 893 932 117 15 0 2155 PERIODS OF CALM(HOURS) 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA. 5 A- 1 3

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02040101 - 02063024 STABILITY CLASS: A DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 5 47 27 0 79 0 0 NNE 3 4 1 0 8 0

0 0 NE 0 5 0 0 5 0 0 ENE 0 3 2 0 5 0

E 1 4 2 0 0 7 0

ESE 2 10 0 0 0 12 0

SE 2 10 2 0 0 14 0

SSE 0 20 15 2 0 _ 37 0

S 6 12 31 12 0 61 0

SSW 2 4 6 4 0 16 0

SW 1 34 16 0 0 51 0

WSW 3 46 29 2 0 80 0

W 3 23 3 0 0 29 0

WNW 3 31 0 0 0 34 0

NW 4 56 4 0 0 64 0

NNW 8 76 21 0 0 105 TOTAL 43 385 159 20 0 0 607 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 0 A2.2-1

SITE. AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02040101 - 02063024 STABILITY CLASS: 3 DT/DZ ELEVATION SPEED:SP1OM DIRECTICN DIR1OM LAPSE DT60M WIND SPEED (CMPH)

WIND DIRECTION 4-7 S-12 13-13 19 - -24 >24 TOTAL N 2 5 4 0 11 NNE 0 0 3 2 0 0 5 00 0 NE 1 3 0 0 4 ENE 0 0 2 0 0 0 2 E 1 0 1 1 0 0 3 ESE 2 0 3 1 0 0 6 SE 1 0 1 0 0 0 2 SSE 3 0 2 2 0 0 7 S 0 0

3 3 2 1 0 9 SSW 0 0 3 3 1 0 7 SW 0 0 9 _ 0 O 14 WSW 3 0 1 2 0 0 6 W 2 0 1 0 PI 0 3 WNW 1 0 1 0 0 0 2 NW4 1 0 4 1 0 0 6 NNW 4 7 0 7 0 0 18 TOTAL_____________2____

TOTAL 29 46 23 2 0 0 105 PERIODS OF CALM(HOURS) 0 VARIA3LE DIRECTION.

HOURS OF MISSING DATA: 0

,-'2 o- 2

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02040101 - 02063024 STABILITY CLASS: C DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 2 12 2 0 0 0 16 NNE 3 2 0 0 0 0 5 NE 0 1 1 0 0 0 2 ENE 0 1 0 0 0 0 1 E 0 1 2 0 0 0 3 ESE 2 5 2 0 0 0 9 SE 0 2 1 0 0 0 3 SSE 1 6 4 0 0 0 11 S 0 1 3 0 0 0 4 SSW 0 5 3 0 0 0 8 SW 1 5 4 1 0 0 11 WSW 1 1 6 1 0 0 9 W 1 0 0 0 0 0 1 WNW 1 3 2 0 0 0 6 NW 0 8 3 0 0 0 11 NNW 5 13 2 0 0 0 20 TOTAL 17 66 35 2 0 0 120 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 0 A2 .2-3

SITE AEP COOK HOURS AT EACH WIND SPEED AND DIRECTICN PERIOD OF RECORD: 02040101 - 02063024 STABILITY CLASS: D DT/DZ ELEVATION: SPEED SP1OM DIRECTICN:DIR1OM LAPSE:DT5OM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 17 563 9 1 0 0 85 NNE 157 1 0 0 3-2 0

NE 7 -3 0 0 0 20 0

ENE 5 6 a 0 0 19 0

E 4 0 0 19 0

ESE 9 23 24 0 0 56 SE 4 is 4 2 0 0 25 SSE 5 20 11 0 0 0 3-5 S 2 14 23 5 0 0 44 SSW 4 13 24 0 0 46

-5 SW 2 9 0 0 37 WSW 2 9 13 1 0 0 25 W 13 34 2 0 0 0 29 WNW 12 13 3 0 0 0 23 NW 11 14 4 0 0 0 29 NNW 25 34 10 0 0 0 69 0 0 TOTAL 137 295 160 i, 0 0 599 PERIODS OF CALM(HCURS) 0 VARIABLE DIRECTION 0 HOURS OF MISSING DATA 0

._ 2-4

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02040101 - 02063024 STABILITY CLASS: E DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT6OM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 9 9 0 0 0 0 18 NNE 8 1 0 0 0 0 9 NE 4 3 1 0 0 0 8 ENE 10 4 1 0 0 0 is E 6 7 1 0 0 0 14 ESE 14 9 3 0 0 0 26 SE 9 17 1 0 0 0 27 SSE 12 25 1 0 0 0 38 S 18 52 22 0 0 0 92 SSW 8 16 6 0 0 0 30 SW 7 19 9 0 0 0 35 WSW 1 13 7 0 0 0 21 W 7 7 0 0 0 0 14 WNW 4 13 0 0 0 0 17 NW 5 4 1 0 0 0 10 NNW 8 4 0 0 0 0 12 TOTAL 130 203 53 0 0 0 386 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 0 A2. 2 -5

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 020 40101 - 02063024 STABILITY CLASS F DT/DZ ELEVATION- SPEED:SP1OM DTRECTICN:DIR10M LAPSE.DTGOM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 3-12 13-13 19-24 >24 TOTAL N 3 0 0 0 0 3 0

NNE 1 0 0 0 0 1 NE 2 0 0 0 0 2 ENE 9 3 0 0 0 0 12 E 2 0 0 0 0 14 ESE 15 0 0 0 0 0 15 0

SE 23 3 0 0 0 0 26 SSE 20 6 0 0 0 0 26 S 22 3 0 0 0 00 30 SSW 5 5 0 0 0 00 L5 SW 5 0 0 0 0 00 WSW 5 2 0 0 0 0 7 W 4 0 0 0 0 4 WNW 3 1 0 0 0 4 NW 2 0 0 0 0 2 NNW 1 0 0 0 0 1

___________132 TOTAL 13 2 30 0 0 0 0 162 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION-0 HOURS OF MISSING DATA:

A2: 2-o

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02040101 - 02063024 STABILITY CLASS: G DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M

_______________________________________WIND__SPEED______P_)

WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 0 1----

N 1 0 0 1 0 0 0 NNE 1 0 0 1 0 0 0 NE 4 0 0 4 0 0 0 ENE 8 0 0 8 0 0 0 E 22 3 0 25 0 0 0 ESE 38 0 0 38 0 0 0 SE 36 0 0 36 0 0 0 SSE 30 1 _ 0 0 31

_ _ 0

_ _ 0 S 31 1 0 32 0

0 0 0 00 SSW 11 1 0 12 0 0 0 SW 8 0 0 8 0 0 0 WSW 2 0 0 2 0 0 0 W 2 0 0 2 0 0 0 WNW 2 0 0 2 0 0 0 NW 2 0 0 2 0 0 0 NNW 1 0 0 1 0 0 0 0_____205______

TOTAL 199 6 0 205 PERIODS OF CALM((HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 0 A2. 2 -7

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD 02040101 - 02063024 STAB ILITY CLASS: ALL DT/DZ ELEVATICN: SPEED:SP1OM DIRECTION DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 9-12 13-18 19-24 >24 TOTAL N 39 131 42 1 0 0 213 NNE 34 25 2 0 0 0 61 NE 18 25 2 0 0 0 45 ENE 34 17 1_ 0 0 0 62 E 46 23 0 0 0 35 ESE 82 50 30 0 0 0 152 SE 75 43 3 2 0 0 133 SSE 71 80 33 0 0 136 S 82 91 31 13 0 0 272 SSW 30 47 42 10 0 0 129 SW 24 80 53 4 0 0 161 WSW 1 1 72 57 4 0 0 150 W 32 45 5 0 0 0 92 WNW 26 62 5 0 0 0 93 NW 25 86 13 0 0 0 124 NNW 52 134 40 0 0 0 226 TOTAL 687 1021 435 41 0 0 2184 PERIODS OF CALM(HOURS) 0 VARIABLE DIRECTION:

HOURS OF MISSING DATA. 0

- I -p

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02070101 - 02093024 STABILITY CLASS: A DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 14 101 0 0 0 0 115 NNE 0 8 0 0 0 0 8 NE 3 14 0 0 0 0 17 ENE 1 22 2 0 0 0 25 E 2 22 0 0 0 0 24 ESE 6 8 0 0 0 0 14 SE 6 9 0 0 0 0 15 SSE 7 13 0 0 0 0 20 S 7 21 15 2 0 0 45 SSW 3 17 12 3 0 0 35 SW 1 54 10 0 0 0 65 WSW 7 48 3 0 0 0 58 W 10 25 0 0 0 0 35 WNW 4 19 0 0 0 0 23 NW 12 20 0 0 0 0 32 NNW 29 65 0 0 0 0 94 TOTAL 112 466 42 5 0 0 625 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION- 0 HOURS OF MISSING DATA: 5 A2 .3 -1

SITE. AEP COOK HOURS AT EACH WIND SPEED AND DIRECTICN PERIOD OF RECORD: 02070101 - 02093024 STABILITY CLASS 3 DT/DZ ELEVATION: SPEED S212M DIRECTION DIRM0M LAPSE:DTS0M WIND SPEED (MPH)

N DID DIRECTION 1-3 4-7 8-12 13-13 19-24 >24 TOTAL N 2 5 0 0 0 0 7 NNE 0 0 0 0 0 0 0 NE 2 0 0 0 0 0 2 ENE 2 1 0 0 0 0 3 E 2 3 0 0 0 0 5 ESE 3 2 0 0 0 0 5 SE 2 0 0 0 0 0 2 SSE 1 0 0 0 0 0 1 S 2 5 4 0 0 0 11 SSW 0 8 6 0 0 0 14 SW 1 10 2 0 0 0 13 WSW 2 0 0 0 0 0 2 N 2 2 0 0 0 0 4 WNW 4 0 0 0 0 0 4 NW 3 0 0 0 0 0 3 N`4qW 3 1 0 0 0 0 4 TOTAL 31 37 12 0 0 0 30 PERIODS OF CA-LM(HCURS): 0 VAILRIABLE DIRECTION 0 HOURS 2? MISSING DATA 5

-'2

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02070101 - 02093024 STABILITY CLASS: C DT/DZ ELEVATION- SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13 -18 19-24 >24 TOTAL 4 7 0 0 11 N 0 0 0 2 0 0 2 NNE 0 0 1 0 0 0 1 NE 0 0 0 1 0 0 1 ENE 0 0 1 0 0 0 1 E 0 0 5 0 0 0 5 ESE 0 0 0 0 0 0 0 SE 0 0 1 0 0 __ _ 0 1 SSE 00 0 2 7 3 0 12 S 0 0 2 5 5 0 12 SSW 0 0 SW 0 4 2 0 6 0 0 WSW 0 2 0 0 2 0 0 0 0 0 0 0 W 0 0 WNW 3 0 0 0 3 0 0 1 0 0 0 1 NW 0 0 2 0 0 0 2 NNW TOTAL 22 28 10 0 0 0 60 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2.3-3

SITE AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD 02070101 - 02093024 STABdILITY CLASS. D DT/DZ ELEVATION: SPEED:SP10M DIRECTION DIR1OM LAPSE DT60M W:ND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-13 19-24 >24 TOTAL

£---

N 27 4 0 53 NNE 10 0 0 9 0 0 19 NE 0 0 2 0 0 9 ENE 0 0 4 4 0 0 a 0 0 4 6 0 0 10 ESE 0 0 2 4 0 0 6 SE 0 0 3 1 0 0 4 SSE 0 0 3 0 _

0 3 S 0 0 0 3 24 6 0 33 SSW 17 0 0 43 37 0 85 SW 0 0 6 21 6 0 33 WSW 0 0 6 1 0 9 0 0 4 0 0 5 WNW 0 0 4 7 0 0 11 NW 0 0 9 2 0 0 11 N'NW 0 0 17 4 0 0 -1

_ 1_ _ _ 1__

_ _ _ _ 54_0-0 TOTAL 110 lol 54 0 0 0 325 PERIOCS OF CALM(HCURS). 0 VARIABLE 2_RECT-ON. 0 HOURS OF MI-SSING DATA.

.A 3-4

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02070101 - 02093024 STABILITY CLASS: E DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 32 4 0 0 0 0 36 NNE 22 9 0 0 0 0 31 NE 25 16 1 0 0 0 42 ENE 18 9 0 0 0 0 27 E 23 7 0 0 0 0 30 ESE 14 2 0 0 0 0 16 SE 8 2 0 0 0 0 10 SSE 11 0 0 0 0 0 11 S 46 93 7 0 0 0 146 SSW 10 33 4 0 0 0 47 SW 7 41 2 0 0 0 50 WSW 9 4 1 0 0 0 14 W 5 8 0 0 0 0 13 WNW 11 3 0 0 0 0 14 NW 5 2 0 0 0 0 7 NNW 9 0 0 0 0 0 9 TOTAL 255 233 15 0 0 0 503 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2.3-5

SITE. AEP CCOK HOURS AT EACH WIND SPEED AND DIRECTION PERICD OF RECORD: 02070101 - 02093024 STB3ILITY CLASS- F DT/DZ ELEV'ATION- SPEED SP1OM DIRE-CTION:DIR1OM LAPSE DT50M WIND SPEED (MPH)

WIND DIRECTION ;-3 4-7 8-12 13-13 19-24 >24 TOTAL N 6 0 0 0 0 0 6 tINE 9 0 0 0 0 0 9 NE 22 0 0 0 0 0 22 ENE 30 0 0 0 0 0 30 E 23 0 0 0 0 0 23 E SE 27 3 0 0 0 0 30 SE 9 0 0 0 0 0 9 SS S 11 0 0 0 0 0 i1 S 31 li 0 0 0 0 42 SSW 17 2 0 0 0 0 9 SW 4 0 0 0 0 0 4

'WS'W 1 1 0 0 0 0 2 W 2 0 0 0 0 0 2 0 0 0 0 0 2 NW 3 0 0 0 0 0 3 1 0 0 0 0 1 TOTAL 198 17 0 0 0 0 215 PERIODS CF CAkLM(HOURS) 0 VARIA3LE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2 3 -rs

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02070101 - 02093024 STABILITY CLASS: G DT/DZ ELEVATION: SPEED:SP10M DIRECTION:DIR1OM LAPSE:DT0M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 4 0 0 0 0 0 4 NNE 7 0 0 0 0 0 7 NE 21 0 0 0 0 0 21 ENE 58 0 0 0 0 0 58 E 81 0 0 0 0 0 81 ESE 71 0 0 0 0 0 71 SE 29 0 0 0 0 0 29 SSE 35 0 0 0 0 0 35 S 47 0 0 0 0 0 47 SSW 21 0 0 0 0 0 21 SW 4 0 0 0 0 0 4 WSW 7 0 0 0 0 0 7 W 4 0 0 0 0 0 4 yNW 1 0 0 0 0 0 1 NW 3 0 0 0 0 0 3 NNW 2 0 0 0 0 0 2 TOTAL 395 0 0 0 0 0 395 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 5 A2.3 -7

SITE. AEP COOK HOURS AT EACH WIND SPEED AuND DIRECTION PERIOD 0CF RECORD 02070101 - 02093024 STA-BILITY CLASS- ALL DT/DZ ELEVAZTON SPEED:SP1OM DIRECTICN D-R'3M LAPSE-DT60M WIND SPEED (MPH)

WIND D7RECT7ON 1-3 4-7 6-12 13-13 13-24 >24 TOTAL N a9 144 4 0 0 0 237 NNE 43 28 0 0 0 0 76 NE 32 1 0 0 0 114 ENJE 113 37 2 0 0 0 152 136 3,8 0 0 0 0 174 ESE 128 19 0 0 0 0 147 SE 12 0 0 0 0 69 SSE 69 13 0 0 0 0 82 S 1383 161 35 2 0 0 336 583 108 64 0 0 233 0

SSW 23 130 22 0 0 175 0

WSW 23 61 5 0 0 94 0

W 27 36 0 0 0 63 0

WNW 29 29 0 0 0 53 0

NVW 36 24 0 0 0 60 0

NNW 63 70 0 0 0 133 TOTAL 1123 942 133 5 0 0 2203 PER-ODS OF CALM(HOURS): 0 V.ARIAB3LE DIRECTION:

HOURS OF MISSING DATA:

A2 '-8

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02100101 - 02123124 STABILITY CLASS: A DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 0 29 1 0 0 0 30 NNE 0 4 0 0 0 0 4 NE 0 16 7 0 0 0 23 ENE 1 10 1 0 0 0 12 E 1 14 4 0 0 0 19 ESE 2 5 0 0 0 0 7 SE 1 5 0 0 0 0 6 SSE 1 16 0 0 0 0 17 S 0 14 186 0 0 0 32 SSW 0 13 2 0 0 20 SW 0 7 19 1 0 0 27 WSW 0 12 11 0 0 0 23 W 2 17 6 0 0 0 25 WNW 1 18 2 0 0 0 21 NW 1 18 0 0 0 0 19 NNW 3 12 3 0 0 0 18 TOTAL 13 202 85 3 0 0 303 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 4 A2.4-1

SITE- AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02100101 - 02123124 STABILITY CLASS 3 DT/DZ ELEVATION SPEED SP1CM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-13 19-24 >24 TOTAL N 0 5 1 0 0 0 6 NNE 0 1 0 0 0 0 1 NE 0 2 1 0 0 0 3 ENE 0 4 1 0 0 0 5 1 4 0 0 0 6

-S 2 2 1 0 0 5 0

SE 0 0 0 0 0 0 0

SSE 1 5 0 0 0 6 0

S 0 4 10 0 0 0 15 SSW 0 0 4 5 0 0 9 SW 0 3 0 0 0 5 WSW 0 6 3 0 0 0 9 W 0 1 3 0 0 0 4 WNW 4 4 3 0 0 0 8 NW 3 4 0 0 0 0 TNIW 2 3 1 0 0 6 TOTAL 13 51 30 1 0 0 35 PERIODS OF CALM (HOURS): 0 VARIABLE DIRECTION 0 HOURS OF MISSING DAT-A 4

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02100101 - 02123124 STABILITY CLASS: C DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL

_________ 0__ -

0- - -

1- - - -

0- - - -

1-0 N 0 0 1 0 0 1 0

NNE 0 1 0 0 0 1 0

NE 3 4 2 0 0 9 0

ENE 1 9 0 0 0 10 0

E 2 2 0 0 0 4 0

ESE 1 1 5 0 0 7 0

SE 1 2 1 0 0 4 0

SSE 0 3 0 0 0 3 S 0 6 4 2 0 0 12 SSW 0 3 7 1 0 0 11 SW 2 5 12 0 0 0 19 WSW 1 5 10 2 0 0 18 W 0 1 4 2 0 0 7 WNW 0 3 1 0 0 0 4 NW 0 3 1 0 0 0 4 NNW 2 4 1 0 0 0 7 TOTAL 13 52 49 7 0 0 121 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 4 A2 .4-3

SITE: AEP COOK HCURS AT EACH WIND SPEED A'ND DIRECTION PERIOD CF RECORD: 02100101 - 02123124 ST.AB3ILITY CLASS: D DT/DZ ELEVATION: SPEED.SP2OM DIRECTION DIR1OM LAPSE:DT6OM WIND SPEED (MPH)

WIND DIRECTICN 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 16 36 20 0 0 0 72 NNE 10 17 10 0 0 0 37 NE 17 53 3 0 0 0 73 ENE 16 36 0 0 0 0 52 20( 23 4 0 0 0 52 ESE 13 17 0 0 0 41 SE 1, 0 0 0 29 7

SSE 20 10 1 0 0 34 3

43 29 8 0 0 83 SSW 2 23 67 14 0 0 106 SW 6 36 6 0 0 61 WSW 3 30 35 35 0 0 103 W 10 51 46 14 0 0 121 WNW 10- ,3 19 0 0 0 102 NW 11 53 19 0 0 0 33 NN`W 17 52 14 0 0 0 83 TOTAL 161 549 344 78 0 0 1132 PERIODS OF CALM (HCURS): 0 V;ABIABLE DIRECTION 0 HOURS OF MISSING DA.TA 4 A2 4-4

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02100101 - 02123124 STABILITY CLASS: E DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE:DT6OM WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 8 1 0 0 0 0 9 NNE 5 4 0 0 0 0 9 NE 7 12 0 0 0 0 19 ENE 9 7 0 0 0 0 16 E 14 0 0 0 0 0 14 ESE 20 9 0 0 0 0 29 SE 13 9 5 0 0 0 27 SSE 13 9 7 1 0 0 30 S 15 48 19 1 0 0 83 SSW 3 20 8 0 0 0 31 SW 1 9 1 0 0 0 11 WSW 3 a 0 0 0 0 11 W 5 9 0 0 0 0 14 WNW 2 1 0 1 0 0 4 NW 2 2 0 0 0 0 4 NNW 5 5 0 0 0 0 10 TOTAL 125 1,3 40 3 0 0 321 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 4 A2 .4-5

SITE: AEP CCOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD 02100101 - 02123124 STABILITY CLASS: F DT/DZ ELEVATION: SPEED:S?10M DIRECTION:DIR1OM LAPSE:DTSOM WIND SPEED (MPH)

WIND DIRECTICN 1-3 4- 7 8-12 13-18 19-24 >24 TOTAL N 0 0 0 0 0 0 0 NNE 2 0 0 0 2 0 0 NE 5 0 0 0 5 0 0 ENE 4 0 0 0 4 0 0

_x 19 0 0 0 13 0 0 ESE 21 0 0 0 21 0 0 SE 9 0 0 0 9 0 0 SSE 20 2 0 0 22 0 0 S 18 25 - - 0 0 43 0 0 SSW 8 4 0 0 0 0 12 0 0 SW 2 1 0 0 3 0 0 WSW 0 0 0 0 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 1 0 0 0 1 0 0 NNW 0 0 0 0 0 0 0 TOTAL 109 32 0 0 141 PERIODS OF CALM(HCURS) 0 VARIA3LE DIREECTION- 0 HOURS OF MISSING DATA: 4

-2 4-5

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02100101 - 02123124 STABILITY CLASS: G DT/DZ ELEVATION: SPEED:SP1OM DIRECTION:DIR1OM LAPSE-DT60M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 0 0 0 0 0 0 0 NNE 0 0 0 0 0 0 0 NE 4 0 4 0 0 0 0 ENE 10 0 10 0 0 0 0 E 19 0 19 0 0 0 0 ESE 9 0 9 1 0 0 0 SE 12 0 13 3 0 0 0 SSE 20 _ _ _ _ _ _ _ 0 23 1 0 0 00 S 8 0 9 0 0 0 0 SSW 1 0 1 0 0 0 0 SW 2 0 2 0 0 0 0 WSW 1 0 1 0 0 0 0 W 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 0 NW 0 0 0 0 0 0 0 NNW 0 0 0 TOTAL 86 5 0 0 0 0 91 PERIODS OF CALM(HOURS): 0 VARIABLE DIRECTION: 0 HOURS OF MISSING DATA: 4 A2 .4-7

SITE: AEP COOK HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: 02100101 - 02123124 STA3ILITY CLASS ALL DT/DZ ELEVATION' SPEED:SP10M DIRECTION DIRlOM LAPSE DT63M WIND SPEED (MPH)

WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N 24 71 23 0 0 0 118 NNE 17 27 10 0 0 0 54 NE 36 87 13 0 0 0 136 ENE 41 66 2 0 0 0 E 75 48 9 0 0 0 ESE 63 28 23 0 0 0 119 SE 43 24 21 0 0 0 88 SSE 58 58 17 2 0 0 135 S 44 141 80 12 0 0 277 SSW 14 59 100 17 0 0 19 0 SW 10 40 71 7 0 0 128 3 61 59 37 0 0 )163 W 17 79 59 16 0 0 )171 WNW 17 99 22 1 0 0 )139 NX1 13 80 20 0 0 0 )118 NTNT i9 0 0 0 )124 TOT.' -L 520 1044 548 92 0 0 )2204 PERIOCS OF CALM(HOURS): 0 VAR-IALE DIRECTION 0 HOURS OF MISSING DATA 4 A' 4-3

OFF-SITE DOSE CALCULATION MANUAL The Off-Site Dose Calculation Manual, PMP 6010 OSD.001, was revised twice during the report period. The reasons for these changes and the Plant Operations Review Committee (PORC) approval, if required, are documented on the Review and Approval Tracking Form. These documents are marginally marked. These changes were determined to maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. Items that are determined to be corrections, in accordance with PMP-2010-PRC-002, Procedure Correction, Change ad Review, do not require approval.

A3.0- I

I .

REVIEW AND APPROVAL TRACKING FORM 7-1= thi-&4h' ! - 4.-. i.

Number: PMP-6010-OSD-001 Rev. 17 Change: 0

Title:

Off-site Dose Calculation Manual rt4~~~~~~~~~~~~~a ti. ..> -- .

>s F Correction (Full Procedure) E CChange (Full Procedure) with Review of Change Only F Correction (Page Substitution) F CChange (Page Substitution) with Review of Change.Only F Cancellation F 1N New Procedure or Change with Full Review n Superseded (list superseding procedures):

_Ets ri--f-s- C I;-

Change Driver/CDI Tracking No(s).: 2J N/A ki -f, - ~,tfit 2 -S'- , _ !,, v2mlt;z(,v .¶m--. I -t- J i-'tk' -

Cross-Discipline Reviews: Programmatic Reviews:

3 Chemistry ] Training l] ALARA F Performance Assurance F Maintenance E Work Control 3 Bus. Services Proc Grp F Reactivity Mgmt Team F NDM [ _ Component Engineering F] SPS (Safety & Health)

C Operations D

[Design Engineering S Surveillance Section F PA/PV F] Emerg Oper Proc Grp F System Engineering F Reg Affairs E] _ Environmental F]_

0 RP EJ None Required . F ISIIIST Coordinator F] None Required E) Cognizant Org Review: Date: ___ou__ 12. __I____/o7 Z Technical Review: __ Date: -631i06/ c2.

__ "-- F-JD4,

-8ilt rren#"e:_ ." z~,4;;

>s_:^ --

I if ,-t ft:,fii-.

i.--W F]1 Ops Mgr Concurrence:

9 Owner Concurrence:

Updated Revision Summary attached'? E] Yes 10 CFR 50.59 Requirements complete? Tracking No.: 2coz-o9 J -oc X Yes F N/A Implementation Plan developed? (Ref. Step 3.4.18) F] Yes 1I N/A Package Complete: Date: S /Hq 1 o PORC Review Required:

Administrative Hold Status: F] Released 3 Yes F]

F] No Reissued Z N/A Mtg. No.:3 3jU1 CR No.:

1 Lt Approval Authority Review/Approval: e , (0 Date: 3 /J2 /0).

Expiration Date/Ending Activity N/A Effective Date: 3 IZto-Periodic Review conducted? (Data Sheet 5 Complete) F] Yes Z No plic V4 !!+s,,1 E 'tit~e! t ; ,* _=l_ b r Commitment Database Updated? F Yes El N/A NDMt notified of new records or changes to records that could affect record retention? F] Yes RI N/A

REVISION SUN\[IMARY Number P\tP-6010-OSD-001 Revision. 17 Change- 0

Title:

Off-site Dose Calculation Manual Mfarginal markings were used.

Replace Revision 16, C2 with Revision 17.

Section or Step Change/Reason For Change j Correction lCriteria 3.2.1 BASES Change: Added description of NRC Commitment 1010. r Reason: This commitment had been cancelled and was restored due to delay in Eberline ESW radiation monitor operability and clarification.

5 2 lg- Change Added reference to previously cancelled NRC r commitment.

Reason- Commitment was reopened.

Attachment 3.2, Change: Corrected wording in last sentence of first r Action 3 paragraph.

Reason: Clarification to ensure delineation between ESW and TRS actions.

Attachment 3.16 Change: Changed Worst Case X/Q and D/Q.

Reason: The 2001 meteorological data indicated higher annual average values than previous. This is a change required by the evaluation required by the ODCMI and implemented through Attachment 3.17.

Attachment 3 19 Change Deleted words 'airborne and' from off-site TLD q station heading Reason: Clarification of station description. These stations only have TLD, not air sampling stations that may have been assumed.

Attachment 3.19 Change: Added words to Ingestion - Milk footnote pertaining to monthly sampling when animals are being fed stored grain.

Reason: The implication with NUREG-13.01 is that when animals are on pasture they are feeding there.

Guidance is given that when the animals are on stored feed then monthly sampling is appropriate This is a change "It ,_l i_~ .. _F. . erv -..

I ..- ""74H iI This is a free-form as called out in PNIP-2010-PRC-002, Procedure Correction.

Change and Review, Rev 9 Page 2 of 2z

JI Rabin PNV-6010-0SD-001 l Rev. 17 l Paue 1 of 84

- I OFF-SITE DOSE CALCIJLATION MANUAL Information I I Effective Date:3 k2Io Doug Foster John Carlson Environmental Writer Owner _ Cognizant Organtion TABLE OF CONTENTS 1 PURPOSE AND SCOPE ............................... , 4 2 DEFINITIONS AND ABBREVIATIONS .............................. 4 3 DETAILS ........ 4 3.1 Calculation of Off-Site Doses ............................................................... 4 3.1.1 Gaseous Effluent Releases ........................................................ 4 3.1.2 Liquid Effluent Releases ............................................... 10 3.2 Limits of Operation and Surveillances of the Effluent Release Points ... 13 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation ............ 13 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation ................ 14 3.2.3 Liquid Effluents .. .................................. 15

a. Concentration Excluding Releases via the Turbine Room Sump (TRS) Discharge ................ ................................. 15

b. Concentration of Releases from the TRS Discharge .................... 15

c. Dose ................................................. 16

d. Liquid Radwaste Treatment System ....................................... 16 3.2.4 Gaseous Effluents .................................... 19

a. Dose Rate ......... ........................................ 19

b. Dose - Noble Gases . ................................................. 19

c. Dose - Iodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form . ................................................. 19

d. Gaseous Radwaste Treatment .............................................. 20 3.2.5 Radioactive Effluents - Total Dose . . ...................22 3.3 Calculation of Alarm/Trip Setpoints ............................ 23 3.3.1 Liquid Monitors ........................... 24

a. Liquid Batch Monitor Setpoint Methodology ............................ 24

b. Liquid Continuous Monitor Setpoint Methodology ..................... 25 3.3.2 Gaseous Monitors ..................................... 27

a. Plant Unit Vent ..................................... 27

b. Waste Gas Storage Tanks ..................................... 29

c. Containment Purge and Exhaust System ............................... 3 0..

d. Steam Jet Air Ejector System (SJAE) ................ .................... 31

e. Gland Seal Condenser Exhaust . ..................................... 31 3.4 Radioactive Effluents Total Dose ......................................... 32 3.5 Radiological Environmental Monitoring Program (REMP) . ....... ......... 32 3.5.1 Purpose of the REMP .............. ..................... 32

AMMIkCAW FUCIRIC PiNP-6010-OSD-00l1 Rev. 17 Page 2 of 84 OFF-SITE DOSE CALCULATION MANUIAIl I nformation I Etfezn e Date.3_/z./o-Doug Foster John Carlson Environmental Writer Owner Cognizant Organization 3.5.2 Conduct of the REMP ....... ....................... .................. 33 35 3 Annual Land Use Census.35 ...... ...... . .35 3.5.4 Interlaboratory Comparison Program ...... ...... .5. ............ ..... 35 3.6 Steam Generator Storage Facility Groundwater Monitoring Program .. 36 3.6.1 Purpose of the Steam Generator Storage Facility Groundwater Radiological Monitoring Program .... .......... ...... ......... .. .. 36 3.6.2 Conduct of the Steam Generator Storage Facilitv Groundwater Radiological TMonitoring Program .............. ................ 36 3.7 Meteorological _Model ............................... 36 3.8 Reporting Requirements ... ....... ... ...... .. .... . ... ... ... ............ ..... 36 3.8.1 Annual Radiological Environmental Operating Report (AREOR) . 36 3.8.2 Annual Radiological Effluent Release Report (ARERR) .37 3.9 10 CFR 50 75 (g) Implementation .... . ........................................... 38 3.10 Reportinga/Management Review ........ . ..................................... 39 4 FINAL CONDITIONS ................ 39 5 REFERENCES ................ 39 SUPPLEMENTS .1 Dose Factors for Various Pathways ...................................... .. Pages 42 - 45 .2 Radioactive Liquid Effluent Monitoring Instruments .... ......... Pages 46 - 47/ .3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements. ........................................... Pages 48 - 49 .4 Radioactive Gaseous Effluent Monitoring Instrunentation .----- Pages 50 - 52 .5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ........................................... Pages 53 - 54 6 Radioactive Liquid Waste Sampling and Analysis PToam . Pages 55 - 56 7 Radioactive Gaseous Waste Sampling and Analysis Program .... Pages 57 - 58 Attaclunent 3 S Multiple Release Point Factors for Release Points.--------- . Page 59 9 Liquid Effluent Release Svstems ...................... .... ..... . Page 60

PMiNfP-6010-OSD-001 Rev. 17 Page 3 of 84 OFF-SITE DOSE CALCULATION MANUTAL Informatio I I Effective Date: 3 __'1a2-Doug Foster John Carlson Environmental Writer Owner - Cognizant Orvanization .10 Plant Liquid Effluent Parameters ............. .............................. Page 61 .1 1 Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors ........................... Page 62 .12 Counting Efficiency Curves for R-19, and R-24 ........................... Pages 63 - 64 .13 Counting Efficiency Curve for R-20, and R-28 ..................................... Page 65 .14 Gaseous Effluent Release Systems ........................................... Page 66 .15 Plant Gaseous Effluent Parameters ................. .......................... Page 67 .16 10 Year Average of 1989-1998 Data ........................................... Pages 68 - 69 .17 Annual Evaluation of Z/Q and D/Q Values For All Sectors ................. Page 70 .18 Dose Factors ........................................... Pages 71 - 72 .19 Radiological Enviroriiental Monitoring Program Sample Stations, Sample Types, Sample Frequencies ............................... Pages 73 - 76 .20 Maximum Values for Lower Limits of DetectionsAX -B .....Pages 77 - 78 .21 Reporting Levels for Radioactivity Concentrations in Environmental Samples ........................... Page 79 .22 On-Site Monitoring Location - REMP ........................... Page 80 .23 Off-Site Monitoring Locations - REMP................................................. Page 81 .24 Safety Evaluation By The Office Of Nuclear Reactor Regulation...................................................................................... Pages - 84 82

Information PMIP-6010-OSD-001 Rev. 17 Pae 4 of 84 OFF-SITE DOSE CALCULATION MALNUJAL 1 PURPOSE ANND SCOPE NOTE: This is an Administrative procedure and only the appropriate sections need be]

performed per PMIP 2010 PRC.003, step 3 2.7.

The Off-Site Dose Calculation Manual (ODCM) is the top tier document for the Radiological Environmental Monitoring Program (RENIP), the Radioactive Effluent Controls Program (RECP), contains criteria pertaining to the previous Radiological Effluent Technical Specifications (RETS) as defined in NUREG-0472, and fully implements the requirements of Technical Specification 6.8.4.

The ODCM contains the methodology and parameters to be used in the calculation of off-site doses due to radioactive liquid and gaseous effluents and in the calculation of liquid and gaseous monitoring instrumentation alarmitrip setpoints.

The ODCM provides flow diagrams detailing the treatment path and the major components of the radioactive liquid and gaseous waste management systems.

The ODCM presents maps of the sample locations and the meteorological model used to estimate the atmospheric dispersion and deposition parameters.

The ODCMv1 specifically addresses the design characteristics of the Donald C. Cook Nuclear Plant based on the flow diaarams contained on the "OP Drawings' and plant 'System Description- documents.

2 DEFINITIONS AND ABBREVLATIONS Term: l Meaning:

S or shiftly At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D or daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W or weeklv At least once per 7 days M or monthly At least once per 31 days Q or quarterly At least once per 92 days SA or semi-annually At least once per 184 days R At least once per 549 days.

S/U Prior to each reactor startup P Completed prior to each release I Sampling evolution Process of changing filters or obtaining grab samples 3 DETAILS 3.1 Calculation of Off-Site Doses 3 1. 1 Gaseous Effluent Releases

Information PMIP-6010-OSD-001 I Rev. 17 i Page 5 of 84 OFF-SITE DOSE CALCULATION MIANUAL

a. The computer program MIDAS (Meteorological Information and Dose Assessment System) performs the calculation of doses from effluent releases. The site-specific parameters associated with MIDAS reside in the following subprograms:

MIDER

MIDEX

M1DEL

MIDEG

MIDEN

b. The subprogram used to enter and edit gaseous release data is called MD 1EQ (EQ). The data entered in EQ can be used to calculate the accumulation of dose to individual land based receptors based on hourly meteorology and release data. The air dose from this data is calculated via the XDAIR subprogram in MIDAS. It computes air dose results for use in Reg. Guide 1.21 reports and 10 CFR 50 Appendix I calculations based on routine releases.

c. The formula used for the calculation of the air dose is generated from site specific parameters and Reg. Guide 1.109 (Eq 7):

Dy, D; air = ' [(M or IV,)

Q

3.1 7E - 8J Where; Dr, Do air =the gamma or beta air dose in mrad/yr to an individual receptor z/ Q the annual average or real time atmospheric dispersion factor over land, sec/m 3 from Attachment 3.16, 10 Year Average of 1989-1998 Data 3

M = the gamma air dose factor, mrad I yr Ci, from Attachment 3.18, Dose Factors N. the beta air dose factor, mrad rn3 I yr PCi, from Attachment 3.18, Dose Factors Q. = the release rate of radionuclide, 'i", in JLCi/yr.

3.17E-8 number of years in a second (years/second).

d. The value for the ground average X IQ for each sector is calculated using equations shown below. Formula used for the calculation is generated from parameters contained in MIDAS Technical Manual, XDCALC (Eq 2).

Information I PIP-6010-OSD-001 I Rev. 17 l Page 6 of 84 OFF-SITE DOSE CALCULATION MANUAL 2.03 Where;

[ 2H Eg= mnitniumfav+ H or 7g=,[3 x = distance downwind of the source, meters. This information is found in parameter 5 of MIDEX.

Uam wind speed for ground release, (meters/second) cr vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 1.111 Fig.1)

R. = building height (meters) from parameter 28 of MIDER.

(Containment Building = 49.4 meters)

Tf = terrain factor (= I for Cook Nuclear Plant) because we consider all our releases to be ground level (see parameter 5 in MIDEX).

2.03 = .2 -I7T 0.393 radians(22.5°)

e. The dose due to gaseous releases, other than the air dose, is calculated by the MIDAS subprogram GASPRO. GASPRO computes the accumulation of dose to individual receptors based on hourly meteorology and release data. Calculations consider the effect of each important radionuclide for each pathway, organ, age group, distance and direction.

f. Calculations are based on the environmental pathways-to-man models in Reg. Guide 1. 109. The program considers 7 pathways, 8 organs, and 4 age groups in 16 direction sectors. The distances used are taken from the MIDEG file.

