RA-05-021, Co. 2004 Radiological Reports

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Co. 2004 Radiological Reports
ML051230412
Person / Time
Site: Maine Yankee
Issue date: 04/27/2005
From: Meisner M
Maine Yankee Atomic Power Co
To:
Document Control Desk, NRC/FSME
References
MN-05-018, RA-05-021
Download: ML051230412 (42)
v • d • e


Text

Maine Yankee 321 OLD FERRY RD. - WISCASSET, ME 045784922 April 27, 2005 MN-05-018 RA-05-021 UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555

References:

(a) License No. DPR-36 (Docket No. 50-309)

(b) Maine Yankee Quality Assurance Program (c) Maine Yankee Off-site Dose Calculation Manual (ODCM)

Subject:

2004 Radiological Reports Gentlemen:

Enclosed as indicated below are radiological reports for 2004 submitted in accordance with the relevant portions of References (b) and (c).

l Report Name j l

MY QAP

~Reference (Ref. b) ll ODCM (Ref. c)

Annual Radioactive Effluent Release Appendix D, IV.B Appendix C, Item 2 Report Appendix E, IV.C Estimated Dose Report For 2003 Not Applicable Appendix C, Item 3 Changes to the ODCM Appendix D, III.C Not Applicable Appendix E, IV.B I We trust we have completed submission requirements for these reports. Should you have questions or comments, please contact John Niles, ISFSI Manager at 207-882-1300.

Sir J. Meisner esident, Chief Nuclear Officer ADO'?~42 Enclosures C: Dr. R. R. Bellamy, NRC Region I Mr. J. T. Buckley, NRC NMSS Project Manager, Decommissioning Mr. P. J. Dostie, State of Maine, Division of Health Engineering Mr. S. J. Collins, NRC Regional Administrator, Region I Mr. W. C. Toppan, Dir., Maine Division of Health Engineering

MAINE YANKEE ATOMIC POWER COMPANY ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January-December 2004

1.0 INTRODUCTION

Tables 1 and 2 summarizes the quantity of radioactive gaseous and liquid effluents, respectively, for each quarter of 2004. Table 3 states that waste was shipped off-site for burial or disposal during the year 2004. Table 4 contains supplementary information.

Appendices A through D, indicate the status of reportable items per the requirements of the Off-site Dose Calculation Manual (ODCM) sections 2.1.5, 2.2.6, 2.3.3, 2.3.4, 2.5 and Appendix C.

Changes to the ODCM made during the year 2004 are summarized in Appendix E.A complete copy of the revised manual is attached as well as the specific pages that changed.

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TABLE 1A Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report First and Second Quarters, 2004 Gaseous Effluents-Summation of All Releases

! ' 'unit. 1:- 1 t 1 -- Est. Total

__; __ _- __ _, _--:- _- _.---l: _ Quarter Quarter Elr,% PW A Fission -and Activation Gases  :- _______ _;--:._.___:

1. Total Release Ci 4.676 N/D 2.50E+1
2. Average release rate for uCi/sec **9.17E-1 NID period
3. Percent of regulatory limit  % **1.6E-2 N/D B. Iodines- - ,, -- -. , ,,;,
1. Total lodine-131 Ci N/D* N/D* 2.50E+1
2. Average release rate for uCi/sec N/D* N/D*

period

3. Percent of regulatory limit _ __ N/D* N/D*

C. Particulates <->_

1. Particulates with T-1/2 > 8 Ci 4.7 E-8 N/D* 3.50E+1 days
2. Average release rate for uCi/sec **5.22E-5 N/D*

period

3. Percent of regulatory limit _ **4.57E-4 N/D*
4. Gross alpha radioactivity Ci N/D* N/D*

D .- T ritium .-  ; -- - ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1. Total release Ci 6.32E-2 2.24E-2 2.50E+1
2. Average release rate for uCi/sec 8.04E-3 2.85E-3 period
3. Percent of regulatory limit  % 1.41 E-4 4.99E-5 .

