Text

Yankee Atomic Electric Company - Submittal of the 2006 Annual Radiological Effluent Release Report and Offsite Dose Calculation Manual
	ML070430316
Person / Time
Site:	Yankee Rowe
Issue date:	01/31/2007
From:	Gerard van Noordennen; Yankee Atomic Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BRY 2007-006
	Download: ML070430316 (172)
	v • d • e

YANKEE ATOMIC ELECTRIC COMPANY Telephone (413) 424-5261 Q9ANNE 49 Yankee Road, Rowe, Massachusetts 01367 January 31, 2007 BRY 2007-006 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-001

Reference:

Subject:

License No. DPR-3 (Docket No. 50-29) 2006 Annual Radiological Effluent Release Report and Offsite Dose Calculation Manual Yankee Atomic Electric Company (YAEC) herewith submits the 2006 Annual Radiological Effluent Release Report (ARERR) and Revision 19 and 20 to the Offsite Dose Calculation Manual (ODCM). The ARERR provides a summary of the quantities of radioactive liquid and gaseous effluent and solid waste released at the Rowe site and also includes a summary of estimated dose commitments from all radioactive liquid and gaseous effluents released in 2006. This information is submitted in accordance with the Yankee Quality Assurance Program (YQAP) and the ODCM.

The enclosed Revision 19 and 20 to the ODCM is being submitted in accordance with Appendix D of the YQAP. The current revision of the ODCM was approved on October 24, 2006.

Should you have any questions regarding this submittal, please contact me at (860)267-3938.

Very truly yours, YANKEE ATOMIC ELECTRIC CO Gerry van Noordennen Regulatory Affairs Manager

Enclosure:

cc:

S. Collins, NRC Region I Administrator J. Hickman, NRC Project Manager bZJSS& /

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (ARERR)

YANKEE NUCLEAR POWER STATION ROWE, MASSACHUSETTS JANUARY 1, 2006 - DECEMBER 31, 2006 DOCKET NO. 50-29 LICENSE NO. DPR-3 YANKEE ATOMIC ELECTRIC COMPANY Rowe, Massachusetts

TABLE OF CONTENTS 1.0 IN T R O D U C T IO N........................................................................................................

1 2.0 DOSE ASSESSMENT..................................................................................................

3 3.0 R E F E R E N C E S.......................................................................................................................

3 Appendix A: Liquid Holdup Tanks..............................................................................

A-1 Appendix B: Radiological Environmental Monitoring Program.................................... B-1 Appendix C: Process Control Program (PCP)............................................................

C-1 Appendix D: Off-Site Dose Calculation Manual (ODCM)...........................................

D-1 Appendix E: Supplemental Information..........................................................................

E-1 Appendix F: Sewage Sludge Disposal.......................................................................

F-1 ii

LIST OF TABLES TABLE 1:

Maximum Off-Site Doses and Dose Commitments to Members of the Public....... 4 TABLE 2:

Solid W aste and Irradiated Fuel Shipments......................................................

5-8

-iii-01/30/07

ANNUAL (2006) RADIOACTIVE EFFLUENT RELEASE REPORT YANKEE NUCLEAR POWER STATION ROWE, MASSACHUSETTS

1.0 INTRODUCTION

The Yankee Quality Assurance Program (YQAP), Appendix D, requires that an Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted (to the NRC) before May 1 of each year. This report includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the site. The material provided is (1) consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) and (2) is in conformance with 10CFR50.36a and Section IV.B.1 of Appendix I to 10CFR50. The ODCM details the specific information to be included in the annual report.

Yankee Nuclear Power Station's last day at any power level was October 1, 1991. The facility is permanently shut down and has completed the final stages of decommissioning.

Due to completion of decommissioning there have been no gaseous and liquid effluent releases during 2006. All spent fuel and radioactive materials have been moved to the on-site dry fuel storage facility with no effluent release points.

In July 2002, the first dry spent fuel storage canister was placed on the Independent Spent Fuel Storage Installation (ISFSI) pad located within the station's protected area. By design, there are no liquid or gaseous effluent release pathways from storage canisters once placed on the ISFSI pad. In June 2003, all transfers of spent fuel and greater than Class C materials requiring storage in the ISFSI were completed. Following the completion of these transfers to dry storage, the Spent Fuel Pool (SFP) water inventory was processed by filtration and demineralization and released to the environment as a controlled liquid waste discharge in accordance with our NPDES permit and the ODCM. The discharge of the SFP water represented the last major source of station process water requiring processing as part of station decommissioning.

No liquid or gaseous radiological releases were made during 2006.

Table 1 summarizes the total dose to the maximally impacted off-site individual from all station related sources for 2006.

All dose measurements for this reporting period are well below the regulatory dose criteria of 10CFR Part 50, Appendix I and 40CFR Part 190. 01/30/07

Appendices A through F indicate the status of reportable items per the requirements of the following documents:

Appendix Title Requirement Reference A

Liquid Holdup Tanks YQAP; Appendix D B

Radiological Environmental Monitoring Program ODCM (Rev. 20)

C Process Control Program (PCP)

PCP D

Offsite Dose Calculation Manual (ODCM)

ODCM (Rev. 20)

E Supplemental Information Regulatory Guide 1.21 F

Sewage Sludge Disposal ODCM (Rev. 20) 01/30/07

2.0 TOTAL DOSE FROM DIRECT EXTERNAL RADIATION The annual total dose or dose commitment to any member of the public due to direct radiation from fixed sources are limited to the EPA's radiation protection standards for the uranium fuel cycle (40CFR190). The dose limits are set to less than or equal to 25 mrem per year to the total body or any organ, except the thyroid, which is limited to less than or equal to 75 mrem per year.

Direct external dose from fixed sources of radioactive materials, such as the on-site ISFSI, was evaluated by comparing the Station's 2006 TLD data for offsite indicator stations versus the control and outer ring (beyond 6 miles) locations. Since there was no distinguishable difference between the indicator measurements at the site boundary or beyond and the control measurements, it was concluded that there is no measurable station-related direct radiation dose for 2006.

Table 1 shows that the total dose to the maximum off-site individual for 2006 is well below the EPA dose limit criteria.

3.0 REFERENCES

1.

YNPS Offsite Dose Calculation Manual (ODCM), Revision No. 20, effective date, October 24, 2006.

2.

Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Release of Reactor Effluents for the Purpose of Evaluating Compliance With 10CFR Part 50, Appendix I," U.S. Nuclear Regulatory Commission, Office of Standards Development, Revision 1, October 1977. 01/30/07

TABLE 1 Yankee Nuclear Power Station, Rowe, Massachusetts 2006 Annual Radioactive Effluent Release Report Maximum Off-Site Doses and Dose Commitments to Members of the Public Source Unit 1 st Quarter 2 nd Quarter 3rd Quarter 4t1h Quarter Year (a)

Direct Dose - (b) (c)

Direct External Dose mnremn 0

0 0

(a)

"Maximum" dose for the year is the sum of the maximum doses for each quarter. This resuits in a conservative yearly dose estimate, but still within the limits of 10 CFR Part 50.

(b) 2006 TLD data for off-site (site boundary) indicator stations and control / outer ring (beyond 6 miles) stations were compared. No statistical difference which could be attributed to station sources was identified.

(c)

Direct dose is equivalent to Total Dose to Maximum Off-Site Individual (4OCFR 190) 01/30/07

TABLE 2 (Sheet 1 of 4)

Yankee Nuclear Power Station, Rowe, Massachusetts 2006 Annual Radioactive Effluent Release Report Solid Waste and Irradiated Fuel Shipments First and Second Quarters A. SOLID WASTE SHIPPED FOR BURIAL OR DISPOSAL (not irradiated fuel) 6-month Est. Total

1. Type of Waste Unit Period Error, %

a.

Dry Active Waste: (D & D Composite Waste:

m3 1.8 E+04 25 demolition rubble, compactable trash, noncompactable Ci (est.)

trash)

Class A Container: metal boxes (intermodal, gondola)

b.

Soil:

m3 2.81 25 Class A Ci (est.)

0.0 Container

metal boxes (B-25 metal box, drums)

2. Estimate of Nuclide Composition > 1% (by type of waste)

a.

Hvdroaen-3 (tritium) 42.3 Iron-55

.77 Cobalt-60 5.3 Nickel-63 24.4 Cesium-137 24 Carbon-14 9

3. Solid Waste Disposition Number of Shipments 227 Mode of Transportation Truck/rail Destination Envirocare, Clive, Utah 01/30/07

TABLE 2 (Sheet 2 of 4)

Yankee Nuclear Power Station, Rowe, Massachusetts 2006 Annual Radioactive Effluent Release Report Solid Waste and Irradiated Fuel Shipments I. First and Second Quarters (continued)

B. IRRADIATED FUEL SHIPMENTS (Disposition): None 01/30/07

TABLE 2 (Sheet 3 of 4)

Yankee Nuclear Power Station, Rowe, Massachusetts 2006 Annual Radioactive Effluent Release Report Solid Waste and Irradiated Fuel Shipments II. Third and Fourth Quarters A.

SOLID WASTE SHIPPED FOR BURIAL OR DISPOSAL (not irradiated fuel) 6-month Est. Total

1. Type of Waste Unit Period Error, %

a.

Dry Active Waste: (D & D Composite Waste: demolition rubble, m3 5.7 E+ 03 25 compactable trash, noncompactable trash)

Ci (est.)

.09 Class A Container: metal boxes (intermodal, B-25, gondola, sealand)

2. Estimate of Nuclide Composition > 1% (by type of waste)

a.

Hydrogen-3 (tritium) 23.2 Iron-55 21.3 Cobalt-60 13.8 Nickel-63 15.0 Cesium-137 25.9

3. Solid Waste Disposition Number of Shipments 75 31 11 Mode of Transportation Truck/rail Truck Truck Destination Envirocare, Clive, Utah Envirocare, Clive, Utah Bear Creek, TN

'One shipment was a combined shipment that went to Bear Creek, TN and then went on to Envirocare, Clive, Utah. 01/30/07

TABLE 2 (Sheet 4 of 4)

Yankee Nuclear Power Station, Rowe, Massachusetts 2006 Annual Radioactive Effluent Release Report Solid Waste and Irradiated Fuel Shipments I1. Third and Fourth Quarters (continued)

B. IRRADIATED FUEL SHIPMENTS (Disposition): None 01/30/07

APPENDIX A Liquid Holdup Tanks Requirement:

Response

The Yankee Quality Assurance Program (YQAP), Appendix D, Section H, limits the quantity of radioactive material contained in any outside temporary tank. With the quantity of radioactive material in any outside temporary tank exceeding the limits of the YDQAP, a description of the events leading to this condition is required in the next Annual Radioactive Effluent Release Report.

The limits of the Yankee Quality Assurance Program were not exceeded during this reporting period.

A-1 01/30/07

APPENDIX B Radioloqical Environmental Monitoring Program Requirement:

Response

The Radiological Environmental Monitoring Program is conducted in accordance with the ODCM. With the Radiological Environmental Monitoring Program not being conducted as specified in the ODCM, a description of the reasons for not conducting the program as required and that plans for preventing a recurrence will be included in the next Annual Radioactive Effluent Report.

The requirements of the ODCM, governing the conduction of the REMP were met for this reporting period.

B-1 01/30/07

APPENDIX C Process Control Program (PCP)

Requirement:

Response

PCP Control 2.0 requires that licensee-initiated changes to the PCP be submitted to the Commission in the Annual Radioactive Effluent Release Report for the period in which the change(s) was made.

The PCP was canceled during 2006 because there is no material that would be required to be shipped under the PCP.

The Yankee Quality Assurance Program Revision 33.a. states, "A PCP will be developed if needed and will contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes will be accomplished to ensure compliance with 10 CFR Parts 20, 61, and 71; state regulations; burial ground requirements; and other requirements governing the disposal of solid radioactive waste. There is no expected need for a Process Control Program due to ISFSI Operations. A PCP will be developed and approved by the ISFSI Operations Manger if the need arises."

C-1 01/30/07

APPENDIX D Off-Site Dose Calculation Manual (ODCM)

Requirement:

Response

The ODCM Control requires that licensee-initiated changes to the ODCM be submitted to the Commission in the Annual Radioactive Effluent Release Report for the period in which the change(s) was made effective.

Revision 19 and 20 to the ODCM were approved for implementation in April 2006 and October 2006 respectively. The changes were made to reflect the completion of decommissioning of the site and to reflect the elimination of any radioactive release pathways. The REMP was reduced to measure only direct dose through the implementation of the Environmental TLD Program.

D-1 01/30/07

APPENDIX E Supplemental Information There were no radioactive liquid or gaseous releases during 2006.

E-1 01/30/07

APPENDIX F Sewaqe Sludqe Dis*osal Requirement:

Response

ODCM, Appendix A requires that for periods in which disposal of septage occurs, the licensee shall report in the Annual Radioactive Effluent Release Report, the volume discharged, liquid and solid fractions, and total activity discharged.

There was one sewage disposal shipment (June) made in 2006.

Volume discharged: 6800 gallons Liquid fraction (by weight) of waste: 0.989 Solid fraction (by weight) of waste: 0.011 Nuclide content in liquid fraction (pCi/gm): ND (a)

Nuclide content in solid fraction :

Cs-137 = 3.20E-08 pCi/gm (wet)

Co-60 = 3.50E-08 pCi/gm Total radioactivity discharged:

Cs-137 = 8.80 E-03 pCi Co-60 = 9.70 E-03 uCi Total.:

1.85 E-02 pCi (a) Not Detected F-1 01/30/07

YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL YANKEE ATOMIC ELECTRIC COMPANY PORC MEETING NO./DATE PREPARED BY/DATE REVIEWED BY/DATE REVISION 8 C. L. Albright Mark Strum Meeting No. 92-72 August 19, 1992 August 19, 1992 August 19, 1992 Edward R. Cumming Mark Strum Meeting No. 93-22 REVISION 9 May 12,1993 May 18,1993 May 18,1993 REVISION 10 R. Brad Harvey Mark Strum Meeting No. 93-28 June 22, 1993 June 22, 1993 June 22, 1993 March Strum G. M. Babineau Meeting No. 96-63 October 22, 1996 November 12,1996 October 31, 1996 R. B. Harvey Mark Strum Meeting No. 97-3 REVISION 12 February 7, 1997 February 7, 1997 January 9, 1997 M. S. Strum John S. Gedutis Meeting No. 99-19 REVISION 13 June 17, 1999 June 17, 1999 June 17, 1999 M. S. Strum G. M. Babineau Meeting No. 00-15 April 13, 2000 April 13, 2000 April 13, 2000 REVISION 15 M. S. Strum G. M. Babineau Meeting No. 01-69 November 19, 2001 December 17, 2001 November 19, 2001 Mark Strum Dave Montt Meeting No. 03-65 REVISION 16 August 6, 2003 August 12, 2003 August 14, 2003 Mark Strum Dave Montt Meeting No. 04-05 REVISION 17 January 28, 2004 January 28, 2004 January 29, 2004 Mark Strum Dave Montt Meeting No. 05-28 REVISION 18 July 13, 2005 July 18, 2005 August 4, 2005 Ellen Heath Greg Babineau ISR Review REVISION 19 Gi, 1

/4yb, March 22, 2006 March 22, 2006

REVISION RECORD Revision Date Description 0

12/01/82 Initial printing. Approved by PORC 11/29/82. Submitted for USNRC approval 12/03/82.

1 03/30/84 Change in environmental monitoring sampling locations based on 1983 land use census. Errors in Table 4.1 corrected. Maps revised.

2 07/30/85 Addition of Intercomparison Program description to Section 4.0. Reviewed by PORC 07/30/85.

3 03/19/86 Addition of a PVS 1-131 inspection limit to demonstrate compliance with Technical Specification 3.11.2.1.b.

4 05/21/86 Change in milk sampling location. Samples no longer available at Station TM-1 1.

5 09/30/86 Change in food product sampling location based on 1986 land use census.

6 02/18/88 Change in liquid dose factors to reflect additional dose pathways. Change in gaseous dose factors to reflect five-year average meteorology. Change in gaseous dose rate factors to reflect a shielding factor of 1.0.

Deletion of food product location TF-12 (samples no longer required after 10/31/86). Update of fence line location and several building names and locations in Figure 4-4.

7 05/21/90 Addition of Appendix A which documents the commitments for disposal of septage as provided in YNPS's Application For Approval to Routinely Dispose of Septage under 1 OCFR Part 20.302, and the NRC's acceptance as transmitted in their Safety Assessment, dated May 17, 1990.

8 08/19/92

a. The following changes were implemented in accordance with NRC Generic Letter 89-01, which provided guidance on the relocation of the Radiological Effluent Technical Specifications to the ODCM:

1.

Addition of List of Controls Page (succeeds Table of Contents);

Revision 19

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REVISION RECORD Revision Date 8

08/19/92 Description

2.

Section 1.0, Introduction updated to reflect the change in scope of the ODCM;

3.

Technical Specifications 3/4.0.1, 3/4.0.2, 3/4.0.3, and 3/4.0.4 listed in Section 1.2, Applicability of Controls and Surveillance Requirements (SR), and now referred to as Controls 1.1, 1.2, 1.3, and 1.4, respectively;

4.

Table 1.6, Definition of Terms, modified to include definitions pertinent to the relocated Technical Specifications;

5.

Tables 1.9, OPERATIONAL MODES, and 1.10, FREQUENCY NOTATIONS, added to Section 1.0;

6.

Technical Specification 3/4.11.1.1, now referred to as Control 2.1, relocated to Section 2.0;

7.

Technical Specifications 3/4.11.1.2, 3/4.11.4, 3/4.11.2.1, 3/4.11.2.2, and 3/4.11.2.3, now referred to as Controls 3.1, 3.2, 3.3, 3.4, and 3.5, respectively, relocated to Section 3.0;

8.

Technical Specifications 3/4.12.1, 3/4.12.2, and 3/4.12.3, now referred to as Controls 4.1, 4.2, and 4.3, respectively, relocated to Section 4.0;

9.

Technical Specification 3/4.3.3.6, now referred to as Control 5.1, relocated to Section 5.0;

10.

Technical Specification 3/4.3.3.7, now referred to as Control 5.2, relocated to Section 5.0 (Existing requirements for explosive gas monitoring instrumentation retained in Technical Specification 3/4.3.3.7);

Revision 19

- iii -

REVISION RECORD Revision Date Description 8

08/19/92

11.

Technical Specifications 3/4.11.1.3 and 3/4.11.2.4, now referred to as Controls 6.1 and 6.2, respectively, relocated to Section 6.0;

12.

Section 7.0 created to contain reporting details for the Annual Radiological Environmental Monitoring Operating (Control 7.1) and Semiannual Effluent Release Reports (Control 7.2), and Major Changes to the Liquid and GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEMS (Control 7.3); and

13.

Corresponding Technical Specification Bases relocated with Technical Specifications to become part of controls.

b. All pages renumbered.

9 05/18/93 Replacement of milk sampling location TM-12 with TM-14 in Table 4.4 and Figure 4-2.

10 06/22/93 Technical Specification 3/4.3.3.3 (now referred to as Control 5.5) and its Bases relocated to Section 5.0; Technical Specification 3/4.3.3.3 nominal sensor elevations revised to reflect actual measurement heights; Technical Specification 3/4.3.3.3 Bases revised to eliminate reference to protective action recommendations.

11 10/31/96 Surveillance and analyses schedules for both the in-plant Gaseous and Liquid Effluent Monitoring Programs and the off-site REMP have been reduced. These reductions reflect changes in plant configuration due to plant dismantlement and decommissioning activities, and the elimination of radioactive source terms due to the cessation of the fission process with the shutdown of power operations.

12 02/07/97 The requirement to submit an annual summary of hourly meteorological data with the Semi-annual Radioactive Effluent Release Report due 60 days after January 1 of each year was eliminated.

Revision 19

- iv -

REVISION RECORD Revision Date Description 13 6/17199 Elimination of the Turbine Building composite sampler and analysis requirements due to change in plant configuration. Elimination of milk sampling location due to changes identified in the annual land use census.

Clarification of actions necessary to perform maintenance and testing on an in---line rad-monitor, plus expanded flexibility of operating Aux. Service Water for Spent Fuel Pit cooling while the effluent rad-monitor is out of service. Correction to block diagram (Figure 6-1) to reflect current liquid waste processing system configuration. Change in the reporting requirement for the Radioactive Effluent Release Report from semiannual to annual based on Technical Specification Amendment No. 151. Editorial changes to improve readability and correct or eliminate unnecessary text.

14 4/13/00 Clarifications to liquid effluent monitoring ACTION STATEMENTS 1 and 4 (Table 5.1) are included to provide clearer guidance on the existing intent of actions needed if the liquid radiation effluent monitor is out of service.

15 11/19/01

1.

Eliminates the requirement to maintain the 200 foot On-Site Meteorological Monitoring System (Section 5.5).

2.

References changed to reflect the relocation of Technical Specification Section 6.7 into the Yankee Decommissioning Assurance Program, Appendix D, Administrative Controls.

3.

Change the auxiliary service water flow rate to 120 gpm to reflect current operating performance and deletes the use of pump and valve curves to estimate flow rate.

Revision 19

REVISION RECORD Revision Date Description 16 8/14/03

1.

Eliminates detection requirements for radionuclides which, by natural decay, are no longer a significant effluent or of environmental concern.

2.

Eliminates milk sampling requirements from REMP.

3.

Adds new liquid waste treatment and discharge conditions for decommissioning of structures including the SFP.

4.

Updates site maps and monitoring programs to recognize the new ISFSI operations.

5.

Eliminates Control limits and measurement requirements for dissolved and entrained noble gases in liquid waste.

6.

Updates surveillance requirements for liquid effluent releases.

17 1/29/04

1.

Elimination of Noble Gas effluent dose and dose rate calculation methodology and Control Limit requirements due to removal of source potential from plant systems.

2.

Elimination of Tritium in gaseous effluent dose and dose rate calculation methodology due to removal of source potential from plant systems.

3.

Elimination of gaseous ventilation exhaust treatment system operability and gaseous monitoring instrumentation surveillance requirements (primary vent stack) due to dismantle of system equipment.

4.

Update and expansion of dose and dose rate conversion factors for use in Method I dose projections of airborne particulates released to the Revision 19

- vi -

REVISION RECORD Revision Date Description atmosphere during building demolition.

5.

Update site boundary figure (Fig. 1-2) to reflect changes in effluent release points.

6.

Update of site specific historical atmospheric dispersion factors to reflect removal of the plant vent stack (mixed mode release height) and accommodate ground level release conditions with no building wake credits.

7.

Elimination of the need for gaseous waste sampling and analysis requirements due to the removal of the plant vent stack release point.

8.

Inclusion of supplemental environmental air particulate monitoring in the vicinity of building demolition activities to validate effluent release model assumptions and modeling.

9.

Minor editorial changes to improve referencing in the ODCM.

18

1.

Elimination of airborne release pathway assessments and dose consequence calculation requirements following completion of site demolition activities that have removed airborne effluents.

2.

Reduction in REMP sampling to eliminate airborne pathways (i.e., air sampling, food product collection, and outer area TLD monitoring) due to the elimination of the airborne effluent potential.

3.

Elimination of annual requirement for the performance of a Land Use Census as a result of the removal of airborne effluent potential.

Revision 19

- vii -

REVISION RECORD Revision Date Description

+

4.

Elimination of liquid effluent monitoring/sampling requirements associated with process pathways and discharge points that no longer exist due to decommissioning activities.

19

1.

Replace Figure 1-1 with Figure 1-1a showing new exclusion area boundary.

2.

Add Figure 1-lb showing portion of site remaining under 10 CFR 50 License.

3.

Remove reference to air particulate sampler.

4.

Change sample frequencies for TLDs to semi-annual, groundwater to semi-annual, sediment to annual, fish to annual.

Revision 19

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LIST OF AFFECTED PAGES Page Rev. No.

Page Rev. No.

Cover ii iii iv v

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xi xii xiii xiv xv xvi xvii xviii xix xx 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 1-1 1-2 1-3 1-4 1-5 1-6 1-7 1-8 1-9 1-10 1-11 1-12 1-13 1-13a 1-14 2-1 2-2 2-3 2-4 2-5 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 18 18 18 18 18 3-1 3-2 3-3 3-4 3-5 3-6 3-7 3-8 3-9 3-10 3-11 3-12 3-13 3-14 3-15 18 18 18 18 18 18 18 18 18 18 18 18 18 18 18 4-1 4-2 4-3 4-4 4-5 4-6 4-7 4-8 4-9 4-10 4-11 4-12 4-13 4-14 4-15 4-16 4-17 4-18 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 19 Revision 19

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LIST OF AFFECTED PAGES (Continued)

Page I Rev. No.

Page Rev. No.

5-1 6-1 6-2 6-3 7-1 7-2 18 18 18 18 18 18 18 18 7-3 8-1 A-1 A-2 A-3 A-4 A-5 A-6 A-7 A-8 A-9 A-10 A-11 A-12 A-13 A-14 A-1 5 A-16 A-17 A-18 A-19 A-20 A-21 A-22 A-23 A-24 A-25 A-26 A-27 A-28 A-29 A-30 A-31 A-32 A-33 A-34 A-35 A-36 A-37 A-38 A-39 A-40 A-41 A-42 A-43 A-44 A-45 A-46 A-47 A-48 B-1 B-2 B-3 B-4 B-5 B-6 B-7 B-8 B-9 B-10 B-11 11 11 11 11 11 11 11 11 11 11 11 Revision 19

DISCLAIMER OF RESPONSIBILITY This document was prepared for use by Yankee Atomic Electric Company ("Yankee"). The use of information contained in this document by anyone other than Yankee, or the Organization for which the document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document.

Revision 19

- Xi -

ABSTRACT The Yankee Nuclear Power Station (YNPS) OFF-SITE DOSE CALCULATION MANUAL (ODCM) contains the methodology and parameters used in the calculation of off-site doses resulting from radioactive effluents and in the conduct of the Environmental Radiological Monitoring Program. The ODCM also contains (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Appendix D, Section B.5, of the Yankee Decommissioning Quality Assurance Program (YDQAP) and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release required by controls 7.1 and 7.2, respectively. With initial approval by the U.S. Nuclear Regulatory Commission and the YNPS Plant Operation Review Committee (PORC) and subsequent revisions reviewed by a knowledgeable individual and approved by the ISFSI Operations Manager, this manual is suitable to show compliance where referred to by the Yankee Decommissioning Quality Assurance Programa) and controls listed in this document.