Information PMP-6010-OSD-001 I Rev. 17 l Page 7 of 84 OFF-SITE DOSE CALCULATION MANUAL

g. The formulas used for the following calculations are generated from site specific parameters and Reg. Guide 1. 109:

1. Total Body Plume Pathway (Eq 10)

Dose (mrem/year)= 3.17E- & (Q. /Q S

DFB,)

Where; Sf shielding factor that accounts for the dose reduction due to shielding provided by residential structures during occupancy (maximum exposed individual = 0.7 per Table E-15 of Reg. Guide 1.109)

DFB, = the whole body dose factor from Table B-1 of Reg.

Guide 1. 109, mrem - m3 per pCi - yr. See Attachment 3.18, Dose Factors.

Q = the release rate of radionuclide Xi>, in pCiJyr

2. Skin Plume Pathway (Eq 11)

Dose (mrem/yr) = 3.1 7E - 8

Sf* i * [X(Q,

1.11

DFT) + Z(,

DFSJI Where; 1.11 = conversion factor, tissue to air, mrem/mrad DF 0' = the gamma air dose factor for a uniform semi-infinite cloud of radionuclide 'i", in mrad m3 4/Ci yr from Table B-1, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

DFS, = the beta skin dose factor for a semi-infinite cloud of radionuclide 'i", in mrem m3I/,Ci yr from Table B-i, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

3. Radionuclide and Radioactive Particulate Doses (Eq 13 & 14)

The dose, Dip in mrem/yr, to an individual from radionuclides, other than noble gases, with half-lives greater than eight days in gaseous effluents released to unrestricted areas will be determined as follows:

DiP (mrem/year) = 3.1 7E - 8

7(R,

PF

Q¢C)

Information PiNIP-6010-OSD-001 l Rev. 17 i Page 8 of 84 OFF-SITE DOSE CALCULATION MANUJAL Where; R = the most restrictive dose factor for each identified radionuclide 'i",Jin m2 mrem sec / yr uCi (for food and ground pathways) or mrem in3 yr uCi (for inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles of the site, use the values of R for these real pathways, otherwise use pathways distance of five miles. See Attachment 3.1, Dose Factors for Various Pathways, for the maximum RI values for the most controlling age group for selected radionuclides. RX values were generated by computer code PARTS, see NUREG-0133, Appendix D.

W = the annual average or real time atmospheric dispersion parameters for estimating doses to an individual at the worst case location, and where W is further defined as:

W. = XI 0 for the inhalation pathway, in sec/n 3

-OR-Wfg = D /Q for the food and ground pathways in 1/rm Q = the release rate of those radioiodines, radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than eight days, in ItCi/yr

h. This calculation is made for each pathway. The maximum computed dose at any receptor for each pathway is selected. These are summed together to get the dose to compare to the limits. Only the maximum of the cow milk or goat milk pathway (not both) is included in the total.

i. In addition to the above routines, the QUICKG routine of the MIDAS system is used to provide data used in the monthly reports due to its ability to use annual average meteorological data rather than real time data, thus shortening the run time involved.

j. Steam Generator Blowdown System (Start Up Flash Tank Vent)

1. The amount of radioiodine and other radionuclides that are released via the start up flash tank and its vent are calculated through actual sample results while the start up flash tank is in service.

2. The following calculation is performed to determine the amount of curies released through this pathway. (Plant established formula )

Information PMP-6010-OSD-001 I Rev. 17 Page 9 of 84 OF'F-SITE DOSE CALCULATION MANUAL Curies = E i

GP.M

time on flash tank (mnin) *3 .785E -3 Where; 3.785E-3 = conversion factor, ml Ci/pCi gal.

3. The flow rate is determined from the blowdown valve position and the time on the start up tank. Chemistry Department performs the sampling and analysis of the samples.

4. This data is provided to the MIDAS computer and dose calculations (liquid and gas) are performed to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. MIDAS uses the formulas given in step 3.1.2, Liquid Effluent Releases, to calculate doses to members of the public.

NOTE: This section provides the minimum requirements to be followed at Donald C.

Cook Nuclear Plant. This would be used if actual sample data was not available each time the start up flash tank was in service.

5. The radioiodine release rate must be determined in accordance with the following equation every 31 day period whenever the specific activity of the secondary coolant system is greater than 0.01 gCi/g dose equivalent I-131.

6. IF the specific activity of the secondary coolant system is less than 0.01 uCi/g dose equivalent I-I3 1, THEN the release rate must be determined once every six months. Use the following plant established equation:

0 =Ci* IPF* R1 Where; Qy = the release rate of I-131 from the steam generator flash tank vent, in pCi/sec Ci = the concentration (gzCi/cc) of 1-131 in the secondary coolant averaged over a period not exceeding seven days IPF = the iodine partition factor for the Start Up Flash Tank, 0.05, in accordance with NUREG-0017 Rsgb = the steam generator blowdown rate to the start up flash tank, in cc/sec

7. Use the calculated release rate in monthly dose projections until the next determination to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. Report the release rate calculations in the Annual Radioactive Effluent Release Report.

Information PMP-6010-OSD-001 I Rev. 17 1 Page 10 of 84 OFF-SITE DOSE CALCULATION MANUAL

8. Steam Generators are sparged, sampled, and drained as batches early in outages to facilitate cooldown for entry into the steam generator. This is repeated prior to startup to improve steam generator chemistry for the startup.

3.1l.2 Liquid Effluent Releases

a. The calculation of doses from liquid effluent releases is also performed by the MIDAS program. The subprogram used to enter and edit liquid release data is called MD IEB (EB).

b. To calculate the individual dose (mrem), the program DSlLI (LD) is used. It computes the individual dose for up to 5 receptors for 14 liquid pathways due to release of radioactive liquid effluents. The pathways can be selected using the NIDEL program and changing the values in parameter 1. D.C. Cook Nuclear Plant uses 3 pathways: potable water, shoreline, and aquatic foods (fresh water sport fishing).

c. The equations used are generated from site specific data and Reg. Guide 1.109. They are as follows:

1. Potable Water (Eq 1)

RapilJOO0* Uw *Q Q D~ "l VAf,*F*123E-3 Where; Rm = the total annual dose to organ "j" to individuals of age groups "a" from all of the nuclides "i"in pathway 'p",

in mrem/year 1100 = conversion factor, yr ft3 pCi I Ci sec L Uap = a usage factor that specifies the exposure time or intake rate for an individual of age group 'a" associated with pathway "p". Given in #29-84 of parameter 4 in MIDEL and Reg. Guide 1.109 Table E-5. See Attachment 3.1, Dose Factors for Various Pathways.

np= the dilution factor at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food). Given in parameter 5 of MIIDEL as 2.6.

F = the circulation water system water flow rate, in gpm, is used for evaluating dose via these pathways as dilution flow 2.23E-3 = conversion factor, fte min I sec gal Qi = the release rate of nuclide 'i" for the time period of the run input via MIDEB, Curies/year

Information PMP-6010-OSD-001 I Rev. 17 ! Pa2e 11 of 84 OFF-SITE DOSE CALCULATION MANUAL Daxpj the dose factor, specific to a given age group "a",

radionuclide "if, pathway "p', and organ "j", which can be used to calculate the radiation dose from an intake of a radionuclide, in mrem/pCi. These values are taken from tables E-11 through E-14 of Reg. Guide 1.109 and are located within the MIDAS code.

i= the radioactive decay constant for radionuclide "i", in hours-'

tp = the average transit time required for nuclides to reach the point of exposure, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This allows for nuctide transport through the water purification plant and the water distribution system. For internal dose, tp is the total elapsed time between release of the nuclides and ingestion of food or water, in hours. Given as #25 of parameter 4 in MIDEL. (tp = 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months )

2. Aquatic Foods (Eq 2)

Rao 1100

FUj 2 3 mVp

F

2.23E 3 3

Z 'B

B

Dawp e-la Dl elt Where, Dip = the equilibrium bioaccumulation factor for nuclide 'i" in pathway "p", expressed as pCi L / kg pCi. The factors are located within the MIDAS code and are taken from Table A-l of Reg. Guide 1.109. See Attachment 3.1, Dose Factors for Various Pathways.

tp the average transit time required for nuclides to reach the point of exposure, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This allows for decay during transit through the food chain, as well as during food preparation. Given as #26 of parameter 4 in MIDEL. (tp = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

MP= the dilution factor at the point of exposure, 1.0 for Aquatic Foods. Given in parameter 5 of MIDEL as 1.0.

Information I PMP-6010-OSD-001 Rev. 17 l Page 12 of 84 OFF-SITE DOSE CALCULATION MANUAL

3. Shoreline Deposits (Eq 3)

Rcapj=))0,000* Up *W -* i T *Dp[At1[,~5 M *F* 223E- 3 EDatp le Where; W = the shoreline width factor. Given as an input of 0.3 when running the program, based on Table A-2 in Reg.

Guide 1.109.

Ti = the radioactive half-life of the nuclide, "i", in days Dan = the dose factor for standing on contaminated ground, in mrem rn2 / hr pCi. The values are taken from table E-6 of Reg. Guide 1.109 and are located within the MIDAS code.

See Attachment 3.1, Dose Factors for Various Pathways.

tb = the period of time for which sediment or soil is exposed to the contaminated water, 1.3 1E+5 hours. Given in iŽVIDEL as item 6 of parameter 4.

tp = the average transit time required for nuclides to reach the point of exposure, 0 hours0 days 0 hours 0 weeks 0 months . Given as #28 of parameter 4 in MIDEL.

110,000 = conversion factor yr ft3 pCi i Ci sec in2 day, this accounts for proportionality constant in the sediment radioactivity model Mp = the dilution factor at the point of exposure (or the point of withdrawal of drinkina water or point of harvest of aquatic food). Given in parameter 5 of MIVDEL as 2.6.

d. The MIDAS program uses the following plant specific parameters, which are entered by the operator.

1. Irrigation rate = 0

2. Fraction of time on pasture = 0

3. Fraction of feed on pasture = 0

4. Shore width factor = 0.3 (from Reg. Guide 1.109, Table A-2)

e. The results of DSILI are printed in LDRPT (LP). These results are used in the monthly report of liquid releases.

f. In addition, the program DOSUMI (DMVI) is used to search the results files of DS1LI to find the maximum liquid pathway individual doses. The highest exposures are then printed in a summary table. Each line is compared with the appropriate dose limit. The table provides a concise summary of off-site environmental dose calculations for inclusion in Annual Radioactive Effluent Release Reports, required by Reg. Guide 1 21.

Information PPMIP-6010-OSD-001 I Rev. 17 Page 13 of 84 OFF-SITE DOSE CALCULATION MANUAL NOTE: The performance of each surveillance requirement must be within the specified time interval with a maximum allowable extension not to exceed 25 % of the specified surveillance interval.

3.2 Limits of Operation and Surveillances of the Effluent Release Points 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation

a. The radioactive liquid effluent monitoring instrumentation channels shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, are operable with their alarm/trip setpoints set to ensure that the limits of step 3.2.3 a, Concentration Excluding Releases via the Turbine Room Sump (TRS) Discharge, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments.

c. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure the limits of step 3.2.3a, Concentration Excluding Releases via the Turbine Room Sump (TRS) Discharge, are met without delay, suspend the release of radioactive liquid effluents monitored by the affected channel and reset or declare the monitor inoperable.

d. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable action shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, with a maximum allowable extension not to exceed 25 % of the surveillance interval, excluding the initial performance.

e. Determine the setpoints in accordance with the methodology described in step 3.3. 1, Liquid Monitors. Record the setpoints.

f. Demonstrate each radioactive liquid effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies shown in Attachment 3.3, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

BASES - LIQUID The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure the alarn/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 specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

Information PIWIP-6010-OSD-001 Rev. 17 l _ Page 14 of 84 OFF-SITE DOSE CALCULATION MANUAL Due to the location of the Westinghouse ESW monitors, weekly sampling is required of the ESW system for radioactivity. This is necessary to ensure monitoring of a CCW to ESW system leak and will continue until the Eberline monitors replace the Westinghouse monitors. The Eberline monitors are in the actual ESW effluent stream so they will monitor for this leakage. [Ref 5.2. lgg]

3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation

a. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation, are operable with their alarm/trip setpoints set to ensure that the limits of step 3.2.4a, Dose Rate, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation.

c. With a radioactive gaseous process or effluent monitoring instrumentation channel alarmltrip setpoint less conservative than a value which will ensure that the limits of step 3.2.4a, Dose Rate, are met, without delay, suspend the release of radioactive gaseous effluents monitored by the affected channel and reset or declare the channel inoperable.

d. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, take the action shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation, with a maximum allowable extension not to exceed 25 % of the surveillance interval, excluding the initial performance.

NOTE: This surveillance requirement does not apply to the waste gas holdup system hydrogen and oxygen monitors, as their setpoints are not addressed in this document.

e. Determine the setpoints in accordance with the methodology as described in step 3.3.2, Gaseous Monitors. Record the setpoints.

f. Demonstrate each radioactive gaseous process or effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Attachment 3.5, Radioactive Gaseous Effluent 'Monitonng Instrumentation Surveillance Requirements.

BASES - GASEOUS

Information PNIP-6010-OSD-001 I Rev. 17 I Page 15 of 84 OFF-SITE DOSE CALCULATION MANUAL The radioactive gaseous effluent instrumentation is provided to monitor and control. as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure 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 specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

3.2.3 Liquid Effluents

a. Concentration Excluding Releases via the Turbine Room Sump (TRS)

Discharge

1. Limit the concentration of radioactive material released via the Batch Release Tankcs or Plant Continuous Releases (excluding only TRS discharge to the Absorption Pond) to unrestricted areas to the concentrations in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, limit the concentration to 2E-4 gCi/ml total activity.

2. With the concentration of radioactive material released from the site via the Batch Release Tanks or Plant Continuous Releases (other than the TRS to the Absorption Pond) exceeding the above limits, without delay restore the concentration to within the above limits.

3. Sample and analyze radioactive liquid wastes according to the sampling and analysis program of Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program. -

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within limits.

b. Concentration of Releases from the TRS Discharge

1. Limit releases via the TRS discharge to the on-site Absorption Pond to the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, limit the concentration to 2E-4 pCi/ml total activity.

2. With releases from the TRS exceeding the above limits, perform a dose projection due to liquid releases to UNRESTRICTED AREAS to determine if the limits of step 3 .2.3c. 1 have been exceeded. If the dose limits have been exceeded, follow the directions in step 3.2.3c.2, as applicable.

3. Sample and analyze radioactive liquid wastes according to the program in Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

Information T PIP-6010-OSD-001 l Rev. 17 l Page 16 of 84 OFF-SITE DOSE CALCULATION MAJNUAL

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within the limits stated above.

c. Dose

1. Limit the dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas during any calendar quarter to

1.5 mrem to the total body and to s 5 mrem to any organ, and during any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.

2. With the calculated release of radioactive materials in liquid effluents exceeding ten times any of the limits in Steps 3.2.3a, 3.2.3b, or 3.2.3c. 1 above, prepare and submit a Written Report, pursuant to 10 CFR 20.2203, within 30 days after learning of the event. This report must describe the extent of exposure of individuals to radiation and radioactive material, including, as appropriate:

a) Estimate of each individual's dose, b) Levels of radiation and concentration of radioactive material involved, c) Cause of elevated exposures, dose rates or concentrations,

-AIND-d) Corrective steps taken or planned to ensure against recurrence, including schedule for achieving conformance with applicable limits.

These reports must be formatted in accordance with PMP-7030.001.002, Licensee Event Reports, Special and Routine Reports, even though this is not an LER.

3. Determine cumulative and projected dose contributions from liquid effluents in accordance with this document at least once per 31 days.

Dose may be projected based on estimates from previous monthly projections and current or future plant conditions.

d. Liquid Radwaste Treatment System

1. Use the liquid radwaste treatment system to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.

2. Project doses due to liquid releases to UNRESTRICTED AREAS at least once per 31 days, in accordance with this document.

Information l PMP-6010-OSD-O01 Rev. 17 i Page 17 of 84 OFF-SITE DOSE CALCULATION MANUAL

e. During times of primary to secondary leakage, the use of the startup flash tank should be minimized to reduce the release of curies from the secondary system and to maintain the dose to the public ALARA.

Operation of the North Boric Acid Evaporator (NBAE) should be done in a manner so as to allow the recycle of the distillate water to the Primary Water Storage Tank for reuse. This will provide a large reduction in liquid curies of tritium released to the environment, as there is approximately 40 curies of tritium released with every monitor tank of NBAE distillate.

Drainage of high conductivity water (Component Cooling Water and ice melt water containing sodium tetraborate) shall be evaluated to decide whether it should be drained to waste (small volumes only), the Turbine Room Sump (low activity water only) or routed without demineralization processing to a monitor tank for release. This is necessary in order to minimize the detrimental affect that high conductivity water has on the radioactive wastewater demineralization system. The standard concentration and volume equation can be utilized to determine the impact on each method and is given here. The units for concentration and volume need to be consistent across the equation:

(CI)(VI) + (C.) (V.) = (GC)(VI)

'Where; Q = the initial concentration of the system being added to Vi = the initial volume of the system being added to Ca = the concentration of the water that is being added to the system V. = the volume of the water that is being added to the system C, = the final concentration of the system after the addition Vt = the final volume of the system after the addition The intent is to keep the:

WDS below 500 pmhos/cc.

TRS below lE-5 [C/cc.

Monitor Tank release ALARA to members of the public.

Wastewater leakage into the liquid waste disposal system will be monitored routinely. In the event the leak rate is determined to be over two gallons per minute (the assumed plant design leakage based on the original 2 gpm waste evaporator), increased scrutiny will be placed on locating inleakage, timeliness of job order activities, and/or activities causing increased production of waste water.

Information PMIP-6010-OSD-001 Rev. 17 Page 18 of 84 OFF-SITE DOSE CALCULATION IANUAL BASES - CONCENTRATION This specification is provided to ensure 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 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than 1) the SectionI1.A design objectives of AppendixI, 10 CFR Part 50, to an individual and 2) the limits of 10 CFR Part 20. The concentration limit for noblegasses is based upon the assumption that Xe-135 is the controlling radionuclide and its Effluent Concentration Unit in air (submersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protection (ICRP) Publication 2.

DOSE This specification is provided to implement the requirements of Sections ll.A, I.A, and IV.A of AppendixI, 10 CFR Part 50. The dose limits implement the guides set forth in SectionII A of AppendixI. The ACTION statements provide the required operating flexibility and at the same time, implement the guides set forth in Section IV.A of Appendix I to assure the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are 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 conformance 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 underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents, will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", April 1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guide 1.109 and 1.113.

This specification applies to the release of liquid effluents from each reactor at the site. The liquid effluents from the shared system are proportioned among the units sharing the system.

LIQUID WASTE TREATMENT 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 be used when specified provide 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 Criteria Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section II.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 dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents.

Information i PMIP-6010-OSD-001 I Rev. 17 l Page 19 of 84l OFF-SITE DOSE CALCULATION MANUAL 3.2.4 Gaseous Effluents

a. Dose Rate

1. Limit the dose rate due to radioactive materials released in gaseous effluents from the site to

500 mnremlyr to the total body and

3000 mnremlyr to the skin for noble gases. Limit the dose rate due to all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than eight days to c 1500 mremlyr to any organ.

2. With the dose rate(s) exceeding the above limits, without delay decrease the release rate to within the above limit(s).

3. Determine the dose rate due to noble gases in gaseous effluents to be within the above limits in accordance with the methods and procedures described in this document.

4. Determine the dose rate due to radioactive materials, other than noble gases, in gaseous effluents to be within the above limits in accordance with the methods and procedures of this document by obtaining representative samples and performing analyses in accordance with the sampling and analysis program in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program.

b. Dose - Noble Gases

1. Limit the air dose in unrestricted areas due to noble gases released in gaseous effluents during any calendar quarter, to < 5 mrad for gamma radiation and ' 10 mrad for beta radiation and during any calendar year, to < 10 mrad for gamma radiation and < 20 mrad for beta radiation.

2. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding ten times any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event.

3. Determine cumulative and projected dose contributions for the total time period in accordance with this document at least once every 31 days.

c. Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form

1. Limit the dose to a MEMBER OF THE PUBLIC from radioiodine, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than eight days in gaseous effluents released to unrestricted areas (site boundary) to the following:

a) During any calendar quarter to less than or equal to 7.5 mrem to any organ b) During any calendar year to less than or equal to 15 mrem to any organ.

Information I PMIP-6010-OSD-001 I Rev. 17 l Page 20 of 84 OFF-SITE DOSE CALCULATION IMANUAL

2. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding ten times any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event.

3. Determine cumulative and projected dose contributions for the total time period in accordance with this document at least once every 31 days.

d. Gaseous Radwaste Treatment

1. Use the gaseous radwaste treatment system and the ventilation exhaust treatment system to reduce radioactive materials in gaseous wastes prior to their discharge when projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. Use the ventilation exhaust treatment system to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas when averaged over 31 days would exceed 0.3 mrem to any organ.

2. Project doses due to gaseous releases to UNRESTRICTED AREAS at least once per 31 days in accordance with this document.

BASES-- GASEOUS EFFLUENTS This specification is provided to ensure that the dose rate any time at the SITE BOUNDARY from gaseous effluents from all units on the site 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 2. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix B., Table 2 of 10 CFR Part 20. For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified instantaneous release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to < 500 mrem/yr to the total body or to < 3000 mrem/yr to the skin. These instantaneous release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to < 1500 mremlyr.

This specification applies to the release of gaseous effluents from all reactors at the site. The gaseous effluents from the shared system are proportioned among the units sharing that system.

DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections Il.B, III A, and IV.A of Appendix 1, 10 CFR Part 50. The dose limits implement the guides set forth in Section II.B of Appendix I.

Information X PMIP-6010-OSD-001 i Rev. 17 l Pa e 21 of 84 OFF-SITE DOSE CALCULATION MANNUAL The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept 'as low as is reasonably achievable".

The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown 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 dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors",

Revision 1, July 1977. The ODCM equations provided for determining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

DOSE, RADIOIODINES, RADIOACTIVE MYIATERIALIN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES This specification is provided to implement the requirements of Sections Il.C, IH.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits are the guides set forth in Section II.C of Appendix I.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept 'as low as is reasonably achievable".

The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown 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 ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide the methodology 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 man, and 4) deposition on the ground with subsequent exposure of man.

Information PMP-6010-OSDAOO1 Rev. 17 1 Page 22 of 84 OFF-SITE DOSE CALCULATION NIAN-UAL GASEOUS WASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventilation exhaust treatment systems 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 assurance 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 Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section ll.D 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 guides forth in Sections ll.B and I.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

3.2.5 Radioactive Effluents - Total Dose

a. The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to 5 25 mrem to the total body or any organ (except the thyroid, which is limited to

75 mrem) over a period of 12 consecutive months.

b. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding one half the annual limits of steps 3.2.3c (Dose), 3.2.4b (Dose - Noble Gases), or 3.2.4c (Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form) during any calendar quarter, perform the following:

Investigate and identify the causes for such release rates;

Define and initiate a program for corrective action;

Report these actions to the NRC within 30 days from the end of the quarter during which the release occurred.

IF the estimated dose(s) exceeds the limits above, and IF the release condition resulting in violation has not already been corrected prior to violation of 40 CFR 190, THEN include in the report a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified information of paragraph 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR 50, as addressed in other sections of this document.

c. Determine cumulative dose contributions from liquid and gaseous effluents in accordance with this document (including steps 3.2.3c

[Dose], 3.2.4b [Dose - Noble Gases], or 3.2.4c [Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form]).

Information PIMP-6010-OSD-001 Rev. 17 l Page 23 of 84 OFF-SITE DOSE CALCULATION MANUAL BASES -- TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action, which should result in the limitations of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to any member of the public from other uranium fuel cycle sources is negligible with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected, in accordance with the provision of 40 CFR 190. 11), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed.

An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation, which is part of the nuclear fuel cycle.

3.3 Calculation of Alarm/Trip Setpoints The alarm and trip setpoints are to provide monitoring, indication, and control of liquid and gaseous effluents. The setpoints are used in conjunction with sampling programs to assure that the releases are kept within the limits of 10 CFR 20, Appendix B, Table 2.

Establish setpoints for liquid and gaseous monitors. Depending on the monitor function, it would be a continuous or batch monitor. The different types of monitors are subject to different setpoint methodologies.

One variable used in setpoint calculations is the multiple release point (NIRP) factor.

The MRP is a factor used such that when all the releases are integrated, the applicable LIMIT value will not be exceeded. The MRP is determined such that the sum of the MRP's for that effluent type (liquid or gaseous) is less than or equal to 1. The value of the MRP is arbitrary, and it should be assigned based on operational performance. The values of the MRP's for each liquid release point are given in Attachment 3.8, Multiple Release Point Factors for Release Points.

The Site stance on instrument uncertainty is taken from HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits, which states the NRC position is the result of a valid measurement obtained by a method, which provides a reasonable demonstration of compliance. This value should be accepted and the uncertainty in that measured value need not be considered.

Information I PIP-6010-OSD-001 I Rev. 17 I Page 24 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.3.1 Liquid Monitors Establish liquid monitor setpoints for each monitor of the liquid effluent release systems. A schematic of the liquid effluent release systems is shown as Attachment 3.9, Liquid Effluent Release Systems. A list of the Plant Liquid Effluent Parameters is in Attachment 3.10, Plant Liquid Effluent Parameters.

The details of each system design and operation can be found in the system descriptions. The setpoints are intended to keep releases within the limits of 10 CFR 20, Appendix B, Table 2, Column 2. Determine setpoints using either the batch or the continuous methodology.

a. Liquid Batch Monitor Setpoint Methodology

1. There is only one monitor used on the Waste Disposal System for liquid batch releases. This monitor is identified as RRS-1000. -Steam Generator Blowdown radiation monitors also can be used to monitor batch releases while draining steam generators. The function of these monitors is to act as a check on the sampling program. The sampling program determines the nuclides and concentrations of those nuclides prior to release. The discharge and dilution flow rates are then adjusted to keep the release within the limits of 10 CFR 20. Based on the concentrations of nuclides in the release, the count rate on the monitor can be predicted. The high alarm setpoint can then be set above the predicted value up to the maximum setpoint of the systemr

2. The radioactive concentration of each batch of radioactive liquid waste to be discharged is determined prior to each release by sampling and analysis in accordance with Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

3. The allowable release flow rates are determined in order to keep the release concentrations within the requirements of 10 CFR 20, Appendix B, Table 2, Column 2. The equation to calculate the flow rate is from Addendum AA1 of NUREG-0133:

F. Cr f1 <Fj

[ LJJVJlTj wfRP Where; C. = the concentration of nuclide "i" in ptCiiml LIMITM = the 10 CFR 20, Appendix B, Table 2, Column 2 limit of nuclide "i" in ILCi/ml f = the effluent flow rate in gpm (Attachment 3.10, Plant Liquid Effluent Parameters)

F = the dilution water flow rate as estimated prior to release. The dilution flow rate is a multiple of 230,000 gpm depending on the number of circulation pumps in operation.

MRP = the multiple release point factor. A factor such that when all the release points are operating at one time the limits of 10 CFR 20 will not be exceeded.

Information PMP-6010-OSD-001 I Rev. 17 l Pae 25 of 84 OFF-SITE DOSE CALCULATION MANUAL

4. This equation must be true during the batch release. Before the release is started, substitute the maximum effluent flow rate and the minimum dilution flow rate for f and F, respectively. If the equation is true, the release can proceed with those flow rates as the limits of operation. If the equation is not true, the effluent flow rate can be reduced or the dilution flow rate can be increased to make the equation true. This equation may be rearranged to solve for the maximum effluent release flow rate (f).

5. The setpoint is used as a quality check on the sampling program.

The setpoint is used to stop the effluent flow when the monitor reading is greater than the predicted value from the sampling program. The predicted value is generated by converting the effluent concentration for each gamma emitting radionuclide to counts per unit of time as per Attachment 3.1 1, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, or Attachment 3.12, Counting Efficiency Curves for R-19, and R-24. The sum of all the counts per unit of time is the predicted count rate. The predicted count rate can then be multiplied by a factor to determine the high alarm setpoint that will provide a high degree of conservatism and eliminate spurious alarms.

b. Liquid Continuous Monitor Setpoint Methodology

1. There are eight monitors used as potential continuous liquid release monitors. These monitors are used in the steam generator blowdown (SGBD), blowdown treatment (BDT), and essential service water (ESW) systems.

2. The Westinghouse monitors (R) are being replaced by Eberline monitors (DRS, WRA) and are identified as:

R-19 or DRS 3100/4100 for SGBD R-24 or DRS 3200/4200 for BDT

R-20 or WRA 3500/4500 for the east ESW system

R-28 or WRA 3600/4600 for the west ESW system The function of these monitors is to assure that releases are kept within the concentration limits of 10 CER 20, Appendix B, Table 2, Column 2, entering the unrestricted area following dilution.

3. Tne monitors on steam generator blowdown and blowdown treatment systems have trip functions associated with their setpoints.

Essential service water monitors are equipped with an alarm function only and monitor effluent in the event the Containment Spray Heat Exchangers are used or the ESW system (Eberline).

Information I PMP-6010-OSD-001 Rev. 17 l Page 26 of 84 OFF-SITE DOSE CALCULATION MANI.NUAL

4. The equation used to determine the setpoint for continuous monitors is from Addendum AA1 of NUREG-0133:

CA Eff

RP

F

SF f

Where; Sp = setpoint of monitor (cpm)

C = 5E-7 pLCi/ml, maximum effluent control limit from 10 CFR 20, Appendix B, Table 2. Column 2 of a known possible nuclide in effluent stream. (The limiting nuclide shall be evaluated annually by reviewing current nuclides against historical ones in order to determine if one with a more restrictive effluent concentration limit than Sr90 is found. The concentration limit shall be adjusted appropnately.)

-OR-if a mixture is to be specified, z Ci T Ci

- LIM~IT, Eff = Efficiency, this information is located in Attachment 3.1 1, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, through Attachment 3.13, Counting Efficiency Curve for R-20, and R-28, for the specific monitors. For Eberline monitors the efficiency is nuclide specific and the calculation changes slightly to:

Y(CI *,Ej,) replacesC

Eff S C, LIM 1 MRP = multiple release point factor. A factor such that when all the release points are operating at one time the limits of 10 CFR 20 will not be exceeded (Attachment 3.8, Multiple Release Point Factors for Release Points). The MRP for ESW monitors is set to 1.

F dilution water (circ water) flow rate in gpm obtained from Attachment 3.10, Plant Liquid Effluent Parameters.

For routine operation, the setpoint should be calculated using the minimum dilution flow rate of 230,000 gpm.

SF = Safety Factor, 0.9.

f = applicable effluent release flow rate in gpm. For routine operation, the setpoint should be calculated using maximum effluent flow rate (Attachment 3.10. Plant Liquid Effluent Parameters).

33 2 Gaseous Monitors

Information I PV-6010-OS0oo01 I Rev. 17 l Page 27 of 84 OFF-SITE DOSE CALCULATION MANUAL For the purpose of implementing Step Error!Reference source not found.,

Error! Reference source not found., and Substep 3.2.4a, Dose Rate, the alarm setpoints for gaseous effluents released into unrestricted areas will be established using the following methodology. In addition, the above steps do not apply to instantaneous alarm and trip setpoints for integrating radiation monitors sampling radioiodines, radioactive materials in particulate form and radionuclides other than noble gases. A schematic of the gaseous effluent release systems is presented in Attachment 3.14, Gaseous Effluent Release Systems. Attachment 3.15, Plant Gaseous Effluent Parameters, presents the effluent flow rate parameter(s).

Gaseous effluent monitor high alarm setpoints will routinely be established at a fraction of the maximum allowable setpoint (typically 10% of the setpoint) for ALARA purposes. Alert alarms will normally be set to provide adequate indications of small changes in radiological conditions.

a. Plant Unit Vent

1. The gaseous effluents discharged from the plant vent will be monitored by the plant vent radiation monitor low range noble gas channel [Tag No. VRS-1505 (Unit 1), VRS-2505 (Unit 2)] to assure that applicable alarms and trip actions (isolation of gaseous release) will occur prior to exceeding the limits in step 3.2.4, Gaseous Effluents. The alarm setpoint values will be established using the following unit analysis equation:

SF *MRP

DLj Fp

TIQ M

DCF,4 Where; Sp = the maximum setpoint of the monitor in iiCi/cc for release point p, based on the most limiting organ SF = an administrative operation safety factor, less than 1.0 MRP = a weighted multiple release point factor (< 1.0), such that when all site gaseous releases are integrated, the applicable dose will not be exceeded based on the release rate of each effluent point. The MRP is an arbitrary value based on the ratio of the release rate or the volumetric flow rate of each effluent point to the total respective flow rate value of the plant and will be consistent with past operational experience.

The MRP is computed as follows:

Compute the average release rate, Qp, (or the volumetric flow rate, fQ) from each release point p.

Compute XQp (or Zfp) for all release points.

Ratio Qp/7ZQp (or fp/Ifp) for each release point.

This ratio is the MRP for that specific release point

Repeat the above bullets for each of the site's eight gaseous release points.

Information I P.NIP-6010-OSD-O1 I Rev. 17 Page 28 of 84 OFF-SITE DOSE CALCULATION MANUAL FP = the maximum volumetric flow rate of release point "p",

at the time of the release, in cc/sec. The maximum Unit Vent flow rate, by design, is 186,600 cfm. for Unit 1 and 143,400 cfmn for Unit 2.

DLj = dose rate limit to organ "j" in an unrestricted area (mrem/yr).

Based on continuous releases, the dose rate limits, DLI, from step 3.2.4a, Dose Rate, are as follows:

Total Body S 5C0 mremlyear

Skin < 3000 mremlyear

Any Organ< 1500 mremlyear z/ Q = The worst case annual average relative concentration in the applicable sector or area, in sec/rn 3 (see Attachment 3.16, 10 Year Average of 1989-1998 Data).

WI = weighted factor for the radionuclide:

X,= Ci cik Where, Ci = concentration of the most abundant radionuclide "i" Cap= total concentration of all identified radionuclides in that release pathway. For

- batch releases, this value may be set to I for conservatism.

DCF,1 = dose conversion factor used to relate radiation dose to organ 'j", from exposure to radionuclide "i" in mnrem in3 / yr jiCi. See following equations.

The dose conversion factor, DCFj, is dependent upon the organ of concern.

For the whole body: DCFj = K Where; K = whole body dose factor due to gamma emissions for each identified noble gas radionuclide in mrem m3 / yr !.Ci. See Attachment 3.1 8, Dose Factors.

For the skin: DCF; = Li + 1.1M Where; L = skin dose factor due to beta emissions for each identified noble gas radionuclide. in mrem m3 / yr pLci. See Attachment 3.18, Dose Factors.

Information l PMP-6010-OSD-001 i Rev. 17 i Page 29 of 84 OFF-SITE DOSE CALCULATION MLAINUAL 1.1 = the ratio of tissue to air absorption coefficient over the energy range of photons of interest. This ratio converts absorbed dose (mrad) to dose equivalent (mrem).

1% = the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad m- / yr ptCi. See Attachment 3.18, Dose Factors.

For the thyroid, via inhalation: DCF, = P.

Where; P. = the dose parameter, for radionuclides other than noble gas, for the inhalation pathway in mrem m3 / yr p.Ci (and the food and ground path, as appropriate).

See Attachment 3.18, Dose Factors.

2. The plant vent radiation monitor low range noble gas high alarm channel setpoint, Sp, will be set such that the dose rate in unrestricted areas to the whole body, skin and thyroid (or any other organ), whichever is most limiting, will be less than or equal to 500 mremlyr, 3000 mremlyr, and 1500 mremlyr respectively.

3. The thyroid dose is limited to the inhalation pathway only.

4. The plant vent radiation monitor low range noble gas setpoint, Sp, will be recomputed whenever gaseous releases like Containment Purge, Gas Decay Tanks and CVCS HUTs are discharged through the plant vent to determine the most limiting organ.

5. The high alarm setpoint, Sp, may be established at a lower value than the lowest computed value via the setpoint equation.

6. Containment Pressure Reliefs will not have a recomputed high alarm setpoint, but will use the normal high alarm setpoint due to their randomness and the time constraints involved in recomputation.

7. At certain times, it may be desirable to increase the high alarm setpoint, if the vent flow rate is decreased. This may be accomplished in one of two ways.

Max Con c(pCicc)* Max Flowrate(cfm) ewg Max cfin New Max Concentration(pCL/cc)

-OR-Max Conc (f Ci/cc)

Max Flowrate(q-m) New MaxuCi/cc New Max Flowrate(cfm)

b. Waste Gas Storage Tanks

1. The gaseous effluents discharged from the Waste Gas System are monitored by the vent stack monitors VRS-1505 and VRS-2505.

Information PMP-6010-OSD-001 I Rev. 17 W Page 30 of 84 OFF-SITE DOSE CALCULATION MANUAL

2. In the event of a high radiation alarm, an automatic termination of the release from the waste gas system will be initiated from the plant vent radiation monitor low range noble gas channel (VRS-1505 or VRS-2505). Therefore, for any gaseous release configuration, which includes normal operation and waste gas system gaseous discharges, the alarm setpoint of the plant vent radiation monitor will be recomputed to determine the most limiting organ based on all gaseous effluent source terms.

Chemical and Volume Control System Hold Up Tanks (CVCS HUT), containing high gaseous oxygen concentrations. may be -

released under the guidance of waste gas storage tank utilizing approved Operations' procedures.

3. It is normally prudent to allow 45 days of decay prior to releasing a Gas Decay Tank (GDT). There are extenuating, operational circumstances that may prevent this from occurring. Under these circumstances, such as high oxygen concentration creating a combustible atmosphere, it is prudent to waive the 45-day decay for safety's sake.

c. Containment Purge and Exhaust System

1. The gaseous effluents discharged by the Containment Purge and Exhaust Systems and Instrumentation Room Purge and Exhaust System are monitored by the plant vent radiation monitor noble gas channels (VRS-1505 for Unit 1, VRS-2505 for Unit 2); and alarms and trip actions will occur prior to exceeding the limits in step 3.2.4a, Dose Rate.

2. For the Containment System, a continuous air sample from the containment atmosphere is drawn through a closed, sealed system to the radiation monitors (Tag No. ERS-1300/1400 for Unit I and ERS-2300/2400 for Unit 2). During purges, these monitor setpoints will give a Purge and Exhaust Isolation signal upon actuation of high alarm setpoints for particulate and noble gas channels. The sample is then returned to containment. Grab sample analysis is performed for a Containment purge before release.

3. The Upper Containment area is monitored by normal range area gamma monitors (Tag No. VRS-1 101/1201 for Unit 1 and VRS-2101/2201 for Unit 2), which also give Purge and Exhaust Isolation Trip signals upon actuation of their high alarm.

4. For the Containment Pressure Relief System, no sample is routinely taken prior to release, but a sample is obtained twice per month.