N/D*= Not Detected N/A*= Not Applicable

    • = Calculated for the time period in which the releases actually occurred. All others are calculated for the standard 91 day period (particulate release is an unscheduled release, which is reported in Table 4) 2

TABLE 1A Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2004 Gaseous Effluents-Summation of All Releases

;3 Est. Total

-;-------Qute:Quarterj :Error, %

A.-Fision and Activation RGases J2.50E+1

_i

1. Total Release Ci N/D N/D
2. Average release rate for uCi/sec N/D N/D period
3. Percent of regulatory limit  %/ N/D N/D B. lod in e s _ _ _ _ _ - : _ _ _ - _ _ _  ;--,

-I-------

1. Total lodine-131 Ci N/D* N/D* 2.50E+1
2. Average release rate for uCi/sec N/D* N/D*

period

3. Percent of regulatory limit N/D* N/D*

-C. Particulates  ; - -- . -- - ; - - --

-;-tm-- - - ---.

1. Particulates with T-1/2 > 8 Ci N/D* N/D* 3.50E+1 days
2. Average release rate for uCi/sec N/D* N/D*

period

3. Percent of regulatory limit _ N/D* N/D*
4. Gross alpha radioactivity Ci N/D* N/D*

D. Tritium -: -;; -___-- -_--- _ ---

1. Total release Ci N/D N/D 2.50E+1
2. Average release rate for uCi/sec N/D N/D period -
3. Percent of regulatory limit  % N/D N/D I N/D*= Not Detected N/A*= Not Applicable 3

TABLE 1B Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report First and Second Quarters, 2004 Gaseous Effluents-Elevated Release Continuous Mode Batch Mode Nuclides Released Unit [ St 2na St 2na Quarter Quarter Quarter Quarter I Fission Gases_ E.+.

Krypton-85 Ci 4.676 N/D NIA* N/A*

Krypton-85m Ci N/D* N/D* N/A* N/A*

Krypton-87 Ci N/D* N/D* N/A* N/A*

Krypton-88 Ci N/D* N/D* N/A* N/A*

Xenon-1 33 Ci N/D* N/D* N/A* N/A*

Xenon-1 35 Ci N/D* N/D* N/A* N/A*

Xenon-1 35m Ci N/D* N/D* N/A* N/A*

Xenon-1 38 Ci N/D* N/D* N/A* N/A*

Unidentified Ci N/D* N/D* N/A* N/A*

Total for period Ci N/D* N/D* N/A* N/A*

UT-1107di Mtios __l II _ _

^N ___

4# A __

lodine-1 31 Ci NIA* N/A* N/A* N/A*

Iodine-133 Ci N/A* N/A* N/A* N/A*

Iodine-135 Ci N/A* N/A* N/A* N/A*

Total for period Ci N/A* N/A* N/A* N/A*

. Particula teHsam Strontium-89 Ci N/D N/D N/A* N/A*

Strontium-90 Ci N/D N/D N/A* N/A*

Cesium-134 Ci N/D N/D N/A* N/A*

Cesium-137 Ci 4.7E-8 N/D N/A* N/A*

Cobalt-60 Ci N/D N/D N/A N/A Barium-Lanthanum-140 Ci N/D N/D NIA* N/A*

Others- - _ _ _

Plutonium-238 Ci N/D N/D N/A* N/A*

Curium-243,244 Ci N/D N/D N/A* N/A*

Uranium-234 Ci NID N/D N/A* N/A*

Uranium-238 Ci N/D NID N/A* NIA*

Thorium-232 Ci N/D N/D N/A* NIA*

Radium-226 Ci N/D N/D N/A* N/A*

N/D*= Not Detected N/A*= Not Applicable 4

TABLE 1B Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2004 Gaseous Effluents-Elevated Release Continuous Mode Batch Mode Nuclides Released l Unit [ 3rd 4th 1 3rd 4th Quarter Quarter Quarter Quarter 2n-5