Revision 19

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TABLE OF CONTENTS REVISIO N RECO RD..................................................................................................

ii LIST O F AFFECTED PAG ES....................................................................................

ix DISCLAIM ER O F RESPO NSIBILITY.............................................................................

xi ABSTRACT....................................................................................................................

xii TABLE O F CO NTENTS................................................................................................

xiii LIST O F CO NTRO LS...................................................................................................

xvii LIST O F TABLES........................................................................................................

xviii LIST O F FIGURES.....................................................................................................

xx 1.0 INTRO DUCTION 1-1 1.1 Summary of Methods, Dose Factors, Constants, Variables, and Defintions...... 1-1 1.2 Applicability of Controls and Surveillance Requirem ents (SR).......................... 1-2 2.0 RADIOACTIVE LIQ UID EFFLUENTS..........................................................................

2-1 2.1 Off-Site Concentrations....................................................................................

2-1 2.2 M ethod to Calculate Off-Site Liquid Concentrations..........................................

2-4 2.3 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway 2-4 2.3.1 Test Tank Pathway (Deleted)................................................................

2-4 2.3.2 Auxiliary Service Water System Pathway (Deleted)..............................

2-5 2.3.3 Rem aining Pathways............................................................................

2-5 3.0 DOSE/DOSE RATE CONTROLS AND CALCULATIONS............................................

3-1 3.1 Dose Due to Radioactive Liquid Effluents.........................................................

3-1 3.2 Total Dose....................................................................................................

3-3 3.3 Dose Rate Due to Radioactive Gaseous Effluents (Deleted)............................

3-5 3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effl uents (Deleted)............................................................................................

3-5 3.5 Dose Due to Radionuclides in Particulate Form With Half-Lives Greater than Eight Days (Deleted).........................................................................................

3-5 3.6 Dose Calculation Concepts...............................................................................

3-6 3.7 Method to Calculate the Total Body Dose from Liquid Releases....................... 3-7 3.7.1 M ethod I................................................................................................

3-7 3.7.2 Method II...............................................................................................

3-8 3.7.3 Basis for Method I.................................................................................

3-8 Revision 19

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TABLE OF CONTENTS (Continued)

Paqe 3.8 Method to Calculate Maximum Organ Dose from Liquid Releases................. 3-12 3.8.1 Method I..............................................................................................

3-12 3.8.2 Method II.............................................................................................

3-13 3.8.3 Basis for Method I...............................................................................

3-13 3.9 Method to Calculate the Total Body Dose Rate from Noble Gases (Deleted).. 3-13 3.9.1 Method I (Deleted)..............................................................................

3-13 3.9.2 Method II (Deleted).............................................................................

3-13 3.9.3 Basis for Method I (Deleted)................................................................

3-13 3.10 Method to Calculate the Skin Dose Rate from Noble Gases (Deleted)............ 3-13 3.10.1 Method I (Deleted)..............................................................................

3-13 3.10.2 Method II (Deleted).............................................................................

3-13 3.10.3 Basis for Method I (Deleted)................................................................

3-13 3.11 Method to Calculate the Critical Organ Dose Rate from Particulates with Half-Lives Greater Than Eight Days (Deleted).......................................................

3-14 3.11.1 Method I (Deleted)..............................................................................

3-14 3.11.2 Method II (Deleted).............................................................................

3-14 3.11.3 Basis for Method I (Deleted)................................................................

3-14 3.12 Method to Calculate the Gamma Air Dose from Noble Gases (Kr-85) (Deleted)............................................................................................

3-14 3.12.1 Method I (Deleted)..............................................................................

3-14 3.12.2 Method II (Deleted).............................................................................

3-14 3.12.3 Basis for Method I (Deleted)................................................................

3-14 3.13 Method to Calculate the Beta Air Dose from Noble Gases (Deleted).............. 3-14 3.13.1 Method I (Deleted)..............................................................................

3-14 3.13.2 Method II (Deleted).............................................................................

3-14 3.13.3 Basis for Method I (Deleted)................................................................

3-14 Revision 19

- xiv -

TABLE OF CONTENTS (Continued)

Page 3.14 Method to Calculate the Critical Organ Dose from Particulates (Deleted)....... 3-14 3.14.1 Method I (Deleted)..............................................................................

3-14 3.14.2 Method II (Deleted).............................................................................

3-14 3.14.3 Basis for Method I (Deleted)................................................................

3-14 3.15 Critical Receptors and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways (Deleted).........................................................

3-14 3.15.1 Critical Receptors (Deleted)................................................................

3-14 3.15.2 Yankee Atmospheric Dispersion Model (Deleted)...............................

3-14 3.15.3 Long-Term Average Dispersion Factors for Critical Receptors (D e leted ).............................................................................................

3-14 3.16 Method to Calculate Direct Dose from Site Operation.....................................

3-14 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING...................................................

4-1 4.1 M o nito ring.........................................................................................................

4 -1 4.2 Land Use Census (Deleted)..............................................................................

4-9 4.3 Intercomparison Program...............................................................................

4-10 4.4 Environmental Monitoring Locations...............................................................

4-10 5.0 INSTRUMENTATION (Deleted)...................................................................................

5-1 5.1 Radioactive Liquid Effluents (Deleted)..............................................................

5-1 5.2 Radioactive Gaseous Effluents (Deleted).........................................................

5-1 5.3 Liquid Effluent Instrumentation Setpoints (Deleted)..........................................

5-1 5.3.1 Method (Deleted)..................................................................................

5-1 5.3.2 Liquid Effluent Setpoint Example (Deleted)...........................................

5-1 5.3.3 Basis (Deleted).....................................................................................

5-1 5.4 Gaseous Effluent Instrumentation Setpoints (Deleted)......................................

5-1 5.4.1 Method (Deleted)..................................................................................

5-1 5.4.2 Gaseous Effluent Setpoint Example (Deleted)......................................

5-1 5.4.3 Basis (Deleted).....................................................................................

5-1 Revision 19

-Xv-

TABLE OF CONTENTS (Continued)

Paqe 6.0 RADIOACTIVE WASTE TREATMENT SYSTEMS, AND EFFLUENT PATHWAYS...... 6-1 6.1 Liquid Radioactive Waste Treatment............................................................

6-1 6.2 Gaseous Radioactive Waste Treatment (Deleted)...................

6-2 6.3 Liquid Effluent Streams and Temporary or Portable Construction Dewatering T reatm ent System s...........................................................................................

6-2 6.4 Liquid Effluent Pathways...................................................................................

6-2 6.5 Gaseous Effluent Pathways (Deleted)..............................................................

6-2 7.0 REPORTING REQUIREMENTS..................................................................................

7-1 7.1 7.2 7.3 7.4 Annual Radiological Environmental Operating Report......................................

7-1 Annual Radioactive Effluent Release Report....................................................

7-2 Major Changes to Liquid Radioactive Waste Treatment Systems (Deleted)...... 7-3 Special Reports................................................................................................

7-3 8.0 R EFER E N C ES.............................................................................................................

8-1 APPENDIX A: DISPOSAL OF SEPTAGE.........................................................................

A-1 APPENDIX B: CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BAC KG RO U N D............................................................................................

B-1 Revision 19

- xvi -

LIST OF CONTROLS Control Title Pace 1.1 Applicability of Controls and Surveillance Requirements 1-2 1.2 Applicability of Controls and Surveillance Requirements 1-2 2.1 Off-Site Concentrations 2-1 3.1 Dose Due to Radioactive Liquid Effluents 3-1 3.2 Total Dose 3-3 3.3 Dose Rate Due to Radioactive Gaseous Effluents (Deleted) 3-5 3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effluents 3-5 (Deleted) 3.5 Dose Due to Radionuclides in Particulate Form With Half-Lives Greater than Eight Days (Deleted) 3-5 4.1 Monitoring Program 4-1 4.2 Land Use Census (Deleted) 4-9 4.3 Intercomparison Program 4-10 5.1 Radioactive Liquid Effluents (Deleted) 5-1 5.2 Radioactive Gaseous Effluents (Deleted) 5-1 6.1 Liquid Radioactive Waste Treatment 6-1 6.2 Gaseous Radioactive Waste Treatment (Deleted) 6-2 7.1 Annual Radiological Environmental Operating Report 7-1 7.2 Annual Radioactive Effluent Release Report 7-2 7.3 Major Changes to Liquid and Gaseous Radioactive Waste Treatment Systems (Deleted) 7-3 7.4 Special Reports 7-3 Revision 19

- xvii -

LIST OF TABLES Table Title Paqe 1.1 Summary of Concentration and Method I Dose Equations 1-3 1.2 Dose Factors Specific to the Yankee Plant for Noble Gas Releases (Deleted) 1-4 1.3 Summary of Radiological Effluent Controls and Implementing Equations 1-5 1.4 Summary of Constants (Deleted) 1-6 1.5 Summary of Variables 1-7 1.6 Definition of Terms 1-9 1.7 Dose Factors Specific to the Yankee Site for Liquid Releases 1-11 1.8 Dose and Dose Rate Factors Specific to the Yankee Plant for Particulate Gaseous Releases (Deleted) 1-11 1.9 Frequency Notation 1-12 2.1 Radioactive Liquid Waste Sampling and Analysis Program 2-2 3.1 Radioactive Gaseous Waste Sampling and Analysis Program (Deleted) 3-5 3.2 Environmental Parameters for Liquid Effluents at Yankee Rowe 3-10 3.3 Age-Specific Usage Factors for Various Liquid Pathways at Yankee Rowe 3-11 3.4 Age-Specific Usage Factors (Deleted) 3-14 3.5 Environmental Parameters for Gaseous Effluents at the Yankee Plant (Deleted)3-14 3.6 Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion Factors (Deleted) 3-14 4.1 Radiological Environmental Monitoring Program 4-4 4.2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 4-5 Revision 19

- xviii -

LIST OF TABLES (Continued)

Table Title 4.3 Detection Capabilities for Environmental Sample Analysis 4.4 Radiological Environmental Monitoring Stations 5.1 Radioactive Liquid Effluent Monitoring Instrumentation (Deleted) 5.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements (Deleted) 5.3 Radioactive Gaseous Effluent Monitoring Instrumentation (Deleted) 5.4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements (Deleted)

Revision 19 Page 4-6 4-11 5-1 5-1 5-1 5-1

- xix -

LIST OF FIGURES Table Title PaQe 1-1a Yankee Nuclear Power Station, SITE BOUNDARY LINES 1-13 1-1b Current 10 CFR Part 50 Licensed Site Boundary 1-13a 1-2 Yankee Atomic Electric Company, Effluent Discharge Points, Site Plot Plan 1-14 4-1 Radiological Environmental Monitoring Locations Within 1 Mile (Waterborne Pathways) 4-13 4-2 Radiological Environmental Monitoring Locations Within 12 Miles (Waterborne Pathways) 4-14 4-3 Radiological Environmental Monitoring Locations Outside 12 Miles (Waterborne Pathways) 4-15 4-4 Radiological Environmental Monitoring Locations at the Restricted Area Fence (Direct Radiation Pathway) 4-16 4-5 Radiological Environmental Monitoring Locations Within 1 Mile (Direct Radiation Pathway) 4-17 4-6 Radiological Environmental Monitoring Locations Within 12 Miles (Direct Radiation Pathway) 4-18 4-7 Yankee Plant Radiological Environmental Monitoring Locations Outside 12 Miles (Direct Radiation Pathway) (Deleted) 4-18 6-1 Liquid Effluent Streams and Radioactive Waste Treatment Systems 6-3 6-2 Gaseous Effluent Streams, Radiation Monitors, and Temporary or Portable Construction Dewatering Treatment System at the Yankee Plant (Deleted) 6-3 Revision 19

-xx-

1.0 INTRODUCTION

According to Definition of Terms (Table 1.6), the OFF-SITE DOSE CALCULATION MANUAL (ODCM) contains the methodology and parameters used in the calculation of off-site doses resulting from radioactive liquid effluents and in the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains: (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Program required by Appendix D, Section B.5, of the Yankee Decommissioning Quality Assurance Program (YDQAP) (Reference j) and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Controls 7.1 and 7.2, respectively. The ODCM forms the basis for site procedures which document the off-site doses due to plant decommissioning activities which are used to show compliance with the numerical guides for design controls of Section II, Appendix I, 10CFR Part 50.

The methods contained herein follow accepted NRC guidance, unless otherwise noted in the text. The basis for each method is sufficiently documented to allow regeneration of the methods by an experienced health physicist.

All changes to the ODCM shall be reviewed by a knowledgeable individual and approved by the ISFSI Operations Manager in accordance with YDQAP, Appendix D, Section B.5.d prior to implementation. Changes made to the ODCM shall be submitted to the Commission for their information in the Annual Radioactive Effluent Release Report for the period in which the change(s) was made effective.

1.1 Summary of Methods, Dose Factors, Limits, Constants, Variables, and Definitions This section summarizes the methods for the user. In addition, the applicability of controls and surveillance requirements are listed in this section. The concentration method is documented in Table 1.1, as well as the Method I dose equations. Where more accurate dose calculations are needed, use the Method II for the appropriate dose as described in Sections 3.7 through 3.14 and 3.16. The dose factors used in the equations are in Table 1.7 and the regulatory limits are summarized in Table 1.3. The constants, variables, special definitions, and FREQUENCY NOTATION used in the Revision 19 1-1

ODCM are in Tables 1.4, 1.5, 1.6, and 1.10, respectively. Lastly, Figures 1-1 and 1-2 depict the Yankee site boundary line and liquid effluent discharge points, respectively.

1.2 Applicability of Controls and Surveillance Requirements (SR)

Control 1.1 The controls and ACTION requirements shall be applicable during conditions specified for each control.

Control 1.2 Adherence to the requirements of the controls and/or associated ACTION within the specified time interval shall constitute compliance with the control. In the event that the control is restored prior to expiration of the specified time interval, completion of the ACTION statement is not required.

SR 1.1 Surveillance requirements shall be applicable during the conditions specified for individual controls.

SR 1.2 with:

Each surveillance requirement shall be performed within the specified time interval

a.

A maximum allowable extension not to exceed 25 percent of the surveillance interval, and

b.

Deleted SR 1.3 Performance of a surveillance requirement within the specified time interval shall constitute compliance with OPERABILITY requirements for a control and associated ACTION statements unless otherwise required by the control.

Revision 19 1-2

TABLE 1.1 Summary of Concentration and Method I Dose Eauations Equation No.

Maximum Equation(a) 2-1 Unrestricted Area, Total Fraction of MPC in C.

Liquids F1 =

MP'i

' i 2-2 Deleted 3-1 Total Body Dose Due to Liquids Dtb (mrem) = K

Q, DFLb 3-2 Maximum Organ Dose Due to Liquids Dorgan (mrem) = K

Qi DFLimo 3-3 Deleted 3-4 Deleted 3-5 Deleted 3-6 Deleted 3-6.1 Deleted 3-7 Deleted 3-8 Deleted 5-1 Deleted 5-3 Deleted 5-4 Deleted Note (a):

Ci

=

Concentration of radionuclide "i", at the point of discharge.

NOTE: Original Page 1-4 combined with this page and pages renumbered.

Revision 19 1-3

DFLitb DFLimo =

K

=

Qi

=

TABLE 1.1 (Continued)

Summary of Concentration and Method I Dose Equations Site-specific, total body dose factor for a liquid release of radionuclide "i".

Site-specific, maximum organ dose factor for a liquid release of radionuclide "i".

Deerfield River flow rate correction factor.

Total release (Curies) for radionuclide "i".

TABLE 1.2 Dose Factors Specific to the Yankee Plant for Noble Gas Releases Deleted NOTE: Original Page 1-6 (Table 1.2) deleted and pages renumbered.

Revision 19 1-4

TABLE 1.3 Summary of Radiological Effluent Controls and Imolementina Eauations Control Category Method Limit 2.1 Off-Site Total Fraction of MPC Eq. 2-1

< 1.0 Concentrations Deleted Deleted Deleted of Liquids 3.1 Dose Due to Total Body Dose Eq. 3-1

_ 1.5 mrem in a qtr.

Liquid Effluents

< 3.0 mrem in a yr.

Organ Dose Eq. 3-2

_ 5.0 mrem in a qtr.

< 10.0 mrem in a yr.

3.2 Total Dose Due Total Body Dose Eq. 3-1

< 25.0 mrem in a yr.

to Liquid and Eq. 3-6 Gaseous Eq. 3-9 Effluents Organ Dose Eq. 3-2

< 25.0 mrem in a yr.

Eq. 3-8 Eq. 3-9 Thyroid Dose Eq. 3-2

< 75.0 mrem in a yr.

Eq. 3-8 Eq. 3-9 3.3 Dose Rate Due Deleted to Gaseous Effluents Deleted Deleted 3.4 Dose Due to Deleted Noble Gases in Gaseous Effluents Deleted 3.5 Dose Due Deleted Particulates in Gaseous Effluents 5.1 Liquid Effluent Deleted Monitor Setpoint Revision 19 1-5

TABLE 1.3 (Continued)

Summary of Radioloqical Effluent Controls and ImDlementina Eauations Control Category Method Limit 5.2 Gaseous Deleted Effluent Monitor Setpoint Deleted 6.1 Liquid Total Body Dose Eq. 3-1

_ 0.06 mrem in a mo.

Radioactive Organ Dose Eq. 3-2

< 0.2 mrem in a mo, Waste Treatment 6.2 Gaseous Deleted Radioactive Waste Treatment Deleted Deleted TABLE 1.4 Summary of Constants Deleted NOTE: Original Page 1-9 (Table 1.4) deleted and pages renumbered.

Revision 19 1-6

TABLE 1.5 Summary of Variables Variable Definition Units 1.00 x 10+6 Number of picocuries per microcurie.

pCi DlCi Deleted Deleted C.

Concentration of radionuclide "i", at the point of pLCi discharge.

cc Deleted Deleted Deleted Deleted Deleted Deleted Deleted Dt=

Dose to the total body.

mrem Deleted Deleted Deleted Deleted Deleted Deleted Deleted DFLimo

=

Maximum organ liquid dose factor for mrem radionuclide "i".

Ci DFLitb

=

Total body liquid dose factor for radionuclide "i".

mrem, Ci Deleted Deleted Deleted Deleted Deleted NOTE: Original Page 1-11 combined with this page and pages renumbered.

Revision 19 1-7

TABLE 1.5 (Continued)

Summary of Variables Variable Definition Units F,

=

Total fraction of MPC in liquid pathways.

Deleted Deleted Deleted Deleted MPCi

=

Maximum permissible concentration of radionuclide 1Cii "i" (10CFR Part 20, Appendix B, Table 2, Column 2, cc see Appendix B of the ODCM).

Q=

Release for radionuclide "i".

Ci Deleted Deleted Deleted Deleted Deleted SF

=

Shielding factor.

Deleted Deleted NOTE: Original Page 1-13 combined with this page and pages renumbered.

Revision 19 1-8

TABLE 1.6 Definition of Terms The defined terms of this section appear in capitalized type and are applicable throughout this document.

ACTION ACTION shall be those additional requirements specified as corollary statements to each principle control and shall be part of the controls.

FREQUENCY-NOTATION The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined in Table 1.9.

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC (for purposes of 10CFR50, Appendix I) shall include all persons who are not occupationally associated with the site. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category, are persons who enter the site to Revision 19 1-9

TABLE 1.6 (Continued)

Definition of Terms service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the site operations or decommissioning of the plant.

OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The ODCM contains the methodology and parameters used in the calculation of off-site doses resulting from radioactive liquid effluents and in the conduct of the Environmental Radiological Monitoring Program. The ODCM also contains (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Appendix D, Section B.5 of the YDQAP and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Controls 7.1 and 7.2, respectively.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. Any area within the SITE BOUNDARY used for residential quarters or recreational purposes shall be considered to be beyond the SITE BOUNDARY for purposes of meeting gaseous effluent dose controls. (Realistic occupancy factors shall be applied at these locations for the purposes of dose calculations.)

NOTE: Original Page 1-16 (Table 1.6) deleted and pages renumbered.

Revision 19 1-10

TABLE 1.7 Dose Factors Specific to the Yankee Site for Liquid Releases Total Body Dose Maximum Organ Factor Dose Factor (mrem_

(mrem' DFLitb ere*

DFLimo i

Radionuclide Ci Ci H-3 5.99 x 104 5.99 x 104 C-14 1.64 x 10"0 8.18 x 10+0 Fe-55 3.46 x 10-2 2.11 x 101 Co-60 2.79 x 10' 9.04 x 10-1 Sr-90 6.97 x 10+1 2.75 x 10+2 Ag-11 Om 2.32 x 10-2 2.21 x 10+0 Cs-134 1.79 x 10+'

2.40 x 10+1 Cs-137 1.07 x 10+1 2.07 x 10+1 Ag-1 08m/Ag-1 08 5.70 x 101 1.81 x 10+1 Table 1.8 Dose and Dose Rate Factors Specific to the Yankee Plant for Particulate Gaseous Releases Deleted NOTE: Original Page 1-18 (Table 1.8) deleted and pages renumbered.

Revision 19 1-11

TABLE 1.9 Frequency Notation Notation Frequency W

At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

P Prior to each release.

N.A.

Not applicable.

Revision 19 1-12

YAEC PropertyBoundary Scale 1:25,000 0.5km 0

-W m

-- m 0.5 mi 0

1,000 ft Figure 1-1a - Yankee Nuclear Power Station Site Boundary Lines Yankee Nuclear Power Station - Rowe, MA ERM Revision 19

Figure 1-lb 4 Current 10 CFR Part 50 Licensed Site Boundary A

10 CFR Par 50 Li[ensed Site 8oundary (12t 83 Impacted Area

/

/-,-

I

/

0 125 250 500 750 Fea t Map N=mb,,

-owO057i March 13, 2005

/

C.)

01 J

Sfzernn

'Pond 11 00.3 Plant Grid N 3.

N (True) 37 Dog _

ISFSI W44eeer BoDiverutenf Figure 12 Yankee Nuclear Power Station Effluent Discharge Points Site Plot Plan 0

100 200 300 A Approximab Scale Revision 19 1-14

2.0 RADIOACTIVE LIQUID EFFLUENTS 2.1 Off-Site Concentrations Control 2.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.5.a), the concentration of radioactive material released to Unrestricted Areas (see Figure 1-2) shall be limited to the concentrations specified in 10CFR Part 20, Appendix B, Table II, Column 2, for all.

Applicability At all times.

ACTION With the concentration of radioactive material released from the site to Unrestricted Areas exceeding the above limits, without delay, take actions to restore the concentration to within the above limits.

Surveillance Requirements SR 2.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 2.1.

SR 2.1.2 The results of radioactive analysis shall be used in accordance with the methods of the ODCM to assure that concentrations at the point of release are maintained within the limits of Control 2.1.

Bases Control 2.1 is provided to ensure that the at any time concentration of radioactive materials released in liquid waste effluents from the site above background (unrestricted areas for liquids is at the point of discharge into Sherman Pond/Deerfield River) will be less than the concentration levels specified in 10CFR Part 20, Appendix B, Table II, Column 2 (Appendix B of the ODCM contains a listing of these values as taken from the regulations). These requirements provide operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration numbers in Appendix B, but still within the annual average limitation of the revised (January 1, 1993 effective date) 1 OCFR, Part 20, regulation. Compliance with the design objective doses of Section II.A of Appendix I to 10CFR, Part 50, assure that doses are maintained ALARA, and that annual concentration limits of Appendix B to 10CFR20.1001-20.2401 will not be exceeded.

NOTE: Original Page 2-2 combined with this page and pages renumbered.

Revision 18 2-1

TABLE 2.1 Radioactive Liquid Waste Sampling and Analysis Progqram Liquid Release Type A. Batch Waste Release Tanks(b)

(including SFP draindown via batch effluent test tanks in batch mode)

Deleted (Pathway Abandoned)

B. Plant Continuous Releases(e)

Deleted Turbine Building (Pathway Abandoned)

Sump C. Continuous Continuous(d)

M/2 Principal Gamma 5.00 x 10-7 Releases(e):

Composite(d)

Emitters(f)

Construction Continuous(d)

M Tritium 1.00 x 10-1 Dewatering Composite(d)

Gross Alpha 1.00 x 10-7 Continuous(d)

Q Sr-90 5.00 x 10.8 Composite(d)

Fe-55 1.00 x 10.6 D. Batch Releases:

P P

Principal Gamma 5.00 x 10-7 Construction Each Batch Each Batch Emitters(O Dewatering(b)

Tritium 1.00 x 10-5 P

M Gross Alpha 1.00 x 10-Each Batch Composite(c)

P Q

Sr-90 5.00 x 10.8 Each Batch Composite(c)

Fe-55 1.00 x 10-6 Revision 18 I

2-2

TABLE 2.1 (Continued)

Notation

a.

The LLD is defined in Table Notation (a) of Table 4.3 of SR 4.1.

b.

A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, each batch shall be isolated and thoroughly mixed to assure representative sampling. For construction dewatering sources, a sample aliquot for every 1000 gallons (or less) of water transferred to temporary holiday tanks or basins not capable of recirculation or internal mixing shall be collected and composited to satisfy representative sampling requirements. Alternately, if three separate samples taken from the dewatering source indicate that the gamma emitter/tritium radioactivity does not vary by more than a factor of two, then the dewatering operation maybe considered as a continuous discharge where a composite sampler on the discharge line will collect a representative sample for analysis following the release.

c.

A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. If there is no effluent discharge during the period, no composite sample of collected waste is required.

d.

Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the average effluent release.

e.

A continuous release is the discharge of liquid wastes of a non-discrete volume; e.g.,

from a volume or system that has an input flow during periods when flow exist through the system.

f.

The principal gamma emitters for which the LLD requirement applies exclusively are the following radionuclides: Co-60, Cs-134 and Cs-137. This list does not mean that only these radionuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above radionuclides, also shall be identified and reported. Radionuclides that are below the LLD for the analyses should not be reported as being present at the LLD level.