Information I PNIP-6010-OSD-001 Rev. 17 1 Page 31 of 84 OFF-SITE DOSE CALCULATION MANUAL

5. The containment airborne and area monitors, upon actuation of their high alarm, will automatically initiate closure of the Containment and Instrument Room purge supply and exhaust duct valves and containment pressure relief system valves. Complete trip of all isolation control devices requires high alarm of one of the two Train A monitors (ERS-1300/2300 or VRS-1101/2101) and one of the two Train B monitors (ERS-1400/2400 or VRS-1201/2201).

d. Steam Jet Air Ejector System (SJAE)

1. The gaseous effluents from the Steam Jet Air Ejector System discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-1900 for Unit 1 and SRA-2900 for Unit 2). The monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the Condenser Air Ejector System monitor will be based on the maximum air ejector exhaust flow rate, (Attachment 3.15, Plant Gaseous Effluent Parameters). The alarm setpoint value will be established using the following unit analysis equation:

SF

MRP

DLj Fp /Q * (W *DCFi,)

Where; SSJAE = the maximum setpoint, based on the most limiting organ, in FiCi/cc and where the other terms are as previously defined

e. Gland Seal Condenser Exhaust

1. The gaseous effluents from the Gland Seal Condenser Exhaust discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-1800 for Unit 1 and SRA-2800 for Unit 2). The radiation monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the GSCE monitor will be based on the maximum condenser exhaust flow rate (1260 CFM for Unit 1, 2754 CFM each for the two Unit 2 vents).

The alarm setpoint value will be established using the following unit analysis equation:

_" SF *MRP

DL, F; *%Q*Z W* DCFg,)

Information I PNIP-6010-OSD-001 i Rev. 17 l Page 32 of 84 OFF-SITE DOSE CALCULATION MAXNUAL Where; Srsca = the maximum setpoint, based on the most limiting organ, in .+/-Cilcc and where the other terms are as previously defined 3.4 Radioactive Effluents Total Dose 3.4.1 The cumulative dose contributions from liquid and gaseous effluents will be determined by summing the cumulative doses as derived in steps 3.2.3c (Dose), 3.2.4b (Dose - Noble Gases), and 3.2.4c (Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form) of this procedure. Dose contribution from direct radiation exposure will be based on the results of the direct radiation monitoring devices located at the REMP monitoring stations. See NUREG-0133, section 3.8.

3.5 Radiological Environmental Monitoring Program (REMP) 3.5.1 Purpose of the REMP

a. The purpose of the REMP is to:

Establish baseline radiation and radioactivity concentrations in the environs prior to reactor operations, Monitor critical environmental exposure pathways, Determine the radiological impact, if any, caused by the operation of the Donald C. Cook Nuclear Plant upon the local environment.

b. The first purpose of the REMP was completed prior to the initial operation of either of the two nuclear units at the Donald C. Cook Nuclear Plant Site. The second and third purposes of the REMP are an on-going operation and as such various environmental media and exposure pathways are examined. The various pathways and sample media used are delineated in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Included is a list of the sample media, analysis required, sample stations, and frequency requirements for both collection and analysis. Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, defines the scope of the REMP for the Donald C. Cook Nuclear Plant.

Information I PMP-6010-OSD-001 I Rev. 17 l Pane 33 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.5.2 Conduct of the REMP [Ref. 5.2.1u]

a. Conduct sample collection and analysis for the REMP in accordance with Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, Attachment 3.20, Maximum Values for Lower Limits of DetectionsA B - REMP, and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples. These are applicable at all times. The on-site monitoring locations are shown on Attachment 3.22, On-Site Monitoring Location - REMP, and the off-site monitoring locations are shown on Attachment 3.23, Off-Site Monitoring Locations - REMP.

1. Perform each surveillance requirement within the specified time interval in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, with a maximum allowable extension not to exceed 25 % of the surveillance interval.

2. If an environmental sample cannot be collected in accordance with step 3.5.2a, submit a description of the reasons for deviation and the actions taken to prevent a reoccurrence as part of the Annual Radiological Environmental Operating Report (AREOR).

Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or malfunction of automatic sampling equipment. If the deviation from the required sampling schedule is due to the malfunction of automatic sampling equipment, make every effort to complete the corrective action prior to the end of the next sampling period.

3. If a Tadionuclide is detected in any sample medium exceeding the limit established in Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples, or if more than one radionuclide is detected in any sample medium and the Total Fractional Level (TFL), when averaged over the calendar quarter, is greater than or equal to 1, based on the following formula:

TFL = C,,, + CaT +... >1 La) La)

Where; C(o) = Concentration of I' detected nuclide C", = Concentration of 2'o detected nuclide Lm = Reporting Level of Pt nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

Information I PMIP-6010-OSD-001 i Rev. 17 l Page 34 of 84 OFF-SITE DOSE CALCULATION MIANUAL La) = Reporting Level of 2i nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

And, if the activity is the result of plant effluents, evaluate the release conditions, environmental factors, or other aspects, which may have contributed to the identified levels for inclusion in the AREOR. If the radioactivity was not a result of plant effluents, describe the results in the AREOR.

4. If a currently sampled milk farm location becomes unavailable, conduct a special milk farm survey within 15 days.

a) If the unavailable location was an indicator farm, an alternate sample location may be established within eight miles of the Donald C. Cook Nuclear Plant, if one is available.

b) If the unavailable location was a background farm, an alternate sample location may be established greater than 15 but less than 25 miles of the Donald C. Cook Nuclear Plant in one of the less prevalent wind direction sectors, if one is available.

c) If a replacement farm is unobtainable and the total number of indicator farms is less than three or the background farms is less than one, perform monthly vegetation sampling in lieu of milk sampling.

BASES - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRANM (RENIP)

The REMP provides measurements of radiation and 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. Thereby, this monitoring program supplements the radiological effluent monitoring program by verifying the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified REMP will be effective for at least the first three years of commercial operation. Program changes may be initiated based on operational experience in accordance with the requirements of technical specification 6.8.4.b.

The detection capabilities, required by Attachment 3.20, Maximum Values for Lower Limits of DetectionsB - REMP, are the state-of-the-art for routine environmental measurements in industrial laboratories.

It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine analysis conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

Information I PiMP-6010-OSD-001 I Rev. 17 l Page 35 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.5.3 Annual Land Use Census [Ref. 5.2.lu]

a. Conduct a land use census and identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 square feet producing fresh leafy vegetables in each of the ten land sectors within a distance of five miles.

b. In lieu of the garden census, grape and broad leaf vegetation sampling may be performed as close to the site boundary as possible in a land sector, containing sample media, with the highest average deposition factor (D/Q) value.

c. Conduct this land use census annually between the dates of June 1 and October 1 by door-to-door survey, aerial survey, or by consulting local agricultural authorities.

1. With a land use census identifying a location(s), which yields a calculated dose or dose commitment greater than the values currently being calculated in this document, make appropriate changes to incorporate the new location(s) within 30 days, if possible.

BASES - LAND USE CENSUS This is provided to ensure changes in the use of unrestricted areas are identified and modifications to the monitoring program are made in accordance with requirements of TS 6.8.4b, if required by the results of the 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 since a garden of this size is the minimum required to produce the quantity (25 kglyr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption of a child. To determine this minimum garden size, the following assumptions were used: 1) that 20% of the garden was used for growing broad leaf vegetation (that is, similar to lettuce and cabbage), and 2) a vegetation field of 2 kg/square meter.

3.5.4 Interlaboratory Comparison Program

a. In order to comply with Reg. Guides 4.1 and 4.15, the analytical vendor participates in an Interlaboratory Comparison Program, for radioactive materials. Address program results and identified deficiencies in the AREOR.

1. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the AREOR.

BASES -- INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate the results are reasonably valid.

Information I PMIP-6010-OSD-001 I Rev. 17 l Page 36 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.6 Steam Generator Storage Facility Groundwater Monitoring Program 3.6.1 Purpose of the Steam Generator Storage Facility Groundwater Radiological Monitoring Program

a. The purpose of the temporary on-site Steam Generator Storage Facility Radiological Monitoring Program is to establish baseline radiological data for the groundwater surrounding the facility prior to the storage of the Unit 2 Steam Generator Lower Assemblies. Thereafter, the purpose is to monitor the groundwater through observation wells with locations as shown in Attachment 3.22, On-Site Monitoring Location - REMP, to determine the radiological impact, if any, caused by the use of the Storage Facility.

3.6.2 Conduct of the Steam Generator Storage Facility Groundwater Radiological Monitoring Program

a. Collect and analyze groundwater samples in accordance with Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Apply the values from Attachment 3.20, Maximum Values for Lower Limits of DetectionsAB - REMP, (excluding I-131) and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples, (excluding I-131).

3.7 Meteorological Model 3.7.1 Three towers are used to determine the meteorological conditions at Donald C.

Cook Nuclear Plant. One of the towers is located at the Lake Michigan shoreline to determine the meteorological parameters associated with unmodified shoreline air. The data is accumulated by microprocessors at the tower sites and normally transferred to the central computer every 15 minutes.

3.7.2 The central computer uses a meteorological software program to provide atmospheric dispersion and deposition parameters. The meteorological model used is based on guidance provided in Reg. Guide 1.111 for routine releases. All calculations use the Gaussian plume model.

3.8 Reporting Requirements 3.8.1 Annual Radiological Environmental Operating Report (AREOR)

a. Submit routine radiological environmental operating reports covering the operation of the units during the previous calendar year prior to May 1 of each year.

b. Include in the AREOR:

Summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the reporting period.

Information PMP-6010-OSD-001 I Rev. 17 l Page 37 of 84 OFF-SITE DOSE CALCULATION MALNUAL

A comparison with pre-operational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The results of the land use censuses required by step 3.5.3, Annual Land Use Census.

If harmful effects or evidence of irreversible damage are detected by the monitoring, provide in the report an analysis of the problem and a planned course of action to alleviate the problem.

Summarized and tabulated results of all radiological environmental samples taken during the reporting period. In the event that some results are not available for inclusion with the report, submit the report noting and explaining the reasons for the missing results. Submit the missing data as soon as possible in a supplementary report.

A summary description of the REMP including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used.

A map of all sample locations keyed to a table giving distances and directions from one reactor.

The results of participation in the Interlaboratory Comparison Program required by step 3.5.4, Interlaboratory Comparison Program.

3.8.2 Annual Radiological Effluent Release Report (ARERR)

a. Submit routine ARERR covering the operation of the unit during the previous 12 months of operation within 90 days after January 1 of each year.

b. Include in the ARERR a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in Reg. Guide 1.21, "Measuring. Evaluating and Reporting 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.

c. Submit in the ARERR 90 days after January I of each year and include a quarterly summary of hourly meteorological data collected during the reporting period.

This summary may be in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction and atmospheric stability.

Include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

Information I PPNIP-6010-OSD-001 Rev. 17 Page 38 of 84 OFF-SITE DOSE CALCULATION MANUAL

Include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary during the reporting period. Include all assumptions used in making these assessments (that is, specific activity, exposure time and location) in these reports.

Use the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) for determining the gaseous pathway doses.

Inoperable radiation monitor periods exceeding 30 continuous days; explain causes of inoperability and actions taken to prevent reoccurrence.

d. Submit the ARERR [Ref. 5.2. 1w] 90 days after January 1 of each year and include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Reg. Guide 1.109, Rev. 1.

e. Include in the ARERR the following information for each type of solid waste shipped off-site during the report period:

Volume (cubic meters),

Total curie quantity (specify whether determined by measurement or estimate),

Principle radionuclides (specify whether determined by measurement or estimate),

Type of waste (example: spent resin, compacted dry waste, evaporator bottoms),

Type of container (example: LSA, Type A, Type B, Large Quantity),

-AND-

Solidification agent (example: cement).

f. Include in the ARERR unplanned releases of radioactive materials in gaseous and liquid effluent from the site to unrestricted areas on a quarterly basis.

g. Include in the ARERR any change to this procedure made during the reporting period.

3.9 10 CFR 50.75 (g) Implementation

Information I PMP-6010-OSD-001 I Rev. 17 Paze 39 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.9.1 Records of spills or other unusual occurrences involving the spread of contamination in and around the site. These records may be limited to instances when significant contamination remains after decontamination or when there is a reasonable likelihood that contaminants may have spread to inaccessible areas, as in the case of possible seepages.

3.9.2 These records shall include any known information or identification of involved nuclides, quantities, and concentrations.

3.9.3 This information is necessary to ensure all areas outside the radiological-restricted area are documented for surveying and remediation during decommissioning. There is a retention schedule file number where this information is filed in Nuclear Documents Management to ensure all required areas are listed to prevent their omission.

3.10 Reporting/Management Review 3.10.1 Incorporate any changes to this procedure in the ARERR.

3.10.2 Update this procedure when the Radiation Monitoring System, its instruments, or the specifications of instruments are changed.

3.10.3 Review or revise this procedure as appropriate based on the results of the land use census and REMP.

3.10.4 Evaluate any changes to this procedure for potential impact on other related Department Procedures.

3.10.5 Review this procedure during the first quarter of each year and update it if necessary. Review Attachment 3.16, 10 Year Average of 1989-1998 Data, and document using Attachment 3.17, Annual Evaluation of x/Q and D/Q Values For All Sectors. The X / Q and D / Q values will be evaluated to ensure all data is within +/- 3 standard deviations of the 10 year annual average data and documented by completing Attachment 3.17, Annual Evaluation of -/Q and D/Q Values For All Sectors, and filed in accordance with the retention schedule.

4 FINAL CONDITIONS 4.1 None.

5 REFERENCES 5.1 Use

References:

5.1.1 "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Off-Site Dose Calculation Manual or to the Process Control Program (Generic Letter 89-01)", United States Nuclear Regulatory Commission, January 31, 1989 5.1.2 12-THP-6010.RPP.601, Preparation of the Annual Radioactive Effluent Release Report

Information I PMP-6010-OSD-001 Rev. 17 Page 40 of 84 L OFF-SITE DOSE CALCULATION MANUAL 5.1.3 12-THP-6010.RPP.639, Annual Radiological Environmental Operating Report (AREOR) Preparation And Submittal 5.2 Writing

References:

5.2.1 Source

References:

a. 10 CFR 20, Standards for Protection Against Radiation b- 10 CFR 50, Domestic Licensing of Production and Utilization Facilities

c. P.MI-6010, Radiation Protection Plan

d. NUREG-0472

e. NUREG-0133

f. Regulatory Guide 1.109, non-listed parameters are taken from these data tables

,. Regulatory Guide 1.111

h. Regulatory Guide 1.113

i. Final Safety Analysis Report (FSAR)

j. Technical Specifications, Appendix A, Sections 6.8.1.e, 6.8.4.a, 6.8.4.b, 6.9.1.6, 6.9.1.7, and 6.14, Off-Site Dose Calculation Manual

k. Final Environmental Statement Donald. C. Cook Nuclear Plant, August 1973

1. NUREG-0017

m. ODCM Setpoints for Liquid Effluent Monitors (Bases), ENGR 107-04 8112.1 Environs Rad Monitor System

n. HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits

o. Watts - Bar Jones (WBJ) Document, R-86-C-001, The Primary Calibration of Eberline Instrument Corporation SPING - 314 Low, Mid, and High Range Noble Gas Detectors

p. WBJ Document, R-86-C-003, The Primary Calibration of Eberline Instrument Corporation DAM-4 and Water Monitor

q. 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations

r. NRC Commitment 6309 (N94083 dated 11/10/94)

s. NRC Commitment 1151

t. NRC Commitment 1217

u. NRC Commitment 3240

v. INRC Commitment 3850

w. NRC Commitment 4859

x. NRC Commitment 6442

y. NRC Commitment 3768

z. DIT-B-00277-00, HVAC Systems Design Flows aa. Regulatory Guide 1.21 bb Regulatory Guide 4.1

Information PMP-6010-OSD-001 Rev. 17 l Page 41 of 84 1 OFF-SITE DOSE CALCULATION MLANUAL cc. 1-2-V3-02-Calc #4, Unit Vent Sample Flow rate for isokinetic particulates and Iodine sampling dd. HPS N13.30-1996, Appendix A Rationalefor Methods of Determining Minimum Detectable Amount (NIDA) and Minumurn Testing Level (MIDL ee. DIT-B-01971-00, Dose Factors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway ff. -DIT-B-01987-00, Ground Plane & Food Dose Factors P. for Radioiodines and Radioactive Particulate Gaseous Effluents gg. NRC Commitment 1010 5.2.2 General References

a. Cook Nuclear Plant Start-Up Flash Tank Flow Rate letter from D. L.

Boston dated January 21, 1997

b. Letter from B.P. Lauzau, Venting of Middle CVCS Hold-Up Tank Directly to Unit Vent, May 1, 1992

c. AEP Design Information Transmittal on Aux Building Ventilation Systems

d. PMP4030.EIS.001, Event-Initiated Surveillance Testing

e. Environmental Position Paper, Fe Impact on Release Rates, approved 3/14/00

f. Environmental Position Paper, Methodology Change from Sampling Secondary System Gaseous Effluents for Power Changes Exceeding 15% within 1 hr to Responding to Gaseous Alert Alarms, approved 4/4/00

Information PMP-6010-OSD-001 Rev. 17 1 Page 42 of 84 OFF-SITE DOSE CALCULATION MNANUAL Attachment 3.1 Dose Factors for Various Pathways .e R Dose Factors PATH WAY Nuclide l Ground Vegetable Meat Cow Milk Goat Milk Inhalation H-3 [ O.OE+OO 4.OE,+03 3.3E+02 2.4E+03 4.9E+03 1.3E+03 C-14 O.OE+00 3.5E+06 5.3E+05 3.2E+06 3.2E+06 3.6E+04 Cr-S l 5.4E+06 1.lE+07 1.SE+06 6.9E+06 8.3E+05 2.lE+04 Nln-54 l 1.6E+09 9.4E+08 2.IE+07 2.9E+07 3.5E+06 2.0E+06 Fe-59 3.2E+08 ! 9.6E+08 1.7E+09 3.1E+08 4.0E+07 1.5E+06 Co-58 j 4.4E+08 6.0E+08 2.9E+08 l 8.4E+07 1.OE+07 1.3E+06 Co-60 1 2.5E+10 3.2E+09 1.OE+09 I 2.7E+08 3.2E+07 8.6E+06 Zn-65 8.5E+08 2.7E+09 9.5E+08 j 1.6E+10 1.9E+09 1.2E+06 Sr-89 2.5E+04 3.5E+10 3.8E+08 + 9.9E+09 2.1E+ 10 2.4E+06 Sr-90 O.OE+00 1.4E+12 9-6E+09 9.4E+10 2.0E+1I l.lE+08 Zr-95 2.9E+08 1.2E+09 1.5E+09 9.3E+OS l.IE+05 2.7E+06 Sb-124 6.9E+08 3.OE+09 I 4.4E+08 7.2E-08 8.6E+07 3.8E+06 1-131 1.OE+07 2.4E+10 I 2.5E+09 4.8E+ I1 5.8E+1l 1.6E+07 1-133 1.5E+06 4.0E+08 6.OE+01 I 4.4E-'09 5.3E+09 3.8E+-06 Cs-134 7.9E+09 2.5E+10 1.IE+09 5.0E+10 1.5E+11 .lE+06 Cs-136 1.7E+08 2.2E+08 4.2E+07 5.1EFo09 1.5E+10 1.9E+05 Cs-137 1.2E+10 2.5E+10 1.OE+09 4.SE+10 1.4E+ 1 9.OE+05 Ba-140 2.3E+07 2.7E+08 5.2E+07 2.1E+08 I 2.6E+07 2.OE+06 Ce-141 1.SE+07 5.3E+08 3.0E 07 8.3E+07 1 l.OE+07 6.IE+05 Ce-144 7.9E+07 1 1.3E+10 3.6E'08 7.3E+08 8.7E+07 1.3E+07 Units for all except inhalation pathway are mn mr sec I yr ILCi, inhalation pathway units are mr i3 I yr PCi.

Ut p Values to be Used For the -MaximumExposed Individual Pathway Infant Child Teen Adult Fruits, vegetables and grain (kg/yr) - 520 630 520 Leafy vegetables (kglyr) _ 26 42 64 Milk (L/yr) 330 l 330 400 310 Meat and poultry (kgJyr) - j 41 65 110 Fish (kg/yr) - j 6.9 16 21 Drinking water (L/yr) 330 j 510 510 730 Shoreline recreation (hr/yr) - 14 67 12 Inhalation (m3lyr) 1400 3700 8000 8000 Table E-5 of Reg Guide 1.109.

Information I PMSP-6010-OSD-001 I Rev. 17 1 Page 43 of 84 OFF-SITE DOSE CALCULATION MANUAL .1 Dose Factors for Various Pathways Pa-es.

I ~42-,1 3p Factors for Aquatic Foods pCi 1/ kg pCi Element Fish Invertebrate H 9.0E-1 9.OE-1 C 4.6E3 9. 1E3 Na 1.OE2 2.0E2 P 1 1.0E5 2.0E4 Cr 2.0E2 2.0E3 Mn 4.0E2 9.0E4 Fe l.OE2 3.2E3 Co 5.OE1 2.0E2 Ni 1.OE2 L.OE2 Cu 5.OE1 4.0E2 Zn 2.0E3 l.OE4 Br 4.2E2 3.3E2 Rb 2.0E3 l.OE3 Sr 3.OE1 I.OE2 Y 2.5E1 1.0E3 Zr 3.3EO 6.7E0 Nb 3.0E4 - 1.0E2 Mo l.OE1 l.OEl Tc 1.5E1 5.OEO Ru 1.OE1 3.OE2 Rh l.OE1 3.0E2 Te 4.0E2 6.1E3 I 1.5E1 5.OEO Cs 2.0E3 1.OE3 Ba t 4.0EO 2.0E2 La 1 2.5E1 1.0E3 Ce 1.OEO 1.0E3 Pr 2.5E1 1.0E3 Nd 2.5E1 1.0E3 W I 1.2E3 1.OE1 Np r l.OE1 4.0E2 Table A-I of Reg. Guide 1.109.

Information I PMP-6010-OSD-001 Rev. 17 Page 44 of 84 OFF-SITE DOSE CALCULATION MANUAL .1 Dose Factors for Various Pathways Pages:

Dup. External Dose Factors for Standing on Contaminated Ground mrem m2 / hr pCi Radionuclide Total Body Skin H-3 I 0 0 C-14 0 0 Na-24 2.5E-8 I 2.9E-8 P-32 0 0 Cr-51 2.2E-10 2.6E-10 Mn-54 5.8E-9 6.8E-9 Mn-56 1.1E-8 1.3E-8 Fe-55 1 0 0 Fe-59 8.OE-9 9.4 E-9 Co-58 7.OE-9 8.2E-9 Co-60 1.7E-8 2.OE-8 Ni-63 0 0 Ni-65 3.7E-9 4.3E-9 Cu-64 1.5E-9 1.7E-9 Zn-65 4.0E-9 4.6E-9 Zn-69 0 0 Br-83 6.4E-1 l 9.3E-11 Br-84 1.2E-8 1.4E-8 Br-85 0 0 Rb-86 6.3E-10 7.2E-10 Rb-88 3.5E-9 4.0E-9 Rb-89 1.5E-8 1.SE-8 Sr-89 5.6E-13 6.5E-13 Sr-91 7.1E-9 8 3E-9 Sr-92 l 9.OE-9 l.OE-8 Y-90 I 2.2E-12 2.6E-12 Y-91m 1 3.8E-9 4.4E-9 Y-91 I 2.4E-11 2.7E-11 Y-92 I 1.6E-9 1.9E-9 Y-93 5.7E-10 7.8E-10 Zr-95 5.OE-9 5.8E-9 Zr-97 5.5E-9 6.4E-9 Nb-95 5.1E-9 6.OE-9 Mo-99 1.9E-9 2.2E-9 Tc-99m[ 9.6E-10 1. lE-9 Tc-101 2.7E-9 3 OE-9 Ru-103 3.6E-9 4.2E-9 Ru-105 4.5E-9 5.1IE-9 Ru-106 1.5E-9 1 8E-9 Ag-4Or I 1.8E-8 2.IE-8 Te-125m 3.5E-11 4.8E-11

Information PNIP-6010-OSD-001 Rev. 17 l Page 45 of 84 OFF-SITE DOSE CALCULATION NLANUAL Atahet31Dose Attahmen 3.142- Factors for Various Pathways Pages:

45 J Radionuclide Total Body Skin Te-127m 1.lE-12 1.3E-12 Te-127 1.OE-11 1.1E-11 Te-129m 7.7E-10 9.OE-10 Te-129 7.1E-10 8.4E-10 Te-131m 8.4E-9 9.9E-9 Te-13 1 2.2E-9 2 .6E-6 Te-132 I 1.7E-9 2.0E-9 1-130 1.4E-8 1.7E-8 1-131 2.8E-9 3.4E-9 1-132 1.7E-8 2.0E-8 1-133 3.7E-9 4.5E-9 1-134 1.6E-8 1.9E-8 1-135 1.2E-8 j 1.4E-8 Cs-134 1.2E-8 1.4E-8 Cs-136 1.5E-8 1 1.7E-8 Cs-137 4.2E-9 1 4.9E-9 Cs-138 2.1E-8 2.4E-8 Ba-139 2.4E-9 2.7E-9 Ba-140 2.1E-9 2.4E-9 Ba-141 4.3E-9 4.9E-9 Ba-142 [ 7.9E-9 9.OE-9 La-140 1.5E-8 1.7E-8 La-142 1.5E-8 1.8E-8 Ce-141 5.5E-10 6.2E-10 Ce-143 2.2E-9 2.5E-9 Ce-144 3.2E-10 3.7E-10 Pr-143 1 0 0 Pr-144 i 2.OE-10 2.3E-10 Nd-147 t 1.OE-9 1.2E-9 W-187 [ 3.1E-9 3.6E-9 Np-239 9.5E-10 1. 1E-9 Table E-6 of Reg. Guide 1.109.

Information PNIP-6010-OSD-001 I Rev. 17 Paze 46 of 84 OFF-SITE DOSE CALCUIATION MANUAL Attachment 3.2 Radioactive Liquid Effluent Monitoring Instruments 4 age:47 Instrument Minimum Applicability Action Channels Operable'

1. Gross Radioactivity Monitors Providing Automatic Release Termination

a. Liquid Radwaste (1)# At times of release 1 Effluent Line (RRS-1001) I

b. Steam Generator (1)# At times of release'* 2 Blowdown Line (R-19, DRS 3/4100 +)

c. Steam Generator (1)# At times of release 2 Blowdown Treatment Effluent (R-24, DRS 3/4200 +)

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination

a. Service Water (1) per At all times 3 System Effluent Line(R-20, train #

R-28, WRA 3/4500 and WRA 3/4600 +)

3. Continuous Composite Sampler Flow Monitor

a. Turbine Building Sump (1) At all times 3 Effluent Line

4. Flow Rate Measurement Devices

a. Liquid Radwaste Line (1) At times of release 4 (RFI-285)

b. Discharge Pipes* (1) At all times NA

c. Steam Generator Blowdown (1) At times of release 4 Treatment Effluent (DFI-352)

Pump curves and valve settings may be utilized to estimate flow; in such cases, Action Statement 4 is not applicable. This is primarily in reference to start up flash tank flow.

OPERABILITY of RRS-1001 includes OPERABILITY of sample flow switch RFS-1010, which is an attendant instrument as defined by Technical Specification 1.6. This item is also applicable for all Eberline liqoid monitors (and the:r respective flow switches) listed here t Since these monitors can be used for either batch or continuous release the appropriate action statement of 1 or 2 should apply (that is, Action 1 if a steam generator drain is being performed in lieu of Action 2).

+ Westinghouse (R) radiation monitors are being replaced by Eberine (DRS & WRA) monitors. Either monitor can fulfill the operability requirement.

a IF an R.MS monitor is inoperable solely as the result of the loss of its control room alarm annunciation. THEN one of the followng actions is acceptable to satisfy the ODCMv action statement comvensatory surveillance requirement:

1. Collect grab samples and conduct laboratory analyses per the specific monitor's action statement,

-OR-

Information I PMP-6010-OSD-01 I Rev. 17 l Page 47 of 84 OFF-SITE DOSE CALCULATION MANUAL Hi Pa-es:

Attachment 3.2 Radioactive Liquid Effluent M'vonitoring11~~ Instruments Pages I 46 -47 _

2. Collect local monitor readings at a frequency equal to or greater than (more frequently than) the action frequeny.

IF the RMS monitor is inoperable for reasons other than the loss of control room annunciation, THEN the only acceptable action is taking grab samples and conducting laboratory analyses as the reading is equivalent to a grab sample when the monitor is functional.

TABLE NOTATION Action 1 With the number of channels OPERABLE less than required by the Mlinimnum Channels OPERABLE requirement, effluent releases may continue, provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Step 3.2.3a and;

2. At least two technically qualified members of the Facility Staff independently verify the discharge valving. Otherwise, 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 30 days provided grab samples are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-7 jiCi/gram:

1. At least once per shift when the specific activity of the secondary coolant is > 0.01 pICi/gram DOSE EQUIVALENT 1-131.

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is S 0.01 pCi/gram DOSE EQUIVALENT 1-131.

Action 3 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per shift, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 IpCi/lrn. To further clarify this for ESW monitors the following is provided:

IF the Westinghouse monitor (R-20 and/or R-28) is fulfilling the applicability requirement, THEN grab samples are only needed if the Containment Spray Heat Exchanger is in service since the Westinghouse ESW monitors are only used for post LOCA leak detection and have no auto trip function associated with them.

OR IF the Eberline monitor (WRA-3/45C0 and/or WRA-3/4600) is fulfilling the applicability requirement, THEN grab sampling is required whenever the monitor is inoperable and the applicable train of ESW is in service since this monitor is located in the system effluent.

Action 4 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 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Compensatory actions are governed by PMP-4030.EIS.001, Event-Initiated Surveillance Testing

Information I P.NIP-6010-OSD-001 I Rev. 17 1 Page of 84 OFF-SITE DOSE CALCULATION NUNUAJL Attachment 3.3 I Radioactive Liauid Effluent' Monitoring Pages:

Instrumentation Surveillance Requirements 48 - 49 Instrument

1. Gross Radioactivity Monitors Providing Automatic Release Termination

a. Liquid Radwaste D* P R(3) Q(5)

Effluent Line (RRS-1001)

b. Steam Generator D* . R(3) Q(1)

Blowdown Effluent Line

c. Steam Generator D* "M R(3) Q(l)

Blowdown Treatment Effluent Line

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination

a. Service Water D M R(3) Q(2)

System Effluent Line

3. Continuous Composite Samplers ._.

a. Turbine Building l D* 1 N/A N/A N/A Sump Effluent Line l _

4. Flow Rate Measurement Devices

a. Liquid Radwaste D(4)* l N/A R Q Effluent

b. Steam Generator D(4)* N/A N/A N/A Blowdown Treatment Line I During releases via this pathway

Information I PMP-6010-OSD-001 I Rev. 17 Paee 49 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 33 Radioactive Liquid Effluent Monitoring Pages:

t Instrumentation Surveillance Requirements I d8 - 49 TABLE NOTATION

1. Demonstrate with the CHANNEL FUNCTIONAL TEST that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarni/trip setpoint.

2. Circuit failure.*

3. Instrument indicates a dowascale failure. *

4. Instrument control not set in operating mode.*

5. Loss of sample flow. #

2. Demonstrate with the CHANNEL FUNCTIONAL TEST that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operating mode.

5. Loss of sample flow. #

3. Perform the initial CHANNEL CALIBRATION using one or more sources with traceability back to the National Institute of Standards and Technology (NIST). These sources permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

4. Verify indication of flow during periods of release with the CHANNEL CHECK. Perform the CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch releases are made.

5. Demonstrate with the CHANNEL FUNCTIONAL TEST that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.**

3. Instrunent indicates a downscale failure.**

4. Instrument control not set in operating mode.*

5. Loss of sample flow.

Instrument indicates, but does not provide for automatic isolation

- Instrunent indicates, but does not necessarily cause automatic isolation. No credit is taken for the automatic isolation on such occurrences.

r Applicable only to Eberline sample flow instrumentation Operations currently performs the routine channel checks and source checks. Maintenance and Radiation Protecuon perform channel calibrations and channel functional tests. Chemistry performs the channel check on the continuous composite sampler.

These responsibilities are subject to change without revision to this document.

Information l PIP-6010-OSD-001 I Rev. 17 Page 50 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation Pages:

I ~~50 -52__

instrument (Instrument 4 Operable' l Ninimum Channels r Action Action

1. Condenser Evacuation System

a. Noble Gas Activity (1) T6 b.

Monitor (SRA-1905/2905)

Flow Rate Monitor (SFR-401, 1/2-MIR-054 and/or SRA- 1910/2910)

(1) j_ _ _ 5

2. Unit Vent. Auimlmary Building Ventilation System

a. Noble Gas Activity (1)6 Monitor (VRS-1505/2505)

b. Iodine Sampler (1)

8 Cartridge for VRA-1503/2503

c. Particulate Sampler Filter (1)

8 for VRA-1501/2501

d. Effluent System Flow Rate (1)

5 Mleasuring Device (VFR-315, IVIR-054 and/or VFR-151025 10)

e. Sampler Flow Rate (1) 5 Measuring Device (VFS-152112521)

3. Containment Purge and Containment Pressure Relief (Vent)

a. Containment Noble Gas Activity Monitor (1) *** 2 3 7 ERS-13/1405 (ERS-23/2405)

b. Containment Particulate Sampler Filter (1) **** 10 ERS-1311401 (ERS-23/2401)

4. Waste Gas Holdup System and CVCS HUT

a. Noble Gas Activity (1) ***1 Alarm and Termination of Waste Gas Releases (VRS-1505/2505)

5. Gland Seal Exhaust

a. Noble Gas Activity Monitor (SRA-1805

b. Flow Rate Molotor (SFR-201, MIR-054 or SFR-1810/2810)

At all tunes

- - - During releases via this pathway

Information PMP-6010-OSD-001 Rev. 17 l Page 51 of 84 OFF-SITE DOSE CALCULATION MANUAL .4 Radioactive Gaseous Effluent Mionitoring, Instrumentation BPages:

50 5 1

TABLE NOTATIONS

1. IF an RMS monitor is inoDerable solely as the result of the loss of it's control room alarm annunciation, TEEN one of the following actions is acceptable to satisfy the ODCM action statement compensator; surveillance requirement:

I, Take grab samples and conduct laboratory analyses per the specific monitor's action statement,

-OR-

2. Take local monitor readings at a frequency equal to or greater than (more frequently than) the action frequency.

IF the RMS monitor is inoperable for reasons other than the loss of control room annunciation, THEN the only acceptable action is taking grab samples and conducting laboratory analyses as the reading is equivalent to a grab sample when the monitor is functional.

2. Consider releases as occurring "via this pathway" under the following conditions:

The Containment Purge System is in operation and Containment integrity is established/required,

-OR-

The Containment Purge System is in operation and is being used as the vent path for the venting of contaminated systems within the containment building prior to completing both degas and depressurization of the RCS.

IF neither of the above are applicable, TEEN the containment purge system is acting as a ventilation system and is covered by Item 2 of this Attachment.

-OR-

A Containment Pressure Relief (CPR) is being performed.

3. For purge (including pressure relief) purposes only. See Technical Specification table 3.3-6 for additional information.

4. For waste gas releases only, see Item 2 (Unit Vent, Auxiliary Building Ventilanon System) for additional requirements.

ACTIONS

5. With the number of channels OPERABLE less than required by the Minimumn Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . After 30 days, IF the channels are not OPERABLE, TEEN continue releases with estimation of the flow rate once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent Release Report.

6. With the number of channels OPERABLE less required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 30 days, IF the channels are not OPERABLE, THEN continue releases with grab samples once per shift and provide a description of why the inoperability was not corrected in the-next Annual Radiological Effluent release Report.

7. With the number of channels OPERABLE less than required by the Minrimu Channels OPERABLE requirements, immediately suspend PURGING or VENTING (CPR) of radioactive effluents via this pathway.

8. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples required for weekly analysis are continuously collected with auxrliary sampling equipment as required in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program. After 30 days, IF the channels are not OPERABLE, THEN continue releases with sample collection by 3uxiliary sampling equipment and provide a description of why the mcperability was not corrected in the next Annual Radiological Effluent Release Report.

Sampling evolutions are not an interruption of a continuous release or sampling period.

9. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that pnor to initiating the release:

a. At least two independent samples of the tank's contents are analyzed and,

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineups; otherwise, suspend release of radioactive effluents via this pathway.

10. See Technical Specification 3.4.6.1.

Compensatory actions are governed by PMP-4030.EIS.001, Event-Initiaed Surveillance Testing.

Information l PMIP-6010-OSD-001 Rev. 17 Page 53 of 84 OFF-SITE DOSE CALCULATION IMANUAL Radioactive Gaseous Effluent Monitoring Pages:

tthmnt 3.5 Instrumentation Surveillance Requirements 53 - 54 Instrument CHANNEL l SOURCE l CHANNEL CHANNEL CBECK CBECK CALIBRATION FUNCTIONAL TEST

1. Condenser Evacuation Alarm Only System

a. Noble Gas Activitv Monitor D** M R(2) Q(l)

(SRA-1905/2905)

b. System Effluent Flow Rate D** NA R. Q (SFR-401, MR-054, SRA-1910t2910)

2. Auxiliary Building Unit Alarm Only Ventilation System

a. Noble Gas Activity Monitor D* l MI R(2) Q(1)

(VRS-1505/2505) _ _I_ _ _ _

b. Iodine Sampler W* NA NA NA (For VRA-150312503) I

c. Particulate Sampler W* NA NA NA (For VRA-150112501)

d. System Effluent Flow Rate D* NA R Q Measurement Device (VFR-315, MR-054, VRS-1510/2510)

e. Sampler Flow Rate D* N/A R Q Measuring Device (VFS-1521/2521)

3. Containment Purge System and Alarm and Trip Containment Pressure Relief

a. Containmnent Noble Gas SP+ R(2) Q Activity Monitor (ERS-13/1405 and ERS-23/2405) _

b. Containment Particulate S** NA R Q Sampler (ERS-13/1401 and ERS-23/2401)

4. Waste Gas Holdup System Including CVCS HUT

a. Noble Gas Activity Monitor Providing Alarm and Termination (VRS-150512505) 5 Gland Seal Exhaust

a. Noble Gas Activity (SRA-1805t2805)

b. System Effluent Flow Rate (SFR-201, MR-054, SRA-1810/2810)

At all times

t Durng releases via this pathway

Information PMIP-6010-OSD-O01 I Rev. 17 Page 54 of S4 OFF-SITE DOSE CALCULATION UMANUAL Radioactive Gaseous Effluent Mvonitonng Pages:

e 3 I Instrumentation Surveillance Requirements 53 - 54 TABLE NOTATIONS I Demonstrate with the CHANNEL FUNCTIONAL TEST that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpont.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

2. Perform the itutial CHANNEL CALIBRATION using one or more sources with traceability back to the NIST. These sources permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

3. Demonstrate with the CHANNEL FUNCTIONAL TEST that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.*

3. Instrument indicates a downscale failure.*

4. Instrument controls not set in operate mode.*

Instrument indicates, but does not provide automatic isolation.