_14yt Co NID _ A

--. - /A*

Krypton-85 Ci N/D N/D N/A* N/A*

Krypton-85m Ci N/D* N/D* N/A* N/A*

Krypton-87 Ci N/D* N/D* N/A* N/A*

Krypton-88 Ci N/D* N/D* N/A* N/A*

Xenon-1 33 Ci N/D* N/D* N/A* N/A*

Xenon-1 35 Ci N/D* N/D* N/A* N/A*

Xenon-135m Ci N/D* N/D* N/A* N/A*

Xenon-1 38 Ci N/D* N/D* N/A* N/A*

Unidentified Ci N/D* N/D* N/A* N/A*

Total for period Ci N/D* N/D* N/A* N/A*

j2. i.._es.

lodine-131 Ci N/A* N/A* N/A* N/A*

Iodine-133 Ci N/A* N/A* N/A* N/A*

lodine-1 35 Ci N/A* N/A* N/A* N/A*

Total for period Ci N/A* N/A* N/A* N/A*

3. At.IIEt! s Strontium-89 Ci N/D N/D N/A* N/A*

Strontium-90 Ci N/D N/D N/A* N/A*

Cesium-134 Ci N/D N/D N/A* N/A*

Cesium-1 37 Ci N/D N/D N/A* N/A*

Cobalt-60 Ci N/D N/D N/A N/A Barium-Lanthanum-140 Ci N/D N/D N/A* N/A*

Others-Plutonium-238 Ci N/D N/D N/A* N/A*

Curium-243,244 Ci N/D N/D N/A* N/A*

Uranium-234 Ci N/D N/D N/A* N/A*

Uranium-238 Ci N/D N/D N/A* N/A*

Thorium-232 Ci N/D N/D N/A* N/A*

Radium-226 Ci N/D N/D N/A* N/A*

N/D*= Not Detected N/A*= Not Applicable 5

TABLE 1C Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report January-December 2004 Gaseous Effluents-Ground Level Release Maine Yankee processed 12 Transportable Storage Canisters in the year 2004, which contained spent fuel element assemblies. The processing of these canisters resulted in 4.676 Curies of Krypton-85 released to the environment for the entire year.

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TABLE 2A Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report First and Second Quarters, 2004 Liquid Effluents-Summation of All Releases

-Unit-Est. Total

- -:; . ->; : - -t . - l S - .u e -Quarter- Erro°--

A. Fission and Activation Products

1. Total Release (not including Ci 9.12E-4 1.70E-2 1.50E+1 tritium, gases,alpha) .
2. Average diluted concentration .uCi/ml 3 20E-7 1.35E-6 during period .
3. Percent of applicable limit  % 1.24E-1 7.84E-1 B. Tritium
1. Total Release Ci 9.38E-1 1.95E+00 1.50E+1
2. Average diluted concentration .uCi/ml 2.14E-4 8.18E-5 during period
3. Percent of applicable limit  % 3.27E-1 1.89E-1 C. Dissolved and Entrained Gases
1. Total Release Ci N/D* N/D* 1.50 E+1
2. Average diluted concentration .uCi/ml N/A* N/A*

during period

3. Percent of applicable limit  % N/A* N/A*

D. Gross Alpha Radioactivity

1. Total release Ci N/D N/D 1.50E+1
2. Average diluted concentration .uCi/ml N/A* N/A during period E. Volume of Waste Released (prior Liters 1.11 E+6 2.73E+6 to dilution) 1.OE+1 F. Volume of Dilution Water Used Liters 1.76E+6 7.57E+6 During Period 1.OE+1 N/D*= Not Detected N/A*= Not Applicable

TABLE 2A Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2004 Liquid Effluents-Summation of All Releases