Revision 18 2-3

2.2 Method to Calculate Off-Site Liquid Concentrations The basis for site procedures that the operator requires to meet Control 2.1, which limits the total fraction of MPC (FI) in liquid pathways at the point of discharge (see ODCM Figure 6-1) is discussed. (F1) is limited to less than or equal to one, i.e.,

- Z Ci i MPCi Evaluation of (F1) is required concurrent with the sampling and analysis program in Table 2.1 of Control 2.1.

Determine the total fraction of MPC in liquid pathways as follows:

I Ci (F1)

= x:-..MPGi (Eq.2-1)

Where:

MPC

=

Maximum permissible concentration of radionuclide "i" (10CFR Part 20, Appendix B, Table 2, Column 2. See Appendix B of ODCM for listing).

Ci

=

Concentration at the point of discharge of radionuclide "i" from any significant sources which may be created during plant decommissioning activities.

2.3 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway 2.3.1 Section Deleted NOTE: Original Page 2-6 combined with this page and pages renumbered.

Revision 18 I

2-4

2.3.2 Section Deleted 2.3.3 Remaining Pathways Ci is determined for each of the remaining pathways as follows:

a. Deleted b, Construction Dewatering: The dismantling of buildings and related structures, including foundation excavations, may fill with either ground water or storm water.

The water-filled excavations, in many cases, must be dewatered to complete the dismantling activity. Construction dewatering may also include water generated during the process of digging new ground water monitoring wells, or other dismantlement/demolition related water and waste water sources. This discharge will be directed to either the existing storm drain systems or process treatment flow path with final release to Sherman Pond or the Deerfield River just below the Sherman Dam. With respect to effluent control, dewatering will be sampled and analyzed to determine the radionuclide content as detailed in Table 2.1. Releases to the environment without treatment will occur only if the projected impact is less than ODCM Control 6.1 dose limits. If higher activity water is found, it will be treated as appropriate (see Figure 6.1) prior to release to reduce the radionuclide inventory (other than tritium).

Revision 18 I

I 2-5

3.0 ; DOSE/DOSE RATE CONTROLS AND CALCULATIONS 3.1 Dose Due to Radioactive Liquid Effluents Control 3.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.5.a), the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the site (see Figure 1-2) to available uptake pathways shall be limited:

a.

During any calendar quarter: less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and

b.

During any calendar year: less than or equal to 3 mrem to the total body and less than or equal to 10 mrem to any organ.

Applicability At all times.

ACTION

a.

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, and if not applicable to 1 OCFR Part 50.73, prepare and submit to the Commission within 30 days, pursuant to Control 7.4, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits.

Surveillance Requirement SR 3.1 Dose Calculations - Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

Bases Control 3.1 is provided to implement the requirements of Sections II.A, IllI.A, and IV.A of Appendix I, 10CFR Part 50. The control implements the guides set forth in Section II.A. The Revision 18 3-1

ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. The surveillance requirement implements the requirements in Section IlI.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. Existing pathways of liquid exposure to MEMBER(S) OF THE PUBLIC which form the basis for calculating liquid doses in the ODCM are described in detail in Yankee Atomic Electric Company's design report, "Supplemental Information for the Purpose of Evaluation of 10CFR Part 50, Appendix I", dated June 2, 1976 (with amendments) (Reference f). The point of exposure from existing pathways for dose calculational purposes is taken downstream of Sherman Dam in the Deerfield River. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from 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 10CFR 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. Also, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in finished drinking water that are in excess of the requirements of 40CFR141. No drinking water supplies from the Deerfield River below the site have been identified.

Revision 18 3-2

3.2 Total Dose Control 3.2 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.5.a), the dose or dose commitment to any real MEMBER OF THE PUBLIC from all station sources is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) over a calendar year.

Applicability At all times.

ACTION

a.

With the calculated dose from the release of radioactive materials in liquid effluents exceeding twice the limits of Controls 3.1.a, 3.1.b calculations should be made including direct radiation contributions from the Independent Spent Fuel Storage Installation (ISFSI) and from outside storage tanks to determine whether the above limits of Control 3.2 have been exceeded. If such is the case, and if not applicable to 10CFR Part 50.73, prepare and submit to the Commission within 30 days, pursuant to Control 7.4, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. The Special Report shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from site sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by the report. It also shall describe levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR1 90.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

Revision 18 3-3

Surveillance Requirement SR 3.2 Dose Calculations - Cumulative dose contributions from liquid effluents shall be determined in accordance with SR 3.1 and in accordance with the ODCM.

Bases Control 3.2 is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46FR18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses from site radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40CFR Part 190 if the individual reactors remain within the reporting requirement level.

The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to a MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11, is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 1 OCFR Part 20, as addressed in liquid and gaseous effluent controls.

Revision 18 I

3-4

3.3 Dose Rate Due to Radioactive Gaseous Effluents Control 3.3 Section Deleted 3.4 Dose Due to Noble Gases Released in Radioactive Gaseous Effluents Control 3.4 Section Deleted 3.5 Dose Due to Radionuclides in Particulate Form With Half-Lives Greater than Eight Days Control 3.5 Section Deleted NOTE: Original Pages 3-6, 3-7, 3-8 (Table 3.1), 3-9, 3-10, and 3-11 deleted/combined with this page and pages renumbered.

TABLE 3.1 Radioactive Gaseous Waste Sampling and Analysis Program Deleted Revision 18 I

I 3-5

3.6 Dose Calculation Concepts The term "dose" for ingested or inhaled radioactivity means the dose commitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases "annual dose" or "dose in one year" then refer to the fifty-year dose commitment from one year's worth of releases. "Dose in a quarter" similarly means a fifty-year dose commitment from one quarter's releases. The term "dose,"

with respect to external exposures, such as to ground phase deposition, refers only to the doses received during the actual time period of exposure to the radioactivity released from the site.

Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment.

The dose calculated by "Method I" equations is not necessarily the actual dose received by a real individual, but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the selection and definition of the critical receptors. The radioisotope specific dose factors in each "Method I" dose equation represent the greatest dose to any organ of any age group accounting for existing or potential pathways of exposure. The critical receptor assumed by "Method I" equations is typically a hypothetical individual whose behavior in terms of location and intake results in a dose which is expected to be higher than any real individual. Method II allows for a more exact dose calculation for real individuals if necessary by considering only existing pathways of exposure with the recorded release.

Revision 18 3-6

3.7 Method to Calculate the Total Body Dose from Liquid Releases Control 3.1 limits the total body dose commitment to a MEMBER OF THE PUBLIC from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year. Control 6.1 requires liquid radioactive waste treatment when the total body dose estimate exceeds 0.06 mrem in any 31-day period. Control 3.2 limits the total body dose commitment to any real MEMBER OF THE PUBLIC from all station sources (including liquids) to 25 mrem in a year.

Dose evaluation is required at least once per 31 days. If a Temporary or Portable Construction Dewatering Treatment System is not being used, dose evaluation is required before each release.

To evaluate total body dose for Control 6.1 add the total body dose from today's expected releases to the total body dose accumulated for the time period of interest.

3.7.1 Method I The total body dose from a liquid release is:

Dtb K-Qi gFLilb (Eq. 3-1)

(mrem) i Where:

DFLit

= Site-specific total body dose factor (mrem/Ci) for liquid release. See Table 1.7.

Q,

=

Total activity (Curies) released to liquids of radionuclide "i" during the period of interest. For i = Fe-55, Sr-90, or H-3, use the best estimates (such as the most recent measurements).

K

=

366/Fd; where Fd is the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in ft3/sec). If Fd cannot be obtained or Fd is greater than 366, K can be assumed to equal 1.0. The value, 366, is the ten-year minimum monthly average Deerfield River flow rate below Sherman Dam (in ft3/sec).

Revision 18 3-7

Equation 3-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

a.

Liquid releases to Sherman Pond or Deerfield River just below the Sherman Power House Dam (Outfalls 003 and 004 as specified in Reference K).

b.

Any continuous or batch release over any time period.

3.7.2 Method II If Method I cannot be applied or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data, and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data, or assumptions are more applicable.

3.7.3 Basis for Method I Method I may be used to show that the controls which limit off-site total body dose from liquids (Controls 3.1, 3.2, and 6.1) have been met for releases over the appropriate periods.

These requirements are based on design objectives and standards in 10CFR Part 50 and 40CFR Part 190. Control 3.1 is based on the ALARA design controls in 1 OCFR Part 50, Appendix I, Subsection II A. Control 6.1 is an "appropriate fraction", determined by the NRC, of the ALARA design control. Control 3.2 is based on Environmental Standards for the Uranium Fuel Cycle in 40CFR Part 190 which applies to direct radiation as well as liquid and gaseous effluents. Method I applies only to the liquid contribution.

Method I was developed such that "the actual exposure of an individual... is unlikely to be substantially underestimated (10CFR Part 50, Appendix I). The definition of a single "critical receptor" (a hypothetical individual whose behavior results in an unrealistically high dose) provides part of the conservative margin to the calculation of total body dose in Method I.

Method II allows that actual individuals with real behaviors be taken into account for any given release. In fact, Method I was based on a Method II analysis for the critical receptor and annual average conditions instead of any real individual. The analysis was called the "base case"; it Revision 18 3-8

was then reduced to form Method I. The base case, the method of reduction, and the assumptions and data used are presented.

The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor (in the form of dose factors in mrem/Ci) for a one Curie release of each radionuclide in liquid effluents was derived. The base case analysis uses the methods, data, and assumptions in Regulatory Guide 1.109 (Equations A-3, A-7, A-13, and A-16, Reference A). Tables 3.2 and 3.3 outline human consumption and environmental parameters used in the analysis. It is assumed that the critical receptor fishes below Sherman Dam and eats the fish caught from this location and consumes leafy vegetables and produce from a farm which is irrigated with water from the Deerfield River below Sherman Dam. It also is assumed that the critical receptor drinks milk and eats meat from cows who drink water from the Deerfield River below Sherman Dam and eat silage from the irrigated farm above.

For any liquid release during any period, the increment in annual average total body dose from radionuclide "i" is:

ADOm = (Qi) (DFLib) where DFLtb is the total body dose factor for radionuclide "i", and Qj is the activity of radionuclide "i" released in Curies.

Method I is more conservative than Method II because it is based on the following reduction of the base case. The dose factors, DFLitb, used in Method I were chosen from the base case to be the highest of the four age groups for that radionuclide. In effect, each radionuclide is conservatively represented by its own critical age group.

Revision 18 3-9.

TABLE 3.2 Environmental Parameters for Liquid Effluents at Yankee Rowe (Derived from Reference A)

Food Grown with Contaminated Water Aquatic Shoreline Leafy Variable Food Activity Vegetables Veg.

Meat Cow Milk MP Mixing Ratio(1) 0.84 0.84 0.84 0.84 0.84 0.84 TP Transit Time (hrs) 24.00 0.00 0.00 0.00 480.00 48.00 YV Agricultural Productivity (kg/m 2) 2.00 2.00 2.00 2.00 P

Soil Surface Density (kg/mr2) 240.00 240.00 240.00 240.00 IRR Irrigation Rate (1/m2/hr) 0.15 0.15 0.15 0.15 TE Crop Exposure Time (hrs) 1440.00 1440.00 1440.00 1440.00 TH Holdup Time (hrs) 1440.00 24.00 2160.00 2160.00 QAW Water Uptake Rate for (l/d) 50.00 60.00 Animal QF Feed Uptake Rate (kg/d) 50.00 50.00 Location of Critical Individual Below Below Below Below Below Below Sherman Sherman Sherman Sherman Sherman Sherman Dam Dam Dam Dam Dam Dam (1)

Listed mixing ratios apply to Method I dose factors. Method II analyses can apply calculated mixing ratios based on river flow and site discharge dilution flow which exist over the period of actual release.

Revision 18 3-10

TABLE 3.3 Acie-Soecific Usaqe Factors for Various Liauid Pathways at Yankee Rowe (From Reference A, Table E-5. Zero where no pathway exists)

Leafy Potable Age Veg.

Veg.

Milk Meat Fish Invert.

Water Shoreline Group (kg/yr)

(kg/yr)

(l/yr)

(kg/yr)

(kglhr)

(kg/yr)

(l/yr)

(hr/yr)

Adult 520.00 64.00 310.00 110.00 21.00 0.00 0.00 12.00 Teen 630.00 42.00 400.00 65.00 16.00 0.00 0.00 67.00 Child 520.00 26.00 330.00 41.00 6.90 0.00 0.00 14.00 Infant 000 0.00 330.00 0.00 0.00 0.00 0.00 0.00 Revision 18 3-11 °

3.8 Method to Calculate Maximum Organ Dose from Liquid Releases Control 3.1 limits the maximum organ dose commitment to a MEMBER OF THE PUBLIC from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year.

Control 6.1 requires liquid radioactive waste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any 31-day period. Control 3.2 limits the maximum organ dose commitment to any real MEMBER OF THE PUBLIC from all station sources (including liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Dose evaluation is required at least once per 31 days. If a Temporary Construction Dewatering Treatment System is not being used, dose evaluation is required before each release.

To evaluate the maximum organ dose for Control 6.1, add the organ dose from the expected releases to the organ dose accumulated for the time period of interest.

3.8.1 Method I The maximum organ dose from a liquid release is:

Dorgan = K-" Q DFLimo (Eq. 3-2)

(mrem) i Where:

DFLimo =

Site-specific maximum organ dose factor (mrem/Ci) for a liquid release. See Table 1.7.

Q

=

Total activity (Curies) released to liquids of radionuclide "i" during the period of interest. For i = Fe-55, Sr-90, or H-3, use the best estimates (such as the most recent measurements).

K

=

366/Fd; where Fd is the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in ft3/sec). If Fd cannot be obtained or Fd is greater than 366, K can be assumed to equal 1.0. The value, 366, is Revision 18 3-12

the ten-year minimum monthly average Deerfield River flow rate below Sherman Dam (in ft3/sec).

Equation 3-2 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

a.

Liquid releases to Sherman Pond or Deerfield River just below the Sherman Power House Dam, as permitted by Reference K.

b.

Any continuous or batch release over any time period.

3.8.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit, or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data, and assumptions in Regulatory Guide 1.109, Revision 1 (Reference a),

except where site-specific models, data, or assumptions are more applicable.

3.8.3 Basis for Method I The methods to calculate the maximum organ dose parallel the total body dose methods (see Section 3.7.3). Only the differences are presented here.

For any liquid release during any period, the increment in annual average dose from radionuclide "i" to the maximum organ is:

ADimo = (Qi) (DFLimo) where DFLimo is the maximum organ dose factor for radionuclide "i", and Q, is the activity of radionuclide "i' released in Curies.

The dose factors, DFLimo, used in Method I were chosen from the base case to be the highest of the set of seven organs and four age groups for each radionuclide. This means that the maximum effect of each radionuclide is conservatively represented by its own critical age group and critical organ.

3.9 Method to Calculate the Total Body Dose Rate from Noble Gases Section Deleted 3.10 Method to Calculate the Skin Dose Rate from Noble Gases Section Deleted NOTE: Original Pages 3-21 3-22, 3-23, 3-24 3-25, and 3-26 deleted/combined with this page and pages renumbered.

Revision 18 3-13

3.11 Method to Calculate the Critical Organ Dose Rate from Particulates with Half-Lives Greater Than Eight Days Section Deleted 3.12 Method to Calculate the Gamma Air Dose from Noble Gases (Kr-85)

Section Deleted 3.13 Method to Calculate the Beta Air Dose from Noble Gases Section Deleted 3.14 Method to Calculate the Critical Organ Dose from Particulates Section Deleted TABLE 3.4 Age-Specific Usage Factors Deleted TABLE 3.5 Environmental Parameters for Gaseous Effluents at the Yankee Plant Deleted 3.15 Critical Receptors and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways Section Deleted TABLE 3.6 Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion Factors Deleted 3.16 Method to Calculate Direct Dose from Site Operation Control 3.2 restricts the dose to the whole body and any organ of any real MEMBER OF THE PUBLIC at and beyond the Site Boundary from all station sources (including direct radiation) to the limit of 25 mrem in a year, except for the thyroid which is limited to 75 mrem in a year.

Estimates of direct exposure above background in areas at and beyond the site boundary can be determined from measurements made by environmental TLDs that are part of Revision 18 3-14

the Environmental Monitoring Program (see Table 4.4). Alternatively, direct dose calculations from identified fixed sources on site can be used to estimate the off-site direct dose contribution where TLD information may not be applicable.

NOTE: Original Pages 3-28, 3-29 3-30, 3-31, 3-32, 3-33, 3-34, 3-35 (Table 3.4),

3-36 (Table 3.5), 3-37, 3-38, 3-39, 3-40, 3-41 (Table 3.6), 3-42 deleted/combined and pages renumbered.

Revision 18 3-15

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 4.1 Monitoring Program Control 4.1 In accordance with Decommissioning Quality Assurance Program (Appendix D, Section B.5.b), the Radiological Environmental Monitoring Program shall be conducted as specified in Table 4.1.

Applicability At all times.

ACTION

a.

With the Radiological Environmental Monitoring Program not being conducted as specified in Table 4.1, prepare and submit to the Commission in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or to malfunction of automatic sampling equipment. If the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

b.

With the level of radioactivity as the result of site effluents in an environmental sampling media at one or more of the locations specified in Table 4.1 exceeding the reporting levels of Table 4.2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the receipt of the laboratory analyses, pursuant to Control 7.4, a Special Report which includes an evaluation of any release conditions, environmental factors, or other aspects which caused the limits of Table 4.2 to be exceeded. When more than one of the radionuclides in Table 4.2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) 1.0 reportinglevel (1) reportinglevel (2)

Revision 19 4-1

When radionuclides other than those in Table 4.2 are detected and are the result of site effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal or greater than the calendar year limits of Control 3.1. This report is not required if the measured level of radioactivity was not the result of site effluents, however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

Surveillance Requirement SR 4.1 The radiological environmental monitoring samples shall be collected pursuant to Table 4.1 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Table 4.1 and the detection capabilities required by Table 4.3.

Bases The Radiological Environmental Monitoring Program required by Control 4.1 provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of MEMBER(S) OF THE PUBLIC resulting from the site operation. The monitoring program implementsSection IV.B.2 of Appendix I, 10CFR Part 50, and thereby, supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. Guidance for the monitoring program is Revision 19 4-2

provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Program changes may be initiated based on operational experience.

The detection capabilities required by Table 4.3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. This does not preclude the calculation of an a posteriori LLD for a particular measurement based upon the actual parameters for the sample in question.

Revision 19 4-3

TABLE 4.1 Radiological Environmental Monitoring Program*

Exposure Pathway Number of Sample and/or Sample Locations Sampling and Collection Frequency Type and Frequency of Analysis AIRBORNE

a.

Deleted DIRECT RADIATION 6***

Semi-annually Gamma dose, at least once semi-annually.

WATERBORNE

a.

Surface 2

Composite sample** collected over a period of Gross beta and gamma isotopic analysis one month at downstream location; monthly of each sample. Tritium analysis of grab sample at upstream control location, composite sample at least once per quarter.

b.

Ground 2

Semi-annually Gamma isotopic and tritium analyses of each sample at least semi-annually.

c.

Sediment from 1"***

Annually Gamma isotopic analysis of each sample Shoreline at least annually.

INGESTION

a.

Fish 2

Commercially and recreationally important Gamma isotopic analysis on edible species. Seasonal or annually, if not seasonal.

portions at least annually.

b.

Deleted Specific sample locations for all media are specified in the ODCM and reported in the Annual Radiological Environmental Operating Report.

Composite samples shall be obtained by collecting an aliquot at intervals not exceeding two hours.

Does not include Restricted Area Fence locations, or those TLDs associated with the ISFSI pad monitoring. Includes 2 of the 4 designated control locations.

One sample from downstream area with existing or potential recreational value.

NOTE: Original Page 4-5 combined with this page and pages renumbered.

Revision 19 4-4

TABLE 4.2 Renortinl Levels for Radioactivity Concentrations in Environmental Samples Water*

COLUMN Fish COLUMN COLUMN Analysis (pCi/i)

DELETED (pCi/kg, wet)

DELETED DELETED H-3 3x 10+4 Mn-54 1 x 10+3 3 x 10+4 Co-58 1 x 10+3 3 x 104 Co-60 3 x 10+2 1 x 10+4 Zn-65 3 x 10+2 2 x 104 Zr-Nb-95 4 x 10+2 Cs-134 3 x 10 1 1x 10+3 Cs-137 5 x 10+1 2 x 10+3 Reporting levels for nondrinking water pathways.

Revision 19 4-5

TABLE 4.3 Detection Capabilities for Environmental Sample Analysis(a) (c)

Water COLUMN Fish COLUMN Sediment Analysis(d)

(pCi/I)

DELETED (pCi/kg. wet)

DELETED (pCi/kg. dry)

Gross beta 4 x 10+0 H-3 2 x 10+3 Line Deleted Co-58, -60 1.5 x 10+1 1.3 x 10+2 Line Deleted Line Deleted Cs-134 1.5 x 10+1 1.3 x 10+2 1.5 x 10+2 Cs-137 1.8 x 10+1 1.5 x 10+2 1.8 x 10+2 Revision 19 4-6

TABLE 4.3 (Continued)

Table Notation

a.

The LLD is the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD (4.66) (Sb)

(E)(V)(2.22)(Y)[Exp(-XAt)]

Where:

LLD

=

A priori lower limit of detection as defined above (microcuries or picocuries per unit mass or volume).

Sb

=

Standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute).

E

=

Counting efficiency (counts per disintegration).

V

=

Sample size (units of mass or volume).

2.22

=

Number of disintegrations per minute per picocurie.

Y

=

Fractional radiochemical yield (when applicable).

A

=

Radioactive decay constant for the particular radionuclide.

At

=

Elapsed time between sample collection and analysis.

Typical values of E, V, Y, and At can be used in the calculation. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include Revision 19 4-7

the typical contributions of other radionuclides normally present in the samples (e.g., Potassium-40 in fish samples).

Analysis shall be performed 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 may render these LLDs unavailable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. This does not preclude the calculation of an a posteriori LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis.

b.

Deleted

c.

If the measured concentration minus the 5 sigma counting statistics is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.

d.

This list does not mean that only these radionuclides are to be considered. Other peaks that are identifiable, together with those of the listed radionuclides, also shall be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Control 7.1.

Revision 19 4-8

4.2 Land Use Census Section Deleted.

Revision 19 4-9

4.3 Intercomparison Program Control 4.3 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.5.b), analyses shall be performed on referenced radioactive materials supplied as part of the quality assurance Laboratory Intercomparison Program.

Applicability At all times.

ACTION With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

Surveillance Requirement SR 4.3 A summary of the results of analyses performed as part of the above required Intercomparison Program shall be included in the Annual Radiological Environmental Operating Report.

Bases The control for participation in the Intercomparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed. The independent checks are completed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I, 10CFR Part 50.

4.4 Environmental Monitoring Locations The radiological environmental monitoring stations are listed in Table 4.4. The locations of these stations with respect to the Yankee site are shown on the maps in Figures 4-1 through 4-7.

Revision 19 4-10

TABLE 4.4 Radiological Environmental Monitoring Stations*

Exposure Pathway and/or Sample Location and Distance From Direction From Sample Designated Code**

the Site (km)

Site

1. AIRBORNE Deleted

2.

WATERBORNE

a. Surface WR-1 1 Bear Swamp Lower 6.3 Downriver Reservoir WR-21 Harriman Reservoir 10.1 Upriver

b. Ground WG-1 1 Site Potable On-Site Well WG-12 Sherman Spring 0.2 NW

c. Sediment From SE-1 1 Number 4 Station 36.2 Downriver Shoreline SE-21 Harriman Reservoir 10.1 Upriver

3. INGESTION Line Deleted

a. Fish FH-1 Sherman Pond 1.5 At Discharge Point FH-21 Harriman Reservoir 10.1 Upriver

4. DIRECT RADIATION GM-1 Furlon House 0.80 SW GM-2 Observation Stand 0.50 NW GM-6 Readsboro Road Barrier 1.30 N

GM-13 Restricted Area Fence 0.08 WSW GM-14 Restricted Area Fence 0.11 WNW GM-15 Restricted Area Fence 0.08 NNW GM-16 Restricted Area Fence 0.13 NNE GM-17 Restricted Area Fence 0.14 ENE GM-18 Restricted Area Fence 0.14 ESE GM-19 Restricted Area Fence 0.16 SE GM-20 Restricted Area Fence 0.16 SSE GM-21 Restricted Area Fence 0.11 SSW GM-22 Heartwellville(a) 12.60 NNW GM-27 Number 9 Road(a) 7.60 ENE GM-29 Route 8A(a) 8.20 ESE GM-31 Legate Hill Road(a) 7.60 SSE GM-40 Readsboro Road

.50 W

NOTE: Original Page 4-14 combined with this page and pages renumbered.

Revision 19 4-11

TABLE 4.4 (Continued)

Radiological Environmental Monitoring Stations*

Exposure Pathway Distance From Direction and/or Sample Sample Location and Designated Code the ISFSI (m)

From ISFSI

5.

DIRECT IF-1 ISFSI Security Fence 20 WNW RADIATION IF-2 Observation Stand 560 NW (ISFSI)

IF-3 ISFSI Security Fence 20 N

IF-4 ISFSI Security Fence 34 NE IF-5 ISFSI Security Fence 28 E

IF-6 ISFSI Security Fence 15 SE IF-7 ISFSI Security Fence 23 S

IF-7 ISFSI Security Fence 38 SW IF-9 Restricted Area Fence (site) 50 SE IF-10 Restricted Area Fence (site) 55 SSE IF-1I Restricted Area Fence (site) 135 SW IF-12 Restricted Area Fence (site) 225 N

IF-18 Near Original C. W. Intake 240 NNW IF-19 Restricted Area Fence (Admin Bldg)***

170 W

IF-20 Restricted Area Fence (Gate House)***

235 WNW IF-40 Readsboro Road 1

700 N

Sample locations are shown on Figures 4-1 through 4-7.

- Station 1X's are indicator stations, and Station 2X's are control stations (excluding the direct radiation stations)

Not included as part of the Radiological Environmental Monitoring Program.

a. Designated control locations for use with fixed radiation sources on-site (minimum of 2 required).