Operations currently perfbrms the routine channel checks, and source checks. Maintenance ar.d Radiation Protection perform channel calibrations and channel functional tests. These rcsponsibilztes are subject to change without revision to this document.

fInformation PIMP-6010-OSD-001 Rev. 17 Paue 55of8 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.6 Radioactive Liquid Waste Sampling and Analysis Program P

[Ref. 5.2.1s]

LIQUID SAMIPLING ILNIUIM TYPE OF LOWTER RELEASE FREQUENCY ANALYSIS ACTIVITY LIMIT OF TYPE FREQUENCY ANALYSIS DETECTION (LLD)

(ACi/ml) a A. Batch Waste P P Principal 5x10'7 Release Tanks c Each Batch Each Batch Gamma Emitters 1-131 1x106 P P Dissolved and Entrained Gases Each Batch Each Batch (Gamma lx10-5

_______ ______ ___ Emitters)

P M H1-3 1x1O7, Each Batch Composite'b Gross Alpha 1x10-P Q Sr-89, Sr-90 5x10 4 Each Batch Compositeb Fe-55 lxlO4 B. Plant W Principal Continuous Daily Composite b Gamma 5x10-7 Releases* d Emitters' 1-13 1 1x104 M M Dissolved and Grab Sample Entrained Gases 1x107 (Gamma Emitters) 11-3 lx10(5 Daily Compositeb Gross Alpha ixiO 7 Q Sr-89, Sr-90 5x10-3 Daily Composite b II Fe-55 l1X0O l

Duri.na releases via this pathway

Information I PNIP-6010-OSD-001 Rev. 17 . Page 56 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.6 Radioactive Liquid Waste Sampling and Analysis Progam Pages:

55 -56 TABLE NOTATION

a. The lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3 20, Maximum Values for Lower Limits of DetectionsAB - REMP.

b. 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.

c. A batch release Ls the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, isolate, recirculate or sparge each batch to ensur- thorough mixing.

d. A continuous release Is the discharge of liquid of a non-dLscrete volume; e.g. ftrom a volume of system that has an input flow during the continuous release.

e. The principal ganma emitters for which the LLD specification applies exclusively are 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. Identify and report other peaks, which are measurable and identifiable, together with the above nuclides.

Information P1IP-6010-OSD-001 I Rev. 17 T Page 57 of 84 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Waste Sampling and Pages:

L Analysis Program 57 -58 Gaseous Release Type Frequency Minimum Type of Lower Limit Analysis Actiifty of Detection Frequency Analvsis (iCsml)

a. Waste Gas Storaze P Principal Gamma Tanks and CVCS HUTs Each Tank Each Tank Emitters' I x 10' Grab Sample .

H-3 I x l04

b. Contairment Purge P P Principal Gamma Each Purge Each Purge Emitters d 1 x 104 Grab Sample CPR (vent)** Twice per Twice per Month Month H-3 1 x 10'

c. Condenser Evacuation W or M M Principal Gamma System Grab Sample Particulate Sample Emitters 1 x 1011 Gland Seal Ethaust* M H-3 1x 104 WV' Principle Gamma 1 x 104 Noble Gas Emitters d M 1-131 Iodine Adsorbing - I x l0-'2 Media Continuous Wg Noble Gases Noble Gas Monitor I xlo-,

d. Auxfliary Building Unit Continuousc Wb 1-131 Vent* Iodine Adsorbing 1 x 10-"-

Media Continuous ' W b Principal Gamma l Particulate Sample Emitters d 1cX 1a" Continuous' M Gross Alpha Composite Pareiculate I x l(Y1 Sample w Wh H-3 Grab Sample H-3 Sample I1 x 104 WI IPrinciple Gamma 1 x io4 Noble Gas Emnitters d Continuous' Sr-89, Sr-90 Composite Particulate I x 10-,

Sample Continuousc Noble Gas Monitor Noble Gases 1 1 x l0o-

e. Incinerated 0lP P Princpal Gamma Each Batch' Each Batch E tters d X iO05

Durmg releases via this pathway

"*Only a twice per month sampling program for containment noble gases and 1H3 is required

Information ' PMIP-6010-OSD-01 I Rev. 17 j Page 58 of 84 OFF-SITE DOSE CALCULATION MANU AL I Radioactive Gaseous Waste Sampling and Pages:

Attachment 3.7 Analysis Proram I Anlyt - 1rgam5 TABLE NOTATION

a. The lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3 20, Maximum Values for Lower Limits of DetectionsAB- REMNIP.

b. Change samples at least once per 7 days and complete analyses wlthin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing. Perform analyses at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days following each shutdown, startup or THERMAL POWER change > 15%S per hour of RATED THERMAL POWER WHEN samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, THEN the corresponding LLDs may be increased by a factor of 10. This requirement does not apply IF (1) analysis shows that DOSEQ 1131 concentration in the RCS has not increased more than a factor of 3; and C2)the noble gas monitor shows that effluent activity has not increased more than a factor of 3. [Ref s lAy]

c. Know the ratio of the sample flow rate to the sampled stream flow rate for the time period covered by each dose or dose rate calculation made in accordance with steps 3.2.4a, 3.2.4b, and 3.2.4c of this document.

Sampling evolutions are not an interruption of a continuous release or sampling period.

d The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133M, Xe-135 and Xe-138 for gaseous emussions 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. Identify and report other peaks, which are measurable and identifiable, together with the above nuclides.

e. Releases from incinerated oil are discharged through the Auxilary Boiler System. Account for releases based on pre-release grab sample data.

f. Collect samples of waste oil to be incinerated from the container in which the waste oil is stored (example: waste oil storage tanks, 55 gal. drums) prior to transfer to the Auxiliary Boiler System. Ensure samples arc representative of container contents.

g. Obtain and analyze a gas marinelli grab sample weekly for noble gases effluent quantification.

h. Take trittum grab samples at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling cavity is flooded.

i. Grab sampling of the Glanxd Seal Exhaust pathway need not be performed if the RMS low range channel (Sha.-1805r2805) readings are less than IE-6 pC/cc. Attach the RMfS daily averages in lieu of sampling. This is based on operatung experience indicating no activity is detected in the Gland Seal Exhaust below this value. Compensatory sampling for out of service monitor is still required in the event 1805/2805 is inoperable.

I Information PMP-6010-OSD-001 I Rev. 17 Page, 59 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.8 Multiple Release Point Factors for Release Points PI-e:

Liquid Factors Monitor DescriDtion Monitor Number MRP #

U 1 SG Blowdown 1R19/24, DRS 3100/3200* 0.35 U 2 SG Blowdown 2R19124, DRS 410014200* 0.35 U 1 & 2 Liquid Waste Discharge l RRS-1000 0.30 Gaseous Factors Monitor Desc:iption Monitor Number Flow Rate (cfrn) MRP #

Unit 1 Unit Vent VRS-15C0 186,600 0.54 Gland Seal Vent SRA-18G0 1,260 0.00363 Steam let Air Ejector SRA-1900 3,600 (b) 0.01 Start Up FT Vent 1,536 0.004 Total 192,996 Unit 2 Unit Vent VRS-2500 143,400 0.41 Gland Seal Vent SRA-2800 5,508 (a) 0.02 Steam Jet Air Ejector SRA-2900 3,600 (b) 0.01 Start Up FT Vent 1,536 0.004 Total 154,044 Either R-19, 24, DRS 3/4100 or 3/4200 can be used for blondowtn morutoring as the Eberline monitors (DRS) are replacing the Westinghouse (R) monitors.

Nominal Values a Two release points of 2,754 cfi each are totaled for this value.

b This is the total design maximum of the Start Up Air Ejectors. This is a conservative value for unit 1.

P.Nr-P-6010-OSD-001 I Rev. 17 Page 60 of S4 SITE DOSE CALCULATION MANUAL Paoe:

Liquid Effluent Release Systems 60 FaEdZE SYSTBG PONTS Wa. Phil

Information P'IP-6010-OSD-001 Rev. 17 Page 61 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.10 Plant Liquid Effluent Parameters Page SYSTEMN COMPONENTS TCAPACITY FLOW RATE TANKS PPUMIPS (EACH) (EACH)*

I Waste Disposal System

+ Chemical Drain Tank 1 1 I600 GAL. 20 GPM

+ Laundry & Hot Shower Tanks 2 1 600 GAL. 20 GPM

+ Monitor Tanks 4 2 21,600 GAL. 150 GPM

+ Waste Holdup Tanks 2 - 25,000 GAL.

+ Waste Evaporators 3 30 GPM

+ Waste Evaporator Condensate 2 2 6,450 GAL 150 GPM Tanks____________ __

II Steam Generator Blowdown and Blowdown Treatment Systems

+ Start-up Flash Tank (Vented)# l 1 _ 1,800 GAL. 580 GPM

+ Normal Flash Tank (Not 1 525 GAL. 100 GPM Vented)

+ Blowdown Treatment System l l 1 60 GPM III Essential Service Water System

+ Water Pumps 4 _ 10,000 GPM

+ Containment Spray Heat 4 _ 3,300 GPM Exchanger Outlet _

IV Circulating Water Pumps junit 3 Ii 230,000 GPM Unit 2 l 1 4 230,000 GPM Nominal Values The 580 gpm value is calculated from the Estimated Steam Generator Blowdown Flow vs. DRV Valve Position letter prepared by M. J. O'Keefe, dated 9/27/93. This is 830 gnpm times the 70% that remains as liquid while the other 30%5flashes to steam and exhausts out the flash tank vent.

Information l PMIP-6010-OSD-001 Rev. 17 T Page 62 of 84 OFF-SITE DOSE CALCULATION MvANUAL .11 Volumetric Detection Efficiencies for Principle Gamma Page:

A Emitting Radionuclides for Eberline Liquid Monitors 62 This includes the following monitors: RRS-1000, DRS 3100, DRS 3200, DRS 4100, DRS 4200, WRA 3500, WRA 3600, W-RA 4500 and WRA 4600. [Ref.5.2.1p]

INUCLEDE EFFICIENCY (cprnl/>Cifcc) 1-131 3.78 E7 Cs-137 3.00 E7 Cs-134 1 7.93 E7 Co-60 1 5.75 E7 Co-58 4.58 E7 Cr-51 3.60 E6 Mn-54 3.30 E7 Zn-65 1.58 E7 Ag-lO1M 9.93 E7 Ba-133 4.85 E7 Ba-140 1.92 E7 Cd-109 I 9.58 E5 Ce-139 [ 3.29 E7 Ce-141 j 1.92 E8 Ce-144 4.83 E6 Co-57 3.80 E7 Cs-136 j 1.07 E8 Fe-59 2.83 E7 Sb-124 5.93 E7 1-133 3.40 E7 1-134 7 23 E7 1-135 3.95 E7 Mo-99 8 68 E6 Na-24 4.45 E7 Nb-95 3 28 E7 Nb-97 3.50 E7 Rb-89 5.G0 E7 Ru-103 3.48 E7 Ru-106 1.23 E7 Sb-122 2.55 E7 Sb-125 3.15 E7 Sn-113 7.33 E5 Sr-85 3.70 E7 I Sr-89 2.88 E3 I Sr-92 3.67 E7 I Tc-99M 3.60 E7 I Y-88 5.25 E7 Zr-95 3.38 E7 Zr-97 3.10 E7 Kr-85 1.56 E5 Kr-85Mv 3.53 E7 Kr-88 4.10 E7 Xe-131M 8.15 E5 Xe-133 7.78 E6 Xe-133M 5.75 E6 Xe-135 3.83 E7

( - I --- ( --- I- - (- -, (- ___ f__ [ (__ - ( --- t - __ t__ __ f --- I- I ____ (--

Informa MPI-60t0-OSD-001 Prtioii Rev. 17 l Page 63 of 84 OFF-SITE DOSE CALCULATION MANUAL Ataclhment 3,12 Counting Efficiency Curves for R-19, and R-24 Pages:

Counting Efficiency Curve for R-19 Efficiency Eactor = 4.2 E6 cpm/uCinml on empirncal daia Iaken during prc vIcr4lional isimlug with Cs 137)

(BAsed 1 OOE+07 I OOE+O.

I OOE+05 1,00E+04 I>

.51 m 1 OOE/03 n1 00E+02 - ,

1 OOE+02 1 OOE+O0) o 9 9 9 9 C0 0 C) C>

Cl 0 000 mlcrocu rles/mI

,,-- - -----1 7 ._-

Iniformjiationi PMP-6010-OSD-001 Rev. 17 Page 6of84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.12 Counting Efficiency Curves 9 for R-19, and R-24 Pajes:

63 - 64 Counting Efficiency Curve for R-24 Efficiency Factor =7.5E6 cpniliCt/ml (tased oil empirical data taketi during preopeauoiail CsUnIg with IYi-54) 1 OOEf07 1OOE+06 1 OOE+05 C

1 OOE+04 n 1.O0E+03 IL I OOEtO2 1 OOE+00 CNJ 0) 99 9 w

0 0 0 r r , - 1 mlcrocurlesIml

( [ I f I ( I I I ( I I I I I I

I - - ( - ( __ (_ - F - __ (_ - F - I __ ( ___ F -__ ( --- (____ [- __ f___ I I - [--_ - I Informnatioii - I PMP-6010-OSD-OO1 l Rev. 17 Page 65 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.13 Counting Efficiency Curve for R-20, and R-28 Page:

Colinting Efflclency Curve for R-20 anld R-28 Efficiency Factor = 4 3 E6 cpm/uCinit ,

(Bascdon empincal dataIoken during pre-opuilionat tsting wilth Co-S8) 1 OOE#o7 1 OOE+06 1 OOE.05 C

i OOE+04 E0

.0 05 Us 1 OOE*03 1 OOEtO2 I 00E*02 1 OOE401 I uu--u

-0 I? 09 9 (.

9n I?

O vu0 Lu 0 hi 0

5J Cu s

8 0 o0 o.

mlcrocuriasstml

Information PMiP-6010-OSD-001 Rev. 17 Page 66 of v4 OFF-SITE DOSE CALCULATION MNA'NUAL .14 Gaseous Effluent Release Systems 66ge

Information PMP-6010-OSD-001 I Rev. 17 1 Page 67 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.15 Plant Gaseous Effluent Parameters Page SYSTEM UNIT EXHAUST CAPACITY FLOW RATE (CFM)

I PLANT AUXILIARY BUILDING 1 186,600 max U`IT AGENT 2 143,400 max!

3 WASTE GAS DECAY TANKS (8) 1 125 4082 FTV 100 psig AND CHEMICAL & VOLUME 28,741 ft3 max CONTROL SYSTEM HOLD UP @ 8rt, 0 level TANKS (3)

+ AUIMLIARY BUILDING 1 72,660 EXHAUST 2 59,400

+ ENG. SAFETY FEATURES 1& 2 1 50,000 VENT

+ FUEL HANDLING AREA VENT 30,000 SYSTEM 0 CONTAINMENT PURGE SYSTEM 1& 2 32,000 CONTAINMENT PRESSURE 1&2 1,000 RELIEF SYSTEM INSTRUMENT ROOM PURGE 1& 2 1,000 SYSTEM II CONDENSER AIR EJECTOR 2 Release Points SYSTEM [ One for Each Unit NORMAL STEAM JET AIR 1& 2 230 EJECTORS STARTUP STEAM JET AIR 1&2 3,600 EJECTORS III TURBINE SEALS SYSTEM T 1 1,260 2 5,508 2 Release Points l for Unit 2 IV START UP FLASH TANK VENT 1 1,536 l _l l 2 l 1,536 l

+ Designates total flow for all fans.

Information l PIP-6010-OSD-001 I Rev. 17 Pane 68 of 0 OFF-SITE DOSE CALCULATION MIAINUAL Attachment 3.16 10 Year Average of 1989-1998 Data Pages:

X/Q GROUND AVERAGE (sec/m3 )

DIRECTION __ DMSTANCE (METERS)

OVIND FROM) 594 22416 4020 5630 l 7240

. _ I _ _ _ _ _ -_

N 3.50E-06 4.23E-07 1 1.97E-07 1.16E-07 j 8.13E-08 NNE 2.69E-06 3.22E-07 1 53E-07 9 16E-08 6.44E-08 NE 3 64E-06 4.5 lE-07 1 2.20E-07 1 33E-07 9.43E-08 ENE 5.94E-06 6.70E-07 3.35E-07 2.07E-07 1.48E-07 E 8.68E-06 1 9.50E-07 4.84E-07 3.03E-07 2.17E-07 ESE 3.45E-06 j 9.36E-07 I 4.75E-07 2.96E-07 2.12E-07 SE 9.7 1E-06 I 1.05E-06 5.38E-07 3.37E-07 2.42E-07 SSE 1.09E-05 I 1.20E-06 6.14E-07 3.86E-07 I 2.77E-07 S 1.16E-05 1 1.30E-06 6.53E-07 4.05E-07 12.89E-07 SSW 5.87E-06 6.70E-07 3.30E-07 2.01E-07 1 1.43E-07 SW 3.66E-06 4.26E-07 2.04E-07 1.23E-07 I 8.64E-08 WSW 2.84E-06 3.14E-07 I 50E-07 1.57E-07 1 6.32E-0s W 3.29E-06 3.69E-07 1.75E-07 1.04E-07 I 7.32E-08 WNW 3.20E-06 3.61E-07 1.69E-07 1.OIE-07 j 7.05E-08 _

NW 2.98E-06 3.33E-07 I 1.58E-07 9.44E-08 1 6.61E-08 NNW 3.41E-06 3.S1E-07 I1.78E-07 1.06E-07 j 7.41E-OSE_.

DERECTION DISTANCE (IMETERS) _

O(IND FROMN) 12067 l 24135 40225 56315 t 80500 N 4.03E-08 I 1.55E-08 7.71E-09 4.93E-09 I 3.09E-09 NNE 3.23E-08 1.26E-O8 6.27E-09 4.OE-09 I 2.52E-09 NE 4.78E-08 1.91E-08 9.52E-09 6.11E-09 i 3.88E-09 ENE 7.59E-08 3.08E-08 1 55E08 9.95E-09 I 6.37E-09 E 1. 12E-07 4.62E-08 2 33E-08 I I.OE-08 I 9.64E-09 ESE 1.lOE-07 4.50E-08 2.27E-08 1.46E-08 1 9.38E-09 SE 1.26E-07 5.20E-08 2.62E-08 1.55E-08 I 1.09E SSE 1.44E-07 5.94E-C8 2.99E-08 1.93E-08 1.24E-08 S 1.30E-07 I 6.09E-08 3.06E-08 1.97E-08 1.26E-08 SSW 7 31E-08 i 2.94E-08 1.47E-O8 9.39E-09 1 5.97E-09 SW 4.35E-08 I 1.72E-08 8.56E-09 5.48E409 13.47E WSW 3 1SE-08 I 1.25E-08 6.22E-09 3.99E-09 t 2.53E-09 W 3.66E-08 1.43E-08 7.07E-09 4.55E-09 2.85E-09 WNW 3.50E-08 1.35E-08 6.70E-09 4.28E-09 2.69E-09 NW 3.30E-08 1.28E-08 6 38E-09 4.09E-09 i 2.57E-09 NNW 3 68E-08 1.43E-08 77 08E-09 4.54E-09 I 2.85E-09 DIRECTION - SECTOR !

N = Lk IE =E S W NNE = B I ESE =F SSW =K WINW = P NE = C SE =G SW =L NW = Q ENE = D ISSE =H IWSW =M I NNW = R Worst Case X/Q = 1 65E-5 seclm3 in Sector A 2001

Information l PMP-6010-OSD-001 Rev. 17 Pa-e 69 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1989-1998 Data Pages:

68 -69 2)

D/Q DEPOSION (W/m i DIRECTION I DISTANCE (METERS)

(WND FROM) 594 2416 2 14020 15630 7240 N 2.46E-08 l238E-09 1.08E-09 l 5.66E-10 3.62E-10 NNE 1.06E-08 I 1.02E-09 4.62E-10 1 2.43E-10 1.55E10 NE 1.31E-08 1.27E-09 5.75E-10 I 3.02E-10 1.93E-i0 ENE 1.62E-08 I 1.56E-09 7.09E-10 13.72E-l0 2.37E-10 E I1.92E-08 I 1.85E-09 8.39E-10 4.4E-10 2.81E-10 ESE } 1.82E-08 I 1.76E-09 7.98E-10 4.19E-10 2.67E-10 SE 1.85E-08 I 1.79E-09 8.09E-10 4 25E-10 2.71E-10 SSE 2.24E-08 2.17E-09 9.84E-10 5.15E-10 3.29E-10 S 3.5E-08 3.38E-09 1.53E-09 8.03E-10 5.13E-10 SSW 2.3 1E-08 2.24E-09 1.01E-09 5.31E-10 3.39E-10 SW 2.14E-08 2.07E-09 9.38E-10 4.91E-10 J 3.14E-10 WSW 2.08E-08 2.01E-09 9.12E-10 4.78E-10 13.05E-10 W 2.13E-08 2.06E-09 9.33E-10 4.9E-10 I 3.13E-10 WNW 1.95E-08 1.89E-09 8.54E-10 4.48E-10 l2.86E-0 NW 1.62E-08 1.57E-09 7.11E-10 3.73E-10 2.33E-10 NNW 2.18E-08 2.11E-09 9.56E-10 5.01E-10 3.2E-10 DIRECTIN DISTANCE (METERS)

VIND FROM) 12067 124135 1 40225 56315 80500 N T1.51E-10 4.91E-11 l 1.81E-11 9.65E-12 4.84E-12 NNE 6.78E-11 2.1E- 11 7.75E-12 4.13E-12 2.07E-12 NE 8.18E-11 2.62E-11 9.64E-12 5.15E-12 2.58E-12 ENE 9.95E-1l 3.23E-11 1.19E-11 6.34E-12 3.18E-12 E 1.16E-10 3.82E-1_1 11.41E-11 7.5E-12 3.76E-12 ESE 1.12E- 10 3.64E- 11 1.34E- 11 7.14E-12 3.58E- 12 SE 1.13E-10 3.68E- 11 1.36E-11 7.24E-12 3.63E-12 SSE 1.37E-10 4.47E-1 1.65E-11 I 8.79E-12 4.41E-12 S 2.14E-10 6.97E-11 2.57E-11 I1.37E-11 6.87E-12 SSW 1.42E-10 4.61E-11 1.7E-11 i 9.06E-12 4.54E-12 SW 1.31E-10 I 4.27E-11 1.57E-11 I 8.38E-12 4.21E-12 WSW 1.27E-10 4.15E-1I 1.53E-1} 8.16E-12 4 09E-12 W 1.3E-10 4.25E- 11 1.56E-11 1.73E-11 4.19E-12 WNW 1.19E-10 3.89E-11I - 1.43E-11 7.64E-12 3.83E-12 NW l_1.78E-10 13.24E-11 1.19E-1 l 6.36E-12 3.19E-12 NNW l_1.34E-10

_ 4.35E-11 I 6E-11 8.55E-12 4.29E-12 DIRECTION - SECTOR N A E =E S =J W =N NNE =B ESE =F ISSW =K WNW = P NE =C SE =G SW =L NW =Q ENE = D SSE = H WSW M NNW = R Worst Case DIQ = 4.46E-08 I/m' inSector A 2001

Information PMP-6010-OSD-001 Rev. 17 Page 70 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment317 lAnnual Evaluation of x/Q and D/Q Values For Page:

1 All Sectors - 70

1. Performed or received annual update of z/Q and D/Q values. Provide a description of what has been received.

l Signature Date Environmnental Department (print name, title)

2. Worst z/Q and D/Q value and sector determined. PMfP-6010.OSD.O01 has been updated, if necessary. Provide an evaluation.

I Signature Date Environmental Department (print name, title)

3. Review nuclide mix for gaseous and liquid release paths to determine if the dose conversion factor of total body is still applicable. Provide an evaluation.

Signature Date Environmental Department (print name, title)

4. Approved and verified by:

Signature Date Environmental Department (print name, title)

5. Copy to NS&A for information.

Signature Date Environmental Department (print name, title)

ma PMP-6010-OSD-1 Rev. 17 Pa-,e 71 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.18 Dose Factors Pages:

DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMA AIR BETA AIR DOSE FACTOR FACTOR DOSE FACTOR DOSE FACTOR T

K (DFBi) Li (DFS.) M; (DF ,) N. (DFi) mrem m3 (mrem m 3 (mrad m3 (mrad m3 RADIONtJCLIDE per pCi yr) per gCi yr) per gCi yr) per pCi yr)

Kr-83m 7.56E-02 l --- 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 l.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 l 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 l 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+002 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03

': The listed dose factors are for radionuclides that may be detected in gaseous effluents, from Reg. Guide 1.109, Table B-l.

IPformation I PP-6010-OSD-001 I Rev. 17 l Page 72 of 84 OFF-SITE DOSE CALCULATION MANUAL Page 1 .18 Dose Factors I 71 - 72 J DOSE FACTORS FOR RADIOIODINNES ANLD RADIOACTIVE PARTICULATE, IN GASEOUS EFFLUENTS FOR CHILD* Ref. 5 2 lee and ff Th, P1 INHALATION FOOD & GROUND PATHWAY PATHWAY (mrem m' (mrem m-' sec RADIONUCLIDE per gCi yr) per pCi yr)

H-3 1.12E+03 1.57E+03' P-32 2.60E+06 7.76E+10 Cr-51 1.70E+04 1.20E+07 Mn-54 1.58E+06 1.12E+09 Fe-59 1.27E+06 5.92E+08 Co-58 1.1 lE+06 5.97E +08 Co-60 7.07E+06 4.63E+09 Zn-65 9-95E+05 1.17E+10 Rb-86 1.98E+05 8.78E+09 Sr-89 2.16E+06 6.62E+ 09 Sr-90 1.O1E+08 1.12E+11 Y-91 2.63E+06 6.72E+06 Zr-95 I 2.23E+06 3 44E+08 Nb-95 I 6.14E+05 4.24E+08 I Ru-103 I 6.62E+05 1.55E+08 Ru-106 1.43E+07 3.01E+08 Ag-lOrm 5.43E+06 1.99E+10 1-131 1.62E+07 l 4.34E+ 11 1-132 1.94E+05 I 1.78E+06 1-133 3.35E+06 3.95E+09 1-135 7.92E+05 1.22E+07 Cs-134 1.O1E+06 4 OOE+10 Cs-136 1.71E+05 3.OOE+09 Cs-137 9.07E+05 3.34E+ 10 Ba-140 1.74E+06 I 1.46E+0O Ce-141 5.44E+05 l 3.31E+07 Ce-144 1.20E+07 l 1.91E+08

's Sr-90, Ru-106 and I-131 analyses are performed, THEN use Pi given in P-32 for nonlisted radionuclides.

3 The ursts for bodi H3 factors are Ehe same. mrem m per 4Ci yr

Information PMIP-6010-OSD-001 Rev. 17 lPa-e 73 of 84 OFF-SITE DOSE CALCULATION MINANUAL A Radiological Environmental Monitoring Progan Pages:

Sample Stations, Sample Types Sample s 73 -76

[Ref. 5 2 Iv, 5.2.1a. 5.2.1tJ SSAMPLE I DESCRIPTIONI SA MPLE SANPLE ANALYSIS I ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY l ON-SITE AIRBORNE AND DIRECT RADIATION (ThD) STATIONS ONS-1 (r-1) 1945 ft @ 1S from Plant Axis Airborne Particulate Weekly (Gross Beta Weekly Weekly Gamma Isotopic Quart. Camup.

Airborne Weekly 1-131 Weekly Radioodirn ,

TLD Quarterly Direct Radiation j Quarterly ONS-2 cT-2) 2338 ft a 4S8 from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Weeldy Gamma Isotopic IQuart. Cornp.__

Airborne Weekly 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-3 (T-3) 2407 ft @ 900 from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Weekly Gamma Isotopic Quart. Corap.

Airborne Weekdy I-131 Weekly Radimoodine TLD Quarterlv Direct Radiation Quarterly ONS-4 (T-4) 1S52 ft. @ I180 from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Weekly Gamma Isotopic Quart Comp.

Airborne Weekly 1-131 Weekly Radlioodine TLD Quarterly Direct Radiation Quarterly ONS-5 (T-5) 1895 ft i 189° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Weekly Ganma Isotopic Quar. Conp.

Airborne Weekly 1-131 Weekly Radioiodine _

TLD Quarterly Direct Radiation Quarterly ONS-6 (T-6) 1917 ft 9 210° from Plant Axis Airbome Particulate Weekly Gross Beta Weekly Weekly Garra Isotopic Quart. Comro.

Airborne Weekly 1-131 Weekly Radiorodine TLD Quarterly Direct Radiation Quarterly T-7 2103 ft 0t 360 from Plant AXus TLD Quarterly Direct Radiation Quarterly T-8 2208 ft 0 82° from Plant Axis 1 TLD Quarterly Direct Radiation Quarterly T-9 1368 ft @ 149° from Plant Axls l TLD I Quarterly Dirxt Radiation Quarterly T-10 1390 ft @ 1270 from Plant Axis l TLD l Quarterly Direct Radiation Quarterly T-1 11969 ft@ 110from Plant A-us I TLD Quarterly Direct Radiation Quarterly T-12 2292 ft @631 from Plant Axis TLD I Quarterly Direct Radiation Quarterly CONTROL AIRBORNE AND DIRECT RADIATION (TLD) STATIONS NBF 15.6 rmles SSW Airborne Particulate Weekly Gross Beta Weekly New Buffalo, MI _ Weekly Gamma Isotopic Quart. Comno.

Airborne Radiomodine Weekly 1-131 Weeklv TLD I Quarterly Direct Radiation Quarterly SBN 26 2 miles SE Airborne Particulate Weekly Gross Beta Weekly South Bend. IN Weekly Gamna =Isotooic Quart. Canop Airborne Radioiodine Weekly 1-131 Weekly

'TLD Quarterly Direct Radiation Quarterly DOW l 24.3 miles ENE Airborne Particulate Weekly Gross Beta Weekly Dowagiac, MI Weekly Gamma Isotopic Quart. Como Airborne Radiucidlne Weekly 1-131 i Weekly TLD Quarterly Direct Radianon I Quarerly COL 18 9 miles NNE Airborne Particulate Weekly IGross Beta IWeekly Coloma. MI _ Weekly Gamma lsotopic IQuart. Corno Airborne Radioiodine Weekly 1-131 1AWeekly TLD [ Quarterly Direct Radiation J Quarterly

Information PMIP-6010-OSD-001 Rev. 17 Page 74 of 84 OFF-SITE DOSE CALCULATION MiIANUAL Atcm n 3.9 Radiological Environmental Monitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 73 - 76 SAMIPLE DESCRIPTION/ SAIMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY OFF-SITE DIRECT RADIATION (TLD) STATIONS OFT-I 4.5 miles NE. Pole #B294-44 TLD Quarterly Direct Radiation Quarterly OFT-2 3.6 miles, 1NE, Stevensville TLD Quarterly Direct Radiation Quarterly Substation OFT-3 5.1 mtles NE. Pole #B296-13 TLD Quarterly Direct Radiation Quarterly OFT4 4.1 miles, E. Pole #B350-72 TLD Quarterly Direct Radiation Quarterly OFT-5 4.2 miles ESE. Pole qB387-32 TLD Quarterly Direct Radiation Quarterly OF1-6 4.9 miles SE, Pole #1426-1 TLD Quarterly Direct Radiation Quarterly OFT-7 2.5 miles S. Bndz=an Substation TLD Quarterly Direct Radiation Quarterly OFT-3 4.0 miles S. Pole #B424-20 TLD Quarterly Direct Radiation Quarterly OFT-9 4.4 miles ESE. Pole #1369-214 TLD Quarterly Direct Radiation Quarterly OFT-10 3.3 miles S. Pole #B422-99 I TLD Quarterly Direct Radiation Ouarterly OFT-il 3.3 miles S. Pole #B423-12 I TLD Quarterly Direct Radiation Quarterly GROUNDWATER (WELL WATER) SAMPLE STATIONS W-l 1969 ft @ 110 from Plant Axis Groundwater Quarterly Gamma Isotopic I Quarterly Tnotiumn Quarterly W-2 2302 ft 9 63° from Plant Axis Groundwater Quarterly Ganmua Isotopic I Quarterly Tritturn - Quarterly W-3 3279 ft (D 1070 from Plant Axis Groundwater Quarterly Gamma Isotopic I Quarterly Tntium I Quarterly W-4 418 ft © 3010 from Plant Axis Groundwater Quarterly Gamma Isotopic j Quarterly W-Si 4ntium. I Quarterly W-5 404 ft @ 2900 from Plant Axis Groundwater Quarterly Gamma Isotopic I Quarterly Tnuum I Quarterly W-6 424 ft Q 2189 from Plant Aus Groundwater Quarterly Gamma Isotopic I Quarterly Trtiunm Quarterly W-97 1895 ft @ 1892 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarerly Tninum Quarterly W-3 1274 ft @ 541 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly

__ _ _ _ _ __ _Trinum

_ Quarterly W-9 1447 ft @ 221 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly

_______ Trntium Quarterly W-10 4216 ft @ 1290 from Plant Axis Groundwater Quarterly GaTna Isotopic Quarterly II Tntaum Quarterly W-11 32106 ft @>1530 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly I I Tntium Quarterly W-12 2631 ft @ 1624 from Plant Axis Groundwater Quarterly Garmna Isotopic Quarterly Tritnum Quarterly W-13 2152 ft c~ 1820 from Plant Axis Groundwater Quarterly Gammura Isotopic Quarterly I__ __I_ __ I_ Tntium IQuarterly W-14 1750 ft ~ 164- from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly J___I___I__ Tritium I Quarterly

Information l PIMP-6010-OSD-001 Rev. 17 Page 75 of 84 OFF-SITE DOSE CALCULATION MIANTUAL At h 319 lRadiological Envirornmental MonitornmgProga l Pages:

Sample Stations, Sample Types, Sample Frequencies 73 - 76 SURFACE WATER Sw-1 Condenser Circulaong Water l Surface Water l Daily 1 Gamma Isotopic Month. Comp.

Intake II _ _ Tntium Quart. Cormp SW'L-2 Plant Site Boundary - South Surface Water 1 Daily Gammia Isotopic Month. Comp

- 500 ft. south of Plant Tntium Quart. Ccmp Cenfterline SVL-; Plant Site Boundary - North Surface Water Daily Gamma Isotopic 2Month. Comp

- 500 ft. north of Plant Trmtum Quart Cormp.

Certerlme J I

SL-4 & 5 are dams collection psnts only not actual REMP samples GROUNDWATER (STEAM GENERATOR STORAGE FACIY) SAMPLE STATIONS SG- 0 8 mi. g 95 from Plant Axis Groundwater Quarterly I Gross Alpha Quarterly I Gross Beta Quarterly I Gamma Iso5tODC Quarterly SG-2 0.7 mi. Q 9V from PlantAxis Groundwater Quarterly Gross Aloha Quarterly Gross Beta Quarterly

_ _ __ _ _ Gamma Isotopic Quarterly SG-4 0.7 ML. 0 93' from Plant Axis Groundwater Quarterly Gross Alpha Quarterly Gross Beta Quarterly 1 Gamma Isotopic Quarterly SG-5 0 7 mn. @ 92° from Plant Axis Grourdwater Quarterly Gross Alpha Quarterly Gross Beta Quarterly Gamma Isotopic Quarterly INGESTION - MILK Indicator Farms' _

Milk 1Once every I-131 per sample l 15 days Gamma Isotopic per sample Milk Once every 1-131 per sample 15 days Gamma Isotopic per satnle I Milk Once every 1-131 per sample

_i S days Gamma Isotooic per sample j

Information P.P-6010-OSD-001 Rev. 17 Pa.e 76 of 84 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

ASample Stations, Sample Types, Sample Frequencies 73 - 76 SASMPLE DESCRON SAPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY INGESTION - ML Background Farms lvAighouse Farm 20 miles S. La Porte. IN MllIk Once every 1-131 per sample

_ 15 days I Gamma Isotopic per sample Wyant Farm 20.7 miles E. Dowagiac Milk Once every 1-131 per sample

_ 15 days Gamma Isotopic per samole INGESTION - FISH ONS-N [l 0.3 mile N. Lake Michigan Fish 2/year I Gamma Isotopic per sample ONS-S 0 4 mile S. Lake Michigan Fish 2/year Gamma Isotopic per sample OFS-N 3.5 mile N. Lake Michigan Fish 2Jyear Gamma Isotopic per sample OFS-S 5.0 mile S. Lake MSichigan Fish 2/year GaIma Isotopic 4per sample INGESTION - BROADLEAF IN LIEU OF MILK l 3 indicator samples of broad leaf veget3ticn Broadleaf .Monthly Gamma Isotoptc Monthly collected at different locations, within eight vegetanon when available I131 when available miles of the plant in the highest annual average D/Q land sector.

I background sample of similar vegetation Broadleaf Monthly Gamma IstopIc Monthly grown 15-25 miles distant in one of vegetation when avadlable 1131 when available the less prevalent wind directions.

Collect composite samples of Drinksng and Surface water at least daily. Analyze particulate sample filters for gross beta activity 24 or more hours following filter removal. This will allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 tunes the yearly mean of control samples for any medium, perform gamma isotopIc analysis on the iidividoal samples.

If at least three indicator milk samples and one background milk sample cannot be obtained, three 'ndicator broad leaf samples will be collected at different locations, within eight miles of the plant. in the land iector with the highest DlQ (refers to the highest annual average DiQ). Also, one background broad leaf sample will be collected 15 to 25 miles from the plant in one of the less prevalent DIQ land sectors.

'he three milk indicator farms will be determined by the Annual Land Use C-nsus and those that are willing to participate. IF it is determined that the milk animals are fed stored feed, THEN monthly sampling is appropriate for that time period.

Information PMP-6010-OSD-001 Rev. 17 Page 77 of 84 OFF-SITE DOSE CALCULATION MANUAL .20 lvMaximum Values for Lower Limits of DetectionsA - REMP 77ages Radionuclides Food Product Water Milk Air Filter Fish Sediment pCi/klg, wet pCill pCijl pCi/m 3 pCl/kg, wvet pCi/kg, dry Gross Beta 4* 0.01 H1-3 2000 Ba-140 60 60 La-140 15 15 Cs-134 60 15 15 j 0.06 130 150 Cs-137 60 18 18 0.06 150 180 Zr-95 30 Nb-95 15 Mn-54 15 130 Fe-59 30 260 Zn-65 30 260 Co-58 15 130 Co60 15 j 130 1-131 60 1 j 1 0.07 _

This Data is directly from our plant-specific Technical Specification.

LLD for dnrnking water

Information PNMP-6010-OSD-001 Rev. 17 Page 78 of 84 OFF-SITE DOSE CALCULATION MANTUAL .20 Maximum Values for Lower Limits of DetectionsA>B - REMP Paes NOTES A. Thie Lower Limit of Detection (LLD) is defined as the smallest concentration of radioactive material in a sample that will be detected with 95 % probability and 3a5% probability of filisely concluding that a blank observation represents a 'real' signal.

For a particular measurement system (Which may include radiochemical separation), the LLD is given by the equation:

LLD = 4.66*

E*V7* 2.22*-)Y *e Where LLD is the a priori lower limit of detection as defined above (as pCi per unit mass or volume). Perform analysis in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering radionuclides, or other uncontrollable circumstances nay render these LLDs unachievable.

S is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per mninute).

E Is the counting efficiency of the detection equipment as counts per transformation (that is, disintegration)

V is the sample size in appropriate mass or volume units 2.22 is the conversion factor from picocunes (pCi) to transformations (disintegrauons) per mnanute Y is the fractional radiochemical yield as appropriate 1 is the radioactive decay constant for the particular radionuclide At is the elapsed tuie between the midpoint of sample collection (or end of sample collection period) and time of counting.

B identify and report other peaks which are measurable and identifiable, together with the radionuclides listed in Attachment 3.20, Maximum Values for Lower Limits ofDetectionsAB - REMP.

A 2.71 value may be added to the equation to provide correction for deviations in the Poisson distr.buton at low count rates, that is, 2.71 - 4.66 x S.

Information l PIP-6010-OSD-001 I Rev. 17 I Page 79 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.21 Reporting Levels for Radioactivity Concentrations Page:

in Environmental Samples 79 Radionuclides Food Product Water Mil Air Filter Fish 3

pCilkg, wet pCi/I pCi/l pci/nm pCL/kg, wet H-3 20000 Ba-140 l 200 300 La-140 200 300 Cs-134 1000 30 60 10 1000 Cs-137 2000 50 70 20 2000 Zr-95 400 Nb-95 400 .

Mn-54 1000 30000 Fe-59 400 i 10000 Zn-65 300 l 20000 Co-58 1000 30000 Co-60 1 300 _ 10000 !

I11100 l 2 l 3 0.90 ll

Information PMP-6010-OSD-001 Rev. 17 l Page 80 of 84 OFF-SITE DOSE CALCULATION MANUAL .22 On-Site Monitoring Location - RvIP Page:

REMP Monitotrirq LocatlonS LEGEND ONS ONS-6: Air Sampling Staton T T-12. TLD Sampling Station W W RMP TiS Groundwater Wells SG-1. SG-2. SG-4, SG-5: REMP Non TVS Groundwater Wells SWL-1. 2. 3: Surface Water Samphng Stations SL-2, SL Sediment Sampling Stauons

Information l PMP-6010-OSD-001 I Rev. 17 1 Page 81 of 84 OFF-SITE DOSE CALCULATION MANUAL -

n 33 .t Page: .23 l Off-Site Monitoring Locations -RENT 81I (I I OrFF=E RElW WMONUlORING WOCAflONS

.. x .. .

, 1.

o-r

-- 1-

... '.II TL'D OFr-M' W~t~ ZAig L~onB.XNUCLZA; 1V"uTMss tla eSTATION

Information I PiIP-6010-OSD-001 I Rev. 17 l Page 82 of 84 OFF-SITE DOSE CALCULATION MAN-UAL Attacment3.24 Safety Evaluation By The Office Of Nuclear Pages:

Reactor Regulation 32 - 84 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO DISPOSAL OF SLIGHTLY CONTAMINATED SLUDGE INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316 [Ref. 5.2.1r]

(This is a 10 CFR 50.75 (g) item)

1. INTRODUCTION By letters dated October 9, 1991, October 23, 1991, September 3, 1993, and September 29, 1993, Indtana Michigan Power Company (I&Mt) requested approval pursuant to 10 CFR 20.2002 for the on-site disposal of licensed maternal not previously considered in the Donald C. Cook Nuclear Plant Final Environmental Statement dated August 1973. Specifically, this request addresses actions taken in 1982 in which approximately 942 cubic meters of slightly contaminated sludge were removed from the turbine room sump absorption pond and pumped to the upper parking lot located within the exclusion area of the Donald C. Cook Nuclear Plant. The contaminated sludge was spread over an area of approximately 4.7 acres. The sludge contmined a total radionuclide inventory of 8.89 minllicuries (mCi) of Cesium-137, Cesium-136, Cesium-134, Cobalt-60 and Iodine-131.

In its submittal, the licensee addressed specific information requested in accordance with 10 CFR 20.2002(3), provided a detailed description of the licensed material, thoroughly analyzed and evaluated information pertinent to the impacts on the environment of the proposed disposal of licensed material, and committed to follow specific procedures to mnnmtLze the risk of unexpected exposures.

2. DESCRIPTION OF WASTE The turbine room sump absorption pond is a collection place for water released from the plant's turbine room sunp. The contamination was caused by a primary-to-secondary steam generator leak that entered the pond from the turbine building sump, a recognized release pathway. Sludge, consisting mainly of leaves and roots mixed with sand, built up in the pond.

As a result, the licensee dredged the pond in 1982. The radioactive sludge removed by the dredging activities was pumped to a containment area located within the exclusion area. The total volume of 942 cubic meters of the radioactive sludge that was dredged from the bottom of the turbine room absorption pond was subsequently spread and made into a graveled road over the upper parking lot area of approximately 4 7 acres.

The principal radionuclides identified in the dredged material are listed below.

TABLE 1 NUCLIDE ACTIVITY (mCi) ACTIVITY (mCi)

(half-life) 1982 1991 36 Cs (13.2 d) 0.03 NA*

134CS (2-1 y) 2.34 0.18 37

' Cs (30.2 y) 5.59 4.57 6OCO (5.6 y) 0.90 0.27 1I (8.04 d) 0.03 NA*

TOTAL: 8.89 5.02

NA: not applicable due to decay

Information PMP-6010-OSD-001 Rev. 17 Page 83 of 84 OFF-SITE DOSE CALCULATION MANUAL Safety Evaluation By The Office Of Nuclear Pages:

A Reactor Regulation 1 82 - 84

3. RADIOLOGICAL IMPACTS The licensee in 1982 evaluated the following potential exposure pathways to members of the general public from the radionuclides in the sludge:

(1) external exposure caused by groundshine from the disposal site; (2) internal exposure caused by inhalation of re suspended radionuchde,

-AND-(3) internal exposure from ingesting ground water.

The staff has reviewed the licensee's calculationa] methods and assumpnons and finds that they are consistent with NUREG-1101 "Onsite Disposal of Radioactive Waste," Volumes 1 and 2, November 1986 and February 1987, respectively. The staff finds the assessment methodology acceptable. Table 2 lIts the doses calculated by the licensee for the maximally exposed member of the public based on a total activity of 8.89 mCi disposed in that year.

TABLE 2 Pathway Whole Body Dose Received by Maximally Exposed Individual (mremlyear)

Groundshine 0.94 Inhalation 0.94 Groundwater Ingestion 0.73 Total 2.61 On July 5, 1991, the licensee re-sampled the onsite disposal area to assure that no significant impacts and adverse effects had occurred. A counting procedure based on the appropriate environmental low-level doses was used by the licensee; however, no activity was detected during the re-sampling'. This is consistent with the original activity of the material and the decay time. The 1991 re-sampling process used by the licensee confirms that the environmental impact of the 1982 disposal was very small. The staff finds the licensee's methodology acceptable.

4. ENVIRONMENTAL FINDING AND CONCLUSION The staff has evaluated the environmental impact of the proposal to leave in place approximately 942 cubic meters of shghtly contaminated sludge underneath the upper parking lot on the Donald C. Cook Nuclear Plant site.

In 1982, the licensee evaluated the potential exposure to members of the general public from the radionuclides in the sludge and calculated the potential dose to the maximally exposed member of the public, based on a total activity of 8.89 mCi disposed in that year, to be 2.61 mremnlyr. .The staff has reviewed the licensee's calculational methods and assumipuons and found that they are consistent with NUREG-I 101, Onsite Disposal of Radioactive Waste, Volumes I and 2, November 1986 and February 1987, respectively. The staff finds the assessment methodology acceptable. For comparison, the radiation from the naturally occurring radionuclides in soils and rocks plus cosmic radiation gives a person in Michigan a whole-body dose rate of about 89 mrem. per year outdoors. Subsequent licensee sampling in 1991 identified no detectable activity. The staff evaluated the licensee's sampling and analysis methodology and finds it acceptable. The results, of the 1,991 re-sampling by the licensee, confirm that the environmental impact of the 1982 disposal was very small.

Based on the above the staff finds that the potential environmental impacts of leaving the contaminated sludge in place are insignmficant. With regard to the non-radiological impacts, the staff has determined that leaving the soil in place represents the least impact to the environment.

Information i PMP-6010-OSD-001 Rev. 17 Page 84 of 84 OFF-SITE DOSE CALCULATION MANUAL i Attachment A 33.24 Safety Evaluation By Regulation Reactor The Office Of Nuclear Pages-82 - 84

5. CONCLUSION Based on the staff's review of the licensees discussion, the staff finds the licensee's proposal to retain the material in its present location as documented in this Safety Evaluation acceptable. Also, this Safety Evaluation shall be permanently incorporated as an appendix to the licensee's Offsite Dose Calculation Manual (ODCMI), and any future modifications shall be reported to NRC in accordance with the applicable ODCM change protocol.

I&M letter from E. E. Fitzpatnck to the NRC Document Control Desk, September 29, 1993 Therefore, the licensee's proposal to consider the slightly contaminated sludge disposed by retention in place in the manner described in the Donald C. Cook Nuclear Plant submittals date October 9, 1991, October 23, 1991, September 3, 1993, and September 29, 1993, is acceptable.

The guidelines used by the NRC staff for onsite disposal of licensed maternal and the staff s evaluation of how each guideline has been satisfied are given in Table 3.

Pursuant to 10 CFR 51.32. the Commission has determined that granting of this approval will have no significant impact on the environment (October 31, 1994, 59 FR 54477).

Prinmpal Contributor J. Minns Date: November 10, 1994 TABLE 3 20.20021 GUIDELINE FOR ONSM STAFF'S EVALUATION DISPOSALV

1. The radioactive material should be disposed of in such 1. Due to the nature of the disposed material, recycling to the a manner that it is unlikely that the material would be general public is not considered lkely recycled.

2. Doses to the total body and any body organ of a 2. This guideline was addressed in Table 2. Although the maximally exposed individuals (a member of the 2.61 mrem/yr is greater than staffs guidelines, the staff general public or a non-occupationally exposed worker) finds it acceptable due to 9 yrs decay following analysis and from the probable pathways of exposure to the disposed the expected lack of activity detected in the 1991 survey.

material should be less than I mrem/year.

3. Doses to the total body and any body organ of an 3. Because the material will be land-spread, the staff considers inadvertent intruder from the probable pathways of the maximally exposed individual scenario to also address exposure should be less than S inrem/year. the intruder scenario.

4. Doses to the total body and any body organ of an 4. Even if recycling were to occur after release from regulatory individual from assumed recycling of the disposed control, the dose to a maximally exposed member of the material at the time the disposal site is released from public is not expected to exceed 1 rirem/year, based on regulatory control from all likely pathways of exposure exposure scenarios considered in this analysis.

should be less than I mrem.

2 E. F. Branagan, Jr. and F. J. Cooigel, 'Disposal of Contaminated Radioactive Wastes from Nuclear Power Plants,"

presented at the Health Physics Society's Mid-Year Symposium on Health Physics Consideration in Decontamination/Deconmnissioning. Knoxville, Tennessee, February 1986, (CONF-860203).

REVIEW AND) APPROVAL TRACKING FORM 1amddu riUmun' A--';

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Number: PMP 6010 OSD.001 Rev. 16 Change: C2

Title:

Off-site Dose Calculation Manual a f- ; C4I 0 Oi,,y *edi k L-'

M Correztion (Full Procedure) n Change (Full Procedure) with Review of Change Only Z Correction (Page Substitution) [] Change (Page Substitution) with Review of Change Only n Cancellation n New Procedure or Change with Full Review n Superseded (list superseding procedures):

, -A!96dittid..Cnfii6riiiioWlhtiact'A `sii eiW:-- -- '

Change Dnver/CDI Tracking No(s).:

elq-A Cross-Discipline Reviews:

n Chemistry El Training E ALARA El Performance Assurance a Maintenance O Work Control 5 Bus. Services Proc Grp F1 Reactivity Mgmt Team D NDM El El Component Engineering Ql SPS (Safety & Health)

O Operations O _ El Design Engineering E Surveillance Section D PA/PV E_ _ E Emerg Oper Proc Grp El System Engineering E Reg Affairs E _ El Environmental El n RP

'1 None Reauired F] ISI/lST Coordinator F1 None Required E Cognizant Org Review: N/A Date: I Technical Review: A/#. I .Tj ,, Date: /z 2-/ oi

,tnuinrlineg;- :,:g~,g nsrzM5'XJgwowlti"t "NA,>-i#

E Ops Mgr Concurrence: N/A Date: _I /_

EJ Owner Concurrence: N/A Date: ___//

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Updated Revision Sunmunary attached? Z Yes 10 CFR 50.59 Requirements complete? Tracking No.:_ E] Yes 3 N/A Implementation Plan developed? \ (Ref. Step 3.4 18) E Yes E)N/A Package Complete: Date: / l /l o -

oprox 'ru' ., i PORC Review Required: El Yes E No Mtg. No.:

Administrative Hold Status: El Re CR No.:_

Approval Authority Review/Approval: Date: I 13 /c_Ž Expiration Date/Ending Activitv Date: I Io ge

'4" Periodic Review conducted? Z No Commitment Database Updated? m] NrA NDM notified of new records or changes to records that could affect reco R N/A -

c: Offirce Information For Form Tracking Only - Not Part of Form IbThis form ;s derived trom the information in PMP-2010-PRC-002, 9-: Procedure Correction, Chan-e, and Review, Rev. 9, Data Sheet 1, Z Review and Approval Trncking Form. Page _ ot -

REVISION SUNMMARY Number: PMfP 6010 OSD.001 Revision. 16 Change: C' Title. Off-site Dose Calculation Manual Mvlarginal markings were used.

Pages 3 of 84, Rev 16 Cl; 50 of 84, Rev 16, 52 of 84, Rev 16 and 53 of 84, Rev 16 should be replaced by pages 3 of 84, Rev 16, C2; 50 of 84, Rev 16 C2; 52 of 84, Rev 16 C2 and page 53 of 84, Rev 16, C2.

Section or Step Change/Reason For Change Correction i- Criteria Attachment 3.4, Change: Corrected action for Containment particulate m Item 3b sampler filter. Added Action 10 to provide correct action.

Reason: Clarification of appropriate actions to take in accordance with Technical Specification Table 3.3-6 and Action for TS 3.4.6.1.

Attachment 3.5, Change: Removed the link of the Containment Monitors m Item 3 to annunciate in the Control Room to be operable.

Reason: The containment radiation monitors are in Technical Specification Table 4.3-3. The definition of Channel Functional Test in Technical Specifications requires verification of operability including alarm and/or trip function.

The containment radiation monitors should not be linked to Control Room annunciation like the effluent radiation monitors since they were not included in the Radioactive Effluent Technical Specifications (RETS). The RETS have all be moved to the ODCM through GL 89-01 implementation, but the containment radiation monitors remain in Technical Specifications.

Change:

Reason:

Change:

Reason: l I - Office O. nformationFor Form Trackdng Only zNot Partof Forni This is a free-torm as called out in PMP-2010-PRC-002, Procedure Correction, Change. and Rzview. Rev. 9. Page 2 of -2l

RK PMyP-6010.OSD.001 Rev. 16 Page l of 84 OFF-SITE DOSE CALCULATION MANUAL Information Effective Date: 4/1A/co, Doug Foster John Carlson Environmental Writer Owner Cogntzant Organization TABLE; OF CONTENTS 1 PURPOSE AND SCOPE .................. 4 2 DEFINITIONS AND ABBREVIATIONS ....... ........... .................................... 4 3 DETAILS .. 4 3.1 Calculation of Off-Site Doses .. 4 3.1.1 Gaseous Effluent-Releases .4 3.1.2 Liquid Effluent Releases .10 3.2 Limits of Operation and Surveillances of the Effluent Release Points ... 13 3 .2.1 Radioactive Liquid Effluent Monitoring Instrumentation . .13 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation . .14 3.2.3 Liquid Effluents .. 15

a. ConcentrationExcluding Releases via the Turbine Room Sump (TRS) Discharge .15

b. Concentration of Releases from the TRS Discharge . 15

c. Dose .16

d. Liquid Radwaste Treatment System .16 3.2.4 Gaseous Effluents .. 19

a. Dose Rate .19

b. Dose - Noble Gases .19

c. Dose - Iodine-l 31, Iodine-133, Tritium, and Radioactive Material in Particulate Form ............................. 19

d. Gaseous Radwaste Treatment ............................. 20 3.2.5 Radioactive Effluents - Total Dose ............................. 22 33 Calculation of Alarm/Trip Setpoints ........................ 23 3.3.1 Liquid Monitors ........................................ 24

a. Liquid Batch Monitor SetpointMethodology ...................................... 24

b. Liquid Continuous Monitor Setpoint Methodology ............................. 25 3.3.2 Gaseous Monitors ...... .................................. 27

a. Plant Unit Vent ....................................... 27

b. Waste Gas Storage Tanks ....................................... 29

c. Containment Purge and Exhaust System ....................................... 30

- d. Steam Jet Air Ejector System (SJAE) .......................... 1.............

1

e. Gland Seal Condenser Exhaust .............. ......................... 31 3.4 Radioactive Effluents Total Dose ......................................... 32 3.5 Radiological Environmental Monitoring Program (REMP) ..................................... 32 3.5.1 Purpose of the REMP............................................. ....................................

3 5.2 Conduct of the REIlAP .............. 33

KICIRICX P1NFP-6010.0SD.00O1 Rev. 16 Pagre 2 of go4 OFF-SITE DOSE CALCULATION hSlA Information l l Effective Date:_o -II Doua Foster John Carlson Environmental Writer Owner Cognizant Organization 3.5.3 Annual Land Use Census ..................................................... 35 3.5.4 Interlaboratory Comparison Program ..................................................... 35 3.6 Steam Generator Storage Facility Groundwater Monitoring Program. . 36 3.6.1 Purpose of the Steam Generator Storage Facility Groundwater Radiological IMonitoring Program ..................................................... 36 3.6.2 Conduct of the Steam Generator Storage Facility Groundwater Radiological Monitoring Program ............................................... ...... 36 3.7 Meteorological Model.................................................................3........................... 36 3.8 Reporting Requirements ................ 36 3.8.1 Annual Radiological Environmental Operating Report (AREOR) ............. 36 3.8.2 Annual Radiological Effluent Release Report (ARERR) .......... ................. 37 3.9 10 CFR 50.75 (g) Implementation ........... ......................................... 38 3.10 Reporting/ManagementReview .................................................... 39 4 FINAL CONDITIONS ................ 39 5 REFERENCES...................................................................................................................39 SUPPLEMENTS .1 Dose Factors for Various Pathways .................................. Pages 42 - 45 .2 Radioactive Liquid Effluent Monitoring Instruments .............. Pages 46 - 47 .3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements .......................................... Pages 48 - 49 .4 Radioactive Gaseous Effluent Monitorng Instrumentation . Pages 50 - 52 .5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ........................................... Pages 53 - 54 .6 Radioactive Liquid Waste Sampling and Analysis Program ..... Pages 55 - 56 .7 Radioactive Gaseous Waste Sampling and Analysis Program ... Pages 57 - 58 .8 IMultiple Release Point Factors for Release Points ........................ Page 59 .9 Liquid Effluent Release Systems . ...................... .. Page 60

. I0 Plant Liquid Effluent Parameters ......................... .................... Page 61 .11 Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors ........................... Page 62 .12 Counting Efficiency Curves for R-19, and R-24 .......... ................ Pages 63 - 64 65 .13 Counting Efficiency Curve for R-20, and R-28 .................................... Page .14 Gaseous Effluent Release Systems ............................................. Page 66 .15 Plant Gaseous Effluent Parameters .. ........................................... Page 67 69 .16 10 Year Average of 1989-1998 Data ............................................ Pages 68 - .17 Annual Evaluation of z/Q and D/Q Values For All Sectors ................. Page 70 Dose Factors.................................................................................. Pages - 72 71 .18 .19 Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies ............................... Pages 73 - 76

- 78 Attachment 3.20 Maximum Values for Lower Limits of DetectionsA B - REM? ..... Pages 77 Attachment 3.21 Reporting Levels for Radioactivity Concentrations in Environmental Samples ............................................. Page 79 Attachment 3.22 On-Site Monitoring Location - REMP ............................................. Page 80

.Page 81 Attachment 3.23 Off-Site Monitorng Locations - REWT Attachment 3.24 Safety Evaluation By The Office Of Nuclear Reactor Regulation....................................................................................... Pages 82 - 84 Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation.. Pages 50 & 52 C2 Attachment 3.5 Radioactive Gaseous Effluent Monitoring Instrunentation Surveillance Requirements Page 53 C2 C2 affects pages 3, 50, 52 and 53 of 84

Information PNIP-6010.OSD.001 Rev. 16 l Page 4 of 834 OFF-SITE DOSE CALCULATION MANUAL 1 PURPOSE AND SCOPE NOTE: This is an Administrative procedure and only the appropriate sections need be performed per P.MP 2010 PRC.003, step 3.27.

The Off-Site Dose Calculation Manual (ODCM1) is the top tier document for the Radiological Environmental Monitoring Program (REMP), the Radioactive Effluent Controls Program, (RECP), contains criteria pertaining to the previous Radiological Effluent Technical Specifications (RETS) as defined in NUREG-0472, and fully implements the requirements of Technical Specification 6.8.4.

The ODCMi contains the methodology and parameters to be used in the calculation of off-site doses due to radioactive liquid and gaseous effluents and in the calculation-of liquid and gaseous monitoring instrumentation alann/trip setpoints.

The ODCM provides flow diagrams detailing the treatment path and the major components of the radioactive liquid and gaseous waste management systems.

The ODCMI presents maps of the sample locations and the meteorological model used to estimate the atmospheric dispersion and deposition parameters.

The ODCM specifically addresses the design characteristics of the Donald C.

Cook Nuclear Plant based on the flow diagrams contained on the "OP Drawings" and plant "System Description" documents.

2 DEFINITIONS AND ABBREVIATIONS Term: Mleaning:

S or shifily At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D or daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W or weekly At least once per 7 days M or monthly j At least once per 31 days Q or quarterly At least once per 92 days SA or semi-annually At least once per 184 days R At least once per 549 days.

S/U Prior to each reactor startup P Completed pnor to each release Sampling evolution Process of changing filters or obtaining grab samples 3 DETAILS 3.1 Calculation of Off-Site Doses 3.1.1 Gaseous Effluent Releases

Information PNIP-6010.OSD.001 I Rev. 16 l Page 5 of 84 OFF-SITE DOSE CALCULATION MANUAL

a. The computer program MIDAS (Meteorological Information and Dose Assessment System) performs the calculation of doses from effluent releases. The site-specific parameters associated with MIDAS reside in the following subprograms:

MIDER

MIDEX

MIDEL

MIDEG

MIDEN

b. The subprogram used to enter and edit gaseous release data is called MID1EQ (EQ). The data entered in EQ can be used to calculate the accumulation of dose to individual land based receptors based on hourly meteorology and release data. The air dose from this data is calculated via the XDAIR subprogram in MIDAS. It computes air dose results for use in Reg. Guide 121 reports and 10 CFR 50 Appendix I calculations based on routine releases.

c. The formula used for the calculation of the air dose is generated from site specific parameters and Reg. Guide 1.109 (Eq 7):

DADo air= X Zf(A 1 or N)* Q* 3. 7E-8j Q

Where; D.,, Do air = the gamma or beta air dose in mradlyr to an individual receptor xIQ = the annual average or real time atmospheric dispersion factor over land, seclM 3 from Attachment 3.16, 10 Year Average of 1989-

- 1998 Data Ml = the gamma air dose factor, mrad mr / yr giCi, from Attachment 3.18, Dose Factors Ni = the beta air dose factor, mrad m3 / yr JLCi, from Attachment 3.18, Dose Factors QL = the release rate of radionuclide, "i", in 1iCi/yr.

3.17E-8 = number of years in a second Cyears/second).

d. The value for the ground average X/ Q for each sector is calculated using equations shown below. Formula used for the calculation is generated from parameters contained in MIDAS Technical Manual, XDCALC (Eq 2).

Information I PNIP-6010.OSD.OO1 I Rev. 16 L Page 6 of 84 OFF-SITE DOSE CALCULATION MALNUAL

- 2.03 Urn, -

Where; minimum of , or2E.gH=

or x = distance downwind of the source, meters. This information is found in parameter 5 of TMIDEX. _

= wind speed for ground release, (meters/second) c , = vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 1.111 Fig.1)

H, = building height (meters) from parameter 28 of MIDER.

(C ontainrment Building = 49.4 meters)

Tf = terrain factor (= I for Cook Nuclear Plant) because we consider all our releases to be ground level (see parameter 5 in MIDEX).

2.03 = - 0.393 radians(22 50)

e. The dose due to gaseous releases, other than the air dose, is calculated by the MIDAS subprogram GASPRO. GASPRO computes the accumulation of dose to individual receptors based on hourly meteorology and release data. Calculations consider the effect of each important radionuclide for each pathway, organ, age group, distance and direction.

f. Calculations are based on the environmental pathways-to-man models in Reg. Guide 1.109. The program considers 7 pathways, 3 organs, and 4 age groups in 16 direction sectors. The distances used are taken from the MIDEG file.

Information I PiNI-P-6010.OSD.001 I Rev. 16 l Page 7 of 84 OFF-SITE DOSE CALCULATION MANUAL

g. The formulas used for the following calculations are generated from site specific parameters and Reg. Guide 1.109:

1. Total Body Plume Pathway (Eq 10)

Dose (rnrem/year) = 3.1 7E -8

E (Q,

XIQ

Sf

DFB,)

Where; Sf = shielding factor that accounts for the dose reduction due to shielding provided by residential structures during occupancy (maximum exposed individual = 0.7 per Table E-15 of Reg. Guide 1.109)

DFB, = the whole body dose factor from Table B-I of Reg. Guide 1.109, mrem -m 3 per pCi - yr. See Attachment 3.1 8, Dose Factors.

Q= the release rate of radionuclide "i", in pCi/yr

2. Skin Plume Pathway (Eq I 1)

Dose (mrem/yr)=3 17E-8* S1 *. T L[Q,*1.11*DF)+Z(Q,

DFS,)]

Where; 1.11 = conversion factor, tissue to air, mrem/mrad DF 7= the gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i", in mrad m3/pCi yr from Table B-i, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

DFS, = the beta skin dose factor for a semi-infinite cloud of radionuclide "i", in mrem m 3 /tCi yr from Table B-I, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

3. Radionuclide and Radioactive Particulate Doses (Eq 13 & 14d)

The dose, Dip in mremlyr, to an individual froom radionuclides, other than noble gases, with half-lives greater than eight days in gaseous effluents released to unrestricted areas will be determined as follows:

Dir (nirem/year)= 3.17E - 8 * .( R.

W

Q.;

Information PNIP-6010.OSD.001 I Rev. 16 l Page 8 of 84 OFF-SITE DOSE CALCULATION ALNUAL Where; R = the most restrictive dose factor for each identified radionuclide "i', in rrC mrem sec / yr pCi (for food and ground pathways) or mremnm3 / yr llCi (for inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles of the site, use the values of R, for these real pathways, otherwise use pathways distance of five miles. See Attachment 3.1, Dose Factors for Various Pathways, for the maximum R1 values for the most controlling age group for selected radionuclides.

Ri values were generated by computer code PARTS, see NUREG-0 133, Appendix D.

W = the annual average or real time atmospheric dispersion parameters for estimating doses to an individual at the worst case location, and where W is further defined as:

W. = SC/ Q for the inhalation pathway, in sec/M3

-OR-Wfg = D / Q for the food and ground pathways in /rM2 Qic = the release rate of those radioiodines,radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than eight days, in [LCi/yr

h. This calculation is made for each pathway. The maximum computed dose at any receptor for each pathway is selected. These are summed together to get the dose to compare to the limits. Only the maximum of the cow milk or goat milk pathway (not both) is included in the total.

i. In addition to the above routines, the QUICKG routine of the MIDAS system is used to provide data used in the monthly reports due to its ability to use annual average meteorological data rather than real time data, thus shortening the run time involved.

j. Steam Generator Blowdown System (Start Up Flash Tank Vent)

1. The amount of radioiodine and other radionuclides that are released via the start up flash tank and its vent are calculated through acnial sample results while the start up flash tank is in service.

2 The following calculation is performed to determine the amount of curies released through this pathway. (Plant established formula.)

Information PMP-6010.OSD.001 Rev. 16 Page 9 of 84 OFF-SITE DOSE CALCULATION MANUAL Curies = ti GPM

time on flash tank (min) *3. 785E - 3 ml Where; 3.785E-3 = conversion factor, ml Cif4+/-Ci gal.

3. The flow rate is determined from the blowdown valve position and the time on the startup tank. ChemistryDepartmentperformsthe sampling and analysis of the samples.

4. This data is provided to the MIDAS computer and dose calculations (liquid and gas) are performed to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. MIDAS uses the formulas given in step 3.1.2, Liquid Effluent Releases, to calculate doses to members of the public.

NOTE: This section provides the minimum requirements to be followed at Donald C.

Cook Nuclear Plant. This would be used if actual sample data was not available each time the start up flash tank was in service.

5. The radioiodine release rate must be determined in accordance with the following equation every 31 day period whenever the specific activity of the secondary coolant system is greater than 0.01 PiCi/g dose equivalent 1-13 1.

6. IF the specific activity of the secondary coolant system is less than 0.01 pCi/g dose equivalentI-131, THEN the release rate must be determined once every six months. Use the following plant established equation:

QY = C!

JPF* Rvb Where; QY = the release rate of I- 131 from the steam generator flash tank vent, in jiCi/sec Ci = the concentration (pCi/cc) of I-131 in the secondary coolant averaged over a period not exceeding seven days IPF = the iodine partition factor for the Start Up Flash Tank, 0.05, in accordance with NUREG-0017 Rsgb the steam generator blowdown rate to the start up flash tank, in cc/sec

7. Use the calculated release rate in monthly dose projections until the next determination to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. Report the release rate calculations in the Annual Radioactive Effluent Release Report.

Information I PNIP4010.OSD.001 I Rev. 16 1 Page 10 of 84 OFF-SITE DOSE CALCULATION MANUAL

8. Steam Generators are sparged, sampled, and drained as batches early in outages to facilitate cooldown for entry into the steam generator.

This is repeated prior to startup to improve steam generator chemistry for the startup.

3.1.2 Liquid Effluent Releases

a. The calculation of doses from liquid effluent releases is also performed by the MIIDAS program. The subprogram used to enter and edit liquid release data is called MD 1EB (EB).

b. To calculate the individual dose (mrem), the program DS ILI (LD) is used.

It computes the individual dose for up to 5 receptors for 14 liquid pathways due to release of radioactive liquid effluents. The pathways can be selected using the MIDEL program and changing the values in parameter l. D.C. CookNuclear Plant uses 3 pathways: potable water, shoreline, and aquatic foods (fresh water sport fishing).

c. The equations used are generated from site specific data and Reg. Guide 1.109. They are as follows:

1. Potable Water (Eq 1)

R~91=1JOO* U39 Q 1 f' F* 2.2 - 3E Q Where; R~vj = the total annual dose to organ "j" to individuals of age groups "a" from all of the nuclides "i" in pathway "p", in mremiyear 1100 = conversion factor, yr ft3 pCi / Ci sec L Uap = a usage factor that specifies the exposure time or intake rate for an individual of age group 'a" associated with pathway "p". Given in #29-S4 of parameter 4 in MIDEL and Reg. Guide I.l109 Table E-5. See Attachment 3.1, Dose Factors for Various Pathways.

Ntp = the dilution factor at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food). Given in parameter 5 of MIDEL as 2.6.

F the circulation water system water flow rate, in gpm, is used for evaluating dose via these pathways as dilution flow 2.23E-3 = conversion factor, ft3 min/ sec gal Q =the release rate of nuclide "i"for the time period of the run input via MIDEB, Curies/year

Information I PMP-6010.OSD.001 I Rev. 16 1 Page 11 of 84 OFF-SITE DOSE CALCULATION MANUAL

_~~

Daipj =the dose factor, specific to a given age group "a",

radionuclide"i", pathway "p", and organ "j", which can be used to calculate the radiation dose from an intake of a radionuclide, in mrem/pCi. These values are taken from tables E-1 I through E-14 of Reg. Guide 1.109 and are located within the MIDAS code.

X, = the radioactive decay constant for radionuclide "i", in hours-'

1p = the average transit time required for nuclides to reach the point of exposure, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This allows for nuclide transport through the water purification plant and the water distribution system. For internal dose, tp is the total elapsed time between release of the nuclides and ingestion of food or water, in hours. Given as #25 of parameter4 in MIDEL. (tp = 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months )

2. Aquatic Foods (Eq 2)

R,,7p=lO0* U." Q,

B,*D,,

Mp*F*2.23E3 - B3p Daipy er' Where, BP= the equilibrium bioaccumulation factor for nuclide "i" in pathway "", expressed as pCi L / kg pCi. The factors are located within the MIDAS code and are taken from Table A-I of Reg. Guide 1.109. See Attachment 3.1, Dose Factors for Various Pathways.

t = the average transit time required for nuclides to reach the point of exposure, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This allows for decay during transit through the food chain, as well as during food preparation. Given as #26 of parameter 4 in MIDEL. (tp

= 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

Mp= the dilution factor at the point of exposure, 1.0 for Aquatic Foods. Given in parameter 5 of MIDEL as 1.0.

Information I PMIP-6010.OSD.O01 Rev. 16 l Page 12 of 34 OFF-SITE DOSE CALCULATION MANUAL

3. Shoreline Deposits (Eq 3)

II 0lOO MWF*a.23E-3 *10 *, Dupj [e i*[1-exifb}

Where; W = the shoreline width factor. Given as an input of 0.3 when running the program, based on Table A-2 in Reg. Guide 1.109.

T, = the radioactive half-life of the nuclide, "i", in days Dajpj= the dose factor for standing on contaminated ground, in mrem m2 / hr pCi. The values are taken from table E-6 of Reg. Guide 1.109 and are located within the MIDAS code.

See Attachment 3.1, Dose Factors for Various Pathways.

tb = the period of time for which sediment or soil is exposed to the contaminated water, 1.31 E+5 hours. Given in MIDEL as item 6 of parameter 4.

tp = the average transit time required for nuclides to reach the point of exposure, 0 hours0 days 0 hours 0 weeks 0 months . Given as #28 of parameter 4 in MIDEL.

110,000 =conversion factor yr fit pCi / Ci sec M 2 day, this accounts for proportionality constant in the sediment radioactivity model MP = the dilution factor at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food). Given in parameter 5 of M1IDEL as 2.6.

d. The MIDAS program uses the following plant specific parameters, which are entered by the operator.

1. Irrigationrate = 0

2. Fraction of time on pasture = 0

3. Fraction of feed on pasture = 0

4. Shore width factor = 0.3 (from Reg. Guide 1. 109, Table A-2)

e. The results of DS I LI are printed in LDRPT (LP). These results are used in the monthly report of liquid releases.

f. In addition, the program DOSUMI (DMl) is used to search the results files of DS IL to find the maximum liquid pathway individual doses. The highest exposures are then printed in 3 summary table. Each line is compared with the appropriate dose limit. The table provides a concise summary of off-site environmental dose calculations for inclusion in Annual Radioactive Effluent Release Reports. required by Reg. Guide 1.21.

Information PMP-6010.OSD.001 I Rev. 16 Page 13 of 84 OFF-SITE DOSE CALCULATION rVLkNUAL NOTE: The performance of each surveillance requirement must be within the specified time interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

3.2 Limits of Operation and Surveillances of the EffluentRelease Points 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation

a. The radioactive liquid effluent monitoring instrumentationchannels shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, are operable with their alarmitrip setpoints set to ensure that the limits of step 3.2.3a, Concentration Excluding Releases via the Turbine Room Sump (TRS) Discharge, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments.

c. With a radioactive liquid effluent monitoring instrumentationchannel alarm/trip setpoint less conservative than a value which will ensure the limits of step 3.2.3a, Concentration Excluding Releases via the Turbine Room Sump (TRS) Discharge, are met without delay, suspend the release of radioactive liquid effluents monitored by the affected channel and reset or declare the monitor inoperable.

d. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable action shown in Attachment 31.2, Radioactive Liquid Effluent Monitoring Instruments, with a maximum allowable extension not to exceed 25% of the surveillance interval, excluding the initial performance.

e. Determine the setpoints in accordance with the methodology described in step 3.3. 1, Liquid Monitors. Record the setpoints.

f. Demonstrate each radioactive liquid effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies shown in Attachment 3.3, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

BASES - LIQUID The radioactive licuid 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 tilese instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure the alarmltrip 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 specified in Secti;o. 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

Information PNiP-6010.OSD.001 Rev. 16 Page 14 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation -

a. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Attachment 3.4, Radioactive Gaseous Effluent Mf onitoring Instrumentation, are operable with their alarm/trip setpoints set to ensure that the limits of step 3.2.4a, Dose Rate, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation.

c. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of step 3.2.4a, Dose Rate, are met, without delay, suspend the release of radioactive gaseous effluents monitored by the affected channel and reset or declare the channel inoperable.

d. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, take the action shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation, with a maximum allowable extension not to exceed 25%

of the surveillance interval, excluding the initial performance.

NOTE: This surveillance requirement does not apply to the waste gas holdup system hydrogen and oxygen monitors, as their setpoints are not addressed in this document.

e. Determine the setpoints in accordance with the methodology as described in step 3.3.2, Gaseous Monitors. Record the setpoints.

f. Demonstrate each radioactive gaseous process or effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Attachment 3.5, Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements.

BASES - GASEOUS The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure 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 specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

Information I PMP-6010.OSD.O01 I Rev. 16 I Page 15 of 84I OFF-SITE DOSE CALCULATION MANUAL 3.2.3 Liquid Effluents

a. Concentration Excluding Releases via the Turbine Room Sump (TRS)

Discharge

1. Limit the concentrationofradioactive material released via the Batch Release Tanks or Plant Continuous Releases (excluding only TRS discharge to the Absorption Pond) to unrestricted areas to the

-concentrations in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, limit the concentration to 2E-4 ,uCi/ml total activity.

2. With the concentration of radioactive material released from the site via the Batch Release Tanks or Plant Continuous Releases (other than e the TRS to the Absorption Pond) exceeding the above limits, without delay restore the concentration to within the above limits.

3. Sample and analyze radioactive liquid wastes according to the sampling and analysis program of Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within limits.

b. Concentration of Releases from the TRS Discharge

1. Limit releases via the TRS discharge to the on-site Absorption Pond to the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, limit the concentration to 2E-4 g+/-Cilml total activity.

2. With releases from the TRS exceeding the above limits, perform a dose projection due to liquid releases to UNRESTRICTED AREAS to determine if the limits of step 3.2.3c.1 have been exceeded. If the dose limits have been exceeded, follow the directions in step 3.2.3c.2, as applicable.

3. Sample and analyze radioactive liquid wastes according to the program in Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within the limits stated above.

Information I PMIP-6010.OSD.001 I Rev. 16 l Page 16 of 84 OFF-SITE DOSE CALCULATION MANUAL

c. Dose

1. Limit the dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas during any calendar quarter to

1 5 mrem. to the total body and to

5 mrem to any organ, and during any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.

2. With the calculated release of radioactive materials in liquid effluents exceeding ten times any of the limits in Steps 3.2.3a, 3.2 3b, or 3.2.3c. 1 above, prepare and submit a Written Report, pursuant to 10 CFR 20.2203, within 30 days after learning of the event. This report must describe the extent of exposure of individuals to radiation and radioactive material, including, as appropriate:

a) Estimate of each individual's dose, b) Levels of radiation and concentration of radioactive material involved, c) Cause of elevated exposures, dose rates or concentrations,

-AND-d) Corrective steps taken or planned to ensure against recurrence, including schedule for achieving conformance with applicable limits.