Revision 19 4-12

Figure 4-1 Revision 19 Radiological Environmental Monitoring Locations Within 1 Mile (Waterborne Pathways) 4-13

Figure 4-2 Revision 19 Radiological Environmental Monitoring Locations Within 12 Miles (Waterborne Pathways) 4-14

Figure 4-3 Revision 19 Radiological Environmental Monitoring Locations Outside 12 Miles (Waterborne Pathways) 4-15

Figure 4-4 Radiological Environmental Monitoring Locations at the Restricted Area Fence (Direct Radiation Pathway)

Revision 19 4-16

Figure 4-5 Revision 19 Radiological Environmental Monitoring Locations Within I Mile (Direct Radiation Pathway) 4-17

NW

-22 WNW n-Reservo' ENE t.E~NLMHGE NT

.pon-4 2 2........

4° o..°

.s*s....

-a.."chusetts W

E

...... -,Lower Reseror*

GM NOT:.r.inl.ag 4-22 (Figure..-7 deee..

SW

." SSW

S

KM

,0 2

4 Figure 4-6 Radiological Environmental Monitoring Locations Within 12 Miles (Direct Radiation Pathway)

NOTE: Original Page 4-22 (Figure 4-7) deleted.

Revision 19 4-18

5.0 INSTRUMENTATION 5.1 Radioactive Liquid Effluents Section Deleted TABLE 5.1 Radioactive Liquid Effluent Monitoring Instrumentation Deleted TABLE 5.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Deleted 5.2 Radioactive Gaseous Effluents Section Deleted TABLE 5.3 Radioactive Gaseous Effluent Monitoring Instrumentation Deleted Table 5.4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Deleted 5.3 Liquid Effluent Instrumentation Setpoints Section Deleted 5.4 Gaseous Effluent Instrumentation Setpoints Section Deleted NOTE: Original Page 5-2 through 5-21 (including Tables 5.1 - 5.4) deleted/combined with this page.

Revision 18 5-1

6.0 RADIOACTIVE WASTE TREATMENT SYSTEMS, AND EFFLUENT PATHWAYS 6.1 Liquid Radioactive Waste Treatment Control 6.1 In accordance with Yankee Decommissioning Quality Assurance Program (Appendix D, Section B.5.a), a Temporary or Portable Construction Dewatering Treatment System shall be used to reduce the radioactive materials in the liquid waste prior to its discharge when the estimated doses due to the liquid effluent from the site (see Figure 1-2) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.20 mrem to any organ.

Applicability At all times.

ACTION

a.

With liquid waste being discharged without processing through appropriate treatment systems as defined in the ODCM and estimated doses in excess of the above limits, and if not applicable to 10CFR Part 50.73, prepare and submit to the Commission within 30 days pursuant to Control 7.4, a Special Report which includes the following information:

1.

Explanation of why liquid radioactive waste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reasons for the inoperability;

2.

Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3.

Summary description of action(s) taken to prevent a recurrence.

Surveillance Requirement SR 6.1 Doses due to liquid releases shall be estimated at least once per 31 days in accordance with the ODCM. No dose estimates are required if a Temporary or Portable Construction Dewatering Treatment System has been continually used to reduce the radioactive materials in liquid waste prior to its discharge or if no liquid discharges have taken place over the appropriate 31-day period.

Bases The control that appropriate portions of a Temporary or Portable Construction Dewatering Treatment System be used when specified provides assurance that the releases of Revision 18 6-1

radioactive materials in liquid effluents will be kept "as low as is reasonably achievable."

Control 6.1 implements the requirements of 10CFR Part 50.36a, General Design Criterion 60 of Appendix A, 10CFR Part 50, and the design objective of Section II.D of Appendix I, 10CFR Part

50. The specified limits governing the use of appropriate portions of a Temporary or Portable Construction Dewatering Treatment System were specified as a suitable fraction of the dose design controls set forth in Section II.A of Appendix I, 1 0CFR Part 50, for liquid effluents.

6.2 Gaseous Radioactive Waste Treatment Section Deleted 6.3 Liquid Effluent Streams and Temporary or Portable Construction Dewatering Treatment Systems Figure 6-1 shows the liquid effluent streams and the appropriate liquid treatment system.

6.4 Liquid Effluent Pathways Figure 6.1 indicates the flow paths prior to release. The primary flow path is the construction dewatering of foundations and sumps of ground water and surface run-off. If treatment were needed to meet the radioactivity requirements of Control 6.1, temporary processing equipment as appropriate for the radionuclide content and water quality of the source shall be used.

6.5 Gaseous Effluent Pathways Section Deleted NOTE: Original Pages 6-3, 6-4, 6.5 and 6-6 deleted/combined with this page and pages renumbered.

Revision 18 I

6-2

FIGURE 6-1 Liquid Effluent Streams and Temporary or Portable Construction Dewaterinq Treatment System 7Temporary Skid -

Misc. Waste Water (evaporation, ion exchange I

rIL and/or filtration)

I I.

I I

I.+

Northwest*

Storm Drains Northeast

+

Storm Drains Outfall IOutfall 003j04 ShemanPon Deerfield River Sherma PondSherman Dam Contingency Path For construction dewatering, process treatment, if needed, will be provided before release to the environment. Treatment for radioactivity reduction could include any combination of evaporation, ion exchange and/or filtration by temporary process equipment as appropriate for the radionuclide content and waler quality of the source Miscellaneous waste water includes, but is not limited to, sources such as construction dewatering that are not derived from enclosed plant systems.

CDsuMUWIG6-1 NOTE: Original Page 6-8 (Figure 6-2) deleted and pages renumbered.

Revision 18 6-3

7.0 REPORTING REQUIREMENTS 7.1 Annual Radiological Environmental Operating Report Control 7.1

a.

An Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted to the NRC prior to May 1 of each year.

b.

The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the site operation on the environment.

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period pursuant to the table and figures in the ODCM. In the event that some results are not available to include in the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The report also shall include the following: a summary description of the Radiological Environmental Monitoring Program with a map of all sampling locations keyed to a table giving distances and directions from the reactor, the results of licensee participation in the Intercomparison Program required by Control 4.3, and a discussion of all analyses in which the LLD required by Table 4.3 was not achievable.

Revision 18 7-1

7.2 Annual Radioactive Effluent Release Report Control 7.2

a.

Before May 1 of each year, a report shall be submitted to the NRC covering the radioactive content of effluents released to unrestricted areas during the previous calendar year.

b.

The Annual Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit as outlined in Regulatory Guide 1.21, Revision 1, June 1974, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the format of Appendix B thereof.

In addition, the Annual Radioactive Effluent Release Report shall include an assessment of the radiation doses due to the radioactive effluents released from the site during the previous calendar year. This report also shall include an assessment of the radiation doses from radioactive effluents to MEMBER(S) OF THE PUBLIC due to the allowed recreational activities inside the SITE BOUNDARY (Figures 1-1 and 1-2) during the previous calendar year. All assumptions used in making these assessments (e.g., specific activity, exposure time, and location) shall be included in the report. Historical average meteorological conditions shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the ODCM.

The Annual Radioactive Effluent Release Report also shall include an assessment of radiation doses to the likely most exposed real MEMBER(S) OF THE PUBLIC from site releases (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40CFR190, "Environmental Radiation Protection Standards for Nuclear Power Operation," if Control 3.2 has been exceeded during the calendar year.

Revision 18 7-2

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to site boundary of radioactive materials in effluents made during the reporting period.

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the ODCM.

7.3 Major Changes to Liquid Radioactive Waste Treatment Systems Section Deleted 7.4 Special Reports Control 7.4 Special Reports shall be submitted pursuant to 10CFR50.4 within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference controls:

a.

Liquid Effluents, Controls 3.1 and 6.1.

b.

Deleted

c.

Total Dose, Control 3.2.

d.

Radiological Environmental Monitoring, Control 4.1.

NOTE: Original Pages 7-4 and 7-5 deleted/combined with this page and pages renumbered.

Revision 18 I

I 7-3

8.0 REFERENCES

a.

Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," U.S. Nuclear Regulatory Commission, Revision 1, October 1977.

b.

Deleted

c.

Deleted

d.

Deleted

e.

Deleted

f.

Yankee Atomic Electric Company Supplemental Information for the Purposes of Evaluation of 10CFR Part 50, Appendix I, Amendment 2, October 1976 (Transmitted by J. L. French - YAEC to USNRC in letters, dated June 2,1976; August 31, 1976; and October 8, 1976).

g.

Deleted

h.

Deleted

i.

Deleted

j.

Yankee Decommissioning Quality Assurance Program (YDQAP), Yankee Atomic Electric Company.

k.

Issuance of NPDES Permit No. MA0004367; Letter to J. A. kay from R. Janson, US EPA, dated July 29, 2003 I.

Deleted Revision 18 I

8-1

APPENDIX A DISPOSAL OF SEPTAGE Revision 7-Date:

MAY 21 199 A-I Approved By:

.i;@

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 MAY 1 7 1990 Docket -No.50-029 Mr. George Papanic, Jr.

Senior Project Engineer - Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398

Dear Mr. Papanic:

SUBJECT:

DISPOSAL OF SEPTAGE - YANKEE NUCLEAR POWER STATION By letter dated April 11, 1990, you requested NRC approval for a proposed disposal of sewage sludge containing very low concentrations of radionuclide according to 10 CFR 20.302.

We have completed our review of your request and our evaluation is enclosed.

We have found that your proposed transfer of the

.sludge by a contracted vendor to a public owned treatment works is acceptable.

Sincerely, Patrick Sears, Project Manager Project Directorate 1-3 Division of Reactor Projects - I/Il Office of Nuclear Reactor Regulation Enclosed:

As stated cc w/encl:

See next page Revision 7 -

Date: MAY 2 11990 A-2 Approved B y:

Mr. George Papanic, Jr.

Yankee Rowe.,

cc:

Dr. Andrew C. Kadak, President and Chief Operating Officer Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street Boston, Massachusetts 02110 Mr. T. K. Henderson Acting Plant Superintendent Yankee Atomic Electric Company Star Route Rowe, Massachusetts 01367 Resident Inspector Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission Post Office Box 28 Monroe Bridge, Massachusetts 01350 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 7th Floor Boston, Massachusetts 02111 Mr. George Sterzinger Commissioner Vermont Department of Public Service 120 State Street, 3rd Floor Montpelier, Vermont 05602 Ms. Jane M. Grant Senior Engineer -

PLEX Licensing Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 Revision 7 Date:

MAY 21 19A0 A-3 Approved By:

SAFETY ASSESSMENT BY TFE OFFICE OF NUCLEAR REACTOR REGULATION YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION DOCKET NO.50-029

1.0 INTRODUCTION

By letter of April 11, 1990, the Yankee Atomic Electric Company (Yankee) submitted, pursuent to 10 CFR 20.302(a), a method for the routine disposal of septic tank waste containing very low levels of licensed material.

Yankee proposed to periodically dispose of accumulated septic waste solids from the plant's sanitary system septic tank by transferring them to a public Sanitary Waste-Water Treatment Facility (SWTF) where they will be mixed with, processed with, and disposed as part of the sanitary waste generated from many sources.

Yankee proposed to make such disposals every one to two years over a period of 30 years.

In the submittal, the licensee addressed specific information requested in accordance with 10 CFR 20.302(a), provided-a detailed description of the licensed material, thoroughly analyzed and evaluated the information pertinent to the effects on the environment of the proposed disposal of the licensed material, and committed to follow specific procedures to minimize the risk of unexpected or hazardous exposures.

Revision 7 -

Date:

MAY21 1990 A-4 Approved By:

2.0 WASTE WATER-STREAM DESCRIPTION 2.1 Physical and.Chemical Properties The waste involved consists of residual septage (the accumulated settled and suspended solids and scum) produced by the sanitary sewerage collection and treatment system at the Yankee plant.

To safely dispose of the plant's sanitary waste stream, the Yankee plant supplements the onsite septic system supplemented with offsite treatment at a SWTF.

The onsite septic system consists of a 7,000-gallon buried septic tank and a subsurface soil-absorption leach field.

In the overall system design, the septic tank collects sludge and scum and partially separates liquids from the incoming sanitary waste.

The septage is retained in the septic tank, and the remaining conditioned waste-water liquid flows into the underground leaching field for treatment.

The leach field is the terminal point of the onsite portion of the plant sanitary waste treatment process.

In the offsite portion of this process, the septage is removed from the septic tank and transported to a SWTF.

Revision 7 -

Date:

MAY 21 1990 A-5 Approved By:

2.2 Radiological Properties The plant's sanitary system septic tank collects waste from the lavatories, showers, and janitorial facilities outside the Radiological Control Area (RCA).

No radioactivity is intentionally discharged to the septic system.

However, plant investigations into the source of low levels of licensed material found in septic tank waste have identified very small quantities of radioactive materials, which are below detection limits for radioactivity releases from the RCA.

It is suspected that these materials are carried out of the control area on individuals and spread to floor areas outside the RCA.

Floor wash water from these areas is poured through a filter bag to remove suspended solids and dirt before the water is released into a janitorial sink.

Although the wash water is returned to the RCA for disposal, if it is known to contain radioactivity, very small quantities can be released to, and accumulate in the sceptic tank.

The following values are estimates of the maximum total activity presently in the septic tank based on measurements of radionuclide concentrations in the liquid and solid phases:

Total Activity Nuclide (uCi)

Co-60 1.94 Mn-54 0.057 Cs-134 0.082 Cs-137 0.248 TOTAL 2.33 Revision 7

-Date:

-MAY.2 1 1990 A-6 Approved By:,4 7

ý

/

3.0 PROPOSED DISPOSAL METHOD Yankee proposes to periodically dispose of accumulated septage from its septic tank by contracting with a septic tank pumper that is approved by the Board of Health, Rowe, Massachusetts and transfer the septage to a Massachusetts SWTF for treatment.

This septic tank pumper will transfer the septage to an SWTF, where it is mixed and diluted with other raw sewage and introduced either into an anaerobic digester or an aeration pond for biological treatment.

The resulting processed sludge from the SWTF is then mixed with sand and disposed of in a sanitary landfill, where it will be covered by clean soil daily.

An alternate disposal means could result in the processed sludge being spread as a fertilizer, though generally for vegetation, such as sod, which is not consumed by humans.

None of the region's SWTFs that receive sewage from local septic tank pumpers incinerate their sludge as a means of treatment.

This method of pumping the tank and transferring the septage to an SWTF is the same method normally applied to septic tank systems,.regardless of the presence of licensed material.

3.1 Septic Tank Waste-Procedural Requirements and-Limits The licensee will perform a gamma isotopic analysis on a representative sample of waste from the septic tank no more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months before a contracted septic tank pumper begins to pump the waste from the tank to transfer to-a SWTF.

The licensee will collect at least two septage samples from the plant's septic tank by taking a volumetric column sample that will allow the licensee to Revision 7 - Date: _MAY 21 1991 A-7 Approved By:

determine the ratio of the solid content to the total content of the tank.

By determining the weight of the percentage of solid content of the collected sample and applying this value to the gamma isotopic analysis, the licensee will be able to estimate the total radioactivity of the contents of the tank.

To document the estimation of radiological effect of septage disposal, the licensee will perform these gamma isotopic analyses of the representative samples at the Technical Specification Environmental Lower Limit of Detection (LLD) requirements for liquids, as required in Technical Specification Table 4.12-1, "Detection Capabilities for Environmental Sample Analysis,"

The radionuclide concentrations and total radioactivity identified in the septage will be compared to the concentration and total curie limits established herein before disposal.

The following limits apply to these analyses:

1.

The concentration of radionuclides detected in the volume of septage to be pumped to a disposal truck shall be limited to a combined sum of fractional Maximum Permissible Concentrations in Water (MPC)

(as listed in 10CFR Part 20, Appendix B, Table II, Column 2),

summed over all nuclides present, of less than or equal to 1.0.

2.-

The total gamma activity that can be released during septage transfer to any SWTF or combination of such facilities in one year (12 consecutive months) is limited to not more than 20 microcuries (equivalent to a maximum whole-body dose of 1 mrem to any individual in the public).

Revision 7 - Date:

MAY 21 7ggm A-8 Approved By:

3.2 Administrative Procedures The licensee will maintain complete records of each disposal.

In addition to copies of invoices with approved septic tank pumpers, these records will include the concentration of radionuclides in the septage, the total volume of septic waste disposed, the total activity in each batch, and the total accumulated activity of the septage pumped in any 12 consecutive months.

For periods in which disposal of septage occurs under this application, the licensee shall report, to the Nuclear Regulatory Commission (NRC) in the plant's Semiannual Effluent Release Report, the volume, liquid, and solid mass fractions, radionuclide concentrations in the liquid and solid fractions, and the total activity disposed.

4.0 EVALUATION OF ENVIRONMENTAL IMPACT The proposed method for disposal of septage is the same as currently used by all facilities designed with septic tanks for the collection of septic waste.

No new structures or facilities need be built or modified, nor any existing land uses changed.

Septage from Yankee will be transported to an existing SWTF, where it will make up a small fraction of the total volume of sanitary waste treated each year.

The normal method of septage handling and treatment Revision 7 -

Date:

2190A-9 Approved By:

would involve dilution of Yankee's septage with other waste-water at a public SWTF.

The processed sludge from the SWTF is usually buried in a sanitary landfill, thus limiting the potential exposure pathways to man.

Otherwise, the sludge is widely dispersed in fertilizer, thereby preventing any buildup of activity from successive annual pumpouts from the plant's septic tank.

This method of disposal will not affect topography, geology, meteorology, hydrology, or nearby facilities.

5.0 RADIOLOGICAL IMPACTS The licensee has evaluated the following potential exposure pathways to members of the general public:

(1) inhalation of resuspended radionuclides, (2) ingestion of food grown on the disposal site, (-3) external exposure to a truck driver or SWTF worker, and (4) external exposure caused by long-term buildup and external exposure from standing on the ground above the disposal site.

The staff has reviewed the licensee's calculational methods "and assumptions, and finds that they are consistent with regulatory Guide 1.109.1 The staff finds the assessment methodology acceptable.

Revision 7 -Date:

MAY 2 1 1990 A-10 Approved B~y:

1Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance Vith 10 CFR Part 50, Appendi 1'," Revisicn 1, October 1977.

Doses calculated in this manner by the licensee for the maximum exposed member of the public were as follows (based on a total activity awaiting disposal of 2.3 pCi, more than 80% of which is Co-60):

Maximally Exposed Individual/Whole Body (Child)

Pathway (mrem/year)

Ground Irradiation 0.099 Inhalation 0.0001 Stored Vegetables 0.0214 Leafy Vegetables 0.O01-Milk Ingestion*

(0.0036)

TOTAL 0.12 The licensee then performed a similar calculation using a concervative upper bound activity of 20 pCi to be discharged in any one year.

Based on this upper bound analysis, the dose to the maximally exposed individual member of the general public was estimated to be 1.1 mrem/year, as shown in the following table:

Revision 7 - Date:

AY 21. 1990 A-11 Approved By:

W Maximally Exposed Individual/Whole Body (mrem/year)

Pathway Ground Irradiation inhalation Stored Vegetables Leafy Vegetables TOTAL 0.980 0.0004 0.13 0.007 1.1 Based on this same total was estimated to be 0.01 of 10 CFR 50, Appendix I fuel cycle activities as activity, the dose to truck drivers and SWTF workers mrem/yr.

These doses are within the design objectives and well within the environmental standards for uranium stated in 40 CFR 190.10(a) and are therefore acceptable.

6.0.

SUMMARY

AND CONCLUSIONS The disposal of septage by transferring it to a public SWTF is in accordance with standard practices for treatment of the type of waste material generated by a septic tank and leach field sanitary waste system.

Periodic pumping of the septic tank is necessary for the maintenance and continued operation of Yankee's sanitary waste system.

Yankee requested approval for disposal of septic waste from the Yankee sanitary system to prevent failure of the sanitary system to adequately handle plant domestic waste.

Revision 7 -

Date:

MAY 21 1990 A-12 Approved An alternate means of disposal would involve the treatment of the septage as radwaste.

Such a disposal would require that the licensee stabilize, solidify, and dispose of the matirial at a licensed burial ground, requiring excessive cost and valuable disposal ground.

The results of the radiological analysis indicate that the public health effects of the biological activity and pathogenic constituents of such sanitary waste far outweigh the concerns related to any radioactivity that is present.

By setting release limits that restrict the exposure for an individual to a maximum value of 1 millirem per year, Yankee ensured that radiological risks from the proposed disposal method are insignificant.

The proposed release limits represent a small fraction of NRC limits permitted for disposal of similar waste by licensed facilities who have their sanitary systems connected directly to a public sanitary sewerage system.

These proposed limits are also well within the plant's allowable release limits for the discharge of normal liquid waste to the environment.

Any resulting dose to any individual in the public is less than exposures caused by natural background radiation.

Based on our review of the proposed disposal of septage, the staff makes the following conclusions:

(1) the radionuclide concentrations in-disposed septage will be a small percentage of permissible standards set forth in 10 CFR Part 20; (2) the radiation risk to workers involved in the disposal would be small compared to the routine occupational exposures at the Yankee Nuclear Power Station; (3) because the proposed action involves such very low levels of radioactivity, it will require no change in the decommissioning aspects of the Revision 7 - Date:

MAY 2 1 190 A-13 Approved By: _

facility and will require only insignificant changes in the handling or transport of radioactive material (septage);

and (4) the licensee's procedures with commitments as documented in the submittal are acceptable, provided that the submittal is permanently incorporated into the licensee's Offsite Dose Calculation Manual (ODCM) as an Appendix, and future modifications will be reported to NRC in accordance with licensee commitments regarding ODCM changes.

Contributors:

J. Minns P. Sears Revision 7 - Date:

MAY 21 1990 A-14 Approved By:

YANKEE A TOMIC ELECTRIC COMPANY Telephone (508) 779-6711 TWX 710-380-7619 CAMKIEE 580 Main Street, Bolton, Massachusetts 01740-1398 April 11, 1990 BYR #90-42 United States Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D.C.

20555

References:

(A) License No.

DPR-3 (Docker No. 50-29)

Subject:

10 CFR 20.302 Application

Dear Sirs:

Pursuant to 10 CFR 20.302, Yankee has prepared the attached application for the routine disposal of septage from Yankee Nuclear Power Station.

This application utilizes guidance contained in NRC regulation 10 CFR 20.303 for the disposal of licensed material into a sanitary sewerage system.

We trust that you will find this submittal satisfactory, however, if you have any questions please contact us.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY George PapaniLc,Jr Senior Project Engineer Licensing Enclosure GP/emd Revision 7 Date MY' 21 1ý0L Approved By:

A-15

YANKEE NUCLEAR POWER STATION APPLICATION FOR APPROVAL TO ROUTINELY DISPOSE OF SEPTAGE UNDER 10CFR20.302 MAY 2 1 1990 A

v Revision 7

-Date:

A-16 Approved By:

TABLE OF CONTENTS TABLE OF CONTENTS....................

LIST OF TABLES......................................................

11 LIST OF FIGURES....................................................

iv

1.0 INTRODUCTION

1 2.0 WASTE STREAM DESCRIPTION..........................................

2 2.1 Physical/Chemical Properties.................................

2 2.2 Radiological Properties.....................................

3 3.0 PROPOSED DISPOSAL METHOD...........................................

5 3.1 Septic Tank Waste Procedural Requirements and Limits........

6 3.2 Administrative Procedures...................................

7 4.0 EVALUATION OF ENVIRONMENTAL IMPACT................................

8 5.0 EVALUATION OF RADIOLOGICAL IMPACT.................................

9 5.1 Septic Tank Sample Analysis Data............................

9 5.2 PathwayExposure Scenarios...................................

10 5.3 Dose Assessments.............................................

11 5.3.1 External Exposure to a Truck Driver/SWTF Worker.....

11 5.3.2 External Exposure Due to Long-Term Buildup..........

12 5.3.3 Garden Pathway Scenario.............................

14 5.3.4 Incineration Pathway Scenario.......................

20 5.4 Maximum Releasable Activity................................... 21 6.0

SUMMARY

AND CONCLUSIONS...........................................

23

7.0 REFERENCES

24 MM 2 1900 4/--

Revision 7 -

Date:

A-17 Approved By:

z7

LIST OF TABLES Number I.

2 3

4 Title Landspreading Ingestion Pathways (Adult)

Landspreading Ingestion Pathways (Teen)

Landspreading Ingestion Pathways (Child)

Landspreading Ingestion Pathways (Infant) 25 26 27 28 RMAY 2 1 1991-Revision 7 -

Date:

A-18 Approved By:

LIST OF FIGURES Luge I1 Yankee Plant Sanitary Waste.Disposal Process 29

-iv-Revision 7 -

Date:

MAY 2, 12, A-19 Approved By:

YANKEE NUCLEAR POWER STATION Application for Approval to Routinely Dispose of Septage Under IOCFR20.302

1.0 INTRODUCTION

Yankee Atomic Electric Company (YANKEE) requests approval, pursuant to 10CFR20.302(a),

of a method proposed herein for the routine disposal (typically, once every one to. two years) of septic tank waste containing very low levels of licensed material over an extended period of time of 30 years.

Yankee proposes to periodically dispose of accumulated septic waste solids from the plant's sanitary system septic tank by transferring it to a public Sanitary Waste-Water Treatment Facility (SWTF) where it will be mixed with, processed, and disposed of, as part of sanitary waste generated from many sources.

This is analogous to other Nuclear Regulatory Commission (NRC) licensed facilities who have their sanitary waste systems connected directly to a municipal sewer line.

Part 20.303 of Title 10 to the Code of Federal Regulations already permits these NRC licensees to discharge licensed material into a sanitary sewerage system.

Routine maintenance of Yankee's septic system is necessary to ensure proper operation of the system.

Periodic pumping of the septic tank to remove accumulated solids is necessary to prevent the carryover of solids,into the subsurface leach field which would inhibit the soil absorption capabilities of the field.

This application addresses specific information requested in 10CFR20.302(a),

and demonstrates that the periodic disposal of septage from Yankee's Sanitary Waste System over an extended periods of time (30 years),

under both normal and unexpected conditions, will not result in significant impacts either to the environment or to individuals in the general public. Revision 7-Date:

MAY 21 A-20 Approved By:

2.0 WASTE WATER STREAM DESCRIPTION 2.1 Physical/Chemical Properties The waste involved in this application consists of residual septage (accumulated settled and suspended solids, and scum) associated with the sanitary sewerage collection and treatment system at the Yankee plant.

The Yankee plant utilizes an on-site septic system supplemented with off-site treatment at a SWTF for the safe disposal of the plant's sanitary waste stream.

Figure 1 is a schematic of the overall sanitary waste disposal process.

The on-site septic system consists of a 7,000 gallon buried septic tank and a subsurface soil.absorption leach field.

Sanitary sewage from the plant flows (estimated 2,600 gallons/day) into the septic tank.

The septic tank function in the overall system design is for the collection of sludge and scum and partial separation of liquids from the incoming sanitary waste. Some of the solid particles settle to the bottom and form a layer of sludge, where greases and oils float to the surface creating a scum layer.

(

The septage is retained in the septic tank and the remaining conditioned waste-water liquid is permitted to flow into the underground leaching field for treatment.

The leach field is the terminal point of the on-site portion of the plant sanitary waste treatment process.

Some of the septage stored in the septic tank is reduced to liquid by bacterial action in the septic tank, but the rest of the septage remains essentially untreated.

This material must be pumped out at regular intervals to prevent it from overflowing the tank and entering the leaching field (References 1, 2, 3, 4, 5,

6, 7,

8, 9, and 10) where it will clog the soil and eventually lead to septic system failure.

In general, septage pumped from septic tanks is discharged to a SWTF for treatment as part of the overall system design (Reference 10).

The septage is then co-treated with other sanitary wastes at the SWTF.

The septage pumped periodically from the Yankee plant has, in the past, been treated and disposed of in this fashion when no licensed material was determined to be present. -

Revision 7 - Date:

MAY 21 1990 A-21 Approved By:

Z.

The removal of the septage from the septic tank and subsequent transportation to a SWTF constitutes the off-site portion of the Yankee plant overall sanitary waste disposal process.

2.2 Radiological Properties The plant's sanitary system septic tank collects waste from the lavatories, showers, and janitorial facilities outside the Radiological Control Area (RCA).

No radioactivity is intentionally discharged to the septic system.

However, plant investigations into the source of low levels of licensed material found in septic tank waste have identified that very small quantities of radioactive materials, which are below detection limits for radioactivity releases from the RCA, appear to be carried out of the control area on individuals and accumulate in the septic tank.

The suspected primary source of the radioactivity (i.e., floor wash water) is now either poured through a filter bag to remove suspended solids and dirt before the water is released into a janitorial sink, or the wash water is returned to the RCA for disposal.

An isotopic analysis, at environmental detection limits, of two composite volumetric sample columns of septage taken from the plant's septic tank identified the following plant-related radionuclides:

Activity Concentration Nuclide (pCi/kg wet +/- I sigma)

West Manhole East Manhole Sample Location Sample Location Co-60 92.4 j 3.9 13.2 + 2.2 Cs-134 5.9 + 1.3 Cs-137 9.2 + 1.5 3.2 + 1.0 After the initial analysis of the composite samples noted above, the samples were subsequently centrifuged into liquid and wet solids portions and reanalyzed.

There was no activation or fission products identified in any of the liquid fraction samples indicating that the detected activity was in a form that had been carried out of solution with the solid fraction of the samples. Revision 7 -Date:

MPY2 90A-22 Approved By:

7.Z#

Analysis of the resulting solid fraction of the septage indicated the following radionuclide concentrations:

Activity Concentration Nuclide (pCi/k& wet +

I-1 sigma)

West Manhole East Manhole Sample Location Sample Location Co-60 1,588 + 42 528 +/- 26 Mn-54 47 + 13 Cs-134 67 + 11

.Cs-137 203 +/-17 100 13 The original septic tank samples were volumetric samples representative of the distribution of solids and liquid from bottom to top of the tank.

The ratio of the weight of the solid fraction sample to the weight of the solid fraction plus liquid fraction sample allows a determination of the percentage of total solids content of the septic tank.

For the waste sample from the west manhole, the solid fraction of the composite sample was found to be 0.024, or 2.4 wt. %.

For the east manhole, the solid fraction of the total sample was 0.046, or 4.6 wt.%.

The principle radionuclide is Cobalt-60, which accounts for approximately 82% of all plant-related activity detected in the septage.

The total radioactivity content of the septic tank can be estimated by calculating the mass of solids present in the 7,000 gallon tank by taking the higher (conservative) solids fraction determined from the sample data.

This is multiplied by the mass of septage in the tank and by the highest activity concentration determined in the solids.

As a result, the estimated maximum total activity is:

Total Activity Nuclide (pCi)

Co-60 1.94 Mn-54 0.057 Cs-134 0.082 Cs-137 0.248 TOTAL 2.33 Revision 7 -Date:

__________A-23 Approved By: -' t I

3.0 PROPOSED DISPOSAL METHOD Upon approval from the U.S. Nuclear Regulatory Commission (NRC),

Yankee proposes to periodically dispose of accumulated septage from its septic tank by contracting with a town-approved (Board of Health, Rowe, Massachusetts) septic tank pumper for the removal and transfer by truck of the septage to a Massachusetts SWTF for treatment.

At the SWTF, the septage would typically be mixed and diluted with other raw sewage and introduced either into an anaerobic digester or aeration pond for biological treatment.

The resulting processed sludge from the SWTF is typically then mixed with sand in a ratio of 50/50 and disposed of in a sanitary landfill, where it would be covered by clean soil daily.

An alternate disposal means could potentially result in the processed sludge being landspread as a fertilizer, though generally for nonhuman-consumed vegetation, such as sod.

None of the regions SWTFs which would be used by local septic tank pumpers were identified as incinerating their sludge as a means of treatment.

This method of tank pumping and transfer to an SWTF is identical to that normally applied to septic tank systems, irrespective of the presence of licensed material.

Once the septage is pumped into the contract vendor's transporting vehicle, the situation is analogous to the handling of licensed material under 10CFR20.303.

Part 20.303 of Title 10 to the Code of Federal Regulations already permits these NRC licensees to discharge licensed material into a sanitary sewerage system if certain conditions are met.

Due to the remoteness of the Yankee plant's location, it is impractical to directly connect sewer lines to a facility to handle sanitary waste.

In this case, a tank truck acts as a sewer line in transferring septage to a SWTF.

The quantity and form (soluble or dispersable) of any licensed material contained in our septage is not affected by the means employed to transfer it to a SWTF for processing.

'Therefore, it would be the same whether the plant was directly connected to a municipal sewerage system or trucked its septage on a periodic basis to a SWTF. R i 27 19D Revision 7 -

Date:

________A-24 Approved By:

3.1 Septic Tank Waste Procedural Requirements and Limits Gamma isotopic analysis of septic tank waste shall be made prior to transfer of the waste by a contracted septic tank pumper to a SWTF by obtaining a representative sample from the tank no more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to initiating-pump-out.

At least two septage samples shall be collected from the plant's septic tank by taking a volumetric column sample which will allow for analysis of the solid's content of the tank.

The weight percent of solid content of the collected sample will be determined and applied to the gamma isotopic analysis in order to estimate the total radioactivity content of the tank.

These gamma isotopic analyses of the representative samples will be performed at the Technical Specification Environmental Lower Limit of Detection (LLD) requirements for liquids (see Technical Specification Table 4.12-i, "Detection Capabilities for Environmental Sample Analysis") in order to document the estimation of radiological impact from septage disposal.

The radionuclide concentrations and total radioactivity identified in the septage will be compared to the concentration and total curie limits established herein prior to disposal.

The limits to be applied are as follows:

1. The concentration of radionuclides detected in the volume of septage to be pumped to a disposal truck shall be limited to a combined Maximum Permissible Concentration of Water (MPC)

(as listed in 10CFR, Part 20, Appendix B, Table II, Column 2) ratio of less than or equal to 1.0.

2.

The total gamma activity which can be released via septage transfer to any SWTF in one year (12 consecutive months) is limited to not more than 20 microcuries (equivalent to a maximum whole body dose of 1 mrem to any individual in the public). Revision 7 - Date:

MAY 21 IS93 A-25 Approved By:

If the total activity limit is met, compliance with the self-imposed dose criteria will have been demonstrated since the radiological impact (Section 5) is based on evaluating the exposure to a maximally exposed hypothetical individual such that his annual whole body dose would be limited to approximately I mrem.

Both the concentration and total activity limits represent a small fraction of the allowable limits permitted under IOCFR20.303 to other NRC licensees who have their sanitary waste systems directly connected to a public sewerage system.

If not for the biological nature of sanitary waste, the above release limits would also allow for the direct discharge of the waste under the plant's existing Technical Specification requirements for release of liquids to the environment.

3.2 Administrative Procedures Complete records of each disposal will be maintained.

In addition to copies of invoices with approved septic tank pumpers, these records will include the concentration of radionuclides in the septage, the total volume of septic waste disposed, the total activity in each batch, as well as total accumulated activity pumped in any 12 consecutive months.

For periods in which disposal of septage occurs under this application, the volume, total activity, and relative nuclide distribution, shall be reported to the NRC in the plant's Semiannual Effluent Release Report. Revision 7 -

Date:

MAY 21 1990 A-26 Approved By:

4.0 EVALUATION OF ENVIRONMENTAL IMPACT The proposed method for disposal of septage is the same as currently used by all facilities designed with septic tanks for the collection of septic waste.

No new structures or facilities need be built or modified, nor any existing land uses changed.

Septage from Yankee will be trucked to an existing SWTF, where it will make up a small fraction of the total volume of sanitary waste treated each year.

As a result, there will be no impact on topography, geology, meteorology, hydrology, or nearby facilities by the proposed method of disposal.

R-8-2 MAY 2 1 1990 f

f Revision 7 -

Date:

________A-27 Approved By:

5.0 EVALUATION OF RADIOLOGICAL IMPACT Radiological evaluations have been performed for the purpose of bounding the dose impact associated with the disposal of septage.

The normal method of septage handling and treatment would provide for dilution of Yankee's septage with other waste-water at a public SWTF.

The processed sludge would typically be buried in a sanitary landfill, thus limiting the potential exposure pathways to man, or widely dispersed if used as a fertilizer, thereby preventing any build-up of activity from successive annual pumpouts from the plant's septic tank.

The dose assessments, however, did consider the maximum potential impact of long-term buildup of activity resulting from 30 years of placing septage waste in the same SWTF, with all the processed sludge assumed to be buried in one landfill disposal cell.

5.1 Septic Tank Sample Analysis Data The analysis of the septic tank's measured radioactivity, and its distribution between liquid and solid fractions, provides the bases upon which a dose assessment of disposal of septage can be made.

The composition. of the septic tank waste determined from the sample analysis is:

Composite Sample East End Manhole Location 3.502 kg 0.087 kg Wt. Liquid Wt. Solid Composite Sample West End Manhole Location 3.460 kg 0.167 kg Solid fraction of the composite sample as collected is equal to:

Solid fraction = Wt. solid/(Wt. solid + Wt. liquid)

The solid fraction for the East End sample was 0.0242, and 0.0460 for the West End.

The activity in the solid fraction was basically found to contain all the detected radioactivity as noted below:

East End Solids Sample (DCi/ka) Wet West End Solids Sample (oCi/ke) Wet Mn-54 Cs-134 Cs-137 Co-60 100 528 47 67 203 1,588 Revision 7 - Date:

A-28 Approved By:

With the septic tank volume taken as approximately 7,000 gallons (26,500 liters),

and assuming the maximum solid fraction (0.046) and maximum radionuclide concentration applies to the total tank's content, the total maximum radioactivity content is estimated to be:

Isotope Half-Life Oe (Ci)

Mn-54 312.2 day 5.73 E-08 Co-60 5.272 yr 1.94 E-06 Cs-134 2.065 yr 8.17 E-08 Cs-137 30.17 yr 2.48 E-07 5.2 Pathway Exposure Scenarios Radiological evaluations were performed for both the expected activities associated with handling, processing, and disposal of septage waste at a SWTF, and a hypothetical event causing undiluted septage release.

The bounding case was determined to be associated with a hypothetical event which lead to the spreading of undiluted septage from Yankee's septic tank directly on a garden area where food crops are grown.

The contracts with town approved septic tank pumpers will direct that septage be disposed of at a SWTF in Massachusetts.

It is not expected that any disposal will occur other than at an SWTF.

It is, therefore, not considered credible that successive bounding case activities could occur which lead to a long-term buildup of activity on a single minimum size garden plot.

In addition, since incineration of septic waste is not a current practice in the local area, the potential exposures associated with incineration are not of current concern.

However, the establishment of a conservative-total whole body dose criteria for release of sanitary waste, via the above-noted garden scenario, assures that the potential resulting whole body dose due to incineration would not be expected to result in significant doses to any individual.

This assessment is further detailed in Section 5.3.4.

The contributing pathways of exposure for the normal SWTF disposal process include:

I.

External exposure to a truck driver.

2.

External exposure to a SWTF worker. Revision 7-Date;'

j]9 A-29 Approved By:

/*

3.

External exposure to an individual standing on the SWTF landfill after 30 years of buildup and decay.

The following garden exposure pathways were addressed for the maximally exposed hypothetical individual:

1. Standing on the ground plane.

2.

Inhalation of resuspended material.

3.

Ingestion of leafy vegetables.

4.

Ingestion of stored vegetables.

5.

Ingestion of milk.

6.

Liquid pathways.

It should be noted that the milk pathway is mutually exclusive to the other food production pathways since it would be impossible to support the grass-cow-milk-man exposure chain if the limited land area is utilized for the growing of food crops for direct human consumption.

The two sets of ingestion pathways have been calculated so that the potential maximum impact can be assessed.

Similarly, radionuclide movement into the ground water pathway would tend to reduce the impact of surface-related exposure paths and is, therefore, considered independently.

5.3 Dose Assessments 5.3.1 External Exposure to a Truck Driver/SWTF Worker The external dose rate from a 3,500-gallon tank truck filled with septage containing the total measured activity in the septic tank (2.33 pCi) was calculated for the purpose of estimating exposures associated with shipping the waste to a SWTF.

A three-dimensional point-kernel shielding code for the determination of direct radiation from gamma radiation emanating from a self-attenuating cylindrical source (DIDOS-IV, Reference 14) was utilized to calculate the external dose rate from the tank truck.

The truck was modeled as a cylindrical radiation source with a radius equal to 1.22 meters and a length of 2.84 meters.

A dose rate of 1.2E-04 mrem per hour for a point one meter from the end of the cylinder along the axis was calculated.

No credit for shielding provided by the tank truck or cab was assumed.

The dose to a Revision 7 - Date:

MAY 2 1990 A-30 Approved By:

truck driver making a 100-mile trip to a treatment facility at an average of 20 miles per hour plus a three-hour waiting period at the SWTF, is estimated to be 9.5E-04 mrem.

It is concluded, based on the total activity limits proposed, that this pathway will not lead to significant exposure of any individual.

It is also concluded that due to the sanitary properties of septage handling, a SWTF employee's direct exposure time is kept to a minimum.

Using the dose rate estimated for thLe truck driver above, and conservatively assuming that it requires an employee at the SWTF a full eight-hour day to process each truckload of waste, and not taking any credit for dilution or increased distance from the waste, a waste processing facility employee's dose is also estimated to be 9.5E-04 mrem.

If the maximum activity content proposed to be disposed of each year were assumed as the source term (20 pCi), the dose to the truck driver/SWTF worker is estimated to be less than 1.OE-02 mrem using the same assumptions as noted above.

5.3.2 External Exposure Due to Long-Term Buildup In order to assess the potential impact from the postulated buildup of activity resulting from 30 years of septage'disposed at the maximum annual allowed activity content, it was conservatively assumed that the entire quantity of accumulated activity at the end of 30 years was buried in a common landfill disposal cell which was then available to the general public for uncontrolled access (8,760 hours0.0088 days 0.211 hours 0.00126 weeks 2.8918e-4 months per year).

For regional SWTFs, waste sludge is typically mixed with sand and placed in landfill disposal cells on a daily basis and covered by a layer of at least six inches of composited material before the end of each working day, as required by Massachusetts Department of Environmental Protection regulations (Reference 16).

The landfill disposal cells range in size from about one acre up to about five acres.

After a cell is full, a final layer of compacted material is required to be placed over the entire surface of the cell to a minimum depth of two feet (Reference 16). Revision 7 - Date:

'AY 21 1990 A-31 Approved By:

2*

Analytically, if Qo is the amount of radioactivity per tank full of septage for a give nuclide, then the total accumulated radioactivity Qe(max) disposed of after 30 pumpouts is given by:

Qe(max)

= Qo (1 + E + E2

+ E3 + E4 +....

+ E2 9 )

= Q0 (1 -

E2 9 )/(l E)

(A) where:

E

= exp(-XA~t)

X

= is the decay constant for the selected nuclide (1/year),

and At = time interval between applications, assumed to be 1 year.

If the maximum total activity of 20 microcuries (with the same relative distribution as determined in the current septic tank analysis) were assumed to be released each year, then the accumulated activity at the end of 30 years is found in the following table:

Co-60 Mn-54 Co-134 Co-137 Half Life 5.27 y 312. d 2.07 y 30.2 y (1/year) 0.1315 0.8109 0.3357 0.023 Q0 (uCi/batch) 16.65 0.49 0.70 20 Qe (max) uCi 132.14 0.88 2.45 46.04 182 Total If the 20 microcuries per year limit is assumed to be all Co-60, then the resulting accumulated total after 30 years would be 159 microcuries, and result in a higher calculated dose than that from the above mix.

Assuming a minimum landfill disposal cell to be one acre in-area, and that the 30-year accumulated activity (159 uCi; Co-60) was disposed of in one year along with SWTF sludge that formed a minimum one foot layer which was placed immediately below the two-foot disposal cap of the cell, the resulting Revision 7 - Date:

MAY 2 1 1930.

A-32 Approved By:

dose rate one meter above the ground surface was calculated to be 6.4E-07 mrem/hour.

If it is also assumed that an individual remained on the landfill for a full year (8,760 hours0.0088 days 0.211 hours 0.00126 weeks 2.8918e-4 months ) without taking any credit for shielding by a residential structure, the total whole body dose would be 5.6E-03 mrem, or about 56% of the truck driver's/SWTF workers calculated exposure.

Since the landfill cap (2' minimum) effectively isolates the vegetation zone of the top 15 cm plow layer, no garden pathways of exposure are included.

However, it is noted that the'30-year accumulated activity concentration spread over a one acre landfill disposal cell would result in an area density of only 3.7E-03 microcuries per square foot.

This is approximately a factor of 11 below the surface area density of the garden pathway scenario in Section 5.3.3 for the bounding case of placing 20 microcuries directly on a 500 ft 2 garden.

Therefore, even if it is postulated that an individual were to dig a cellar hole for a new home on the landfill site after closure, the resulting dose impact would still be bounded by the garden scenario as described below.

It is, therefore, concluded that for normal handling, processing, and disposal of septage at a SWTF, the maximum annual dose is received by the truck driver or SWTF worker handling the annual batches of septage pumped for disposal, and not the result of accumulated activity buildup over extended time periods.

5.3.3 Garden Pathway Scenario The radiological impact associated with an event which place undiluted septage directly on a garden was carried out using the dose assessment models in Regulatory Guide 1.109 (Reference 13), and in a manner consistent with the methodology employed by the plant's ODCM.

Special consideration was given to the following:

1. The computation of an effective self-shielding factor to account for the effect provided by the soil after the waste is plowed or mixed in the top 15 cm surface layer. Revision 7 -

Date:

A-33 Approved By:

2.

The definition of an annual activity release rate, which following a year's time of continuous release, would yield the ground deposition expected to prevail after a tank pump-out and spreading on the 500 ft 2 garden.

3.

The definition of an effective atmospheric dispersion factor to represent the resuspended radioactivity.

4.

The proper representation of partial occupancy factors and usage data.

Landspreading. Resuspension. and Occupancy Factors If it is assumed that the garden plot is limited to a surface area of 500 ft 2, then the land deposited radioactive material Se (Ci/m 2 ) following landspreading will be equal to:

Se = Qe (Ci)I(500 ft 2

0.0929 m2/ft 2 )

(B)

The denominator of this equation is equivalent to the (D/Q) deposition factor normally employed in the airborne impact assessment of deposited radionuclides; that is:

(D/Q)

= 1/(500 ft 2

0.0929 m2/ft 2 )

=-2.15E-02 (m-2 )

(C)

Following the application of undiluted septage on the garden, some of the radioactivity may become airborne as a result of resuspension effects.

The model used to estimate the radionuclide concentration in air above the disposal plot was taken from WASH-1400, Appendix VI.

According to that model, the relationship between the airborne concentration Ae (Ci/m3 ) and the surface deposition is:

Ae= Se (Ci/m 2 ) x K (1/m)

(D) 7 MRY 21. 1990-A o

y Revision 7 -Date:

A-34 Approved By:

where:

K is the resuspension factor and is taken to be equal to 1.OE-06 (1/m)

(Reference 11) which is believed conservative due to the limited surface area involved and the irrigation provided to a garden which minimizes airborne dust.

The 500 ft 2 garden size was selected based on the minimum surface area necessary to include a garden as part of the land-use census as required by Yankee's Technical Specification 3/4.12.2.

This is the minimum area which could be expected to produce sufficient food to support the uptake assumption on food consumption noted below.

In addition, by limiting the garden surface area to 500 ft 2 (a circle with a 3.85 m radius) the concentration of radioactivity in the garden is maximized since the concentration for any given surface area is physically limited by the total activity available in the septage.

For direct radiation estimates from standing on the ground plane, a commonly used assumption of an infinite plane source (which can be approximated by a circle with a radius of 15 m) would in fact undercalculate the surface dose rate from that of a 500 ft 2 garden by a factor of about 8 due to the dispersal of the fixed quantity of activity available to be spread.

For use with the garden pathways of exposure, it is assumed that the septage is mixed in the top cultivated 15 cm of soil with no additional clean soil cover placed over it.

As for the occupancy factors for direct exposure to the ground deposition and for immersion in the resuspended radioactivity, 360 hours0.00417 days 0.1 hours 5.952381e-4 weeks 1.3698e-4 months was used for the radiological impact analysis.

The 360-hour interval is believed to be a reasonably conservative time frame a gardener would spend each year on a plot of land or garden during the growing season in the northeast (average two hours a day for six months).

Garden pathway data and usage factors as applicable to the area in the vjcinity of the plant are shown below.

These are the same factors as used in the plant's ODCM assessment of the off-site radiological impacts due to routine releases from the plant, with the following exceptions: Revision 7 - Date:

A-35 Approved By:

1. The soil exposure time was changed from 15 years to I year to account for the discrete application of septage on a garden plot.

2.

The fraction of stored vegetables grown in the garden was.

conservatively increased from 0.76 to 1.0.

3.

The crop exposure time was changed from 2,160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months to 0 hours0 days 0 hours 0 weeks 0 months to reflect the condition that no radioactive material would be dispersed directly on crops for human or animal consumption, the deposition on crops of resuspended radioactivity being insignificantly small; that is, crop contamination is only through root uptake.

USAGE FACTORS Individual Adult Teen Child Infant Vegetables (kg/vr) 520 630 520 Leafy Veg.

(kg/yr) 64 42 26 Milk Inhalation*

(liters/yr)

-(mLyr) 310 400 330 330 329 329 152 58

Inhalation rates have been modified to reflect an annual occupancy factor of 360 hours0.00417 days 0.1 hours 5.952381e-4 weeks 1.3698e-4 months .

VEGETABLE PATHWAY Stored Vegetables Leafy Vegetables Agricultural productivity *kg/m 2 )

Soil surface density (kg/m )

Transport time to user (hours),

Soil exposure time (hours)

Crop exposure time to plume (hours)

Holdup after harvest (hours)

Fraction of stored vegetables grown in garden Fraction of leafy vegetables grown in garden 2.0 240.0 0.0 8,766.0

.0 1,440.0 2.0 240.0 0.0 8,766.0

.0 24.0 1.0 1.0 Revision 7 - Date:

MAY 21 1990 A-36 Approved By:

COW-MILK PATHWAY Pasture Feed Stored Feed Agricultural productivity *kg/m 2 )

.7

.2.0 Soil surface density (kg/m )

240.0 240.0 Transport time to user (hours) 48.0 48.0 Soil exposure time (hours) 8,766.0 8,766.0 Crop exposure time to plume (hours)

.0

.0 Holdup after harvest (hours)

.0 2,160.0 Animals daily feed (kg/day) 50.0 50.0 Fraction of.year on pasture

.5 Fraction pasture when on pasture 1.0 As noted above, liquid exposure pathways are considered independent from those associated with garden exposures.

Since the laboratory analysis data of septic tank waste shows that all the activity is associated with the suspended or settled solids fraction, and not dissolved in the liquid portion, transport of activity through groundwater would not be expected to lead to drinking water supplies being impacted by septage placed on farm lands.

It is, therefore, not anticipated that the groundwater pathway could result in doses comparable to the direct surface exposure pathways.

As confirmation of this, however, a methodology for groundwater analysis, as developed by Kennedy, et al. (1990)

(Reference 12), was used as a check.

This model assumes that the radionuclides on the ground are leached into the water table with a leach rate based on continuously saturated soil.

Once into the water table, the radionuclides are immediately available for consumption.

The volume of water used for dilution is limited to the quantity used by one person in one year (91,250 liters).

No credit is taken by soil retardation of thenuclides, either during the leaching process or during groundwater movement.

Consumption of water is assumed to be 2 liters/day.

The resulting dose factors, by radionuclide, are listed in Table 3.4 of Reference 12.

Of the radionuclides detected in the septage, Co-60 is the dominant nuclide, and has the highest dose factors.

The total effective dose equivalent from drinking water is 4.4E-6 mrem/yr for I pCi of disposed Co-60.

The maximum organ dose is 1.9E-5 mrem/year per pCi, with the organ being the LLI wall.

These results are several orders of magnitude below the direct surface exposure doses as detailed below.

The groundwater pathway is, therefore, not significant. --

7 Revision 7 -

Date:

MAY-_1119_

A-37 Approved By:

9

Direct Ground Plane Exposure To account for the gamma attenuation provided by the soil, it was necessary to carry out an appropriate shielding calculation.