These reports must be formatted in accordance with PMP-7030.001.002, Licensee Event Reports, Special and Routine Reports, even though this is not an LER.

3. Determine cumulative and projected dose contributions from liquid effluents in accordance with this document at least once per 31 days.

Dose may be projected based on estimates from previous monthly projections and current or future plant conditions.

d. Liquid Radwaste Treatment System 1.Use the liquid radwaste treatment system to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site when averaged over 31 days, would exceed 0.06 imrem to the total body or 0.2 mnrem to any organ.

2. Project doses due to liquid releases to UNRESTRICTED AREAS at least once per 3 I days, in accordance with his document.

Information I PNIP-6010.OSD.OO1 I Rev. 16 1 Page 17 of 84 OFF-SITE DOSE CALCULATION MANUAL

e. During times of primary to secondary leakage, the use of the startup flash tank should be minimized to reduce the release of curies from the secondary system and to maintain the dose to the public ALARA.

Operation of the North Boric Acid Evaporator (NBAE) should be done in a manner so as to allow the recycle of the distillate water to the Primary Water Storage Tank for reuse. This will provide a large reduction in liquid curies of tritium released to the environment, as there is approximately 40 curies of tritium released with every monitor tank of NBAE distillate.

Drainage of high conductivity water (Component Cooling Water and ice melt water containing sodium tetraborate) shall be evaluated to decide whether it should be drained to waste (small volumes only), the Turbine Room Sump (low activity water only) or routed without demineralization processing to a monitor tank for release. This is necessary in order to minimize the detrimental affect that high conductivity water has on the radioactive wastewater demineralization system. The standard concentration and volume equation can be utilized to determine the impact on each method and is given here. The units for concentration and volume need to be consistent across the equation:

(CX)( YV)+(Ca)( Va)=(Ct)(Yt)

Where; C; = the initial concentration of the system being added to V, = the initial volume of the system being added to Cl = the concentration of the water that is being added to the system Va = the volume of the water that is being added to, the system C, = the final concentration of the system after the addition V, = the final volume of the system after the addition The intent is to keep the:

WDS below 500 pnmhos/cc.

TRS below 1E-5 pC/cc.

Monitor Tank release ALARA to members of the public.

Wastewater leakage into the liquid waste disposal system will be monitored routinely. In the event the leak rate is determined to be over two gallons per minute (the assumed plant design leakage based on the original 2 gpm waste evaporator), increased scrutiny will be placed on locating inleakage, timeliness of job order activities, and/or activities causing increased production of waste water.

Information I PiNIP-6010.OSD.001 I Rev. 16 l Page 18 of 84 OFF-SITE DOSE CALCULATION MANUAL BASES - CONCENTRATION This specification is provided to ensure 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 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than 1) the Section I.A design objectives of Appendix 1, 10 CFR Part 50, to an individual and 2) the limits of 10 CFR Part 20. The concentration limit for noble gasses is based upon the assumption that Xe-135 is the controlling radionuclide and its Effluent Concentration Unit in air (submersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protection (ICRP) Publication 2.

DOSE This specification is provided to implement the requirements of Sections II.A, IlI.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits implement the guides set forth in Section Ml.A of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time, implement the guides set forth in Section IVMA of Appendix I to assure the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable".

Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCNM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents, will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977) and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix 1", April 1977. NUREG-0 133 provides methods for dose calculations consistent with Regulatory Guide 1.109 and 1.1 13.

This specification applies to the release of liquid effluents from each reactor at the site. The liquid effluents from the shared system are proportioned among the units sharing the system.

LIQUID WASTE TREATMENT 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 be used when specified provide 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 Criteria Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section II.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 dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents.

Information I PMP-6010.OSD.O0l I Rev. 16 l Page 19 of 84 OFF-SITE DOSE CALCULATION MLkNUAL 3.2.4 Gaseous Effluents

a. Dose Rate

1. Limit the dose rate due to radioactive materials released in gaseous effluents from the site to < 500 mremlyr to the total body and

<3000 mremlyrto the skin for noble gases. Limit the dose rate due to all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than eight days to < 1500 mrem/yr to any organ.

2. With the dose rate(s) exceeding the above limits, without delay decrease the release rate to within the above limit(s).

3. Determine the dose rate due to noble gases in gaseous effluents to be within the above limits in accordance with the methods and procedures described in this document.

4. Determine the dose rate due to radioactive materials, other than noble gases, in gaseous effluents to be within the above limits in accordance with the methods and procedures of this document by obtaining representative samples and performing analyses in accordance with the sampling and analysis program in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program.

b. Dose - Noble Gases

1. Limit the air dose in unrestricted areas due to noble gases released in gaseous effluents during any calendar quarter, to < 5 mrad for gamma radiation and 5 10 mrad for beta radiation and during any calendar year, to < 10 mrad for gamma radiation and

20 mrad for beta radiation.

2. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding ten times any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event

3. Determine cumulative and projected dose contributions for the total time period in accordance with this document at least once every 31 days.

c. Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form

1. Limit the dose to a MEMBER OF THE PUBLIC from radioiodine, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than eight days in gaseous effluents released to unrestncted areas (site boundary) to the following:

a) During any calendar quarter to less than or equal to 7.5 mrem to any organ b) During any calendar year to less than or equal to 15 mrern to any organ.

Information I PMP-6010.OSD.001 I Rev. 16 page 20 of 84 P

OFF-SITE DOSE CALCULATION LMANUAL

2. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding ten times any of the above limits, prepare and submit a Wrntten Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event.

3. Determine cumulative and projected dose contributions for the total time period in accordance with this document at least once every 31 days.

d. Gaseous Radwaste Treatment I. Use the gaseous radwaste treatment system and the ventilation exhaust treatment system to reduce radioactive materials in gaseous wastes prior to their discharge when projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. Use the ventilation exhaust treatment system to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas when averaged over 31 days would exceed 0.3 mrem to any organ.

2. Project doses due to gaseous releases to UNRESTRICTED AREAS at least once per 31 days in accordance with this document.

BASES - GASEOUS EFFLUENTS This specification is provided to ensure that the dose rate any time at the SITE BOUNDARY from gaseous effluents from all units on the site 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 2. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix B., Table 2 of 10 CFR Part 20. For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary The specified instantaneous release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to S 500 mremlyr to the total body or to < 3000 mrem/yr to the skin. These instantaneous release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to < 1500 mrem/yr.

This specification applies to the release of gaseous effluents from all reactors at the site. The gaseous effluents from the shared system are proportioned among the units sharing that system.

DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections lI.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits implement the guides set forth in Section II.B of Appendix 1.

Information I PMP-6010.OSD.001 I Rev. 16 l Page 21 of 84 OFF-SITE DOSE CALCULATION MANUAL The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown 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 dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide I .I I I, "Mviethods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. The ODCM1 equations provided for determining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0 133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.1 1 1.

DOSE, RADIOIODINES, 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 dose limits are the guides set forth in Section fl.C of Appendix I.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown 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 ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",

Revision 1, October 1977 and Regulatory Guide 1.1 I1,"Mlethods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide the methodology 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 man, and

4) deposition on the ground with subsequent exposure of man.

Information l PMIP-6010.OSD.001 I Rev. 16 l Page 22 of 84 OFF-SITE DOSE CALCULATION MANUAL GASEOUS WASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventilation exhaust treatment -

systems 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 assurance 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 Section I 1.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section II.D ofAppendix. 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 guides forth in Sections II.B and ILC of Appendix 1, 10 CFR Part 50, for gaseous effluents.

3.2.5 Radioactive Effluents - Total Dose

a. The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to c 25 mrem to the total body or any organ (except the thyroid, which is limited to

75 mrem) over a period of 12 consecutive months.

b. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding one half the annual limits of steps 3.2.3c (Dose), 3.2.4b (Dose - Noble Gases), or 3.2.4c (Dose - Iodine-1 31, Iodine- 133, Tritium, and Radioactive Material in Particulate Form) during any calendar quarter, perform the following:

Investigate and identify the causes for such release rates; Define and initiate a program for corrective action; Report these actions to the NRC within 30 days from the end of the quarter during which the release occurred.

IF the estimated dose(s) exceeds the limits above, and IF the release condition resulting in violation has not already been corrected prior to violation of 40 CFR 190, THEN include in the report a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified information of paragraph 190.1 I(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR 50, as addressed in other sections of this document.

c. Determine cumulative dose contributions from liquid and gaseous effluents in accordance with this document (including steps 3.2.3c [Dose],

3.2.4b [Dose - Noble Gases], or 3 2.4c [Dose - lodine-l 3l1,odine-133, Trntiurn, and Radioactive Svaterial in Particulate Form]).

Information I PMP-6010.OSD.001 I Rev. 16 l Page 23 of 84 OFF-SITE DOSE CALCULATION Mv4ANUAL BASES -- TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action, which should result in the limitations of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to any member of the public from other uranium fuel cycle sources is negligible with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected, in accordance with the provision of 40 CFR 190.1 1), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation, which is part of the nuclear fuel cycle.

3.3 Calculation of Alarn/Trip Setpoints The alarm and trip setpoints are to provide monitoring, indication, and control of liquid and gaseous effluents. The setpoints are used in conjunction with sampling programs to assure that the releases are kept within the limits of 10 CFR 20, Appendix B, Table 2.

Establish setpoints for liquid and gaseous monitors. Depending on the monitor function, it would be a continuous or batch monitor. The different types of monitors are subject to different setpoint methodologies.

One variable used in setpoint calculations is the multiple release point (NMRP) factor. The MRP is a factor used such that when all the releases are integrated, the applicable LIMIT value will not be exceeded. The MRP is determined such that the sum of the MRP's for that effluent type (liquid or gaseous) is less than or equal to 1. The value of the MRP is arbitrary, and it should be assigned based on operational performance. The values of the MRP's for each liquid release point are given in Attachment 3.8, Multiple Release Point Factors for Release Points.

The Site stance on instrument uncertainty is taken from HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits, which states the NRC position is the result of a valid measurement obtained by a method, which provides a reasonable demonstration of compliance. This value should be accepted and the uncertainty in that measured value need not be considered.

Information I PMIP-6010.OSD.001 I Rev. 16 1 Page 24 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.3.1 Liquid Monitors Establish liquid monitor setpoints for each monitor of the liquid effluent release systems. A schematic of the liquid effluent release systems is shown as Attachment 3.9, Liquid Effluent Release Systems. A list of the Plant Liquid Effluent Parameters is in Attachment 3.10, Plant Liquid Effluent Parameters.

The details of each system design and operation can be found in the system descriptions. The setpoints are intended to keep releases within the limits of 10 CFR 20, Appendix B, Table 2, Column 2. Determine setpoints using either the batch or the continuous methodology.

a. Liquid Batch Monitor Setpoint Methodology

1. There is only one monitor used on the Waste Disposal System for liquid batch releases. This monitor is identified as RRS-1000. Steam Generator Blowdown radiation monitors also can be used to monitor batch releases while draining steam generators. The function of these monitors is to act as a check on the sampling program. The sampling program deternines the nuclides and concentrations of those nuclides prior to release. The discharge and dilution flow rates are then adjustedto keep the release within the limits of 10 CFR 20. Based on the concentrations of nuclides in the release, the count rate on the monitor can be predicted. The high alarm setpoint can then be set above the predicted value up to the maximum setpoint of the system.

2. The radioactive concentrationof each batch of radioactive liquid waste to be discharged is determined prior to each release by sampling and analysis in accordance with Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

3. The allowable release flow rates are determined in order to keep the release concentrations within the requirements of 10 CFR 20, Appendix B, Table 2, Column 2. The equation to calculate the flow rate is from Addendum AA 1 ofNUREG-0 133:

[s *f <F+f L LSl#T, CVRP Where; C, = the concentration of nuclide "i" in flCimnl LI' Y1Tj =the 10 CFR 20, Appendix B, Table 2, Column 2 limit of nuclide 'i" in pLCi/mi f = the effluent flow rate in gpm (Attachment 3.10, Plant Liquid Effluent Parameters)

F =the dilution water flow rate as estimated prior to release.

The dilution flow rate is a multiple of 230,001) gpm depending on the number of circulation pumps in operation.

MRP = the multiple release point factor. A factor such that when all the release points are operating at one time the limits of I0 CFR 20 will not be exceeded.

Information I PNIP-6010.OSD.001 I Rev. 16 l Page 25 of 84 OFF-SITE DOSE CALCULATION MANUAL

4. This equation must be true during the batch release. Before the release is started, substitute the maximum effluent flow rate and the minimum dilution flow rate for f and F, respectively. If the equation is true, the release can proceed with those flow rates as the limits of operation. If the equation is not true, the effluent flow rate can be reduced or the dilution flow rate can be increased to make the equation true. This equation may be rearranged to solve for the maximum effluent release flow rate (f).

5. The setpointis used as a quality check on the sampling program. The setpoint is used to stop the effluent flow when the monitor reading is greater than the predicted value from the sampling program. The predicted value is generated by converting the effluent concentration for each gamma emitting radionuclide to counts per unit of time as per Attachment 3. 11, Volumetric Detection Efficiencies for Principle Ganmma Emitting Radionuclides for Eberline Liquid Monitors, or Attachment 3.12, Counting Efficiency Curves for R-19, and R-24.

The sum of all the counts per unit of time is the predicted count rate.

The predicted count rate can then be multiplied by a factor to determine the high alarm setpoint that will provide a high degree of conservatism and eliminate spurious alarms.

b. Liquid Continuous Monitor Setpoint Methodology

1. There are eight monitors used as potential continuous liquid release monitors. These monitors are used in the steam generator blowdown (SGBD), blowdown treatment (BDT), and essential service water (ESW) systems.

2. The Westinghouse monitors (R) are being replaced by Eberline monitors (DRS, WRA) and are identified as:

R-19 or DRS 3100/4100 for SGBD R-24 or DRS 3200/4200 for BDT

R-20 or WRA 3500/4500 for the east ESW system

- R-28 or WRA 3600/4600 for the west ESW system The function of these monitors is to assure that releases are kept within the concentration limits of 10 CFR 20, Appendix B, Table 2, Column 2, entering the unrestricted area following dilution.

3. The monitors on steam generator blowdown and blowdown treatment systems have trip functions associated with their setpoints. Essential service water monitors are equipped with an alarn function only and monitor effluent in the event the Containmrent Spray Heat Exchangers are used or the ESW system (Eberline).

Information I PVIP-6010.OSD.001 I Rev. 16 l Paae 26 of 34 OFF-SITE DOSE CALCULATION MANUAL

4. The equation used to determine the setpoint for continuous monitors is from AddendumrAAl of NUREG-0133:

<C

Eff *MRP

F

SF Sp- f Where; Sp = setpoint of monitor (cpm)

C = SE-7 pCi/nil, maximum effluent control limit from 10 CFR 20, Appendix B, Table 2, Column 2 of a known possible nuclide in effluent stream. (The limiting nuclide shall be evaluated annually by reviewing current nuclides against historical ones in order to determine if one with a more restrictive effluent concentration limit than Sr9O is found. The concentration limit shall be adjusted appropriately.)

-OR-if a mixture is to be specified, E C.

E C, LLflfT, Eff = Efficiency, this information is located in Attachment 3. 11, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, through Attachment 3.13, Counting Efficiency Curve for R-20, and R-28, for the specific monitors. For Eberline monitors the efficiency is nuclide specific and the calculation changes slightly to:

(C,

Eff) replaces C

Eff C,

LIM1T, MRP = multiple release point factor. A factor such that when all the release points are operating at one time the limits of 10 CFR 20 will not be exceeded (Attachment 3. 8.

Multiple Release Point Factors for Release Points). The MvRP for ESW monitors is set to 1.

F = dilution water (circ water) flow rate in gpr obtained from Attachment 3.10, Plant Liquid Effluent Parameters. For routine operation, the setpoint should be calculated using the minimum dilution flow rate of 23 0,000 gpm.

SF = Safety Factor, 0.9.

f = applicable effluent release flow rate in gpm. For routine operation, the setpoint should be calculated using maximum effluent flow rate (Attachment 3.10, Plant Liquid Etfluent Parameters).

Information PMP-6010.OSD.001 I Rev. 16 l Page 27 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.3.2 Gaseous Monitors For the purpose of implementing Step 3.2.2, Radioactive Gaseous Effluent MonitoringInstrumentation, and Substep 3.2.4a, Dose Rate, the alarm setpoints for gaseous effluents released into unrestricted areas will be established using the following methodology. In addition, the above steps do not apply to instantaneous alarm and trip setpoints for integrating radiation monitors sampling radioiodines, radioactive materials in particulate form and radionuclides other than noble gases. A schematic of the gaseous effluent release systems is presented in Attachment 3.14, Gaseous Effluent Release Systems. Attachment 3.15,'Plant Gaseous Effluent Parameters, presents the effluent flow rate pararneterT(s).

Gaseous effluent monitor high alarm setpoints will routinely be established at a fraction of the maximum allowable setpoint (typically 10% of the setpoint) for ALARA purposes. Alert alarms will normally be set to provide adequate indications of small changes in radiological conditions.

a. Plant Unit Vent

1. The gaseous effluents discharged from the plant vent will be monitored by the plant vent radiationmonitor low range noble gas channel [Tag No. VRS-1 505 (Unit 1), VRS-2505 (Unit2)] to assure that applicable alarms and trip actions (isolation of gaseous release) will occur prior to exceeding the limits in step 31.2.4, Gaseous Effluents. The alarm setpoint values will be established using the following unit analysis equation:

SF* AlP *DLJ sp FP*YQ*L(W, *DCFO)

Where; Sp = the maximum setpoint of the monitor in iCi/cc for release point p, based on the most limiting organ SF = an administrative operation safety factor, less than 1.0 MRP a weighted multiple release point factor (S5 1.0), such that when all site gaseous releases are integrated, the applicable dose will not be exceeded based on the release rate of each effluent point. The MRP is an arbitrary value based on the ratio of the release rate or the volumetric flow rate of each effluent point to the total respective flow rate value of the plant and will be consistent with past operational experience. The MRP is computed as follows:

Compute the average release rate, Qp, (or the volumetric flow rate, f.) from each release point p.

Compute ZQp (or Ifp) for all release points.

Ratio Qp/IQp (or fp/Zfp) for each release point. This ratio is the MRP for that specific release point

Repeat the above bullets for each of the site's eight gaseous release points.

Information I PNIP-6010.OSD.001 I Rev. 16 l Page 28 of 84 OFF-SITE DOSE CALCULATION MlANUAL Fp = the maximum volumetric flow rate of release point "p", at the time of the release, in cc/sec. The maximum Unit Vent flow rate, by design, is 186,600 cfm for Unit 1 and 143,400 cfm for Unit 2.

DLj = dose rate limit to organ "j" in an unrestricted area (mrem/yr).

Based on continuous releases, the dose rate limits, DL, from step 3.2.4a, Dose Rate, are as follows:

Total Body < 500 inrem/year

Skin < 3000 mremlyear

Any Organ <1500 mremnyear

/0 = The worst case annual average relative concentration in the applicable sector or area, in sec/M3 (see Attachment 3.16, 10 Year Average of 1989-1998 Data).

W1 = weighted factor for the radionuclide.

_ C, ZCk Where, C, = concentration of the most abundant radionuclide "i" Ck = total concentration of all identified radionuclides in that release pathway. For batch releases, this value may be set to I for conservatism.

DCF11 = dose conversion factor used to relate radiation dose to organ .';j, from exposure to radionuclide "i" in mrem m / yr p.Ci. See following equations.

The dose conversion factor, DCF1 j, is dependent upon the organ of concern.

For the whole body: DCFuJ = K; Where; K, = whole body dose factor due to gamma emissions for each identified noble gas radionuclide in mrem m3 / yr 1iCi. See Attachment 3.13, Dose Factors.

For the skin: DCF j1= L, + 1.lM Where; L, = skin dose factor due to beta emissions for each identified noble gas radionuclide, in mremn m3 / yr gCi. See Attachment 3.18, Dose Factors.

Information PMP-6010.OSD.001 Rev. 16 Page 29 of 84 OFF-SITE DOSE CALCULATION MANUAL 1.] = the ratio of tissue to air absorption coefficient over the energy range of photons of interest. This ratio converts absorbed dose (mrad) to dose equivalent (mrem).

M= the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad m3 / yr ILCi. See Attachment 3.18, Dose Factors.

For the thyroid, via inhalation: DCFj =P.

Where; P. = the dose parameter, for radionuclides other than noble gas. for the inhalation pathway in mrem ml I yr tCi (and the food and ground path, as appropriate).

See Attachment 3.18, Dose Factors.

2. The plant vent radiation monitor low range noble gas high alarm channel setpoint, Sp. will be set such that the dose rate in unrestricted areas to the whole body, skin and thyroid (or any other organ),

whichever is most limiting, will be less than or equal to 500 mrem/yr, 3000 mrem/yr, and 1500 mremlyr respectively.

3. The thyroid dose is limited to the inhalation pathway only.

4. The plant vent radiation monitor low range noble gas setpoint, Sp, will be recomputed whenever gaseous releases like Containment Purge, Gas Decay Tanks and CVCS HUTs are discharged through the plant vent to determine the most limiting organ.

5. The high alarm setpoint, Sp, may be established at a lower value than the lowest computed value via the setpoint equation.

6. Containment Pressure Reliefs will not have a recomputed high alarm setpoint, but will use the normal high alarm setpoint due to their randomness and the time constraints involved in recomputation.

7. At certain times, it may be desirable to increase the high alarm setpoint, if the vent flow rate is decreased. This may be accomplished in one of two ways.

Max Conc ( .tCi/cc)* Max Flowrate(cfin) - JVev Max cfm New Max Concentration(uCi/cc)

-OR-Max Conc ( L.Ci/cc)

Max Flowrate (cfm) _ Aew Max p.Ci/cc New Max Flo wrate (cfin)

b. Waste Gas Storage Tanks The gaseous effluents discharged from the Waste Gas System are monitored by the vent stack monitors VRS-l 505 and VRS-2505.

Information PN1P-6010.OSD.001 Rev. 16 Page 30 of 34 OFF-SITE DOSE CALCULATION MANUAL

2. In the event of a high radiation alarm, an automatic termination of the release from the waste gas system will be initiated from the plant vent radiation monitor low range noble gas channel (VRS-1 505 or VRS-2505). Therefore, for any gaseous release configuration, which includes normal operation and waste gas system gaseous discharges, the alarm setpoint of the plant vent radiation monitor will be recomputed to determine the most limiting organ based on all gaseous effluent source terms.

Chemical and Volume Control System Hold Up Tanks (CVCS HUT), containing high gaseous oxygen concentrations, may be released under the guidance of waste gas storage tank utilizing approved Operations' procedures.

3. It is normally prudent to allow 45 days of decay prior to releasing a Gas Decay Tank (GDT). There are extenuating, operational circumstances that may prevent this from occurring. Under these circumstances, such as high oxygen concentration creating a combustible atmosphere, it is prudent to waive the 45-day decay for safety's sake.

c. Containment Purge and Exhaust System

1. The gaseous effluents discharged by the Containment Purge and Exhaust Systems and InstrumentationRoom Purge and Exhaust System are monitored by the plant vent radiation monitor noble gas channels (VRS-1505 for Unit 1, VRS-2505 for Unit 2); and alarms and trip actions will occur prior to exceeding the limits in step 3.2.4a, Dose Rate.

2. For the Containment System, a continuous air sample from the containment atmosphere is drawn through a closed, sealed system to the radiation monitors (Tag No. ERS-1300/1400 for Unit I and ERS-2300/2400 for Unit 2). During purges, these monitor setpoints will give a Purge and Exhaust Isolation signal upon actuation of high alarm setpoints for particulate and noble gas channels. The sample is then returned to containment. Grab sample analysis is performed for a Containment purge before release.

3. The Upper Containment area is monitored by normal range area gamma monitors (Tag No. VRS- 1101/1201 for Unit 1 and VRS-2101/2201 for Unit 2), which also give Purge and Exhaust IsolationTrip signals upon actuation of their high alarm.

4. For the Containment Pressure Relief System, no sample is routinely taken prior to release, but a sample is obtained twice per month.

5. The containrmentairborne and area monitors, upon actuation of their high alarm, will automatically initiate closure of the Contairrnent and Instrument Room purge supply and exhaust duct valves and contaimnentpressure relief system valves. Complete trip of all isolation control devices requires high alarm of one of the two Train A monitors (ERS- 1300/2300 or VRS-1 1l0/2101) and one of the two Train B monitors (ERS- 1400/2400 or VRS-1201/2201)

d. Steam Jet Air Ejector Syster (SJAE)

1. The gaseous effluents from the Steam Jet Air Ejector System discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-1900 for Unit I and SRA-2900 for Unit 2). The monitor will alarm prior to exceeding the limits of step 31.2.4a, Dose Rate. The alarm setpoint for the Condenser Air Ejector System monitor will be based on the maximum air ejector exhaust flow rate, (Attachment 3.15, Plant Gaseous Effluent Parameters).

The alarm setpoint value will be established using the following unit analysis equation:

SF

MRP

DL, FI/Q Fp * *Z (Wi* DCF,)

Where; SsjAE = the maximum setpoint, based on the most limiting organ; in p.Ci/cc and where the other terms are as previously defined

e. Gland Seal Condenser Exhaust 1.- The gaseous effluents from the Gland Seal Condenser Exhaust discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-1 800 for Unit l and SRA-2800 for Unit 2). The radiation monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the GSCE monitor will be based on the maximum condenser exhaust flow rate (1260 CFM for Unit 1, 2754 CFM each for the two Unit 2 vents). The alarm setpoint value will be established using the following unit analysis equation:

SF*MRP* DL, FP* X/ *Z-(W

DCFu)

Information I PNIP-6010.OSD.001 I Rev. 16 1 Page 32 of 84 OFF-SITE DOSE CALCULATION NAINUAL Where; SGSCE = the maximum setpoint, based on the most limiting organ, in ptCi/cc and where the other terms are as previously defined 3.4 Radioactive Effluents Total Dose 3J.4.1 The cumulative dose contributions from liquid and gaseous effluents will be determinedby summing the cumulative doses as derived in steps 3 2 3c (Dose).

3.2.4b (Dose - Noble Gases), and 3.2.4c (Dose - Iodine- 131, Iodine-1 33 Tritium, and Radioactive Material in Particulate Form) of this procedure. Dose contribution from direct radiation exposure will be based on the results of the direct radiation monitoring devices located at the REIMP monitoring stations.

See NUREG-0 133, section 3.8.

3.5 Radiological Environmental Monitoring Program (REMP) 3.5.1 Purpose of the REMIP

a. The purpose of the RENIP is to:

Establish baseline radiation and radioactivity concentrations in the environs prior to reactor operations,

Monitor critical environmental exposure pathways, Determine the radiological impact, if any, caused by the operation of the Donald C. Cook Nuclear Plant upon the local environment.

b. The first purpose of the REMP was completed prior to the initial operation of either of the two nuclear units at the Donald C. Cook Nuclear Plant Site.

The second and third purposes of the REMP are an on-going operation and as such various environmental media and exposure pathways are examined. The various pathways and sample media used are delineated in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Included is a list of the sample media, analysis required, sample stations, and frequency requirements for both collection and analysis. Attachment 3.19, Radiological Enviromnental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, defines the scope of the R-EMP for the Donald C. CookNuclear Plant.

J

Information I PMP-6010.OSD.OO1 l Rev. 16 Page 33 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.5.2 Conduct of the REM? [ReS. 5.2. Iu]

a. Conduct sample collection and analysis for the REMP in accordance with Attachment 3 19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, Attachment 3.20, Maximum Values for Lower Limits of DetectionsAB - REMP, and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples. These are applicable at all times. The on-site monitoring locations are shown on Attachment 3.22, On-Site Monitoring Location - REMP, and the off-site monitoring locations are shown on Attachment 3.23, Off-Site Monitoring Locations - REMP.

1. Perform each surveillance requirement within the specified time interval in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, with a maximum allowable extension not to exceed 25%

of the surveillance interval.

2. If an environmental sample cannot be collected in accordance with step 3.5.2a, submit a description of the reasons for deviation and the actions taken to prevent a reoccurrence as part of the Annual Radiological Environmental Operating Report (AREOR).

Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or malfunction of automatic sampling equipment. If

-the deviation from the required sampling schedule is due to the malfunction of automatic sampling equipment, make every effort to complete the corrective action prior to the end of the next sampling period.

3. If a radionuclide is detected in any sample medium exceeding the limit established in Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples, or if more than one radionuclide is detected in any sample medium and the Total Fractional Level (TFL), when averaged over the calendar quarter, is greater than or equal to 1, based on the following formula:

TIFF = C(1) + C/2) f, Lo~ L(2 Where; C(l) = Concentrationof IS"detectednuclide C(2) = Concentration of 2 nd detected nuclide L(l) = Reporting Level of 15' nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

Information PMP-6010.OSD.001 Rev. 16 Page 34 of 84 OFF-SITE DOSE CALCULATION MANUAL L(2) = Reporting Level of 2nd nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

And, if the activity is the result of plant effluents, evaluate the release conditions, environmental factors, or other aspects, which may have contributed to the identified levels for inclusion in the AREOR. If the radioactivity was not a result of plant effluents, describe the results in the AREOR.

4. If a currently sampled milk farm location becomes unavailable, conduct a special milk farm survey within 15 days.

a) If the unavailable location was an indicator farm, an alternate sample location may be established within eight miles of the Donald C. Cook Nuclear Plant, if one is available.

b) If the unavailable location was a background farm, an alternate sample location may be established greater than 15 but less than 25 miles of the Donald C. Cook Nuclear Plant in one of the less prevalent wind direction sectors, if one is available.

c) If a replacement farm is unobtainable and the total number of indicator farms is less than three or the background farms is less than one, perform monthly vegetation sampling in lieu of milk sampling.

BASES - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (RENMP)

The REMIP provides measurements of radiation and 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. Thereby, this monitoring program supplements the radiological effluent monitoring program by verifying the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

The initially specified REMIP will be effective for at least the first three years of commercial operation. Program changes may be initiated based on operational experience in accordance with the requirements of technical specification 6.S.4.b.

The detection capabilities, required by Attachment 3.20, Maximum Values for Lower Limits of DetectionsA' 3 - REMP, are the state-of-the-art for routine environmental measurements in industrial laboratories.

It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (afler the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine analysis conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

Information I PNIP-6010.OSD.001 - Rev. 16 l Page 35 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.5.3 Annual Land Use Census [Ref. 5 2.1u]

a. Conduct a land use census and identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 square feet producing fresh leafy vegetables in each of the ten land sectors within a distance of five miles.

b. In lieu of the garden census, grape and broad leaf vegetation sampling may be performed as close to the site boundary as possible in a land sector, containing sample media, with the highest average deposition factor (D/Q) value.

c. Conduct this land use census annually between the dates of June 1 and October 1 by door-to-door survey, aerial survey, or by consulting local agricultural authorities.

1. With a land use census identifying a location(s), which yields a calculated dose or dose commitment greater than the values currently being calculated in this document, make appropriate changes to incorporate the new location(s) within 30 days, if possible.

BASES -- LAND USE CENSUS This is provided to ensure changes in the use of unrestricted areas are identified and modifications to the monitoring program are made in accordance with requirements of TS 6.8.4b, if required by the results of the 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 since a garden of this size is the minimum required to produce the quantity (25 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption of a child. To determine this minimum garden size, the following assumptions were used: 1)that 20% of the garden was used for growing broad leaf vegetation (that is, similar to lettuce and cabbage), and 2) a vegetation field of 2 kg/square meter.

3.5.4 Interlaboratory Comparison Program

a. In order to comply with Reg. Guides 4.1 and 4 15, the analytical vendor participates in an Interlaboratory Comparison Program, for radioactive materials. Address program results and identified deficiencies in the AREOR

1. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the AREOR.

BASES -- INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate the results are reasonably valid.

Information I PIMP-6010.OSD.001 Rev. 16 i Page 36 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.6 Steam Generator Storage Facility Groundwater Monitoring Program 3 6.1 Purpose of the Steam Generator Storage Facility Groundwater Radiological Monitoring Program

a. The purpose of the temporary on-site Steam Generator Storage Facility Radiological Monitoring Program is to establish baseline radiological data for the groundwater surrounding the facility prior to the storage of the Unit 2 Steam Generator Lower Assemblies. Thereafter, the purpose is to monitor the groundwater through observation wells with locations as shown in Attachment 3.22, On-Site Monitoring Location - RENMP, to determine the radiological impact, if any, caused by the use of the Storage Facility.

3.6.2 Conduct of the Steam Generator Storage Facility Groundwater Radiological Monitoring Program

a. Collect and analyze groundwater samples in accordance with Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Apply the values from Attachment 3.20, Maximum Values for Lower Limits of DetectionsA"3 - REMiP, (excluding 1-13 1) and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples, (excluding I-13 1).

3.7 Meteorological Model 3.7.1 Three towers are used to determine the meteorological conditions at Donald C.

Cook Nuclear Plant. One of the towers is located at the Lake M\ichigan shoreline to determine the meteorological parameters associated with unmodified shoreline air. The data is accumulated by microprocessors at the tower sites and normally transferred to the central computer every 15 minutes.

3.7.2 The central computer uses a meteorological software program to provide atmospheric dispersion and deposition parameters. The meteorological model used is based on guidance provided in Reg. Guide 1. I I for routine releases.

All calculations use the Gaussianplume model.

3.8 Reporting Requirements 3.8.1 Annual Radiological Environmental Operating Report (AREOR)

a. Submit routine radiological environmental operating reports covering the operation of the units during the previous calendar year prior to Mvlay I of each year.

b. Include in the AREOR:

Summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the reporting period.

Information PMP-6010.OSD.001 Rev. 16 Page 37 of 84 OFF-SITE DOSE CALCULATION MANUAL

A comparison with pre-operational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The results of the land use censuses required by step 3.5.3, Annual Land Use Census.

If harmful effects or evidence of irre ersible damage are detected by the monitoring, provide in the report an analysis of the problem and a planned course of action to alleviate the problem.

Summarized and tabulated results of all radiological environmental samples taken during the reporting period. In the event that some results are not available for inclusion with the report, submit the report noting and explaining the reasons for the missing results. Submit the missing data as soon as possible in a supplementary report.

A summary description of the REMP including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used.

A map of all sample locations keyed to a table giving distances and directions from one reactor.

The results of participation in the Interlaboratory Comparison Program required by step 3.5.4, Interlaboratory Comparison.Program.

3.8.2 Annual Radiological Effluent Release Report (ARERR)

a. Submit routine ARERR covering the operation of the unit during the previous 12 months of operation within 90 days after January 1 of each year.

b. Include in the ARERR a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released frorn the units as outlined in Reg. Guide 1.21, "Measuring, Evaluating and Reporting 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

c. Submit in the ARERR 90 days after January I of each year and include a quarterly summary of hourly meteorological data collected during the reporting period.

This summary may be in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form ofjoint frequency distributions of wind speed, wind direction and atmospheric stability.

Inciude an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

Information l Pi1P-6010.OSD.001 I Rev. 16 l Page 38 of 84 OFF-SITE DOSE CALCULATION MIANUAL

Include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary during the reporting period. Include all assumptions used in making these assessments (that is, specific activity, exposure time and location) in these reports.

Use the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) for determining the gaseous pathway doses.

Inoperable radiation monitor periods exceeding 30 continuous days; explain causes of inoperability and actions taken to prevent reoccurrence.

d. Submit the ARERR [Ref. 52. Iw] 90 days after January I of each year and include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Reg. Guide l .109, Rev.1.

e. Include in the ARERR the following information for each type of solid waste shipped off-site during the report period:

Volume (cubic meters),

Total curie quantity (specify whether determined by measurement or estimate),

Principle radionuclides (specify whether determined by measurement or estimate),

Type of waste (example: spent resin, compacted dry waste, evaporator bottoms),

Type of container (example: LSA, Type A, Type B, Large Quantity),

-AiND-

Solidificationagent (example: cement).

f. Include in the ARERR unplanned releases of radioactive materials in gaseous and liquid effluent from the site to unrestricted areas on a quarterly basis.

g. Include in the ARERR any change to this procedure made during the reporting period.

39 10 CFR 50.75 (g) Implementation

PMP-6010.OSD.001 Rev. 16 Page 39 of 84 Information OFF-SITE DOSE CALCULATION rVLAkNUAL 3.9.1 Records of spills or other unusual occurrences involving the spread of contamination in and around the site. These records may be limited to instances when significant contamination remains after decontamination or when there is a reasonable likelihood that contaminants may have spread to inaccessible areas, as in the case of possible seepages.

3.9.2 These records shall include any known information or identification of involved nuclides, quantities, and concentrations.

3.9.3 This information is necessary to ensure all areas outside the radiological-restricted area are documented for surveying and remediation during decommissioning. There is a retention schedule file number where this information is filed in Nuclear Documents Management to ensure all required areas are listed to prevent their omission.

3.10 Reporting/ManagementReview 3.10.1 Incorporate any changes to this procedure in the ARERR.

3.10.2 Update this procedure when the Radiation Monitoring System, its instruments, or the specifications of instruments are changed.

3.10.3 Review or revise this procedure as appropriate based on the results of the land use census and REMP.

3.10.4 Evaluate any changes to this procedure for potential impact on other related Department Procedures.

3.10.5 Review this procedure during the first quarter of each year and update it if necessary. Review Attachment 3.16, 10 Year Average of 1989-1998 Data, and document using Attachment 3.17, Annual Evaluation of x/Q and D/Q Values For All Sectors. The XI Q and D I Q values will be evaluated to ensure all data is within +/- 3 standard deviations of the 10 year annual average data and documented by completing Attachment 3.17, Annual Evaluation of x/Q and D/Q Values For All Sectors, and filed in accordance with the retention schedule.