This was accomplished through use of the DIDOS computer code which computed the radiation levels from a cylindrical volume source with a radius of 3.85 m and a height of 0.15 m, with the receptor located along the axis, 1 m above the source.

The source density was set equal to 1.6 g/cc, which is equivalent to the Regulatory Guide 1.109 value of 240 kg/m 2 for the effective surface density of soil within a 15 cm plow layer.

If the total activity content of the septic tank, as listed earlier, were assumed to be uniformly distributed in the source disk, the volume source dose rate is equivalent to a dose rate of 2.8E-04 mrem/hr.

The total dose from standing on the garden area for 360 hours0.00417 days 0.1 hours 5.952381e-4 weeks 1.3698e-4 months each year is seen to be 0.099 mrem from the total activity content measure in the septic tank (2.33 pCi) being placed on the garden.

Garden Pathway Total Dose The maximum individual ingestion/inhalation exposure assessments resulting from garden crops or pasture grass grown on a septage disposal plot were added to the direct ground plane doses discussed above.

This results in a bounding estimate of dose to a hypothetical maximum exposed individual.

The whole body and critical-organ radiation exposures after a tank pump-out and spreading on a garden at a concentration level equivalent to the measured concentrations in septic waste are as follows:

Radiation Exposure Individual/Organ Maximum Exposed Individual 0.122 mrem/yr Child/Whole Body 0.157 mrem/yr Child/Liver Revision 7 -

Date:

MA*Y 2 1 l90 A-38 Approved By:

The individual pathway contributions to the total dose are as follows:

Pathway-Dependent Critical Organ Doses Maximally Exposed Maximally Exposed Individual/Organ Individual/Whole Body (Child/Liver)

(Child)

Pathway (mrem/year)

-(mrem/year)

Ground Irradiation 0.099 0.099 Inhalation 0.0003 0.0001 Stored Vegetables 0.055 0.0214 Leafy Vegetables 0.0028 0.0011 Milk Ingestion*

(0.019)

(0.0036)

TOTAL 0.157 0.122 Tables I through 4 detail the internal dose breakdown by radionuclide and pathway of exposure.

As can be seen in the results, the whole-body and maximum exposed organ dose are appropriately equivalent.

This is due to the dominance of the external ground plane exposure pathway controlling the dose to both the organs and whole body.

5.3.4 Incineration Pathway Scenario At the present time, there are no known facilities for the incineration of septage in the vicinity of the Yankee plant.

For completeness, however, we have addressed the radiological impact expected from incineration.

This will preclude the necessity of revising this application request if such a facility becomes available in the future.

The basis for the radiological assessment of incineration is a report by Murphy, et al. (1989)

(Reference 15),

in which they calculated individual and population dose impacts from low level waste disposal scenarios.

This report used a radionuclide distribution that was based on extensive studies of

As described above, the milk pathway is mutually exclusive to the vegetable ingestion pathway; and, therefore, not added into the total. Revision 7 - Date:

MAY 21 i

A-39 Approved By:

power reactor low level wastes.

This distribution was similar to the measured distribution in the Yankee septage in that Co-60 and Cs-137 were the predominant gamma emitters.

The results of their analyses show that the transport worker receives the highest dose from the incineration scenario.

The transport worker dose is approximately a factor of 5 higher than either the maximum incinerator worker or the maximum disposal site operator, and is several orders of magnitude higher than the maximum individual doses to the general public.

The dose to the transport worker has been discussed above (Section 5.3.1) for the off-site disposal of septage from Yankee.

This transport worker dose will not change if the septage is incinerated, since it was conservatively assumed that the worker spends 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months traveling to the disposal site.

Therefore, the dose to the individual landowner, from the garden scenario, will still be controlling for all disposal options, including incineration.

5.4 Maximum Releasable Activity The above analysis for landspreading on a garden the measured activity levels detected in the septic tank indicates that over 80% of the total whole body dose received by the hypothetical individual is due to direct external exposure to the ground plane.

Of this direct dose component, Co-60 accounts for about 96% of the exposure.

In determining a practical means by which any future detectable levels of licensed material can be limited to ensure that the controlling hypothetical individual's annual dose is limited to approximately I mrem or less, the sum of all measured gamma emitting nuclides can be assessed as Co-60 to determine the quantity of gross activity that, if released in septage, would limit the dose to 1 mrem.

Repeating the above controlling analysis for the event which placed the septage shipment directly on a garden plot, and assuming that the activity available is all Co-60, the total activity which relates to the annual dose Revision 7 - Date:

MAY 21 30 A-40 Approved By:

)

limit criteria of 1 mrem is determined to be approximately 20 microcuries.

The breakdown by exposure pathway for this scenario, assuming an activity release of 20 microcuries in the form of Co-60 is as follows:

Maximum Exposed Individual/Whole Body Pathway (mrem/year)

Ground Irradiation 0.980 Inhalation 0.0004 Stored Vegetables 0.13 Leafy Vegetables 0,0068 TOTAL 1.1 All other scenarios for the normal treatment and disposal of septage, including postulated accumulation and build-up of activity at a single SWTF for a 30-year period (at 20 microcuries/year), result in radiological impacts to individuals which are approximately a factor of 100 or more below the whole body dose for the garden pathway.

The following summary compares the calculated whole body doses associated with normal handling of septage with the I mrem bounding event garden scenario.

This demonstrates that by limiting the annual quantity of activity in septage to 20 microcuries, the expected dose impact for disposing of septage at a SWTF will in fact be well below a dose criterion of 1 mrem/year:

Maximum Whole Body Annual Dose Scenario (mrem)

(a)

Septic truck driver/SWTF worker.

1.OE-02 (20 uCi Co-60 per year)

(b)

SWTF landfill after closure.

5.6E-03 (30-year accumulation; 159 uCi Co-60) e o

7 ae A-41 Approved By:

6.0

SUMMARY

AND CONCLUSIONS The disposal of septage by transferring it to a public SWTF is in accordance with standard practices for treatment of the type of waste material generated by a septic tank/leach field sanitary waste system.

Periodic pumping of the septic tank is necessary for the maintenance, and continued operation of Yankee's sanitary waste system.

Approval for disposal of~septic waste from the Yankee sanitary system is requested to prevent failure of the sanitary system to adequately handle plant domestic waste.

Alternate means of disposal of the septage would involve the treatment of it as radwaste, with the subsequent need to stabilize, solidify, and dispose of the material at a licensed burial ground at excessive cost and a loss in valuable disposal ground volume.

The radiological analysis results indicate that the public health effects due to the biological activity and infectious constituents of such sanitary waste far outweigh the concerns due to any radioactivity which is present.

By setting release limits which restrict the exposure to a maximum hypothetical individual of 1 mrem per year, it is ensured that radiological risks from the proposed disposal method are of no significance.

The proposed release limits represent a small fraction of NRC limits permitted for disposal of similar waste by licensed facilities who have their sanitary systems connected directly to a public sanitary sewerage system.

These proposed limits are also within the plant's current allowable release limits for discharge of normal liquid waste to the environment, with any resulting dose to any individual in the public being far less than committed exposures due to natural background radiation. Revision 7 Date: MA 21 !32 A-42 Approved By:

V

7.0 REFERENCES

1.

"Design Manual - On-Site Waste-Water Treatment and Disposal Systems,"

U.S. Environmental Protection Agency, EPA-625/l-80-012, October 1980.

2.

"Septage Management," U.S. Environmental Protection Agency, EPA-600/8-80-032, August 1980.

3.

"Handbook - Septage Treatment and Disposal," U.S. Environmental Protection Agency, EPA-625/6-84-009, October 1984.

4.

"Septic Tank Care, U.S. Department of Health," Education, and Welfare, U.S. Public Health Service, 1975.

5.

"Manual of Septic Tank Practice," U.S. Public Health Service, Publication No. 526, 1957.

6.

"Your Septic System," Prepared for the Massachusetts Department of Environmental Quality Engineering, Publication No. 10043-32-625-12-77-CR, January 1978.

7.

"Septic System", Massachusetts Metropolitan Area Planning Council, 1981.

8.

"Septic Systems," Massachusetts Division of Water Pollution Control, Publication No. 12551-24-300-9-81-CR, 1981.

9.

Clark, J. W., W. Viessman, and M. J. Hammer, "Water Supply and Pollution Control," International Textbook Company, 1971.

10. Metcalf & Eddy, Inc., "Waste-Water Engineering:

Treatment, Disposal, and Reuse," McGraw-Hill, 1979.

11.

Cember, H., "Introduction to Health Physics," Page 321, Pergamon Press, 1969.

12. Kennedy, W. E.,

Peloquin, R. A.,

"Residual Radioactivity Contamination From Decommissioning," NUREG/CR-5512, January 1990 (Draft'Report for Comment).

13.

Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR, Part 50, Appendix I, USNRC, Revision 1, 1977.

14.

J.

N.

Hamawi, "DIDOS-Ill - A Three-Dimensional Point-Kernel Shielding Code for Cylindrical Sources, ENTECH Engineering, Inc., Technical Report PIOO-R2, December 1982 (updated to Version DIDOS-IV, October 1989, Yankee Atomic Electric Company).

15.

Murphy, E. S., Rogers, V.

C.,

"Below Regulatory Concern Owners Group:

Individual and Population Impacts From BRC Waste Treatment and Disposal,"

EPRI NP-5680, Interim Report, August 1989.

16.

Massachusetts Department of Environmental Protection Regulations 310 CMR 19.15 (Disposal of Solid Waste in Sanitary Landfills). Revision 7 - Date: MY2 A-43 Approved By:

TABLE 1 LANDSPREADING INGESTION PATHWAYS (ADULT)

(2.33 UCI TOTAL ACTIVITY)

(CREM)

PATHWAY BONE INHALATION 54 MN 60 CO 134 CS 137 CS LIVER 2.93E-06 2.11E-05 7.22E-05 1.39E-04 KIDNEY 7.28E-07 Q.OOE+0O 2.44E-05 4.98E-05 LUNG 1.04E-04 1.09E-02 8.31E-06 1.68E-05 GI-LLI 5.72E-06 5.21E-04 8.85E-07 1.88E-06 THYROID VHOLE BODY 0.00E+00

0. OOE+O0 3.17E-05 1.07E-04 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 4.66E-07 2.71E-05 6.19E-05 9.58E-05 TOTAL FOR PATHWAY STORED VEGETABLES 54 MN 60 CO 134-Cs 137 CS TOTAL FOR PATHWAY 1.39E-04 2.35E-04 7.49E-05 1.11E-02 5.30E-04 O.OOE+00 1.85E-04 O.0OE+00 O.OOE+00 2.24E-03 9.25E-03 3.10E-04 1.78E-03 5.33E-03 1.27E-02 9.21E-05 0.OOE+00 1.72E-03 4.29E-03

0. OOE+O0 O.0OE+00 5.72E-04 1.43E-03 9.48E-04 3.34E-02 9.32E-05 2.45E-04 O.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 5.91E-OS 3.92E-03 4.35E-03 8.29E-03 1.15E-02 2.01E-02 6.11E-03 2.00E-03 3.47E-02 0.00E+00 1.66E-02 LEAFY VEGETABLES 54 MN 60 CO 134 CS 137 CS TOTAL FOR PATHWAY COW MILK 54 M*

6O CO 134 CS 137 CS TOTAL FOP PATHWAY O.OOE+00 O.OOE+00 2.91E-04 1.14E-03 4.34E-05 2.24E-04 6.92E-04 1.56E-03 1.29E-05 O.OOE+00 2.24E-04 5.31E-04 O.OOE+O0 O.OOE+00 7.4,E-05 1.76E-04 1.33E-04 4.20E-03 1.21E-05 3.03E-OS 0.00E+00 0.00E+00 O.OOE+O0 O.OOE+00 8.29E-06 4.93E-04 5.66E-04 1.02E-03 1.43E-03 2.52E-03 7.68E-04 2.51E-04 4.38E-03 O.OOE+00 2.09E-03 O.OOE+00 0.00E+00 8.11E-04 3.31E-03 2.39E-06 5.33E-05 1.93E-03 4.53E-03 7.10E-07 0.00E+00 6.25E-04 1.54E-03 0.00E+00 O.0OE+00 2.07E-04 5.11E-04 7.31E-06 1.00E-03 3.38E-05 8.77E-05 0.00E+00 O.OE+00 0.00E+00 O.OOE+00 4.55E-07 1.18E-04 1.58E-03 2.97E-03 4.12E-03 6.51E-03 2.16E-03 7.18E-04 1.13E-03 0.00E+00 4.66E-03 25 Revision 7 Date MAY 21 M"T A-44 Approved By:

TABLE 2 LANDSPREADING INGESTION PATHUAYS (TEEN)

(2.33 UCI TOTAL ACTIVITY)

(,REJ4)

PATHUAY BONE INHALATION 54 1M 6O CO 134 CS 137 CS LIVER 3.78E-06 2.77E-05 9.60E-05 1.90E-04 KIDNEY 9.41E-07 O.OOE+00 3.19E-05 6.80E-05 LUNG 1.47E-04 1.60E-02 1.25E-05 2.70E-05 GI-LLI 4.94E-06 4.75E-04 8.31E-07 1.90E-06 THYROID WHOLE BODY O.OOE+0O O.OOE+0O 4.28E-05 1.50E-04 O.00E+00 O.OOE+00 0.00E+00 O.OOE+00 6.21E-07 3.63E-05 4.67E-05 6.96E-05 TOTAL FOR PATHWAY STORED VEGETABLES 54 MN 60 CO 134 CS 137 Cs.

TOTAL FOR PATHWAY 1.93E-04 3.17E-04 1.01E-04 1.622-02 4.82E-04 O.OOE+00 1.53E-04 O.OOE+O0 0.0OE+00 3.65E-03 1.57E-02 4.84E-04 2.83E-03 8.59E-03

-2.10E-02 1.44E-04 0.OOE+00 2.73E-03 7.13E-03 O.OOE+O0 O.OOE+00 1.04E-03 2.77E-03 9.93E-04 3.69E-02 1.07E-04 2.98E-04 O.OOE+O0 O.OOE*00

0. OOE200 O.OOE+00 9.60E-05 6.37E-03 3.98E-03 7.30E-03 1.94E-02 3.29E-02 1.00E-02 3.81E-03 3.83E-02 0.00E+00 1.78E-02 LEAFY VEGETABLES 54 MR 60 CO 134 CS 137 CS TOTAL FOR PATHWAY COW MILK 54 MN 60 CO 134 CS 137 Cs TOTAL FOR PATHWAY 0.00E+00 O.OOE+00 2.57E-04 1.05E-03 3.68E-OS 1.93E-04 6.05E-04 1.40E-03 1.10E-05 0.00E+00 1.92E-04 4.77E-04 O.OOE+00 O.OOE+00 7.342E-05 1.85E-04 7.55E-05 2.51E-03 7.52E-06 1.99E-05 0.00E+00 0.00E+00 O.OOE+00 O.OOE+O0 7.30E-06 4.34E-04 2.81E-04 4.88E-04 1.31E-03 2.24E-03 6.80E-04 2.59E-04 2.61E-03 0.00E+00 1.21E-03 0.00E+00 O.OOE+00 1.41E-03 6.00E-03 3.98E-06 9.03E-05 3.31E-03 7.99E-03 1.19E-06 0.00E+00 1.05E-03 2.72E-03 O.OOE+00 O.OOE+00 4.02E-04 1.06E-03 8.15E-06 1.18E-03 4.12E-05 1.14E-04 0.00E+00 0.00E+00 0.002+00 0.002+00 7.88E-07 2.03E-04 1.54E-03 2.78E-03 7.41E-03 1.14E-02 3.77E-03 1.46E-03 1.34E-03 0.00E+00 4.52E-03 26 Revision 7 Date MAY 2 1 1990 A-45 Approved By:

TABLE 3 LANDSPREADING INGESTION PATHWAYS (CHILD)

(2.33 UCI TOTAL ACTIVITY.)

(HREH)

PAT HWAY BONE LIVER KIDNEY INHALATION 54 MR 60 CO 134 CS 137 CS LUNG 1.17E-04 1.29E-02 1.03E-05 2.33E-05 GI-LLI 1.69E-06 1.76E-04 3.27E-07 8.10E-07 THYROID WHOLE BOOY 0.00E+00 0.001+00 5.54E-05 2.03E-04 3.17E-06 2.40E-05 8.63E-05 1.85E-04 7.41E-07 0.OOE+00 2.81E-05 6.32E-05 0.O0E+0O 0.00E+00 O.OOE+00 O.OOE+00 7.03E-07 4.15E-05 1.91E-05 2.87E-05 TOTAL FOR PATHWAY STORED VEGETABLES 54 MN 60 CO 134 CS 137 Cs TOTAL FOR PATHWAY 2.58E-04 2.98E-04 9.20E-05 1.31E-02 1.79E-04 0.00E+00 9.00E-05

0. OOE+00 O. OOE+00 8.42E-03 3.80E-02 7.25E-04 4.40E-03 1.38E-02 3.63E-02 2.03E-04 0.00E+00 4.28E-03 1.18E-02 0.00E+00 0.00E+00 1.54E-03 4.26E-03 6.08E-04 2."E-02 7.45E-05 2.27E-04 0.00E+00 0.0OE+00 0.00E+00 0.OOE+00 1.93E-04 1.30E-02 2.91E-03 5.36E-03 4.64E-02 5.53E-02 1.63E-02 5.80E-03 2.53E-02 0.OOE+00 2.14E-02 LEAFY VEGETABLES 54 WN 6O CO 134 CS 137 CS TOTAL FOR PATHWAY COW MILK 54 MN 60 CO 134 CS 137 CS TOTAL FOR PATHWAY 0.00E+00

0. 00E+00 4.45E-04 1.90E-03 4.13E-05 2.25E-04 7.30E-04 1.82E-03 1.16E-05 0.00E+00 2.26E-04 5.94E-04 0.00E+00 0.00E+00 8.11E-05 2.14E-04 3.47E-05 1.24E-03 3.93E-06 1.14E-05 0.DOE+00 0.00E+00 0.00E+00 0.00E+00 1.10E-05 6.62E-04 1.54E-04 2.69E-04 2.35E-03 2.82E-03 8.32E-04 2.95E-04 1.29E-03 0.00E+00 1.10E-03 O.00E+00 0.00E+00 3.25E-03 1.45E-02 5.95E-06 1.40E-04 5.+33E-03 1.38-02 1.67E-06 O.OE+00 1.65E-03 4.51E-03 O.OOE+00 O.OOE+00 5.93E-04 1.62E-03 4.99E-06 7.77E-04 2.87E-05 8.67E-05 O.OOE+00 0.00E-600 0.00E+00 O.0E0+00 1.58E-06 4.613E-04 1.12E-03 2.04E-03 1.77E-02 1.93E-02 6.171E-03 2.22E-03 8.971E-04 O.OOE+00 3.58E-03 27 MAY 2 1 1990 Revision "7 Date A-46 A-46 ApprovedBy

TABLE 4 LANDSPREADING INGESTION PATHWAYS (INFANT)

(2.33 UCI TOTAL ACTIVITY)

(MREI)

PATHWAY BONE LIVER KIDNEY LUNG GI-LLI THYROID WHOLE BODY INHALATION 54 MN 60 CO 134 CS 137 CS O.OOE+00 0.00E+00 3.37E-05 1.23E-04 1.87E-06 3.69E-07 1.47E-05 O.OOE+O0 5.98E-05 1.62E-05 1.37E-04 3.85E-05 7.39E-05 5.22E-07 O.O0E+00 3.69E-07 8.25E-03 5.84E-05 0.OOE+00 2.16E-05 6.78E-06 1.14E-07 O.OOE+00 6.34E-06 1.59E-05 2.99E-07 O.OOE00 1.02E-05 8.35E-03 5.94E-05 O.00E+00 3.84E-05 TOTAL FOR PATHWAY 1.57E-04 2.13E-04 5.51E-05 STORED VEGETABLES 54 MN 60 CO 134 CS 137 CS TOTAL FOR PATHWAT LEAFY VEGETABLES 54 MR 60 CO 134 CS 137 CS TOTALFOR PATHWAY COW MILK 54MR 60 CO 134 CS 137 CS TOTAL FOR PATHWAY 0.00E+00 O.OOE+0O O.OOE+O00 O.OOE+0O

0. DOE+0O O.OOE+00 O.OOE+00 O.OOE+O0 0.0OE+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.OOE00 O.00E+00 0.OOE+00 O.00E+O0 0.OOE+00 O.OOE+00 0.00E+00 O.00E+O0 O.OOE+00 O.OOE+00 O.OOE+0O O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.0OE+00 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.00E+O0 0.00E+00 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00

0. OOE+00 O.OOE+00 O.OOE+00 0.002+00 O.OOE+O0 0.002+00 O.OOE+00 O.OOE+00 O.OOE+00 O.0OE+00

0. OOE+00 2.51E-06 6.76E-04 9.85E-04 1.92o-03 3.589-03 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+O0 5.23E-03 2.31E-02 1.11E-05 2.86E-04 9.76E-03 2.70E-02 2.45E-06 O.O0E+O0 2.51E-03 7.25E-03 0.0OE+O0 4.06E-06 O.OOE+00 O.00e+o0 6.81E-04 O.ooe+00 1.03E-03 2.65E-05.

0.00E+00 2.94E-03 8.45E-05 0.00E+00 2.83E-02 3.71E-02 9.77E-03 3.97E-03 7.96E-04 0.OOE+O0 28 Revision 7 Date PAY 2 1 i99o A-47 Approved By:

0 co 0

4 I...............

I.........

I.................................

SOIL ABSORPTION, LEACH FIELD SEPTIC TANK.

LIQUID (PRETREATMENT)

T

~TREATMENT YA N KEE PLA N T.... i I......

D.......

WASTEVWATER.

...... D-G E**::......

SEPTAGE r...............................

TREATMENT ON-SITE PROCESS I-(See Note)

(LIQUIDS)

WASTEWATER OFF-SITE PROCESS (SEPTAGE)

FACILIT I.........

Note:

Septage Haulerfrank Truck Pipe Line YANKEE PLANT SANITARY WASTE DISPOSAL PROCESS FIGURE I

Appendix B Concentrations in Air and Water. Above Natural Background (10CFR20.1-20.602, Appendix B)

Revision 11 R12\\120 B-1

Appendix B APPENDIX B TO §§ 20.1-20.602--CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND.

(See footnotes at end of Appendix B]

Isotope Tati Element (atomic number)

Col. 1-Air

(_A_/M_)

Actinium (89)..

Americium (95)

AC 227...................

Ac228.................

Am 241.............

Am 242m...............

S S

S 2X10-12 3X10 "

8X10-4 2X10-S Am 242................

Am243.....

Am 244 S

S S

An~mon..........

4Sb 1

S I

Argon (18)

Arsenic (3).

Sb 125 A 37...........

A 41 As 73.................

Sub2 Sub S

S I

6X10-12 6X10-1=

3x,10-10 4x10-1 5X10-6 6X10-10 4x10-'

2X105-2x10-'

1X10-7 2x10-'

2x10-1 5X 10-3x10-4 6x10-3 2x10-6 2x10-6 4X10"7 3X10-1 I x10-1 1X10-1 1.X10-:"

5X10-2 4X10-'

7X10-3X10-4 1 X10-'

4X10-'

1 x10-7 4X10-4 9X 1.0-1X10-7 1X10-7 I1X10-11 6X10-i 1X10-6 2x10-1 1 X10-7 2X10-'

2x 10-7 2X10-6 6X10-,

Il Tabk Col. 2-Cof. 1--Air Water (I.Ci/ml) 6X10-1 8X10-11 9XI10-9X10-"

3X10-3 3x10~1 3x10-1 6X10-10 1x10-4 2x10-1 8X10-,

4X10"-1 1X10-4 2X10-"

3X10'-

9X10-4X10-3 1X10-9 4X10-=

2X10-1 1xl0-1 2X10-"

8X10- 4 4X10-lxl0-1 1X10~1 i x16-'

1x10-7 8x10-4 6X101' 8x10- 4 5x10~1 7x10-4 5x10-1 7x10-4 7xI0-10 3x10-3' 2x10'=

3x10-3 9x10-10 1X10-'

1x10-7x10-$

1X10-1 1X10-0 2x10-2 1XI0'S 2x10-3 4X10-9 6x10-4 4X10'9 6XI0-4 3X10--

2x10-2 2X10-5X10-9 2X1O-10 2X10-3 1X10-9 5X10-3 4X)10-'

5X10- 3 1X10- 8 8X10-4 4X10'-

7X10-4 1X10-9 2X10-2 3x10" 2X10-2 4X10-9 6X10-3 5X10-9 6X10-3 4X10-1 5X10-1 2x10-1 5x10'-

4X10-6 1X10-3 6X10-1 1X10-3 5X10-,

2x10-1 6X10-2x10-3 5X10'-1 1X10-1 2X10.

eII Col. 2-Water (JCi/ml) 4X10-:

3x10-s 4X10-.

9X10-3 1X10-4 1X10-1 4X10-1 3X10-5 5X10-1 5X10-1 3X10-6 3X10-s 2X10-5 2x10-5 1X10-1 1X10-4 5X10-4 5X10-6 5X10-1 2X10-S

2x1O-5 8X10-5 2X10-*

7X10-3

.2X 10-4 2X10-4 3x10-5 6 X10-4 2X10-1 2XI0-1 2x10-3 2)x10-'

4X10-s 4X10-s 6X10-s 6X10-5 4X10-s 2x10-6 3x10-1 9X10-3 9X10-S As 76-......

S As 77-......

Astatine(85)

At 211..

Barium (56)..

Ba 131.............

Ba 140..

.1 S

S S

I S

S S

Berkelium (97)

Bk 249.........

8k 250...................

Beryllium (4)....

Be 7.................

Bismuth (63)........................

R I i 206 Bi 207..........

ER Revision I I B -2

Nuclear Regulatory Commission Pt. 20 [ §§ 20.1--20.602]1, Ap p. -B APPENDIX B TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued

[See footnotes at end of Appendix 81 Isotope Table I Table II Element (atomic number)

Col I Air CoL 2-C 1-Air Col. 2-water Water (lOm/ml)

( (nO/ml)

(Ci/ml)

(aCi/mlm Bi 212....................