4 FINAL CONDITIONS 4.1 None.

5 REFERENCES 51 Use

References:

5.1.1 "Implementation of Programmatic Controls for Radiological Effluent Technical Soecifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Off-Site Dose Calculation Manual or to the Process Control Program (Generic Letter 89-01)", United States Nuclear Regulatory Commission, January 31, 1989 5.1.2 12-THP-60 10.RPP 601, Preparation of the Annual Radioactive Effluent Release Report

Information iPMP-6010.0SD.001 I Rev. 16 l Page 40 of 34 OFF-SITE DOSE CALCULATION MANUAL 5.1.3 12-THP-6010.RPP.639, Annual Radiological Environmental Operating Report (AREOR) Preparation And Submittal 5.2 Writing

References:

5 2.1 Source

References:

a. 10 CFR 20, Standards for Protection Against Radiation

b. 10 CFR 50, Domestic Licensing of Production and Utilization Facilities

c. PMI-60 10, Radiation Protection Plan

d. NUREG-0472

e. NUREG-0 133

f. Regulatory Guide 1.109, non-listed parameters are taken from these data tables

g. Regulatory Guide 1.111

h. Regulatory Guide 1.113

i. Final Safety Analysis Report (FSAR)

j. Technical Specifications, Appendix A, Sections 6.8.1.e, 6.3.4.a, 6.8.4.b, 6.9.1.6, 6.9.1.7, and 6.14, Off-Site Dose Calculation Manual

k. Final Environmental Statement Donald. C. Cook Nuclear Plant, August 1973

1. NUREG-0017

m. ODCM Setpoints for Liquid Effluent Monitors (Bases), ENGR 107-04 8112.1 Environs Rad Monitor System

n. HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits

o. Watts - Bar Jones (WBJ) Document, R-86-C-001. The Primary Calibration of Eberline Instrument Corporation SPING - 3/4 Low, Mid, and High Range Noble Gas Detectors

p. WBJ Document, R-86-C-003, The Primary Calibration of Eberline Instrument Corporation DAM-4 and Water Monitor

q. 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations

r. NRC Commitment 6309 (N94083 dated I 1/1 094)

s. NRC Commitment 1151

t. NRC Commitment 1217

u. NRC Commitment 3240

v. NRC Commitment 3850

w. NRC Commitment 4859

x. NRC Commitment 6442

y. NRC Commitment 3768

z. DIT-B-00277-00, HVAC Systems Design Flows aa. Regulatory Guide 1.21 bb. Regulatory Guide 4 1

cc. 1-2-V3-02-Calc #4, Unit Vent Sample Flow rate for isokinetic particulates and Iodine sampling dd. HPS N 13.30-1996, Appendix A Rationalefor Methods of Determining Minimum Detectable Amount (MDA) and Minumum Testing Level (MDL ec. DIT-B-01971-00, Dose Factors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway ff. DIT-B-01987-00, Ground Plane & Food Dose Factors P. for Radioiodines and Radioactive Particulate Gaseous Effluents 5.2.2 General References a Cook Nuclear Plant Start-Up Flash Tank Flow Rate lctter from D. L.

Boston dated January 21, 1997

b. Letter from B.P. Lauzau, Venting of Middle CVCS Hold-Up Tank Directly to Unit Vent, May 1, 1992

c. AEP Design Information Transmittal on Aux Building Ventilation Systems

d. PMiP4030.EIS.001, Event-Initiated Surveillance Testing

e. Environmental Position Paper, Fe Impact on Release Rates, approved 3/14100

f. Environmental Position Paper, Metbodology Change from Sampling 15%

Secondary System Gaseous Effluents for Power Changes Exceeding within 1 hr to Responding to Gaseous Alert Alarms, approved 4/4/00

Information PMIP-6010.OSD.O01 Rev. 16 Page 42 of 34 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.1 Dose Factors for Various Pathways 42 Pg45 R, Dose Factors PATHVAY Nuclide Ground l Vegetable Meat Cow Milk Goat Milk I Inhalation H-3 0.OE+00 4.0E+03 3.3E+02 2.4E+03 4.9E+03 1.3E+03 C-14 0.OE+00 3.5E+06 5.3E+05 3 2E+06 3 2E+06 j 3 6E + 04 Cr-51 5.4E+06 1.1E+07 1.5E+06 6.9E+06 8.3E+05 I 2.IE+04 NIn-54 1.6E+09 9.4E+08 2.1E+07 2.9E+07 3.5E+06 2 OE+06 Fe-59 3.2E+08 9.6E+08 1.7E+09 3.1E+08 [ 4.0E+07 1.5E+06 Co-58 4.4E+08 6.OE+08 2.9E+08 8.4E+07 1.OE+07 1.3E+06 Co-60 2.5E+10 3.2E+09 1.0E+09 2.7E+08 3.2E+07 8.62+06 Zn-65 8.5E+08 2.7E+09 9.5E+08 1.6E+ 10 1.9E+09 1.2E+06 Sr-89 2.5E+04 3.5E+ 10 3.8E+08 9.9E+09 2.1E+10 2.4E+06 Sr-90 O.OE+00 1.4E+ 12 9.6E+09 9.4E+10 2.OE.11 1.lE+08 Zr-95 2.9E+08 1.2E+09 1.5E+09 9.3E+05 1.1E+05 2.7E+06 Sb- 124 6.9E+08 3.0E+09 4.4E+08 7.2E+08 8.62+07 3.8E+06 I-131 1.0E+Q7 2-4E+10 2.5E+09 4.8E+ 11 5.3E+11 1.6E+07 1-133 1.5E+06 4.0E+08 6.0E+01 4.4E+09 5.3E+09 3.8E +06 Cs-134 7.9E+09 2.5E+ 10 1.1E+09 5.0E+10 1.5E+11 1.1E+06 Cs-136 1.7E+08 2.2E+08 4.2E+07 5.1E+09 1.5E+ 10 1.9E+05 Cs-137 1.2E+10 2.52+10 1.0E+09 4.5E+10 1.4E+/- 11 9.OE+05 Ba-140 2.3E+07 2.7E+08 5.2E+07 2.1E+08 2.6E+07 2.0E+06 Ce-141 1.5E+07 5.3E+08 3.0E+07 8.3E+07 1.02+07 6.1E+05 Ce-144 7.9E+07 1 1.3E+10 3.62+/-08 7.3E+08 8.7E+07 1.3E+07 Units for ail except inhalation pathway are mZ rr sec I yr l.iCi, inhalation pathway units are mr m3 / yr jiCi U0 , Values to be Used For the Maximum Exposed Individual Pathway Infant Child Teen Adult Fruits, vegetables and grain (kg/yr) - 520 630 520 Leafy vegetables (kg/yr) I - 26 42 64 Milk (L/yr) 330 330 400 310 Meat and poultry (kg/yr) - 41 f 65 110 Fish (ka/yr) - 6.9 16 21 Drinking water (L/yr) 330 510 510 730 Shoreline recreation (hrlyr) l 14 67 12 Inhalation (m3/yr) 1400 [ 3700 0 1 8000 Table E-5 ofReg. Guide 1.109.

I PNIP-6010.OSD.001 Rev. 16 I Page 43 of 84 Information OFF-SITE'DOSE CALCULATlON MANUAL I i age_

Pages:

Attachlment 3.1 l Dose Factors for Various Pathways 42 - 45 Bp Factors for Aquatic Foods pCilIkgpCi Element Fish Invertebrate H 9.OE- I 1 9.OE-1 C 4.6E3 1 9.1E3 Na *1-1-.0E2 2.0hb2.

p LO f 2.OE4 I 2.0E3 I Cr Nn 4.0E2 9E4 Fe L.0E2 3.2E Co 5.OEI 2.0E2 Ni 1.0E2 I .0E2 Cu 5.OEI I 4.0E?2 Zn I2.0E3 Ij 1.0E4 Br 4.2E2 1 33E Rb 2.0E3 I 1.0E3 Sr 3O I 1.0E2

-Y 2.5EI1 i.OE3 T 3.3E0 6.7E0 Nb 3.0E4 1.0E2 Mo lOEI 1.OEl Tc 1.5El I 5.OEO Ru . 1. l 3.0E2 Rh 1.OEl I 3.0E2 Te 4.OE2 6.1E3 I 1.5El 5.OEO 2.0E3 1 .0E3 Ba 4.OEO 2.0E2 La 1 2.SEl l .E3 Ce 1.0EU 1.0E3

¢ Pr 2.5E1 1.0E3 Nd 2.51 1.0E3 Iw

~W 12E3 1.0El t l.OEI 4.0E2 Table A- I of Reg. Guide 1.109.

Information I PNIP-6010.OSD.001 Rev. 16 R j Page 44 of 84 OFF-SITE DOSE CALCULATION MANL-UAL .1 Dose Factors for Various Pathways Pages:

D2,p; External Dose Factors for Standing on Contaminated Ground mrem m 2 /hrpCi Radionuclide l Total Body Skin H-3 j0 O C-14 0 0 Na-24 2.5E-8 2.9E-8 -

P.-32 0 0 Cr-SI 2.2E-10 2.6E-10 Mn-54 5.8E-9 6.SE-9 Mn-56 1.1 E-8 1.3E-8 Fe-55 0 0 Fe-59 8.0E-9 9.4E-9 Co-58 7.OE-9 8.2E-9 Co-60 I.7E-8 2.OE-8 Ni-63 0 0 Ni-65 3.7E-9 4.3 E-9 Cu-64 1.5E-9 I .7E-9 Zn-65 4.OE-9 4.6E-9 -

Zn-69 0 0 Br-83 6.4E-11 9.3E-11 Br-84 1.2E-8 1.4E-8 Br-85 0 0 Rb-86 6.3E-10 7.2E-10 Rb-88 3.5E-9 4.OE-9 Rb-89 1.5E-8 1.8E-8 Sr-89 5 6E-13 6.5E-13 Sr-91 7.IE-9 , 8.3E-9 Sr-92 9.0E-9 ILOE-8 Y-90 2.2E-12 2.6E-12 Y-91m 3.8E-9 4.4E-9 Y-91 2.4E,- I 2.7E-11 Y-92 1.6E-9 1.9E-9 Y-93 5.7E-10 7.8E-10 Zr-95 5.OE-9 5.3E-9 Zr-97 5.5 E-9 6.4E-9 Nb-95 5.1 E-9 6.OE-9 Mo-99 1.9E-9 j 2.2E-9 Tc-99m 9.6E-10 I 1.1E-9 Tc-101 2.7E-9 3.OE-9 Ru-103 3.6E-9 4.2E-9 Ru-105 4 5E-9 5.iE-9 Ru-106 15E-O 1.8E-9

.Ag-IIOmrn 1.8E-8 2.1E-8 re-125m 33 SE-1 I 4.3E- 1I

Radionuclide Total Body Skin Te-127m 1.lE-12 J 1.3E-12 Te- 127 1.OE- i1 1.1E-11 Te-129m 7.7E- 10 9.OE- 10 Te-129 7.1E-10 8.-E-10 Te- 13 lm 8.4E-9 9.9E-9 Te-131 l 2.2E-9 2.6E-6 Te-132 1.7E-9 2.0E-9 1-130 1.4E-8 1.7E-8 1-131 2.8E-9 3.4E-9 1-132 1.7E-8 2.OE-8 1-133 3.7E-9 4.5E-9 1-134 1.6E-8 1.9E-8 1-135 1.2E-8 1.4E-8 Cs-134 1.2E-8 1.4E-8 Cs-136 1.5E-8 1.7E-8 Cs-137 4.2E-9 4.9E-9 Cs-138 2.1E-8 2.4E-8 Ba-139 2.4E-9 2.7E-9 Ba-140 2.1E-9 2.4E-9 Ba-141 4.3E-9 4.9E-9 Ba-142 7.9E-9 9.OE-9 La- 140 1.5E-8 1.7E-8 La-142 - 1.5E-8 1.8E-8 Ce-141 5.5E-10 6.2E-1i Ce- 143 - 2.2E-9 2.5E-9 Ce-144 3.2E-10 3.7E-10 Pr- 143 I 0 0 Pr-144 2.OE-10 2.3E-10 Nd-147 1.OE-9 1.2E-9 W-187 3.lE-9 3.6E-9 Np-_39 9.5E-10 I_.lE-9 Table E-6 of Reg Guide 1.109.

Information l PNP-6010.OSD.OO1 Rev. 16 Page 46 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.2 l Radioactive Liquid Effluent Monitoring Instruments Pages:

46 - 47 j

Instrument Minimum Applicability Action Channels Oper able'

1. Gross Radioactivity Monitors Providing Automatic Release Termination a Liquid Radwaste \ (1)# At times of release 1 EffluentLine (RRS-1001) l

b. Steam Generator (1)4 At times of release+* 2 Blowdown Line (R-19, DRS 3/4100 +) I

c. Steam Generator (1)# At times of release 2 BlowdownTreatment Effluent (R-24, DRS 3/4200 +)

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination

a. Service Water (1) per At all times 3 System Effluent Line(R-20, train i R-28, WRA 3/4500 and WRA 3/4600 +)

3. Continuous Composite Sampler Flow Monitor

a. Turbine Building Sump (1) At all times 3 Effluent Line 4 Flow Rate Measurement Devices

a. Liquid Radwaste Line (RFI-285)

b. Discharge Pipes*

c. Steam Generator Blowdown Treatment Effluent (DFI-352) applicable. This is Pump curves and valve settings may be utilized to estimate flow; in such cases, Action Statement 4 is not primarily in reference to start up flash tank flow.

instrument as g OPERABILITY of RRS-1001 includes OPERAIL1TY of sample flow switch RES-1010, which is an attendant by Technical Specification 1 6. This item is also applicable fcr all Eberline liquid monitors (and their respective flow defined switches) listed here.

should apply v* Since these monitors can be used for either batch or continuous release the appropriate action statement of I or 2 (that is, Action I if a stearn generator drain is being performed in lieu of Action 2).

can fulfill the Westinghouse cR) radiation monitors arc beng replacd by Ebertine (DRS & NRA) .nonitors Either monitor operability requirement.

THEN a IF an RMS monitor is inoperable solely as the result of the loss of its control room alarm annunciation, surveillance one of the following actions is acceptable to satisfy the 00CMI action statement compensatory requirement:

I: Collect grab samples and conduct laboratory analyses ier the specific monitor's action statement,

-OR-2 Collect local monitor readings at a frequency equal to or -reater than (more frequently than) -he action frequency.

Information I P 6IPo010.OS.00 I Rev. 16 i Pae 47 of 84 OFF-SITE DOSE CALCIJLATION IMANTUAL Radioactive Liquid Effluent Monitoring Instruments I Pages:-

Attachment 3.2 46- 4 I-than) the action

2. Collect local monitor readings at a frequency equal to or greater than (more frequently frequency.

annunciation, TE'N' the only IF the RIMS monitor is inoperable for reasons other than the loss of control room samples and conducting laboratory analyses as the reading is equivalent to a acceptable action is taking grab grab sample when the monitor is functional.

TABLE NOTATION OPERABLE Action 1 With the number of channels OPERABLE less than required by the Minimum Channels requirement, effluent releases may continue, provided that prior to initiating a release:

and:

1. At least two independent samples are analyzed in accordance with Step 3.2.3a verify the

2. At least two technically qualified members of the Facility Staff independently valving Otherwise, suspend release of radioactive effluents via this pathway.

discharge OPERABLE Action 2 With the number of channels OPERABLE less than required by the Minimum Channels grab samples requirement, effluent releases via this pathway may continue for up to 30 days provided 10-7 pICi/grarn:

are analyzed for gross radioactivity (beta or garruna) at a limut of detection of at least

> 0.01 jiCi/gram

1. At least once per shift when the specific activity of the secondary coolant is DOSE EQUIVALENT 1-131.

is

0.01 piCUgram

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant DOSE EQUIVALENT 1-131.

OPERABLE Cl Action 3 With the number of channels OPERABLE less than required by the Minimum Channels that at least requirement, effluent releases via this pathway may continue for up to 30 days provided gross radioactivity (beta or gamma) at a once per shift, grab samples are collected and analyzed for lower limit of detection of at least 10-7 piCi/ml in accordance with the following:

THEN IF the Westinghouse monitor (R-20 and/or R-28) is fulfilling the applicability requirement,since the Spray Heat Exchanger is in service grab samples are only needed if the Containment have no auto trip Westinghouse ESW monitors are only used for post LOCA leak detec:-on and function associated with them.

OR requirement, IF the Eberline monitor (WRA-314500 and/or WRA-3/4600) is fulfilling the applicability is inoperable and the applicable train of ESW THEN grab sampling is required whenever the monitor is in service since this monitor is located in the system effluent.

OPERABLE Action 4 With the number of channels OPERABLE less than required by the Minimrnum Channels the flow rate requirement, effluent releases via this pathway may continue for up to 30 days provided is estimated at leas= once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Comnensatory acuons are governed by PMP403O.E]S.001, Evena-initiated Surveillance Tesung

l PMP-6010.OSD.001 Rev. 16 Page 48 of 84:

Information OFF-SITE DOSE CALCULATION MANUAL I

Radioactive Liquid Effluent Monitoring Pages:

Attachment 3.3 Instrumentation Surveillance Requirements 48 - 49 Instrument . CHANNEL SOURCE CHMNNEL CHANNEL CHECK CHEiCK CALIBRA-TION FUN9CTIOlNAL

1. Gross Radioactivity Mvonitors Providing Automatic Release Termination

a. Liquid Radwaste D l P R(3) Q(5)

Effluent Line (RRS-1001)

b. Steam Generator DM R(3) Q(l)

Blowdown Effluent Line

c. Steam Generator D* M R(3) Q(M)

Blowdown Treatment Effluent Line

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination a Service Water D MI R(3) Q(2)

System Effluent Line 3 Continuous Composite Samplers D* NIA NA N/A

a. Turbine Building _

Sump Effluen Lin

4. Flow Rate Measurement Devices a~Liquid Radwaste D(4)* NIARQ Effluent D(4)* N11A Ni/A NIA

b. Steam Generator Blowdown Treatment, Line 0 During rcases via this pathway

PNIP-6010.OSD.OO1 I Rev. 16 Page 49 of 84 Information OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Attachmen 3 48 - 49 Instrumentation Surveillance Requirements TABLE NOTATION isolation of this pathway and control I Demonstrate with the CHANNEL FUNCTIONAL TEST that automatic room alarm annunciation occurs if any of the followmgconditions exists.

1. Instrument indicates measured levels above the alarn/trip setpoint.

2. Circuitfailure.*

3. Instrument indicates a downscale failure.*

4 Instrumentcontrolnot set in operatmgmode *

5. Loss of sample flow #

alarm annunciation occurs if any

2. Demonstrate with the CHANNEL FUNCTIONAL TEST that control room of the following conditions exists:

I Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operating mode.

S. Loss of sample flow. 9 with traceability back to the 3 Perform the initial CHANNEL CALIBRATION using one or more sources calibratingthe system over its National Institute of Standards and Technology(NIST). These sources permit CALIBRATION, sources that intended range of energy and measurementrange. For subsequent CHANNEL have been related to the initial calibrationmay be used.

Perform the CHANNEL

4. Verify indication of flow during periods of release with the CHANNEL CHECK. batch releases are made.

continuous, periodic or CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which isolation of this pathway and control

5. Demonstrate with the CHANNEL FUNCTIONAL TEST that automatic exists:

room alarm annunciation occurs if any of the following conditlions

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.**

3. Instrumentindicates a downscale failure **

4. Instrument control not set in operatingmode.*

5. Loss of sample flow.

Instrument indicates, but does not provide for automatic isolation is taken for the automatic

K Instrument indicates, but does not necessarily cause automatic isolation. No credit isolation on such occurrences.

Applicable only to Eberline sample flow instrumentation Operations currently performs the mutine channel checks and source checks. Maintenance and Radiation Protection perform functional tests Chemistry performs the channel check on the continuous comnpOaite sampler.

channel calibrations and channel These responsibilitlies are subject to change without revision to this document.

Information PMP-6010.0SD.00l1 Rev. 16 l Page 50 of 84 OFF-SITE DOSE CALCULATION MAINUAL Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation P a ge Instrument (Instrument #) Operable' Minimum Action Channels Action

1. Condenser Evacuation System

a. Noble Gas Activity (1) 6 lonitor (SRA-1905r2905)

b. Flow Rate Monitor (SFR-401, (1) I 5 1/2-MR-054 and/or SRA- 1910/2910)

2. Unit Vent. Auxiliary Building Ventilation System

a. Noble Gas Activity (1)

6 Monitor (VRS-1505/2505)

b. Iodine Sarmpler (1) *3 Cartridge for VRA-1503/2503

c. Particulate Sampler Filter (1) for VRA-1501/2501

d. Effluent System Flow Rate (I)

Measuring Device (VFR-315, MR-054 andlor VFR-1510/2510)

e. Sampler Flow Rate ( 5 Measuring Device (VFS-1521/2521)1

3. Containmnent Purge and Containment Pressure Relief (Vent)

a. Containment Noble Gas Activity Monitor (1) ****2 3 7 ERS-13/1405 (ERS-23/2405)

7

b. Containment Particulate Sampler Filter (1) 10 C2 ERS-1311401 (ERS-23/2401)

4. Waste Gas Holdup System and CVCS HUT

a. Noble Gas Activity (1) 1 ***4 9 Alarm and Termination of Waste Gas Releases (VRS-1505/2505) _ _

5. Gland Seal Exhaust ]

a. Noble Gas Activity Monitor (SRA-1805/2805)

b. Flow Rate Monitor (SFR-201, MR-054 or SFR-1910/2810)

At all times

1*1 During releases via this pathway

TABLE NOTATIONS

1. IF an RMS monitor is inoperable solely as the result of the loss of its control room alarm annunciation, THEN one of the following actions is acceptable to satisfy the ODCMI action statement compensatory surveillance requirement.

1. Take grab samples and conduct laboratory analyses per the specific monitor's action statement,

-OR-

2. Take localmonitorreadings at a frequencyequal to or greaterthan (more frequertlythan) the action frequency the only IF the RMS monitor is inoperable for reasons other than the loss of control room annunciation, THEN to a grab acceptable action is taking grab samples and conducting laboratory analyses as the reading is equivalent sample when the monitor is functional.

2. Consider releases as occurring "via this pathway" under the following conditions The ContainmentPurge System is in operationandContainmnentintegrityis established~required,

-OR-

The Containment Purge System is in operation and is being used as the vent path for the venting of contaminated systems within the containment building prior to completing both, degas and depressurizationof the RCS.

system and IF neither of the above are applicable,THEN the containment purge system is acting as a ventilation is covered by Item 2 of this Attachment-

-OR-

A ContainmentPressure Relief (CPR) is being performed

3. For purge (including pressure relief) purposes only. See Technical Specification table 3 3-6 for additional information.

for additional

4. For waste gas releases only, see Item 2 (Unit Vent, Auxiliary Building Ventilation System) requirements ACTIONS

5. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirementeffluent releases via this pathway may ccntinue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . After 30 days, IF the channels are not OPERABLE, THEN continue releases with corrected in estimationof the flow rate once per4 hours and providea descriptionof why the inoperabilitywas not the next Annual RadiologicalEffluent Release Report requirement,

6. With the number of channels OPERABLE less required by the Minimum Channels OPERABLE least once per effluentreleases via this pathway may continue for up to 30 days provided grab samples are taken at are not shift and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 30 days, IF the channels why the continue releases with grab samples once per shift and provide a description of OPERABLE, THEN inoperabilitywas not correctedin the next Annual RadiologicalEffluentrelease Report.

7. With the number of channels OPERABLE less than required by the Mvlinunum Channels OPERABLE requirements, immediately suspend PURGING or VENTING (CPR) of radioactive effluents via this pathway.

effluent S With the number of channels OPERABLE less thdn required by the vininmum Chamels OPERABLE requiremert, up to 30 days provided samples required for weekly analysis are releases via the affected pathway may continue for continuously collected with auxiliary sampling equipment as required in Attachment 3.7. Radioactive Gaseous Waste Sampling and Analysis Program. After 30 days. IF the channels are not OPERABLE, THEN continue releases withthe sample next collection by auxiliary sampling equipment and provide a descnpuion of why the inoperahility was not corrected in Annual Radiological Effluent Release Report.

Sampling evolutions are not an interruption of a continuous release or sampling pcncd.

9. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, die contents of the tank(s) may be released to the environment for up to 14 days provided that pnor to initiating the release:

a. At least two independent samples of the tank's contents are analyzed and,

b. At least two technically qualified members of the Facility Staff independently verify the release rate calcularnons and discharge valve lineups; otherwise, suspend release of radioactive effluents via this pathway.

Co

10. See Technical Specification 3.4.6.1.

Compensatory actions are governed by PMP-4030.EIS.001. Event-Initiated Surveillance Testing.

Information I PNpP-6010.OSD.001 I Rev. 16 Page 53 of 84 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitorng Pages Attachment .5 1 Instrumentation Surveillance Requirements 53 - 54 System D*M R(') QM a Noble Gas Activity Monitor (SRA-1905/2905)

Dt NA R Q b System Effluent Flow Rate (SFR-401, MR-054, SRA-1910/2910)

2. Auxiliary Building Unit Alarm Only Ventilation System

a. Noble Gas Activity Montor D R(2) Q(l)

(VRS- 1505/2505)

W'Y NA NA NA

b. Iodine Sampler (For VRA-1503/2503)

W* NA NA NA

c. Particulate Sampler (For VRA-1501/2501)

D NA R Q

d. System Effluent Flow Rate Measurement Device (VFR-315, MR-054, VRS-15 10/2510)

D N/A R Q

e. Sampler Flow Rate Measuring Device (VFS-1521/2521)

3. Contamnment Purge System and Alarm and Trip Containment Pressure Relief S** P R(2) Q C2

a. Contairnient Noble Gas Activity Monitor (ERS-13/1405 and ERS-23J2405)

S** NA R Q b Containment Particulate Sampler (ERS-13/1401 and ERS-23/2401) .-

4. Waste Gas Holdup System Alarm and Trip Including CVCS HUT P* l R(2) Q(3)

a. Noble Gas Activity Monitor Providing Alarm and Termination (VRS-1 505/2505) 5 Gland Seal Exhaust Alarm Only

a. Noble Gas Activity (SRA-1805/2805)

D** M l R(2) 1 I

Q(l)

D* NA Q

b. System Effluent Flow Rate (SFR-201. MR-054, SRA-1810/2810)

At all tunes

- During releases via this pathway

Information l PMP-6010.OSD.001 I Rev. 16 l Page 54 of 84 OFF-SITE DOSE CALCULATION NMNUAL l Radioactive Gaseous Effluent Monitoring Pages:

l Atachment35D Instrumentation Surveillance Requirements 53 - 54 TABLE NOTATIONS 1 Demonstrate with the CHANNEL FUNCTIONAL TEST that control room alarm annunciation occurs if any of the following conditions exists-

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument contro Is not set in operate mode.

2. Perfo rm the initial CHANNEL CALIBRATION using one or more sources with traceability back to the NIST These sources permit calibrating the system over its intended range of energy and measurement range For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

3. Demonstrate with the CHANNEL FUNCTIONAL TEST that automatic isolation of this pathway and control room alarm annunciationoccurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarmn/trip setpoint.

2. Circuit failure.*

3 Instrumentindicates a downscale failure.*

4. Instrumentcontrolsnot set in operate mode *'

Instrument indicates, but does not provide automatic isolation.

Operations currently performs the routine channel checks, and source checks Maintenance and Radiation Protection perfcrrn channel calibrations and channel functional tests These responsibilities are subject to change without revision to this documeat.

_J

Information PMP-6010.OSD.001 Rev. 16 f Page 55 of 84 OFF-SITE DOSE CALCULATION TMANUAL Attachment 3.6 Radioactive Liquid Waste Sampling and Analysis Program Pages:

55 - 56 jRef 5.2.1ls I LIQUID SAMPLING VIININIMUM TYPE OF LOWER RELEASE FREQUENCY ANALYSIS ACTIVITY LIMIT OF TYPE FREQUENCY ANALYSIS DETECTION (LLD)

(p.Ci/ml) 2 A. Batch Waste P P Principal Gamma 5x1-Release Tanks ' Each Batch Each Batch Emitters' 1-131 lxO-1 P P Dissolved and Entrained Gases Each Batch Each Batch (Gamma 1x10 Emitters) p M H-3 1x10 51 Each Batch Composite b Gross Alpha lx10 7 P Q Sr-89, Sr-90 5xlQ0-Each Batch Composite b Fe-55 1x10 6 B. Plant W Principal Gamma Continuous Daily Composite b Emitters 5x10 Releases d 1-131 1xl04 M M Dissolved and Grab Sample Entrained Gases 1XO-(Garnma Emitters)

M H-3 lxlO-Daily Composite b Gross Alpha 1x10 7 Q Sr-89, Sr-90 5x10 4 Daily Composite b Fe-55 ix1O6

During releass via this pathway

TABLE NOTATION of

a. The lower limit ofdetection (LLD) is defined inTable Notation A of Attachment 3.20, Maximum Vaiues or Lower Limits DetecuonsAB- REMSIP and

b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged in which the method ofsampling employed results in a specimen which is representative of the liquids released.

c. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, isolate, recirculate or sparge each batch to ensure thorough mixing.

flow

d. A continuous release is the discharge of liquid of a non-discrete volume; e g. from a volume of system that has an input during the continuous release.

Fe.

e. The principal garnma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, 59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-1 44 This list does notmean that only these nuclidesnuclides are to be detected and reported. Identify and report other peaks, which are measurable and identifiable, together with the above

Information PIP-6010.OSD.001 j Rev. 16 Page 57 of 84 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Waste Sampling and Pages:

A Analysis Program 57 - 58 Gaseous Release Type Frequency Mylinimum Tpe of Lower Limit Analysis Activity of Detection Frequency Analysis (PiCi~m]) a

a. Waste Gas Storage Tanks P Principal Gamma and CYCS HUTs Each Tank Each Tank Emitters d 1 lx Grab Sample

b. ContamnmentPurge P Principal Gamma Each Purge Each Purge Emitters d x 10o4 Grab Sample CPR (vent)* Twice per Twice per Month Month H-3 I x 1F

c. Condenser Evacuation W or M M Principal Gamma System Grab Sample Particulate Sample Emitters d I x 10o-Gland Seal Exhaust* M H-3 1 x 106 W e Principle Gamma I x 104 Noble Gas Emitters d M 1-13i Iodine Adsorbin- Ix 1o01 Media Continuous W g Noble Gases Noble Gas Monitor I x 104

d. AuxiliaryBuildingUmt Continuous' 'b 13 Vent* lodineAdsorbin, 1 x 10.12 Media Continuous'c W Principal Gamma ParticulateSample Emittersd I x 10-1 Continuous' M Gross Alpha Composite Particulate I x 01 Sample W W il17-3 Grab Sample H-3 Sample I x 104 W PrincipleGamma I x 104 Noble Gas Emitters d Continuous' Q Sr-89, Sr-90 CompositeParticulate I I x 10"-

Sample Ccntinuous: Noble Gas Monitor l Noble Gases I x 0IV

e. incinerated~il' P Princiual Gamma I Each Batch' Each Batch l Emitters l 5 x l0-

"During releases via this pathway

- Only a twice per month sampling program for containment noble gases and HI is required

Information PrYIP-6010.OSD.OO1 Rev. 16 t Page 53 of 84 OFF-SITE DOSE CALCULATION MAkNUAL Radioactive Gaseous Waste Sampling and Pages.

Attachment 3.7 Analysis Program 57 - 53 TA13LE NOTATION

a. The lower limit of detection (LLD) is defuled in Table 'otation A of Attachment 3 20, Maximum Values for Lower Lmits of DetecnonsA.IB - REMP.

b Change samples at least once per 7 days and comolete analyses withn 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing Perforrn analyses at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days following each shutdown, startup or THERMAL POWER change > 15% per hour of RATED THERMAL POWER. WHEN samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, THEN the corresponding LLDs may be increased by a factor of

10. This requirement does not apply IF (I) analysis shows that DOSEQ 1131 concentration in the RCS has not increased more than a factor of 3, and (2) the noble gas monitor shows that tffluent activity has not increased more than a factor of 3 [Rer. .Jryl

c. Know the ratio of the sample flow rate to the sampled stream flow rate for the time period covered by each dose or dose rate calculation made in accordancc with steps 3 2 4a, 3.2 4b, and 3 2 4c of his document.

Sampling evolutions are not an interruption of a continuous release or sampling period.

d The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-97, Kr-88, Xe-133, Xe-133M, Xe-135 and Xe-1338for gaseous emissions and in-54, Fe-59. Co-53, Ca-60, Zn-6S. Mo-99, Cs.13-4, 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.

- Identify and report other peaks, which are measurable and identifiable, together with the above nuclides.

e. Releases from incinerated oil are discharged through the Auxiliary Boiler System Account for releases based on pre-release grab sample data f Collect samples of waste oil :o be incinerated from *he container in which the waste oil Lsstored (example: waste oit stcrage tanks.

55 gal drums) prior to transfer to Lhe Auxiliary Boiler System. Ensure samples are represntative of container contents.

g. Obtain and analyze a gas marinelli grab sample weekly for noble gases effluent quantification h Take intium grab samples at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the resueling cavity is flooded

i. Grab sampling ofthe Gland Seal Exhaust pathway need not be perforned if the RMS low range channel (SRA-180512505) readings are less than IE-6 pClcc. Attach the R.MIS daily averages in lieu of sampling. This is based on operating experience indicating no activity is detected in the Gland Seal Exhaust below this value. Compensatory sampling for out of service monitor is still required in the event 1805/2805 is inoperable.

Information ) PMP-6010.OSD.0O1 Rev. 16 Page 59 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.8 Multiple Release Point Factors for Release Points l Page Liquid Factors Mvlonitor Descnption Mvlonitor Number l RP j U I SG Blowdown IR19t24, DRS; 300/3200l 0 35 U 2 SG Blowdown 2R19F24,DRS 4lGOI4200* 035 U I &2Liquid WasteDischarge RRS-1000 0.30 Gaseous Factors Monitor Description Monitor Number Flow Rate (cfm) MRP ,

Unit I Unit Vent VRS-1500 186,600 054 Gland Seal Vent SRA-1800 1,260 0.00363 Steam Jet Air Ejector SRA-1900 3,600 (b) 0.01 StartUp FTVent 1,536 0 004 Total 192,996 Unit 2 Unit Vent VRS-2500 143,400 041 Gland Seal Vent SRA-2800 5,508 (a) 0 02 Steam Jet Air Ejector SRA-12900 3,600 (b) 0.01 Start Up FT Vent 1,536 0.004 Total 15i,044 Either R-19, 24, DRS 314100 or 3/4200 can be used for blowdown monitoring as the Eberline monitors (DRS) ar replacing the Westinghouse (R) monitors.

Nominal Values a Two release points oal2,754 cfm each are totaled for this value.

b This is the total design maximum of the Stan Up Air Ejectors. This is a conservative value for unit 1.

Informantion I PMIP-6O01.OXSD.001 I Rev. 16 l Page 60 of 84 OFF-SITE DOSE CALCULATION MANUAL Attacluent 3.9 Liquid Effluent Release Systems l 60

Information l PMP-6010.OSD.001 Rev. 16 Page 61 of 84 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.10 Plant Liquid Effluent Parameters 61 SYSTEM COMPONENTS CAPACITY i FLOW RATE TANKS lPUMPS (EACH) (EACH)*

I Waste Disposal System

+ Chemical Drain Tank 1 1 600 GAL. j 20 GPM

+ Laundry & Hot Shower Tanks 2 1 600 GAL. 20 GPM

+ MonitorTanks 4 2 21,600,GAL.l 150 GPM

+ Waste Holdup Tanks 2 20 GAL.

+ Waste Evaporators 30 GPM

+ Waste Evaporator Condensate 2 2 6,450 GAL 150 GPM Tanks II Steam Generator B lowdown and Blowdown Treatment Systems

+ Start-up Flash Tank (Vented)#

I

_ _ _____GP_

,800 GAL. [ 580 GP l

+ Normal Flash Tank (Not 1 525 GAL. 100 GPM Vented) _

+ Blowdown Treatment System 1 60 GPM III Essential Service Water System

+ Water Pumps l 4 10,000 GPM

+ Containment Spray Heat 4 3,300 GPMvI Exchanger Outlet IV Circulating Water Pumps Unit I 1 230,000 GPMl Unit 2 4 230,000 GPM Nominal Values The 580 gpm value is calculated from the Estimated Steam Generator Blowdown Flow vs. DRV Valve as Position letter prepared by M. J. O'Keefe, dated 9127193. This is 830 gpm times the 700,% that remains licuid while the other 30% flashes to steam and exhausts out the flash tank vent.

Information t PNIP-6010.OSD.001 I Rev. 16 7 Page 62 of 84 OFF-SITE DOSE CALCULATION MANUAL .11 Volumetric Detection Efficiencies for Principle Gamma Page:

I Emitting Radionuclides for Eberline Liquid Monitors 62 This includes the following monitors: RRS- I 000. DRS 3100, DRS 3200, DRS 41 0, DRS 4200, WRA 3500, WRA 3600, WRA 4500 and WRA 4600. (Ref.5 2 .1p]

NUCLIDE EFFICIENCY (cpnimLCiicc) 1-131 3.78 E7 Cs-137 3.00 E7 Cs- 134 7.93 E7 Co-60 5.75 E7 Co-58 4.58 E7 Cr-51 3.60E6 I n-54 3.30 E7 Zn-65 153E7 Ag-I IOM 9.93 E7 Ba-133 4.35 E7 Ba-140 1.92 E7 Cd-109 9.55 ES Ce-139 3.2 E7 Ce- 141 1.92 ES Ce- 144 4.33 E6 Co-57 3.80 E7 Cs-136 I.07ES Fe-59 2.33 E7 Sb-124 I 93 E7 1-133 3.40 E7 1-134 723 E7 1-135 3.95 E7 hlo-99 8.63 E6 Na-24 4.45 E7 Nb-95 3.23 E7 Nb-97 3.50 E1 Rkb89 5.t0 E7 Ru-103 3.48 E7 Ru-106 1.23 E7 Sb-122 2.55 E7 Sb-125 3.15 E7 Sn- 113 7.33 E5 Sr-85 3.70 E7 Sr-89 2.3S E3 Sr-92 3.67 El Tc-99M 3 60 E7 Y-88 5_25 E7 Zr-95 3.38 E7 Zr-97 3.10 E7 Kr-85 156 E5 Kr-85M 3 53 E7 Kr-83 4.10 E7 X-i J-1311 3.15 E5 Xe- 133 7.73 E6 Xe-133My 5.75 E6 j Xe-135 3 S3 E7

F ( - ( [F F r (I1 [ -

PNP-6010.OSD.001 I~v 6 1ag3 ofL Informato OF'-SITi, DOSE CALCIJLATION MANUAL Counting Efficiency Curves for R-19, and R-24 Pages3 At~achnent 3.12 3-6 Coulntilig Efficiency Curve for R-19 Efficiency Factor = 4.2 E6 cpm/uCilml d tAlsikcn during pre operacionol testing widb Cs-137)

(Bascd on cmnpuical 1 ofE if7 1 oOE406 1 ooE#oS 0

1 OOE+04

.n 0

.4 . -0 I ODE4O3 t,

E I o0E+02 i ooEto1 1 DOE40C 9 9 9 s W

UIIJ si LIULL _ _ 8_

Im___ . __ _ I microcurles/ml

Information PMP'-6010.0SD.001 ev. 16 Page 64 of 84 COFF-SITE DOSE CALCULATION MANUAL 3CPages:

AttachrnenL 3.12 Counting Efficiency Curves for R-19, and R-24 633- 64 Counting Efficiency Curve for R-24 Efficiency Factor =7 5E6 cpm/uCilnil (Bascd on cmpmu. d a taw duiulg pe-opcramoul wLcs, li Mn-54) wih 1 OOEiO7 I OOE-06 1 OOEtf5 0) t,)C I OMEI04 x

.0 1 OOEtO3 (U

1 OOE+02 I OOE+0l 0

1 DOEtOO U)

C-) c'J 0 U:. n 0 0 Lii 9' III uJ 0) 00 0 0 0 0D 0 8

microcuries/ml I f I t I (( l[ I I I

r- -- I - I -- I--- -

( (

MP6EN10.SD.O1 Rev. 16 Information OFF-SITE DOSE CALCULATION MANUMX Attachnment 3 13 Counting Efficiency Curve for R-20, and R-28 Counting Efficitncy Curve for R-20 and R-28 Efficiency Factor = 4 3 E6 cpmluCi/ml

.ca (Bgscdon o.p.r B..u iaks.odoing pr.-opera.1oeI lostsal widhCo 5J) 1 ODE*07 1 OOE 05 - -. __ _____- _ ._

0 1 OOEO4 04 -

I OOE+03 1 OOE+02

-rEs0i II 9 0 .. 0 9 1m u,ooEo... ll XU.!

mnicrocurlesiml

Information I PMP-6010.OSD.OO1 I Rev. 16 l Page 66 of 84 OFF-SITE DOSE CALCULATION MANUAL Paae: 14 Gaseous Effluent Release Systems Page 66

.da.rg I L1

Information PMIP-6010.OSD.O01 l Rev. 16 I Page 67 of 84 OFF-SITE DOSE CALCULATION MANUAL Pa-e 67 Attachment 3.1 5 _Plant Gaseous Effluent Parameters 67 SYSTEM UNIT EXHAUST CAPACITY FLOW RATE (CF

____ MI)

I PLANT AUXILIARY BUILDING 1 186,600 max UNIT VENT 2 143,400 max WASTE GAS DECAY TANKS (8) AND 1 125 4082 FT G100 psig CHEMICAL & VOLUME CONTROL 28,741 t3 max SYSTEM HOLD UP TANKS (3) @ S#, 0 level

+ AUXILIARY BUILDING 1 72,660 EXHAUST 2 9,400

+ ENG. SAFETY FEATURES I &2 50,000 VENT FUEL HANDLING AREA VENT 1 30,000 SYSTEM_

CONTAINMENTPURGE SYSTEM 1&2 f 32,000 CONTAINMENT PRESSURE 1& 2 1.000 RELIEF SYSTEM INSTRUMENT ROOM PURGE 1&2 1,000 SYSTEM II CONDENSER AIR EJECTOR SYSTEM 1 2 Release Points One for Each Unit NORMAL STEAM JET AIR 1 &2 230-EJECTORS START UP STEAM JET AIR 1&2 3,600 EJECTORS.