Bromine (35).

r 82...............

Cadmium (48)...........

Calcium (20).....

Califonium (8 Cd 115m..............

Cd 115.................

ca 45_._...____..

Ca 47..................

Cf 249........

Cf 250M.__

O a252....

Cf 252 Cf 254 C0 2......--....

Co 141 Ce 143...

Ce 144...

S S

S s

SS I

SI S

IS I

SI S

I S

Sub SI S

Cefium (58)_.

6X10-*

1X10-7 2X10' 1X10-'

2xIO-7 5X10-6

.7X10-'

4X10-1 4X10-,

2X10-2x10-7 3X10-,

Ix10-7 2x10-1 2X10-1 2x10-1 1X10-10 5x10"- 2 1 X10- "1 2xlO-12 1XlO-10 6X10-12 3X10-"

8X10-10 8XJO-12 5X10-12 4x10-4 5X10-S 4X 10-2X10-2 3x10"'

2x10-1 1X10-6 6X10-1 3X10-4 4X107s 6X10'-

4X10-.

1X10-6 5X10'-

9X10-'

-4X10-7 2x10-7 6X10'-

1X10-6 4x10- 7 2X1076 3X10-6 2X10-1, 1 X110-S 2x10-6 3X10-'

2XIO-7 2x10-'

9X10-1 8X10-7 5X10-6 3X10-'

9X10-9 2x10-1 2X10-1 1X10-1o 1X10-1 IX 10-2 1 X10-1 1X10-1 5X 10-3 5X10-3 7X10-4 7X10-4 1X10-1 3 X10-3 15XIO-3 1X1O-'

1X10-1 7 x10-4 4X10-4 7X10-4 1 xlO"'

8X10-4 2x10-'

2x10-1 2X10-1 4x10-'

4x10-4 4X10-'

4X10-'

2x10-1 3X10'-

3X10-3 1X10-3 1 XIO-4 3X10-'

3x10-'

7X10-1 3x10-2 2X10-'

3X10-,

axw 3 X10-3 3X10-1 7x1Q'3 2x 10--

2x10-3 4XIO-4 4X10-'

2X10-'

2x10-1 1X10-2 2X10-1 SX10-2 5X10- 2 2x10-.

1X 10-2 8X10-2 6X10-1 4x10-1 3x10-1 1X10-1 1 X10-3 I1X10-1 6X10-1 7X10-2x10-'*

3X10-,

7x10-1 6x10-1 2x10-1 3X10-1 IX1i0-1 X 10-'

8x10-,

6X10-1 1X10'-

4x10-9 6x10-t 6X10-,

5X10-,1 3X10-1=

2x.10-12 3xlO-11 6X 10-"

3XIO-12 3x10-"

3x10-"

2x10"13 2X 10-'

1X10-7 IX 10-11 2x10-6 5X10-*

9X10-$

7X10-3x10-10 2x10-10 4x10-1 1X10-7 1X10-1 2x10-7 A X10"9 4x110-10 2x10-'

3x10-9 1 xIO-6X 10-'

2X10-2 5X10-10 1X10-1 8X10-,o 9X10-9 7x10-4 4x10-1 8X10-'

I xlO-'

OX10-I 6x I0-*

3x10-'

3x10-'

2x10-'

Ix 10-'

3X10-7 6x 10-9 6x10-7 4X10-1 4X10-1=

4x10-'

4x10-1 3x10-1 4x10-S 2xIO-'

2x10-4 3x10-1 3x10-$

x10 1-S 9X10-'

2x10-1 SX10-3 3xIO-5 4x10-6 2x10'-

2x10-'

3X10-'

4x10-4 3xlO-S 7x10-6 7X10'-

1 X10-4 1x10-'

1X10-,

8X10-7 9X10-1 9X10-5 4X10-3 4x10-3 1 xIO-s I X10-'

2X10-6 2x10-2 9X10-'

6x10-3 9X10-'

4x10-s 1X10-4 2X10-4 6x10-S

2xlO-S 6X10-S 2x10-3 4X 10-8X10-$

6x10-'

5x 10-'

4X10-2 4x10-4 2x10-3 AX10-4 4x10-1 3x10-1 2 x10-3 9X10-1 3x10-1 3X10-,

2x10-1 2X10-*,

Cesium (55)

Cs 131 S

Cs 134m.....

Cs134....

Cs 135_L---

Cs 136.-.._-.---

S S

SI I

Cs 137....-

-l Chlorine (17)C--I.......................... ] Cl 36........

Chromium (24)..................................... -C 51 Cobalt (27)..............................................

C6 57....................

S S

SI S

Co 58m................

Co 58 Co 60......................

Cu 64......................

Cm 24 Copper (29).-----------------

Curium (96)................................................

IS Cm 242 Is Revision 11 B-3

Pt. 20 [§§ 20.1-20.602], App. B 10 CFR Ch. I (1-1-93 Edition)

APPENDIX B TO.§§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of Appendlix 3]

Isotope Table I Table II Element (atomic number)

Col.

1

-Air Col. 2-Col. 1 -Air Col. 2-Water Water (f/ml) i/m)

(fkCilml)

Cm 243.............

S Cm 244.........

IS Cm 245.................

Cm 246.................

Cm 247..................

S S

S Cm 248.................

S Cm 249 yarosiuy m(66)...................................

Dy 165............

Einsteinhim (99)..............................

Es 253..............

Es 25m.

ES 254___.._......

Europium (63)...........................

Er169........

El 171............

Eu 152.............

(T/2=9.2 hrs)--....

Eu 152...........

(T/2= 13 yrs)5....

Eu15......

2X10-,0 6X10-"

9X10-0 5x 10-12 5xlO-u 5X10-12 lX10-10 6xIO-11 IX10-1 0 1XI10-'

1X10-5 3x1O-'

2X10-'

8X10-10 6x10-t" 5x10-9 6x10-*

2x10-"

6X10-10 1xlO-"

5X10-10 4X10-16 6X10'-

4X10-'

7.Xl0-'

6x10'-

4Xl0-7 3X10-7

1 xlO-e 2x10-'

AX10-9 7x10-9 7x10-s 7X10-4 2x10-1 X10-'

3X 10-'

2x10-8 5x 10-'

3X10-*

2x10-6 9x10-6 5x10-1

-2x10-1 2X10-Ixia-'

6XlO-1 6X10-'

6X10-'

3x10-7 2X 10-1X10-1 8x10-1 4.X 10-'

7x10-1 2x10-1 7x10-1 7x10-1 2x10-'

2XI0-4 8X10-'

lx 10-4 OX10-'

1X 10-8X10-1 6X10-1 6X10-5 1 xIO-t Ax10-$

6X 10-2 6X10-2 lX10-2 1X10-3 lx10-2 7X10-4

'7x10-4 5XI0-4 5x10-1 5x 10-':

4x10-4 4x10-'

OX10-'

8X10-'

3x 10-'

3x10-3 3x10-3 3x10--

2x10-2 2x10-'

2x10-2x10-3 6x10-'

6X10-'

6x10-3 6x10-3 4X10-3 4x10-3 I X10-3 IX10-3 3x10-5 3X10-1 2XO-2 1X10-'

6X10-3 6x10-3 2x10-3 2x10-2 Ix 10-3 lX 10-5X10- 2 5x10-1 2x10-1 21 x0-'

1XiO-'

4X10-1 4x10-1 2x10-1 9x10-1 6X 10-"1 2x10-',

3X 10-"

3x10-"

3X10-,2 2x10-"

2x10-"

2 xl 0o-Ax 10-"=

2x10-3 4X 10-"2 2X10-,"

4X10-"

4x10-1 AX10-'

9X10-4 7x10-*

8X10-9 7x10-'

3X10-"

2x10-"

2x10-10 2x10-16 6X10-13 4x10"12 2X10-"

2x10-"

1X10-o 2x10-4 2x10-'

1X10-1 1X10-1 4X10-1o 6x10-10 Ix 10- "o 2x10-1 3X10-9 2X10-1 2x10-1 2x10-9 4x10-10 1 x1O-'t 6x10-"

2X107-9X10-1 8x10-9 2x10'-

2X10-'

1X10-1 OX10-9 6x10-1 4XIO-1 2x10-1 4X10-s 2x10-'

AX10-1 4x10-1 3X10-4 3x 10.4-3x10-1 7x10-'

2X10-1 5X10-'"

2X10-7)<10-1 3X10-'

4X10-4; 3x10-1 4X10-'

3x 10-s 4X10-'

2X10-5 4X 10-'

1X 10-"

2X10-'

2x10-1 4x 10-4 4x10-1 4X 10-4X10-'

2X10-5 2x10-'

2x10-3 2X10-5 1X10-S 1xl10-S 3xlO-$

3xlO-6 9X10-5 9X10-6 iX10-1 1xlO-4 6x 10-'

6X10-5 6X10"-'

8X!0-=

2X10-'

2xl0-'

2X10-5 2X10-1 61X10-'=

1X10-4 3X10-4 1X10-s 3x10-5 9X10-7 2X10-'

2X10-4 I2x10-'*

2x10-4 2x10-4 4x 10-'

4X.10-1 2x10-1 2x10-3 Ox 10-'

1 X10-4 5x10-1 2x 10-'

2x10-1 2x710-1 7x10-5 1 x 10-1 2x 10-1 k

x (100)

FM 254_______

Fm 255____L....

)

F......

F 6.....

Gadolirgnm (64).......................

G-

-53 Gd 159..............

Gallium (31)............

Ga 72..................

S S

SI S

I S

S S

s Germanium (32)..................................

Ge 71................

Gold (79)

Au 196...............

Au 198..............

Au 199...................

Hafnium (72)....................

Hl 1............

Holm ium (67)

I Ho 166...................

Revision I I B-4

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.602], App-B APPENDIX B TO §§ 20.1-20.602--CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of Appendix 0]

Isotope' Table I Element (atomic number) o.A Col. 2-SCol. 1-Air Water (J.Cj/ml)

C ml Table II Hydrogen (1)...........................

Indium (49)........................................

H3....

In 113m................

In 114m..................

In 115m.................

Sub S

S II S.

In 115................... J I

lodine, (s3)....................................................

1 125....

I 126........... _IS I

131..................... i I

S I

S 135................... IS I

Iuldium (T7)...........

jrS..............II 9

I If Is 2x,10-I 5X10-4 5X10-1 2x10-.

.8X10-'

7x10-1 1X10-1 2x10-'

2x10-6 2X10-,

2xlO-7 3X10-9 5X10-9 2x10-7 8X10-,

3x10-1 2x10-7x10-'

9X10-3X10-"

2X10-"

9X10"7 3X10-,

2X10 8-5X10-7 3X10-1 4x10'7 1X10-6 4x10-7 1X1O-'

3x10-4 aX 10-a 2x10-7 2x 101 9X10-7 1X10-6 1X10-1 5X10-4 6x10-'

1 X 10-'=

1XlO-4 2x10-1 1X10-1 3x10-'

2x10-4 1X10-10 2x10-1 2x10-4 2X10'-

6x10-1 5X10-7 2x10-'

1X10-7 4X10-1 4X1.0-8 8X10-7 5X10-,

7X 10 1X10-,

3X 10-7X10'-

1 X10-7 9X10-1 1x10-'

IX10-'

4xlO-1 4x10-z 4x10-'

6 X 10-'

5X10-,

1X10-1 lx10-1 3X10-,

3X10-1 4x10-6x 10-1 5X 10-5 ax 10-?

1X10-*

6X10-1 6x10-6XI0-3 2X10-3 5X10-,

2X10-4 1X10-1 4x10-3 2x10-1 7x10-'

2X10-3 6x 10-3 5X10-2 1XI0-3 1X10-3 1X10-3 9X10-4 2X10-2 7x10- 2 2xlO-Z 2X10-1 7x10-'

7X10-4 1X10-2 1XIO-2 4x10-'

5X10-3 6x10-1 SX10-4 3X 10-.

3X10-3 1 X10-3 9X10-1 4x10-1 3X 10-4x10-1 3x-10-1 6x10-1 SX10-1 9X10-1 1X10-2 5X10-1 3X10-?

Cl1-AlCal Cal.

-Air Waler 6xý10' 3X10-1 2x10'1 3X10-1 2x10-'

3X 4X10-1.............

3x10-'

1x10-1 2X10-1X10-1 4X10-9 2X10-1 7X10-'*

2x10-1 8x16--

4x10-6x10O1 4x10-'

9x10-2 9X1O-5 8xl0f1 2x10-1 6x10-1 2x10-9x10"1 3x10-7 lxiO'1 9x10-,

W0x11 6X10-1 2x10-'

2x1O-1 1x10-1 3x10-1 iX10-6 6X10-3 3X10-,

8X10-4 3x10-4 2X10-1 4X10-1 1X10'1 7x10-'

4x10-5 6x10-9 2x10-1 1x10-,

6x 10-4 4x10-1 2x10-1 1x10-1 2x10--

4X10'9 4x10-'

9X10-11 4XII0-8x10-1 3X1O-3 5x10-9 3x10-3 3x10-6 8X10'1 3Xi0-4 2x 10-3 SXiO-9 6x10-1 2x1079 SX10-5 3x 10-1 2x10-6 210 5x10-9 2x10-5 9x10-ý 4x 10-4 6X10-4 4x10-4 4x10'"

ixia-7 BxlO-12 2x10-1 6x10-10 2X10-3 7x1O-10 2x10--1 2x101 1X10-4 2x10-,

1 x10'

7x107, 3xI10s 5xlOT' 3x10-1 2XIO-1x10-1 3XIO-1 2x 10-4 3x-101 2x 10-4 4X10-'

3X10-4 9x10-1 5X10-2X10-9 2x10-1 4X10-I X lO If 194..................... S Iron (26)..

I Fe S

Fe 59................... S Krypton Lanttianum(S).......

Kr 8...........

Kr88 Sub Sub S ub Sub S

Lead (62)

I Pb 203_____.___

Pb 210......

S Lutetium (71).......................

I Lu 1............

Is Manganese (25)..

M n 52...................

S I

S M n 54................

M n 56....................

M ercury (80)...............................................

Hg 197m................ IS Hg 197...................

Hg 203...................

SS Revision 11 B-5

Pt. 20 [§§ 20.1-20.602], App. B 10 CFR Ch. 1 (1-1-93 Edition)

APPENDIX B TO §§ 20.1--20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of Appendix B]

Isotope Table I Table II Element (atomic number) tl 1

0o4 2;- I Col. 2-Col.

Im Water wCo. I-Air Water I

JCM)I(/*.CUI/ mll)

(IGi/MI)

Molybdenun (42)

IMo99.............

Neodymium (60)

I Nd 144...................

S SI S

S Nd 147...................

Nd 149..................

'Neptunium (93)...............................

Np 237.............. S I

Np 239...................

Nickel (28)...

INi 59...................

Niobium (Columbium).(41).......................

Ni 63......................

Ni 65.....................

Nb 93m..............

Nb 95..

Nb 97...................

Osium (76)........

Os 185..........

0Os 191m...............

OS191................

Os 193................

7X10-1 2x10"7 2X10-1 3X10-4XI0-7 2x10'7 2X 10-1 2x 10-"

1X 10-"

4lxl0-ý,

1X10- 1 8x10-1 5X10-7 5X10-7 8x10 1-6X 10-"

3X10-7 9X10-7 5X10-7 7x10- 1 2x10'-

5X10- 1 lX10-1 6X10-4 5XlO-41 5X10-1 5x10'-

2xlO-*

9X 10-'

4x10 7-4X10-7 3X 10-Ix10-4 7X10-7 6X10-7 4X10-7 8X10-1 6X10-I 7x10-4 1X10-4 3X10-'

6X10'-

5X10-1 8X 10-6X10-7 2XlO-11 2x 10- 2 2x 10-"2 2x 10-9X10-"1 4X1 0-'

2 xl 0" 4X10-2X10-'

4X10-'*

2x10"-

5X10-"

2X10-"

3X10-11 5X10- 1o 2X10-o 5X10-2 IX10-1 2x10-3 2XI0-3 8X 10-3 8X.10-'

9X10-9X10-4 4X10-3 4X10--1 6X 10-6x 10-'

8X10-1 2X10-2 4X10-3 3 x10-'

1 X10-'

1 X10-2 3X10-2 3X10-1 3x10-'

3X10-2x10'-

2X10-3 7X 10-2 7x10-2 5X10-3 5X 10-2X10-3 X 10-2 8X10'-

3X10-2X10-'

5X10-'

7X 10-4 ax 10-'

3X10-3 3X10-2 3X10-2 5X10-2 3X10-:

3X 10-7 4X<10-3 3 x10-3 1)<10-4 SX10-1 1X10-"I 1X10-4 ax10o-'

6XIO-1 7X10-1 4x`10-1 1X10-4 9X10-1 1 X10-1X10-1 1X10-1 3X10'-

2x10-8xl0-'

ax 10-'

7x10*'

3xO-10 lxl40-"i 1 x10-1 6X10-I 6X 10-1 5X10-1 4X10-11 2x10-1 2X10-1 3X10-1 2x10-9 I1X10-1 3X10- I 2X10-'

4X1I7-1 5X10-*

2X10-:

3X10-1 2 X10-2X1 0-2X10-'

2x10-'

6X10-7 3X10-7 4XI0-1 I1X10-1 1X10-4 9X10-1 5X10-0 3X10-6 2x10-1 1X10-6 2XI0-9 SX10-9 a x10-1 2X10-4 2X10-1 2X10-4XIO-9 1X10-1 2XI0-1 2XI0-1 3X10-4 2X 7x10-1 I1X10-11 6X10-11 2 xl 0' 6xiO1' aX10-1 3x10-12 1XlO-9=

3x10-'1 1 X10-9x10'-

6 X10-'

8X10-1 6X10-'

2xlO-1 ax 10X-'1 7x10'*

2xt0'1 4X10-4 4X10-s 7x10-'

7XI0-3 8x10'-

6X 10-=

6X 10-'

3x10'-

3X 10'-

3x10-'*

3x10-3 1X10-1 2x10-1 3x10-'

7X 10-1X 10-'

1X10-"

4x10-4 9X 10-9x10-=

7x10'-

2x10-'

2x10'-

2xlO-S 5X10-1 3x10'-

3X10-4 9X10-3 2x10-3 2X10-1 2X10-1 1X10-3 1 X 10-1 xlO-'

1 xl0-'

9X10-4 2 X10-3 1X10-'

9X10-1X10-4 5X10-4 3X10-'

5X10-'

3X10-5X10-6 3X10-1 2x 10' 1X10-1 5X10-1 3x10G-3X10-'

3x 40' 7x 10-'

3X10-'

Palladium (46).....................................

Pd103......._

Pd 109..........

Phosp (15)......

P32.......

Platinum (78)..

....................... :........ IPt 191..................

Pt 193m........

Pt 193....................

Pt 197m...............

Pt 197....................

Plutonium (94)

Pu 238..........

Pu 239...............

Pu 240................ IS Pu 241....................

Pu 242....................

Pu 243....................

Pu 244....................

S S

S Polonium (&4).....................................

...... IPo 210..........

Revision 1 l B-6

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.602), App. B APPENDIX 1 TO §§ 20.1-20.602--CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of Appendix 81 Isotope Table I Table l1 Element (atomic number)

Cot, 1-Air Col. 2-Waler (pCiml)

Col. 1--Air (pCilml)

Cot. 2-Water (pCi/ml)

Potassium (19)..........................

1K42 PraseodymiurD (59)..................

Pr 142.

Pr 143.....................

Promethium (61)..................................

Pm 147........

Pm 149.........

S S

S Protoactinium (91).......................................

Pa 230....................

Pa 231................

Pa 233....................

Radium (88)..................................................

I Ra 223...................

Ra 224.............. IS Rhenium (75).-..........................................

Ra 226...................

Ra 228...................

Rn 220...................

Rn 2221................

Re 183...............

Re 186..............

Re 187..............

Re 188...................

S S

2X10',

2xI10' 2X10-'

3x10-7 2x10-1 6X10-1 2x10-1 2xI0-'

1 x10' 2xI107 2x10-9 SXlO-1 7 xI 10 3X10-1 5X10-1 3X10-1 3X10-4 25<10-1 2*10-1 4X10-'

2x10-1 8x10-S 6Xl0--

8XIO-7 5x10-7 3xl0-'*

5X 10-1 7x10-6 2XIO-4 2X10'6 5x10- 1 8X10-6 7X10-"

5X O10-8X10-1 6x10-*

7XIO10" 3x1-"*

6x10-6 1x 10-7 5)(10-7 4XIO-7 2x10-1 2x10-6X10'7 5X10',

2x10-1 lX O10 lX O10 lX 10-'

SX O10 6 x 10' I X 10-'1 7x 10'3 I X10-3X10-78X 10-'

7x 10-3 3x10-1 a X 10-4x10-1 9X10-,

2X10-1 7X 10ý 8X 10-3 3x101 IX10-3 4X10-2 4x10-3 4x10-2 2X10-3 7XIO10 3X10-1 2x10..'

4X10-4 I X10-1 3X10-1 8X10-1 8 X 10-2 I X10-3 9X 10-3 7 x10-4X to-'

7x10-'

5 X t0-I X 10-6X 10-,

2x10-1 3x 101 II x10 SX10-1 6XIO1' 3X 10-1 4 X 10-4 X 10-ý 2x101 6X10-9 6XOx 10 8xIO1" 2x 10-l 2x10'.

3xIO1" 2xIO1" 2x10"1 Ixi10-12 1 xlO1 3x10-1 9x 10-6 5x10',

2x10-1 8x10'1 3x10'1 2x10'6 1 xI10-6xI0-'

3X 10-'

2xI0-1' 3x io-,

2X1-'!

lx 0-,

2xI0'9 2X10-4 2 X10-9 8x 10-6' 6x10-4 2x10-4 ax to-'

2x10'-.

2x10-'

3XIO-1 2x10-1 2x10-1 9X10-12 2x10'9 SX10-1 2x10-1 1 x10-1 8X10-1 6X10-2x10'1 2 xl O 6x 10':

5)x 10-1 AX 10-AX 10'1 3X O10 2 x10-1 3xl 10 3 x10-5x10-1 Ax 10-Ax O10 2x 10-'

42X10-92X10-'

2x10-1 1 X 10-1 1 xlO-1 7x10-7 4X 10-'

2x10-1 5x10-1 3X10-4 3X10-1 3x 10-5 6x10-4 3xI10' 9xl0-1 5 xl 0' 3XIO-3 2X10' 6x10'1 3x10-5 lx 10ý 1 X 10-'

1 xlO-1 1x10'1 7XI0-s 2x10' 1x10-1 2x10'1 Ax 10-4 ax 10-'

8x10' 8X10-5 lx10I-'

I xlO-'

1 x10-1 IX 10-1 6XOx1-5 7x10-1 4X10-1 4x 10-1 8X10-1 8X10-1 4x 10-1 4x 10-1 9X10-1 9 X 1O0-3X 10-1 3x10-1 3x 10-1 3X1iT' Rhodium (45)...............

Rh 103m................

Rh 105...................

Rubidium (37)..............................................

IRb 86.....................

Rb 87.....................

Ruthenium (44)................................I Ru 97....................

Ru 103..............

Ru 105...................

Ru 106..................

Samarium (62)............................................

Sm 147..............

S SS s

S S

S SS Sm 151.............

Sm 153..................

Scandium (21)......................

Sc 46.

Sc 47 Sc 48...............

Selenium 34).........

Revision 11 I..............I SC 75._.

B-7

Pt.-20 [§§ 20.1-20.6021, App. B 10 CFR Ch. I (1-1-93 Edition)

APPENDIX B To §§ 20.1 -20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUNO---Co--ttntiUed (See footnotes at end of Appendix B]

Isotope' Table I Table It CoI2-Element (atomic number)

Col. 1 -Air Wte2

(!ICi/ml)

(pCi/mI)

o. -ArICcl.

2--

Go*!Il.-Ai Water (imI)

(ur;/ml)

Silicon (14).

Si31.....................

Silver (47).....................................................

Ag 105....................

S S

I I

Sodium (11).

Strontium (38)............................................

Ag 110m................

Ag III...................

Na 22...............

Na 24.....................

Sr e5m...................

S& 85...................

Sr 89.......................

Sr 9o......................

S S

I S

Sr 91....................

Sr92.........

S Sulfur (1S)........

35....................

S Ta tau (3)..

ITa 182................

S Technetium (43)...........................

Tc 96m.

S S

6xI10' 1 )< 10-6X 10-7 2xI0' 1lX 10-3XI10' 2x10-7 2x10'1 9X to-'

1Ix 10 1x10'7 4XIO-S 3xI103 2x O10 1 X 10-7 3x 10-6 AXI106 I10x 1' 5X 10-9 AXIO-0 3x10'7 4X10-7 3x10-7 3x10'1 3x O1O 4X1lO1 2X10-1 8X io-5 3XIO-5 Ox IO-1 2x10'7 2x10'1 2X10-1 1X10-5 3X10-1 4XIO-5 lx10I-1 2 x10-1 6x1O-1 4x:10-1 lX 10' 1 X10-7 4x 10' 2X 10-6 9 x10-7 Ox10-0 3X 10-1 5 XI10' 4XIO10' 4x-107 2x10-,

2X O10 1 X10-7 IX10-1 3x-109 3x 00-'

IXI10' 2x10'1 9XIO-7 8x10-7 2XIO10 6x10-1 3XIO10.

3X tO-2 6x O10 3x10-3 3xI0-1 9 Xl 0 9x10-1 1X10'1 1 X10o-1 xlO-'

SX10-1 6x10-1 8X10'1 2xl 10 2x10'1 ax 10-'

5x10'3 3 X10-4 8XOx 04 1lX 10-'

1lX 10-'

2>10" lX 10' 2x10-3 2x 10' 2x10'3 8X10' 1 X 10-'

1 x10-4x10'1 3x10'1 3x10'1 I X 10-:

1 X to0 5X10-3 5x10-2 2X10-2 2xI101 Sx 10-2 1 x10-2 5xI0'1 5XJO03 3XI1-3 2 X10-3 2x 10-3 6Ox 10-5XIO-3 lx 10-'

6x 10-i 2x10-2 2XIO02 2X10-1 1I x

9 xl0-1 6Ox 10-IX 10-1 lx10I-'

I X to-7xl10' 9 X1I0'-

5x10-'

4X10-3 2x:10-1 3x10'-

2x 1O-'

2xl0-'

9X10-1 3xl0' 2x10-1 2xIt06 IXI101 3Xi10-1 X10-3xiO-1 3x10-5 I3X10-:

4XI0-1 ix10-1 4XIO-3 6X1O-9 4X10-'

3xiO-I*

3x101

.4x10-6 2xI101 5x10'9 3xl0-'

1X10'1 7x10'1 1x10'1 7X10-3 4XI1-1 2x10-1 31x10' 3x10-1 3x10-1, 3x10'7 2x10'9 7x10-3 9x10-9 5x10-5 2x10-6 7x10'5 1x10-1 6xI105 9x10-9 6x10-5 9x10'*

3x10-1 7xO10" 4XIO0' 3X10'4 IxlO-'

IX10O' Ix10-2 SX109 5X10-5 5X10-4 4XIO-4 5 X107' 2x10' lxlO-8X10'4 1x10'1 6xI10' 5x10-3X10'3 7x>,10-6 3X10-4 2XIO-2 2x110-1 x10-

. 2x10-4 4)<10-9 lxlO-1 5X10',

6x101s 1IlO-9 SX10-3 6x<10-6 3x10'1 3><10-'

2X10-1 lx10-1 2x10'5 2x10' 8 10l-4 1 X10-'

8xI10' 1 x10'1 6x10'5 6xiO-1 4xI101 7X10-1 3X10-5

.4xI0-'

2x 10-5 3x10-1 4X10-5 9XO10 4x10-1 4XIO-1 2x 10-7 X10-1 3x10-'

3xIO-6 2xl10' 3XIO-1 1(10-4 8xIO01 7x10-'

2x10'I 1x10-9X10*"'

6x10-'

Tc 97m.......

....... -S I

Tc 97......................

Tc 99m...................

Tc 99..................

S IS s.

S S

Tellurium Te 125m Te 127m...............

Te 127.................

S S

Te 129m...............

Te 129..................

S I

Te 131m................-

S I

Te 132................ IS Terbium (65).........................

Tb 160....... :............ s I

Thallium (81)

TI 200...............

S TI 201................... S TI 202................... S TI 204...................

S I

Revision 11 B -8

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.602], App. B APPENDIX B TO §§ 20.1-20.602--CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued

[See footnotes at end of Appendix 8]

Isotope T

Table I Table II Element (atomic number)

Col 1--ir Col. 2-C Col. I-Ai CoL 2-Water 2

Cot I-Air Watef (I.Ciml)pCi/)

(jCd*m/)

(tC/m)

Thorium (90)..............................

Th 227.

Th 228..................

Th 230..................

Th 231.............

Th 232...................

Th natural.............

Th 234.................

Thuliurn(69)...................=....=..........=....... I Tmn 170..........

Tm 171.............

TinSn 13...

jn 3

Tungstn (ottram) (4)..---- -- -------

W 185................

Uranium (92)

W187...

U230.

U 232..............

3 xl -1 2x10"1 6x10-1 2xlo10 1XIO1" I X 10-'

lX O1' 3xlOat.

3X10t*

6X10-1 6X10-4 3x10-8 4x10-6 3 x10'4 lX 10-7 2x10-1 4 X10,7 sXlO-'

1 x10-7 8X10-4 2x10-4 1I xO-1 8X10-1 1)<10-7 4X 10-'

3 X10-1 Ux10-1 1X10-16 1X10-w ax IO-1w 5xl0-t o lX1 0-"0 6 x10-1 lX 10-to 5X10-14 lx 10-.to 6 x10-10 1XIO10" 7X10-11 1 X10-1 2x10t" 2x10-1 lX101 lX 10-to 2 XI-10 6X106 2x10-5 1Ix 10 1Ixi-6 4XIO-'

7X10 t?

6X10-1 I X 10-7 2 x 10-2x10-1 3x10',

2XIO-1 1 X10-'

5x10'4 5xi0-'

2x10-1 4xI1O4 5xi0-1 9X10',

7xI0'2 7x10-3 5 x 10-1 x10ý 6x 10 6xI0-1 5X10-1 5X10-4 1Ix 10 1 xlO-1 JX 10-2 2X10-3 2x10-1 5x 10-'

6XlO'1 IX 10-'

IX10-t 4 x 10 3 XIO0' 2 x102 2XIO-3 1X 10-'

3X10-1 2x10-1 4x10'1 lx 10-2 X 10-:2 2x 10-:

82X10-11 Ix 10-2x10' 31IO01 4X10'1 2X 109 3 x109 Sx 10-1 4x10-9

.210xl6 1 X 10" U 233...-...............

S U 234...................

U 236...........

U 238...........

U 240..................

U-natural..............

S 4

S S4 S.

I S4 I

va adum (3)

IV 48...........

Xe 131m..............

Xe 133....................

Xe 133m....

Xe 135...............

Yb 175...............

Ytterbium (70)............

S 4

S Sub Sub Sub Sub S

S S

S SS I 10-1 4x10-12 SX10'4 3X10-1t 8X10'4 9x10-1 9XIO04 2x10-11 9X10-4 2X10-0 '

9X10-4 4X10-1 SXlO-4 2x10-11 SX10-4 4X10-12 I X10-3 2x10-11 lX10-3 4XIO12 1 X10-3 3X10-1x10-3 SxIO-12 1Xl0-3 SxIOtt 9X10'3 6X10-9 1X1-3xIO17 8X10-4 2XIO-1 axO-x10' 0

x10'1 6x1............

3X10-1

ixia, 2x1O0" 3x10-1 2x10-6 6xa 104 1X10-'

6XIO-4 31x0-6X10-1 1x10-1 2x10-1 l1XO-8x10",

6X10',

8XI0-1 SX10-1 2x10- 1 2X10-'

7X10-6 1 x10-=

2X10-3x10-$

2x10-"

2x10-4 2X10'-

4X10-2 2X10-2X10-1 2X10-5 2X10 2-5x1O-,

5X10" 5x10-4 5X10-4 9X 10-2X10-1 2X10-1 2X10-'

4X10-4 3x10-1 1X10-4 1 x10-4 17xlO-'

7x10-1 5X10-'

6xIO-3 5X10-'

3X10-'

3x10-'

3X10-5 3x10-5 3Xi0-3 3X10-6 3x10"5.

4x10-5 3xI0-3.

4X10-3 3x10-3 3x10-5 3xIO-5 3xlO-S 3x10-5 1X1O-1 1XI0-1 2x10-s 2x10-5 3x10-1 3X10-1 3X1o-s 6xi0-1 6x10-1 3x10-1 3X 10-Yttrium (39).......................

I Y g0................

Y 91m..............

Y 91.......................

Y 92................

Y 93 --

Revision I I B -9

Pt. 20 [§§ 20.1-20.6021, App. B 10 CFR Ch. I (1-1-93 Edition)

APPENDIX B TO §§ 20.1-20.602-CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND-Continued (See footnotes at end of Appendix B]

Isotope Table I Table II Element (atomic number) 1 Co 1Air Col 2-CoL

-Air C0l. 2-C I-Air Water Water (0xfml jCi/ml)

(/JLCi/ffl)

(VuCi/ml)

Zinc (30).........................................

Zn65

.......... S X1I0-3X10-3 4X10-lx 10X-I 6x10-4 5x10-'

2x10-9 2x10-'

Zn69m...................

S 4X10-7 2X10-3 1X10-7x10-3X10- 7 2x10-'

tX10-6x10-1 Zn69..........

S 7X10"'

5X10-2 2x10-1 2x10-1 I

9X10-'

5X10-3x 10-2X10-3 Zirconium (40).................Zr93..........

S 1X10-'

2X10'-

4xi0-8x10-'

I 3X10-7 2X10-7 1X10"0 8XI0-Zr 95 S

1X10-1 2x10-1 4XIO-9 6x10-I 3x10-6 2x10-3 1x10-9 6x10-s Zr 97..................

S 1X10-7 5x10-1 4x10-1 2x10-1 I

9X10-'

5x10-'

3x10-1 2x10-s Any single radionuclide not listed above...............................

Sub IX107.......

3x10-1...............

with decay mode cther than alpha emission or spontaneous fission and with radioactive half-ife less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Arny single radionuclide not lsted above 3X10-9 9X10.5 1X10-11 3X10-1 with decay mode other than alpha emission or spontaneous fission and with radioactive half-life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Any single radionuclide not listed.............................

6x10-"3 4x10-7 2X104 3x10-6 above. which decays by alpha emis-sion or spontaneous fission.

'Soluble (S); Insoluble (I).

Sub" means that values given are for submersion in a semispherical infinite cloud of airbome material-MTese radon concentrations are appropriate for protection from radon-222 combined with its shodrived daughters.

A.temtively. the value in Table I may be repiaced by one-third (%) "woddng level." (A "working level" is defined as any combination of short-lived radon-222 daughters. polonium-218, lead-214, bismuth-214 and polonium-214. in one rfler of air.

without regard to the degree of equilibrium, that will result in the ultimate emission of 1.3X 10 3 MeV of alpha particle energy.)

The Table l1 value may be replaced by one-thirtieth (Vzo) of a "working. leveL" The limit on radon-222 concentrations in restricted areas may be based on an annual average.

'For soluble mixtures of U-238. U-234 and U-235 in air chemical toxicity may be the limiting factor. If the percent by weight-enrichment) of U-235 is less than 5. the concentration value for a 40-hour workweek, Table I, is 0.2 milligrams uranium per cubic meter of air average. For any enrichment, the product of the average concentration and time of exposure durng a 40-hour workweek shall not exceed 8xi0-o SA j.Ci-hr/mI. where SA is the specific activity of the uranium inhaled. The concentration value for Table II is 0.007 milligrams uranium per cubic meter of aki. The specificactivity for natural uranium is 6,77x10-2 curies per gram U. The specific activity for other mixtures of U-238, U-235 and U-234. if not known, shall be:

SA=36x 10- curies/gram U U-depleted SA=(0.4+0.38 E+0.0034 E ) 10-'

E>-0.72 where E is the percentage by weight of U-235. expressed as percent NOTE: In any case where there is a mixture in air or water of more than one radionuclide. the limiting values for purposes of this Appendix should be determined as follows:

1. If the identity and concentration of each radionuclide in the mixture are known, the limiting values should be derived as follows: Determine. for each radionuclide in the mixture, the. ratio between the quantity present in the mixture and the limit otherwise established in Appendix B for the specifid radionudide when not in a mixture. The sum of such ratios for all the radionuclides in the mixture may not exceed -I" (Le.. "inity")

EXAMPLE: If radionuctides A, 8, and C are present in concentrations CA, C8, and Cc, and if the appticabte MPC's, are MPG,.

and MPCa, and MPCc respectively, then the concentrations shall be limited so that the following relationship exists:

(CA/MPC,.)+(CV/MPC.) +(C¢/MPCJ) _'

.2. If either the identity or the concentration of any radionuclide in the mixture is not known, the limiting values for purposes of Appendix B shall be:

a. For purposes of Table I. Col. 1 -6x 10- 3 b-For purposes of Table I, Col. 2-A x 10- 7

c. For purposes of Table II. Col. 1-2 x 10- 14

d. For purposes of Table II. Col. 2-3 x 10- 1

3. If any of the conditions specified below are met, the corresponding values specified below may be used in lieu of those specified in paragraph 2 above.

a-If the identity of each radionuclide in the mixture is known but the concentration of one or more of the radionuclides in the mixture is not known the concentration limit for the mixture is the limit specified in Appendix "B" for the radionuclide in the mixture having the lowest concentration limit: or

b. If the identity of each radionuclide in the mixture is not known, but it is known that certain radionuclides specified in Appendix "6" are not present in the mixture, the concentration limit for the mixture is the lowest concentration limit specified in Appendix "8" for any radionuclide which is not known to be absent from the mixture. or Revision I I B - 10

Nuclear Regulatory Commission Pt. 20 [§§ 20.1-20.6021, App. C Table I Table It

c. Element (atomic number) and isotope Col C-Air Col. 2--

Cot. I--

Col. 2--

Water Air (0Ci/

Water ml)

(1.Cilml) ml)

(jaCa/ml)

If it is known that Sr 90.

125. I 126. I 129.1 131 (I 133, Table II only). Pb 210. Po 210. At 211. Ra 223. Ra 224. Ra 226. Ac 227. Ra 228. Th 230. Pa 231. Th 232. Th-nat, Cm 248, C_ 254. and Fm 256 are not present..................................

9x10-3x10-'

If it is known that Sr 90. 1 125. I 126. 1 129 (1 131, 1 133, Table I1 only). Pb 210. Po 210. Ra 223, Ra 226. Ra 228. Pa 231. Th-nat. Cm 248. Cf 254.

and Fm 256 are not present.................................................................................

6X 10 2x 10-If it is known that Sr 90.1 129 ( 125. t 126.1 131. Table II only). Pb 210. Ra 226. Ra 228, Cm 248. and Cf 254 are not present....................

2 6x 10-If it is known that (1 129. Table It only). Ra 226. and Ra 228 are not present..................

3-10-'

x 10-1 If it is known that alpha-emitters and Sr 90.1 129. Pb 210. Ac 227. Ra 228.

Pa 230. Pu 241. and Bk 249 are not present.........

3X10'...........1X10..................

If it is known that alpha-emitters and Pb 210. Ac 227. Ra 228. and Pu 241 are not presenL.....................

3x10 0

-I 1x10-............

If it is known that alpha-emitters and Ac 227 are not present 3x10-,

xl 10........-..........

If it is known that Ac 227. Th 230. Pa 231. Pu 238, Pu 239. Pu 240. Pu 242.

Pu 244. Cm 248. Cf 249 and Cf 251 are not presente.._t._

_._.]

3x10-9'...........

1x10-.

4. It a mixture of radionuclides consists of uranium and its daughters in ore dust prior to chemical separation of the uranium from the ore. the values specified below may be used for uranium and its daughters through radium-226. instead of those from paragraphs 1. 2, or 3 above.

a. For purposes of Table I, Col. 1 -1x10-o jLCi/m gross alpha activity;, or 5 x10-

/Ci/ml natural uranium or 75 micrograms per cubic meter of air natural uranium.-

b. For xuposes of Table II. Col. 1-3 x 10-1fCi/ml gross alpha activity;, 2x 10- -

aCi/ml natural uranium. or 3 micrograms per cubic meter of air natural uranium.

5. For purposes of this note. a radionuclide may be considered as not present in a mixture if (a) the ratio of the concentration of that radionuclide in the mixture (CA) to the concentration lm for that radionuclide specified in Table 11 of Appendix -B" (MPCF, does not exceed Vo. (i.e. CIMPA<-_I1O) and (b) the sum of such ratios for all the radionuclides considered as not present in the mixture does not exceed % i.e.

(CA/.PCA + C/MPC.... +(

Y

'/4).

Revision I I B - I I

YANKEE NUCLEAR POWER STATION INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

OFF-SITE DOSE CALCULATION MANUAL TABLE OF CONTENTS I 1.0 IN T R O D U C T IO N 3

1.1 Definitions I 2.0 RADIOLOGICAL ENVIRONMENTAL MONITORING..................................................

7 2.1 Monitoring Program 2.2 Environmental Monitoring Locations 2.3 Dose/Dose Rate Controls and Calculations 3.0 REPORTING REQ UIREM ENTS...............................................................................

13 3.1 Annual Radiological Environmental Operating Report 3.2 Annual Radioactive Effluent Release Report I 4.0 R E FE R E N C E S.......................................................................................................

.. 14 Revision 20 1

LIST OF FIGURES Fiqure Title Paqe 1.1a YNPS Site Boundary Lines 5

1.1b Current 10CFR50 Licensed Site Boundaries 6

2.1 Radiological Environmental Monitoring Locations On-Site 10 (Direct Radiation Pathway) 2.2 Radiological Environmental Monitoring Locations within 1 Mile 11 (Direct Radiation Pathway) 2.3 Radiological Environmental Monitoring Locations 12 (Direct Radiation Pathway)

Revision 20 2

1.0 INTRODUCTION

The purpose of this document is to provide a method for demonstrating compliance with the dose limits for MEMBERS OF THE PUBLIC and contains the guidance for submittal of the annual reports required by 10 CFR Part 50. In addition, the document provides the locations and type of monitoring required for the Radiological Environmental Monitoring Program (REMP).

In accordance with the requirements of 40CFR Part 190, the dose to a MEMBER OF THE PUBLIC for radioactive material in effluents and direct radiation from an Independent Spent Fuel Storage Installation (ISFSI) is limited to 25 mrem/yr to the whole body, 75 mrem/yr to the thyroid and 25 mrem/yr to any other critical organ as a result of exposure to planned discharges of radioactive materials to the environment, direct radiation from the ISFSIand any other radiation from uranium fuel cycle operations within the region.

Under normal operations, experience has shown that the ISFSI will be operated at a small fraction of the above dose limits. This is primarily due to the design of the Independent Spent Fuel Storage Installation, which prevents the release of radioactive materials in liquid and particulate form and there are no other uranium fuel cycle operations within 5 miles of the YAEC site. Therefore, the dose equations from 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 50, Appendix I are not necessary for inclusion in the ODCM. The remaining dose component to be considered is from direct radiation. 40 CFR 190 establishes this dose limit as 25 mrem/yr for MEMBERS OF THE PUBLIC.

Figure 1 - la shows the site boundary lines for the site, and Figure 1-1b shows the current 1 OCFR Part 50 licensed site boundary.

1.1 DEFINITIONS Member(s) of the Public MEMBER(S) OF THE PUBLIC (for the purposes of 10CFR50, Appendix I) shall include all persons who are not occupationally associated with the site.

This category does not include employees of the utility, its contractors, or vendors.

Also excluded from this category, are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the site operations or decommissioning of the plant.

Offsite Dose Calculation Manual (ODCM)

The ODCM contains the methodology and parameters used in the calculation of off-site doses in the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by the Yankee Quality Assurance Program (QAP) and (2)

Revision 20 3

descriptions of the information that should be included in the Annual Radiological Environmental Operation and Annual Radioactive Effluent Release Reports.

Site Boundary The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. Realistic occupancy factors shall be applied at these locations for the purposes of dose calculations.

Revision 20 4

Revision 20 5

4, Figure 1-1 b Current 10 CFR Part 50 Licensed Site Boundarj

.i I.

.4

*1~~

.1 C. IA A~

.$(t

~C C.-,

I Revision 20 6

2.0 RADIOLOGICAL ENVIRONMENTAL MONITORING The Radiological Environmental Monitoring Program (REMP) for the ISFSI monitors for direct radiation exposure only.

There are no radioactive gaseous or liquid effluent pathways associated with the ISFSI.

2.1 Monitoring Program The type and number of radiological environmental monitoring stations including collection and analysis frequencies are shown below.

Number of Exposure Pathway Locations Collection Frequency Type and Frequency of Analysis DIRECT RADIATION 7

Semi-annual Gamma dose, at least once per six months 2.2 Environmental Monitoring Locations The radiological environmental monitoring stations are listed below. The locations of these stations with respect to the Yankee ISFSI are shown on the maps, Figures 2.1, 2.2

& 2.3.

Monitoring Location and Distance From Direction From Exposure Pathway Designated Code the ISFSI (km)

ISFSI DIRECT RADIATION GM-15 On-Site Perimeter 0.24 NW GM-16 On-Site Perimeter 0.22 NNW GM-17 On-Site Perimeter 0.13 NNE GM-21 On-Site Perimeter 0.17 WSW GM-2 Observation Stand 0.50 NW GM-6 Readsboro Road Barrier 1.30 N

GM-27*

Number Nine Road 7.60 ENE

Designated control location. Two TLD's at this sample location.

Revision 20 7

2.3 Dose/Dose Rate Controls and Calculations By design, there are no liquid or gaseous effluents associated with the operation of the ISFSI. With the completion of site remediation activities that required period dewatering of construction excavations, along with the removal of all systems or operations that generated, contained or processed waste gas or airborne particulates, there are no longer any gaseous or liquid effluent releases from site operations. Therefore, requirements for control, sampling, analyzing, monitoring or dose impact assessment for radioactive liquids or gases are not needed.

2.3.1 Total Dose Control 2.3.1 In accordance with Yankee Quality Assurance Program (QAP), the dose or dose commitment to any real MEMBER OF THE PUBLIC from all site sources is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem) over a calendar year.

Applicability At all times.

ACTION With the calculated or projected dose from direct radiation contributions from the Independent Spent Fuel Storage Installation (ISFSI) determined to be, or projected to be, above the annual (calendar) limits of Control 2.3.1, prepare and submit to the commission with 30 days, pursuant to 10CFR50.4, a Special Report that defines the corrective action to be taken to reduce subsequent exceedences to prevent recurrence of exceeding the above limits and include the schedule for achieving conformance with the above limits. The Special Report shall include an analysis that estimates the radiation exposure (dose) to a member of the public from site sources for the calendar year covered by the report. It also shall describe levels of radiation and concentrations of radioactive material, if any, involved and the cause of the exposure levels or concentrations.

If the estimated dose(s) exceeds the above limits, and if the exposure condition resulting in violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR1 90. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

Revision 20 8

SURVEILLANCE REQUIREMENTS SR 2.3.1 Dose calculations - Cumulative dose contributions from direct radiation shall be determined semi-annually in accordance with Section 2.3.2 of the ODCM.

Bases Control 2.3.1 is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46FR18525. The control requires the preparation and submittal of a Special Report whenever the calculated or projected doses from the site exceed the dose limits of 40CFR Part 190. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to a MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFRPart 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20.

2.3.2 Method to Calculate Direct Dose from ISFSI Operations Control 2.3.1 restricts the dose to the whole body and any organ of any real MEMBERS OF THE PUBLIC at and beyond the Site Boundary from all site sources (including direct radiation) to the limit of 25 mrem in a year, except for the thyroid which is limited to 75 mrem in a year.

Estimates of direct exposure above background in areas at and beyond the site boundary can be determined from measurements made by environmental TLDs that are part of the Environmental Monitoring Program (sections 2.1 and 2.2). A net response is determined by subtracting the average TLD value of the control stations from the semi-annual off-site TLD measurements. A positive net exposure is assumed if the net value is greater than the propogated uncertainty of the TLD indicator and control measurements. Alternatively, direct dose calculations from identified fixed sources on-site can be used to estimate the off-site direct dose contribution where TLD information may not be applicable.

Revision 20 9

Figure 2.1 Radiological Environmental Monitoring Locations On-Site (Direct Radiation Pathway)

Revision 20 10

Figure 2.2 Radiological Environmental Monitoring Locations within 1 mile (Direct Radiation Pathway)

Revision 20 11

NW A-22 WNW an Reservo ENE IN F1G,.RE 4.K

...... L

'N..-----

V erm..' t

-r'---nPoe-

-f27-----

-~YN~ A-I-assachusetts W

E Bea.- Swamp...........

9-.._

"".Lower Reseroir.

"":/1.ES E...

d avik

,"ai

ý "SE jsw SS SS "a

2 4

Figure 2.3 Radiological Environmental Monitoring Locations (Direct Radiation Pathway)

Revision 20 12

3.0 REPORTING REQUIREMENTS 3.1 Annual Radiological Environmental Operating Report

a.

An Annual Radiological Environmental Operating Report covering the operation of the site during the previous calendar year shall be submitted to the NRC by May 1 of each year.

b.

The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the site operation on the environment.

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental monitoring during the report period pursuant to the table and figures in the ODCM. In the event that some results are not available to include in the report, the report shall be submitted noting and explaining the reasons for the missing results: The missing data shall be submitted as soon as possible in a supplementary report.

The report also shall include the following: a summary description of the Radiological Environmental Monitoring Program with a map of all monitoring locations keyed to a table giving distances and directions from the ISFSI.

3.2 Annual Radioactive Effluent Release Report

a.

By May 1 of each year, a report shall be submitted to the NRC covering the radioactive content of effluents released to unrestricted areas during the previous calendar year.

b.

The Annual Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, Revision 1, June 1974, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the format of Appendix B thereof.

In addition, the Annual Radioactive Effluent Release Report shall include an assessment of the radiation doses due to the radioactive effluents released from the site during 2006.

This report also shall include an assessment of the radiation doses from radioactive effluents to MEMBER(S) OF THE PUBLIC due to the allowed recreational activities Revision 20 13

inside the SITE BOUNDARY during 2006. All assumptions used in making these assessments (e.g., specific activity, exposure time, and location) shall be included in the report. Historical average meteorological conditions shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the ODCM.

The Annual Radioactive Effluent Release Report also shall include an assessment of radiation doses to the likely most exposed real MEMBER(S) OF THE PUBLIC from site releases (including doses from primary effluent pathways and direct radiation) for 2006 to show conformance with 40CFR190, "Environmental Radiation Protection Standards for Nuclear Power Operation," if Control 2.3.1 has been exceeded during the calendar year.

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to site boundary of radioactive materials in effluents made during the reporting period.

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the ODCM.

4.0 REFERENCES

a.

Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," U.S. Nuclear Regulatory Commission, Revision 1, October 1977.

b.

Yankee Atomic Electric Company Supplemental Information for the Purposes of Evaluation of 10CFR Part 50, Appendix I, Amendment 2, October 1976 (Transmitted by J. L. French - YAEC to USNRC in letters, dated June 2, 1976; August 31, 1976; and October 8, 1976).

c.

Yankee Quality Assurance Program (QAP), Yankee Atomic Electric Company.

d.

Issuance of NPDES Permit No. MA0004367; Letter to J. A. Kay from R. Janson, US EPA, dated July 29, 2003 Revision 20 14

ML070430316

Text

Reference:

Subject:

Enclosure:

1.0 INTRODUCTION

3.0 REFERENCES

0.0 Container

Response

Response

Response

Response

Response

1.0 INTRODUCTION

8.0 REFERENCES

Dear Mr. Papanic:

SUBJECT:

1.0 INTRODUCTION

SUMMARY

References:

Subject:

Dear Sirs:

1.0 INTRODUCTION

SUMMARY

7.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

7.0 REFERENCES

1.0 INTRODUCTION

4.0 REFERENCES

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