III TURBINE SEALS SYSTEM I r 1,260 2 Release Points 2 5,508

_ _for Unit 2 IV START UP FLASH TANK VENT 1 1,536 l _l 12 1,536 _ _l

+ Designates total flow for all fans.

GRO

^X/Q IJND AVERAGE (sec/mrn)

DIRECTION l DISTANCE (5METERS)

(WIND FROrM) 594 12416 4020 5630 J7240 -

N 3.50-0 4.23E-07 1.97E-07 l. 16E-07 S.13E-08 NNE 2.69E-06 1 3.22E-07 ji.3 -07 9.16E-03 .44E-NE 3 o4E- 4.51E-07 2.1 0E-7 1.33E-0T 7 9,43E-0S ENE 94E-06 S 6.70E-07 13 35E-07 2.d7E-07 I 48E-07 68E- 9.SE-0 4.84E-07 1 3.03E-07 j 2.17E-07 ESE S.45E-06 9.36E-07 1.75E-07 I2.96E-07 3.31£-01 2.12E-07 2 42E-07 j SE 5E-06 91 I.3O5E-06 5.38E-07 SSE 1 I9E-O5 i20E-06 6 14E-07 .3.86E-07 2.77E-07 j S_ 1. 6E-0 130E-06 o53E-0-1 4.05E-07 .39E-07 SSW 5.S7E-06 6.70E-07 13.30E-07 2.IIE-07 1.43E-07 SW 3.66E-06 4.26E-0 12 04- .23E-07 j64E-08 2 84E-06 3 14E-07 1.50E-0 1l 57E-07 6 632E-08 WSW 3 29E-06 3 69E-07 175E0 1.04E-07 7.32E-08 WNW 3. 6 3.61E-0 1.69- 7 OIE-07 7E NW - T2.98E-Oo 3T33E-07 .5SE-07 44E-08 6.61E-0 8 3.SIE-7 1.78E-07 I 1.06E-07 l 741E-08_

jNNW 3E-0 DIRECTION DISTANCE (METERS) 12067 24135 40225 56315 80500 (WIND FROM) 1

.55E-08 7.71 E-09 4 93E-09 3.09E-09 4.03 E-N NNE 3 23E-08 1-26E-08 6.27E-u9 4.01E-09 12.2E 09 NE 1 1E-OS 1.9 9.52E-09 l6.1 IE-09 3.88E-09 T 7.59E-08 3.08E-08 1.55E-08 9.Y5 E-09 97E-09 ENE EII -07 4.62E-08 2.33E-08 1.50E-08 j 9.64E-__9_1 ESE j .LOE-07 4.30E-08 2.27E-08 I 1.46E-08 9.38E-09 SE T2F1 5.0£0x.6£0 l .s2E0E.-08 l .9 8 j

Sl 44E-0I 3.94E-08 2.99E-08 F_1 93 -8 1I1.24E-08 S 1.50E-07 6 .099-08 3.06E-08 'I 97E-O j ;_I 26E-08_j SSW SW WSW 7l1.3E-08 44.,5E-08 3.18E-08 2 294E-08 17?2E-08 11 25E-O8 l.47E:48

.56E-9 I6.22E-09 9 939E-09 l i.48E-09 l3.99E-09 I 5.Y7E-O9 3.47E-09 2.53E-09 W i 3 66E-08 F 143£-08 l 7.07E-09 j4 S5E-09 T 2.85E-0

%VNW l_3.SOE-08 l__TY3 6.70E-09 4.28E-09 l _2.69E-09 NW j3.30E-OS I 1.28E-08 638E-09 4.0YE-09 2.57E-09 NNW i6E-08 1.43E-l 7 08E-09 l4.54E-09 l2.35E-09 Worst Case ^X/Q = I 54E-5 sec/m 3 in Sector A 1993

Information I PIMP-6010.0SD.OOI I Rev. 16 Page 69 of 84 OFF-SITE DOSE CALCULATION MAINUAL Attachment 3.16 10 Year Average 10 of 1989-1998 Data 6Pag-es:-69 2

D/Q DEPOSITION (1/rm )

DlRECTION DISTANCE (METERS)

(lIND FROM) 594 2416 4020 5630 I 7240 2 3SE-09 E 5 66E-I L) 3.62E-10 IN 1-2.46E-08 1.j5E-10 NNE 1 06E-0S 1.02E-09 4.62E-10 2 43E-10 NE 1.31E-08 1727£-09 5 75E-10 3 02E-10 1.93E-10 1.56E-09 7.09E-10 3 72E-10 237E-10 ENE E l.92E1 I.B5E-09 8.39E-10 4.4E-10 2 SIE-1O ESE 1.82E- I 76E-09 7.98E-10 4.19E-10 1 267E-10 SE 1.85E-OS 1.79E-09 8 09E-10 4.95E-10 2.71E-10 SSE 2.24E-0 2 17E-09 9.84E-10 5.ISE-10 3.29E-10 S 3.5E-OS 3.38E-09 J.53E-09 S.03E-I0 5.1,E-10 SSW 2.31E-08 2.24E-09 1.01E-09 5.31E-10 3.39E-10 SW 2.14E-0S 2.07E-09 9.38E-10 4.91E-10 3.14E- I0 WSW 2 08E-08 2.01E-09 9.12E-10 4.78E-10 .05E-10 W 2.13E-08 T2.06E-09 9.33E-10 4.9E-10 3.13E-10 1 96E-08 j 58E-09 85E144S-0 26-0 NW 1.62E-08 I 1.57E-09 7-I IE-10 3.73E-10 2.38E-lO 2.1 1E-09 056E-I 5.0E-10 3.2E-10 NNW I DIRECTION DISTANCE (METERS)

(WIND FROM) 12067 24135 40225 56315 80500 Il.SIE-10 j4.91E-11 L.81E-11 9 65E-12 4.84E-12 N

-6 78E-1I j 2 IE-11 7.75E-12 l 4.13E-12 l 2.07E-12 NNE 18.18E-11 2.62E-I1 964E-12 l5.1SE-12 2.58E-12 NE 9995E-1I 3.23E-I1 1.19E-1 6.34E-12 3.18E-12 ENE 1.16E-10 3.82E- I1 1.41E-I1 75E-12 l 3.76E-12 E

ESE l 1.12E-10 3.64E-11 1 .34E-11 7.14E-12 3.58E-12 SE l.13E-lo 3.68E-1I 1.36E- I l 7 '4E-12 l 3.63E-12 SSE 1.37E-10 4.47E-11 1.65E-11 8.79E-12 1 441E-12 S l 2.14E-10 6.97E-1 l 2.57E-I I 1.37E-1 1 1 6 87E-12 l1.42E-10 14.6 1E-1II 1.7E- II 906E-12 4.54E-12 SSW SW I 1 31E-10 1427E-11 -T1.57E-1 II 8.38E-12 [ 4.21 E-12 WSW r1.27E-I0 1 4.15E-11 1.53E-1 1.56E-11 1 89.16E-12 1.73E-1 1

4.09E-12 4.19E-12 W 1.3E-10 4.25E-1 2

1.19E-10 l3.89E-1 i .43E-1 7.64E-12 j 3.83 IIWNWj NlW 1.71SE-I0 l 324E- II 1.19E-1II 636E-12 I 3.19E-12 NNW 1.34E-10 4 35E-11 1.6E-II 8 55E-12 4.29E-12 DIRECTION - SECTOR IN A E = E S =N NNE =B ESE =F SSW =K W =P lNE =C SE G SW =L NW = Q ENE =D I SSE =H WSWV = M N-NW = R Worst Case DIQ = 4 41 E-OR l/rn in Sector A 1990

Information I PMIP-6010.OSD.O01 Rev. 16 R l Page 70 of 84 OFF-SITE DOSE CALCULATION IANUAL Attachment 3.17 Annual Evaluation of x/Q and D/Q Values For Page.

Atahet31 All Sectors 70 I Performed or received annual update of ; /Q and D/Q values. Provide a description of what has been received. I Signature Date Environmental Department (print name, title)

2. Worst /Q and D/Q value and sector determined. PMLP-6010.OSD.001 has been updated, if necessary. Provide

, an evaluation. I Signature Date Environmental Department (print name, title)

3. Review nuclide mix for gaseous and liquid release paths to determine if the dose conversion factor of total body is still applicable- Provide an evaluation.

Signature Date Environmental Department (print name, title)

4. Approved and verified by:

Signature Date Environmental Department (print name, tide)

5. Copy to NS&A for information. /

Signature Date Environmental Department (print name, title)

Information PMP-6010.OSD.001 I Rev. 16 Page 71 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.18 Dose Factors Pages:

I 7 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMAL-k AIR BETA AIR DOSE FACTOR FACTOR DOSE FACTOR DOSE FACTOR X, (DrB,) Li (DFS,) Mi (DF11) Nz, (DP',)

mrem m' (mrem m' (mrad ml (mrad m 3 RADIONUCLIDE per ptCi yr) per piCi vr) per tCi yr) per AiCi yr)

Kr-83m 7.56E-02 --- 1.93E+01 2.88E102 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E-.-03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73 E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.733E+04 1.06E+04 Kr-90 i.56E+04 7.29E+03 1 63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.II E+03 Xe-133m J 2.5 1E -02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m. 3.12E-:-03 7.1 1E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83EE+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E4--031 2.69E+03 9.30E+03 3.28E+03

The listed dose factors are for radionuclides that -nay be detected in gaseous effluents, from Reg. Guide 1.109, Table B-l

Information I PNIP-6010.0SD.001 I Rev. 16 1 Page 72 of 34 OFF-SITE DOSE CALCULATION MANLNUAL .13 Dose Factors Pages. -

1 71 - 72 DOSE FACTORS FOR RADIOIODINES AND RAD IOACTIVE PARTICULATE, IN GASEOUS EFFLUENTS FOR CHILD* Ref 52.leeandTfj P, _ .

INHALATION FOOD & GROUND)

P.ATHWAY PATHVAY (mrem m3 (mrem m 2see RADIONUCLIDE per gCi yr) per gCi yr)

H-3 1.12E+03 1.57E+03 P-32 2.60E+06 7.76E-t-I0 -

Cr-51 1.70E+04 1.20E+07 Mn-54 1.58E+06 1,12E+09 Fe-59 1 27E+06 5.92E+08 Co-58 _1.I1E+06 5.97E+O3 Co-60 7 07E+06 4 63E+09 Zn-65 9.95E+05 1.17E+]0 Rb-86 1.98E+05 88.73E+09 Sr-39 2.16E+06 6.62E+09 Sr-90 1.01E+08 1.12E+II Y-91 2.63E+06 6.72E+06 Zr-95 2.23E+06 3.44E+08 Nb-95 6.14E-t-O5 4.24E÷08 Ru-103 6.62E+05 1.55E+08 Ru-106 1.43E+07 3.O1E+08 Ag-tlOm 5.48E+06 1.99E+1I 0 I-131 1.62E+07 4.34E+I1 1-132 1.94E+05 1.78E+06 1-133 3.85E-06 3.95E+09 1-135 7.92E-05 I 22E+07 Cs-134 1.01E+06 400E4'10 Cs-136 1.71E--5 3.OOE+09 Cs-137 9.07E+05 334E+10 Ba-140 1.74E 06 1.46E+08 Ce-141 5.44E+05 3.31E+07 Ce-144 1.20E+07 l 1.91E- 08

'AsSr-90, Ru-106 and 1-131 analyses are perforned.TIIEN use P.given in ?-32 Fornonhsted radionuclides

' The units for both H3 factors are the same, irern mn'per J.LCyr

Information l PMP-6010.0SD.001 I Rev. 16 Page 73 of 84 OFF-SITE DOSE CALCULATION MANUAL Attahmet 3.19 Radiological Environmental Monitoring Program Pages:

Sample Stations, Sample Types. Sample Frequencies 73 - 76 rRef S 2.1v, 5 2.1x 52 It1 SANIPLE I DESCRIPTION/ SAMPLE SAIPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY ON-SITE AIRBORNE AND DIRECT RADL'\TION (TLD) STA-TIONS ONS-1 (1-1) 1945 fltl 18 from Plant Axis Airvonic Partitulatc Weekly Uross ileta j Wekly

[ Wcekly Gamma Isotopic I Quart Comp Airborne Radioiodire Weekly 1-131 I Weekly TLD Quarterly DirectrRadiation I Quarterly ONS-2 (T-2) 233 8 ft 48° from Plant .Axis Airborne Parrtculatc Weekly j Gross Beta l Weekly Weekly Gamma isotopic I Quart Com:

Air orneRadiotodine Weekly 1-131 I WeekLy TLD Quartcriy Direct Radiation Quarterly ONS-3 (r-3) 2407 It@ 90Y from Plant Axis AirborneParticulate Weekly Gross Beta Weekly Weekly I Gamma Isotopic Quart Comp Airborne Radioiudine Weekly 1-131 Weekly TLD Quartcrly Direct Radiation Quarterly ONS-4 tT-4) 1852 fLt I 13 from Plant Axis Airborne Particulate Weekly Gross Beea We:lky Weekly Gamma Isatopic Quart Comp Airbome Radwotod~nc WeckLy 1-131 Weekly TLD Quarterly Direct Radiatnon Quarterly ONS-5 (T-5j 1895 tt @ 189' trom Plant Axis Airborne Particulate Weekly Gross Beta Weekly Weekly Gamma Isotopic Quart Comp Airborne Radiotedinc Weekly 1-131 Weekly ONS-6 (T-6) 1917 ft@210 fromPlant Axis TLD Airboure Particulat:

Quarterly Weekly Weekly II Direct Radiation Gross Beta Gamm3 Isotopic Quarterly Weekly Quart Comp Airborne Radiosodinc Weekly 1-131 Weekly TLD Quarterly Direct Radiation Quarterly T-7 2103 ft@ 360 from Plant Axis TLD Quarterly Direct Radiation Quartcrly T-8 220S ft @ 820 frorn Plant Axis TLD Quarterly Direct Radiation QuanIerly T-9 1368 ft @ 1490 from Plant Axis TLD Quarterly j Direct Radiation Quarterly T-10 1390 ft ai 1270 from P!ant Axis TLD Quarterly Direct Radiation Quarterly T-1I 1969 it@ 11° from Plant Axis TLD Quarxry Direct Radiation Quarterly T- 12 2292 ft @ 63° from Plant Axis TLD Quarterly Direct Radiation Quarterly CONTROL AIRBORNE AND DIRECT RADIATION (TLD) STATIONS NBF 15 b miles SSW Airbome Particulate Weekly l Gross Beta Weekly New Buffalo, Nl Weekly Gamma Isotopic Quart Comp.

Atrborne Radioiodine Weekly I 1-131 Weekly TLD Quarterly I Direct Radiation IQuarterly SBN 262 Miles SE Airborne Particulate Weekly I Gross Beta I Weekly lSouth Bend, IN - Wcekly Gamma Isotopic C I QuarL Comp.

Airborne Radioiodine Weckly 1-131 I Weekly TLD Quartcrly Direct Radiation T Quarterly DOW 243 miles ENE Airbome Particulate Weekly I Gross Beta - Weekly Dowagac, Ml Weekly j Gamma Isotopic l Quart Comp.

Airborne Radioiodine Weekly 1.131 I Weekly TLD Quarterly I Dircct Radiation Quarterly COL jIS9 miles NNt Aircorre Particulate Weekly Gross Beta Weekly Colomar MI Weekly I Gamma Isotopic Quart Comp AirroOrne Radioiodine Weekly 1-131 Weekly I_____________ TLD l Quarterl- Drect Radiatin lQuarterly

Information I PMP-6010.0SD.OO I Rev. 16 I Page 74 of S4 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.19 Radiological Environmental Mlonitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 73 - 76 SAMPLE DESCRIPTION/ I SANIPLE SAMPLE ANALYSIS I ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY OFF-SITE AIRBORNE AND DIRECT RADIATION JTLD) STATIONS OfT-I I 45 milcs NE, Pole NI3294-44 7 TLD I Quartlv l Direct Raiatiun juarterly OFT-2 3 6 miles, NE, Stevensyille TLD 1 Quarteriy l Direct Radiation Quanerly Substation I I I OFT-3 Si Jmles NE, Pole iE32%-l3 TLDJ Quanerly Direct Radiation Quarterly OF1-4 41 milcs, B. Pale #B3S0-72 j TLD Quarerly Direct Radiation IQuarterly

_Fl_ 4 2 mides FSE. Pole 4B387-32 TLD Quarterly Direct Radiation -Quarterly CF-T-6 49 milSE£ Pole 1342h-l I TLD Quarterly Duect Radiation Quarerly OFT-? 2.5 mileS S. 13rdgmnn Subseaton TLD I Quarterly Direct Radsatio 7 QuarterLv OFT-8 4 0 miles S. Pole #B1424-20 TLD Quarterly Direct Raiation l Quanerly OFT-9 44 miles ESE, Polce3B369-214 TLD Quartcrly Direct Rania Ion Quarterly OFT-Lu J g miles S. Polc #B422-99 TLD Quarterly Direct Radiation I Quartcrly OFT-I I 1 3.8 miles S, Pole YB423-12' TLD Quanerly Direct Radiation l Quanerly GROUNDWATER (WELL WATER) SAMPLE STATIONS WY-I 1969 ft 8 I' from Plant .Axis Groundwater Quarterly Gamma Isotopic Quarterly I I_ Trtum Quarterly W-2 2302 ft J 630 from Plant AxiS Groundwater Quarterly Gamma Isotopic Quarterly I ___ Ttiumr Quarterly W-I 3279 R kg 107' from Plant Axis Groundwater Quarterly Gamma Isotopic Qunrterly Triciumr Quarterly W-4 418 ft@ 301' from Pl ant Axis Groundwater Quarterly Gamma Isotopic Quarterly I__ Tnritium IQuarterly W-S 404 ft 2900 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly

_ _ __ _ _ nrtium -- Quarterly W-6 424 it j 2730 trom Plant A-as Groundwatcr _ Quarterly Gamma Isotopic Quarterly tritium Quarterly W-7 1895 ft i1 189' from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-8 1274 it {3 541 from Plant Axis Groundwater i Quarterly Gamma Isotopic Quarterly IIITritium QuarscrlyI W-9 l1441 fIt 22- from Plant Axis Groundwater Quarterly Gam 1st - Quartly ITritium Quarterly W-l I 4216 ft 153° trom Plant Axis Groundwater

_416_topicI Quarterly Quarterly Gamma Isotopic j Quarterly W-1I Il 320 f~i 5' fro

-Plant Axis ntrium Quarterly Gondwater Quarerly Gam sooi Quarterly W-12 2631 ft r0 162' from PlantAxis Groundwater Quarterly Gamma Isotopic IQuarterly I __I_ ITritium Quarterly W-13 2152 ft IS"2 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly ITr__ _. I Tnlium I Quanerty W-14 1 S80ft 8 164' from Plant Axis Groundwater Quarterly Gamma Isotopic l Quarterly t I _ I I Tnhium Quans HyI

Information I PMP-6010-OSD.001 I Rev. 16 1 Page 75 of 84 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Prop-ram Paces:

Attachment 3. 19 M Sample Stations, Sample Types. Sample Frequencies 73) - 76 SANI'PLE DESCRIPTION! SAMPLE SAMPLE ANALYSIS ANALYSIS STATION I LOCATION TYPE FREQUENCY TYPE FREQUENCY DRINKING WATER SI St. Josech Public Intake Sta Drinkingo attr Dail) Gros Bcta 1I-day Comp 9 mL NE Gamma Isotopic I 14 day Comp.

1-131 1L4 day Comp Tritium Quart Comp L rw Lake 1wp Public Intake Sta. Drinkine w iter Daily Gross Beta 14 day Comp 06rmi S Gamma Isotopic 14 day Comp 1-131 Tritium 14 day Cmp, j Quart Comp 3 SURFACE WATER SWL-I Condenser Circulating Water l Surface Watcr Daily Gamma Isotopic Month. Comp l Intake I Tntium Quart Camp SWL-2 Plant Site Boundar - South Surtace Watcr Daily Gamma Isotopic Month Camp

- 5a5 ft south of Plant Centerine ITritium Quan Ccmp I SWL-3 Plant Site Bouneary - North Surface Water Dady Gamma Isotopic Month Comp 500 ft north of Plant C__terln_ _

Tritium Quart Comp I SEDIMENT SL-2 Plant Site Boundary - South Sediment Semi-Ann. Gamma Isotopic Semi-Annual

- 500 ft south of Plant Centerline i SL-3 Plant Site Boundary - North Sediment Semi-Ann. Gamma Isotopic Semi-Anr.ual

- 500 fL north of Plant Centerline I SL-4 Plant Site Boundary - South Sediment Quartcrly Gamma Isotopic Quarterly South storm drain culvert to lake SL-5 Plant Site Boundary - North Sediment Quarterly Gamma Isotopic Quarterly North storm dratin culvert to lake SL4 & i are data collection points only not actual REMP samples GROUNDWATER (STEAM GENERATOR STORAGE FACILITY) SAMPLE STATIONS SG_- 0a 8mi.95' from Plant Axis Groundwater Quarterty Gross Alpha Quarterly lGross Beta Quarterly

_ _ _ Gamma Isotopic I Quarterly SG-2 0.7 mi @ 92° from Plant Axis Growid waetr Quarterly Gross AlDha t Quarterly Gross Beta Quanerly

_ l_Garnrna Isotopic I Quarterly SG-4 0 7 imn. 931 from Plant Axus Groundwater Quarterly Gross Alpha Quarterly I Gross Beta Quarterly

_ _ l l Gamma Isotopic Quarterly SG-5 0.7 mi. @ 92° from Plant Axis Groundwater Quarterly Gross Alpha Quarterly Gross Beta Quarterly I

_ l l Gamma Isotopic Quarterly INGESTION - MILK Indicator Farmns Mi k Onceevery i 1-131 persample l _l_15 days Gamma Isotopic Tper sample Milk nce evcry 1-13I per sample

_ 15 days i Gamma Isotopic per sample I I l ilk Once every 1-131 per sample

_ _ _ l IIi days jGamrna Isotopic j per sample

Information I PvIP-6010.OSD.001 I Rev. 16 1 Page 76 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3. 19 Radiological Environmental IMlonitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 73 - 76 SAMIPLE j DESCRIPTIONI SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY INGESTION - MILK Background Farms avinyloure FPam 'ml es S, La Porte, IN j Mifk l Unce ev.ery 1-11 pcr samile Ili days Gamma Isotopic per sample Wyant Farm 20 7 mnles E. Dowaglac Milk Once evry j 1-131 per sample

_ Ii Jays Gamma Isotopic I per sample INGESTION - FISH ONS-N 0 3 mile N, Lake Michigan l Fish 2/ycar Gamma Isotopic l per sample ONS-S OFS-N 0 4 mile S. Lake Michiganr 3 5 mile N. Lake Michig3nr Fish 2/year Gamma Isotopic persamnple 1 Fish 2/year Gamma Isotopic I per sample S- _ 5 0 mile S, Lake Michigan SO Fish 3'year GGamma Isotopic l per sample INGESTION - FOOD PRODUCTS On Site ONS-G Nearest sample to Plant in the CTmpes At time ot Gamma Isotopic At time of highest D/Q land sector harvest harvest containing media O Broadleaf At time ot Gamma Isotopic At time of

_ _ _ vegetation Jharvcst harvest IOtf Site OFS-G In a land sector containing Grapes At lime ot Garm a Isotopic At tune of grapes, approximately 2.0miles harvest harvest from the plant, in one of the

____ _ less prevalent DIQ land sectors INGESTION - EROADLEAF IN LIEU OF MILK 3 indicator samples of broad leaftvegetation Bro3dleat Monthly Gamma Isotopic 1 Mtonthly collecteJ at different locations, within eight vegetation when available 1131 when 3vUalaDle miles of the plant mnthe highest annual average D/Q land sector.

I baLkground sample ot similar vegetatzoa I roadleaf lNtonthly Gamma Isotopic Monthly grown 15-25 miles distant in one of vegetation when available ]131 when available the less prevaJent-Aind directions Collect coimpcsite samples of Drinking and Surf3ce watcr at least daily Analvze particulate sample filters for gross beta activiry 24 or more hours following filter removal Thts will allow for radon and thoron daughter decay. It gross bea activity in air or water r greater than 10 times he yearly mean of control samples for any medium, perform gamma isotopic analysis on the individual samples If at least thre: indicator milk samples and one background milk sampie cannot be obtained, three indicator broad leaf samples will be collected at different locations, within eight miles of the plant, In the land sector with the highest D/Q (refers to the highest annual average D/Q) Also, one background broad leaf sample will be collected 15 to 25 miles frum the plant n cne of the less prevalent DIQ land sectors.

The Jdreamilk indicator farms will be determined by the Annual Land Use Census and those that are willing to participate

Information I PiP-6010.OSD.001 I Rev. 16 l Page 77 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachrent 3.20 Maximum Values for Lower Limits of Detections A3S-REMP 77age78

[Ref 5 2 Ivi Radionuclides Food Product Water IMilk Air fFilter Fish Sediment pCi/kg-,wet p Cil pCill pCi/M 3 pCi/kg, wet pCilkg, drv Gross Beta 4 i 0 01 H-3 2000 _

Ba-140 La- 140 Cs-134 60 60 5

15 60 15 15 0.06 130 I 150 Cs-137 60 18 18 0.06 150 180 Zr-95 30 Nb-95 15 1 Mn-54 15 130 Fe-59 30 260 Zn-65 1 30 l 260 Co-58 15 130 Co-60 I5 130 1-131l 60 1 1 0.07 This Data is dircctly from our plant-specific Tcchnical Specification

' LLD for drink:ng water

Information I PMP-6010.OSD.001 j Rev. 16 1 Page 73 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.20 Maximum Values for Lower Limits o lDetections-AR2.ES9IP Pages:

77-7 NOTES

.A Te Lower Limit of Detecticn (I,LD) s detiried is the smallest conccntration ol radioacttie nmate:al in a camole that vill be detected %ith 95%,1 probability and 5%N proodbilitv of falsely concluding that a blank obs-rvation represents a "real" signoll.

For a pu.tiular measurement sysem (which miy include radochemijcal scparation), thc LLD is gixen by the equatioit.

LLD = 4.66a

S E*V* 2.22*Y*e(~&'h)

Where LLD is the a Drion lower limit of detection as defined above (as pCi per unit mass or volume) Perform analysis in such a manner that the stated LLDs will be achieved under routine conditions Occasionally background fluctuatons, unavoidably small sample sizes the presence of interfering radionuclides, or other uncontrollable circumstances may render these LLDs unachievable.

S 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 efficiency of the detection equipment as counts per transfornianon (that is, disintegration)

V is the sarnple size in appropriate mass or volume units

2. 22 is the conversion factor from picocuncs (pCi) to transformations (disintegrations) per minute Y is the fractional radiochemical yield as appropriate X is the radioactive decay corstant for ibe particular radionuclide At is the elapsed time between the midpoint of sample collection (or end of sample collection period) and time of counting.

B. Identify and report other peaks which are meassarable and identifiable, together with the radionuclides listed in Attachment 3 20, Maximum Values for Lower Limits of DetecticnsA,3 - REMP.

a A 2 71 value may be added to the equation to provide correction for deviations in the Poisson distribution at low count rates,

-hat is, 2.71 + 4.66 x S

Information } PMP-6010.OSD.001 Rev. 16 l Page 79 of 84 OFF-SITE DOSE CALCULATION MANUAL Attchmnt .21Reportin ConcentrationsLevels for RadioactivirlyPae in Environmental Samples 79 Radionuclides Food Product Water Milk Air Filter Fish 3

l pCilkg, wcet pCi/ l pCi/m w pCifkg, et H-3 _ 20000 Ba-140 200 300 La- 140 200 300 Cs-134 1000 30 60 10 1000 Cs-137 2000 50 70 20 2000 Zr-95 400 Nb-95 400 Mn-54 1000 30000 Fe-59 400 10000 Zn-65 300 20000 Co-58 1000 30000 Co-60 300 10000 I-131 l 100 2 3 0.90

Information I PiMIP-6010.OSD.001 I Rev. 16 l P3we 80 of 84 OFF-SITE DOSE CALCULATION MANUAL .22 On-Site Monitoring Location - RENIPPage:

I 8o LEGEND ONS-I - CNS Air Sampling Station Tr-O - T-12: TLDSamphngStaticn W-l - W-14: REMP T/S Gmundwacer Wells SG-I, SO-2, SG-4, SGC-. RE.MP Non T/S Groundwater Wells SWL-1 .2, 3 Surbace Water Sampling Stations SL-2. SLt-: Sediment Sampling Stations

Information I PMP-6010O.OSD.001 I Rev. 16 1 Page 81 of 84 OFF-SITE DOSE CALCULATION MNANUAL .23 Off-Site Monitoring Locations - REMIP Page:

I I SI

Information I PNIP-6010.OSD.001 Rev. 16 I Page 82 of 34 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.24 Safety Evaluation By The Office Of Nuclear Paaes:

Reactor Regulation 32 - 34 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO DISPOSAL OF SLIGHTLY CONTAMIINATED SLUDGE INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLAINT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316 (Ref 5 2 Irl (This is a 10 CFR 50.75 (g) item)

1. rNTRODUCTION By letters dated October 9, 1991, October 23, 1991, September 3, 1993, and September 29, 1993, Indialia Michigan Power Company (I&M) requested approval pursuant to 10 CFR 20 2002 for the on-site disposal of licensed material not previously considered in the Donald C. Cook Nuclear Plant Final Environmental Statement dated August 1973. Specifically, this request addresses actions taken in 1982 in *which approximately 942 cubic meters of slightly contarrunated sludge were removed from the turbine room sump absorption pond and pumped to the upper parking lot located within the exclusion ara of the Donald C.

Cook Nuclear Plant The contaminated sludge was spread over an area of approximately 4 7 acrcs. The sludge contained a totalmdLonucfide inventory of 8 89 millicunes fmCi) of Cesium-137, Cesium- 136, Cesium-134, Cobalt-6O and Iodine- 131.

In its submittal, the licensee addressed specific information requested in accordance with 10 CFR 20.2002(a), provided a detailed description of the licensed material, thoroughly analyzed and evaluated information pertinent to the impacts on the environment of the proposed disposal of licensed material, and committed to follow specific procedures to minimize the risk of unexpected exposures.

I DESCRIPTION OF WASTE The turbine room sump absorption pond is a collection place for water released from the plant's turbine room sump. The contamination was caused by a prirnary-to-seccndary steam generator leak that entered the pond from the turbine building sump, a recognized release pathway. Sludge, consisting mainly of leaves and roots mixed with sand, built up in the pond. As a result, the licensee dredged the pond in 1982 The radioactive sludge removed by the dredging activities was pumped to a containment area located within the exclusion area. The total volume of 942 cubic meters of the radioactive sludge that was dredged from the bottom of the turbine room absorption pond was subsequently spread and made into a graveled road over the upper parking lot area of approximately 4 7 acres The principal radionucldes identified in the dredged material are listed below.

TABLE I NUCLIDE ACTIVITY (mCi) ACTIVITY (mCi)

(half-life) 1982 1991

"'Cs (13.2 d) 0.03 1 NA 4

'" Cs(2.1 y) 2.34 0.18

'j'Cs (30.2 y) 5.59 4.57 50Co (5.6 y) 090 0.27 l (8.04 d) 0.03 NA*

TOTAL: 8.89 5.02 NA not applicable due to decay

Information I PMP-6010.OSD.OO1 I Rev. 16 1 Page 83 of 84 OFF-SITE DOSE CALCULATION MANUAL l Safety Evaluation By The Office Of Nuclear Pages:

tacen3.4Reactor Regulation 82 - 84 3 RADIOLOGICAL IMPACTS The licensee in 1982 evaluated the following potential exposure pdthways to members of the general publik from the radionuclides in the sludge (1) exterM31 exposure caused by g-oundshine from the disposal site, (2) internal exposure caused by inhalation of re suspended radionuclide;

-AND-(3) internal exposure from ingesting ground %vater The staff has reviewed the licensee's calculational methods and assumptions and finds that they are consistent with NUREG-I 101, "Onsite Disposal of Radioacmi',e Waste," Volumes I and 2, November 1986 and February 1987, respectively The staff finds the assessment methodology acceptable Table 2 lists the doses calculated by the licensee for the maximally exposed member of the public based on a total activity of 8 89 mCi disposed in that year TABLE 2 Patb, ay Whole Body Dose Received by Maximally Exposed Individual (mremlyear)

Groundshin 0.94 Inhalation0.94 Groundwater Ingestion 0.73 Total 2.61 On July 5, 1991, the licensee re-sarmpled the onsite disposal area to assure that no significant impacts and adverse effects had occurred. A counting procedure based on the appropriate environmental lcw-level doses was used by the licensee; however. no activity was detected during the re-samplingi. This is consistent with the original activity of the material and the decay time.

The 1991 re-sampling process used by the licensee confirms that the environmental impact of the 1982 disposal was very small.

The staff finds the licensees methodology acceptable 4 ENVIRONMENTAL FINDING AND CONCLUSION The staff has evaluated the environmental impact of the proposal to leave in place approximately 942 cubic meters of slightly contaminated sludge underneath the upper parking lot on the Donald C. Cook Nuclear Plant site In 1982. the licensee evaluated the potential exposure to members of the general public from the radionuclides in the sludge and calculated the potential dose to the maximally exposed member of the public, based on a total activity of 3 39 mCi disposed in that year, to be 2.61 mremLyr. The staff has reviewed the licensee's calculational methods and assumptions and found that they are consistent with NUREG-I 101, Onsite Disposal of Radioactive Waste, Volumes I and 2, November 1986 and February 1987, respectively The staff finds the assessment methodology acceptable. For companson. the radiation from the naturally occurring radionuclides in soils and rocks plus cosmic radiation gives a person in Michigan a whole-body dose rate of about 89 mrem per year outdoors Subsequent licensee sampling in 1991 identified no detectable activity. The staff evaluated the licensee's sampling and analysis methodology and finds it acceptable. The results, of the 1991 re-sampling by the licensee, confirm that the environmental impact of the 1982 disposal was very small Based on the above the staff finds that the potential environmental impacts of leaving the contaminated sludge in place are insignificant With regard to the non-radiological impacts, the staff has determined that leaving the soil in place represents the least impact to the environment.

Information } PNIP-6010.OSD.O1 I Rev. 16 l Page 84 of 84 OFF-SITE DOSE CALCULATION .LANUAL Safety Evaluation By The Office Of Nuclear Pages.

Attachment 3.24 Reactor Regulation 32 -

5 CONCLUSION Based on tcheataff's review of the licensees dtscusbion, the stall finds the licens.e's proposal to Tctain 'he material in its present location as documented in this Safety Evaluation acceptable. Also, this Safety Evaluation shall be aermancntly incorporated as an appendix co the iccrnsee's Ot~site Dose Calculation Manual (ODC%), and any future modifications shall be reported to NRC in accordanct with the applicable ODCM[ change protocol I&M letter from E E. Fitzpatrick to the NRC Document Control Desk, September 29, 1993 Therefore, the licensee's proposal to consider the slightly contaminated sludge disposed by retention in place in the manner described in the Donald C. Cook Nuclear Plant submittals date October 9, 1991, October 23, 1991, September 3, 1993, and September 29, 1993, is acceptable.

The guidelines used by the NRC staff for onsite disposal of licensed material and the staffs evaluation of how each guideline has been satisfied are given in rable 3.

Pursuant to 10 CFR 51 32, the Commrinssion has determined that granting of this approval will have no significant impact on the environment (October31, 1994, 59 FR 5447).

Principal Contributor 1. Minns Date: November 10, 1994 TABLE 3 20.2002 GUIDELINE FOR ONSITESTAFF'S EVALUATION DISPOSAL'

1. The radioactive material should be disposed of in such a I Due to the nature of the disposed material, recycling to the manner that it is unlikely that the matenal would be general public is not considered likely.

recycled.

2. Doses to the total body and any body organ of a 2 This guideline was addressed in Table 2 Although the maximally exposed individuals (a member of the general 2.61 mrel/yr is greater than staff's guidelines, the staff finds it public or a non-occupationally exposed worker) from the acceptable due to 9yrs decay following analysis and the probable pathways of exposure 'o the disposed material expected lack of activity detected in the 1991 survey.

should be less than I mnrem/year.

3. Doses to the tctal body and any body organ of an 3. Because the material will be land-spread, the staff considers inadvertent intnider from the probable pathways of the maximally exposed individual scenano to also address the exposure should be less than 5 mirerniyear. intruder scenario.

4 Doses to the total body and any body organ of an 4. Even if reycling, were to occur after release from regulatory individual from assumed recycling of the disposed control, the dose to a maxJmally exposed nember of the material at the time 'he disposal site is released from public is not expected to exceed I miremyear, based on regulatory control from al likely pathways of exposure exposure scenarios considered in this analysis.

should be less than I mrem.

Z E. F Branagan. Jr. and F J. Congel, "Disposal of Contaminated Radioactive Wastes from Nuclear Power Plants." presented at the Health Physics Society's Mid-Year Symposium on Health Physics Consideration in Deconiamination/Decommissioning, Knoxville, Tennessee, February 1986, (CONF-360203).

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