---.Unit 3rd - 4th -Est. Total

-_-_-;__IQu arter] Quarter Error %

A. Fission and Activation Products

1. Total Release (not including Ci 3.46E-2 3.14E-5 1.50 E+1 tritium, gases,alpha)
2. Average diluted concentration .uCi/ml 2.1OE-6 1.16E-7 during period
3. Percent of applicable limit /0 2.57E+00 8.56E-3 B. Tritium
1. Total Release Ci 1.17E+00 3.31E-3 1.50 E+1
2. Average diluted concentration .uCi/ml 6.61 E-5 1.33E-6 during period
3. Percent of applicable limit  % 2.27E-1 1.37E-3 C. Dissolved and Entrained Gases
1. Total Release Ci N/D* l N/D* l 1.50 E+1
2. Average diluted concentration .uCi/ml N/A* N/A*

during period

3. Percent of applicable limit _ N/A* N/A*

D. Gross Alpha Radioactivity

1. Total release Ci N/D N/D 1.50 E+1
2. Average diluted concentration .uCi/ml N/A N/A during period E. Volume of Waste Released (prior Liters 1.44E+6 2.41 E+6 1.OE+1 to dilution)

F. Volume of Dilution Water Used Liters 1.OE+l During Period 3.71 E+6 0 N/D*= Not Detected N/A*= Not Applicable 8

TABLE 2B Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report First and Second Quarters, 2004 Liquid Effluents Continuous Mode Batch Mode Nuclides Released Unit OlSt 2na St 2nd Quarter Quarter Quarter Quarter Strontium-89 Ci N/A* N/A* N/D* N/D*

Strontium-90 Ci N/A* N/A* N/D* 1.68E-3 Cesium-1 34 Ci N/A* N/A* N/D* N/D*

Cesium-137 Ci N/A* N/A* 7.64E-5 4.63E-4 lodine-131 Ci N/A* N/A* N/D* N/D*

Cobalt-58 Ci N/A* N/A* N/D* N/D*

Cobalt-60 Ci N/A* N/A* 8.32E-4 1.27E-2 Iron-59 Ci N/A* N/A* N/D* N/D*

Zinc-65 Ci N/A* N/A* N/D* N/D*

Manganese-54 Ci N/A* N/A* N/D* N/D*

Chromium-51 Ci N/A* N/A* N/D* N/D*

Zirconium-Niobium-95 Ci N/A* N/A* N/D* N/D*

Molybdenum-99 Ci N/A* N/A* N/D* N/D*

Technetium-99m Ci N/A* N/A* N/D* N/D*

Barium-Lathanium-140 Ci N/A* N/A* N/D* N/D*

Cerium-141 Ci N/A* N/A* N/D* N/D*

Others- Iron-55 Ci N/A* N/A* N/D 2.16E-3 Antimony-1 25 Ci N/A* N/A* 4.20E-6 6.84E-6 Unidentified Ci N/A* N/A* N/D* N/D*

Total for period (above) Ci N/A* N/A* 9.12E-4 1.70E-2 Xenon-1 33 Ci N/A* N/A* N/D* N/D*

Xenon-1 35 Ci N/A* N/A* N/D* N/D*

N/D*= Not Detected N/A*= Not Applicable 9

TABLE 2B Maine Yankee Atomic Power Station Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2004 Liquid Effluents Continuous Mode Batch Mode Nuclides Released Unit 3rd 4th 3rd 4th Quarter Quarter Quarter Quarter Strontium-89 Ci N/A* N/A* N/D* N/D*

Strontium-90 Ci N/A* N/A* 2.48E-4 N/D*

Cesium-134 Ci N/A* N/A* N/D* N/D*

Cesium-137 Ci N/A* N/A* 2.82E-3 1.73E-5 lodine-131 Ci N/A* N/A* N/D* N/D*

Cobalt-58 Ci N/A* N/A* N/D* N/D*

Cobalt-60 Ci N/A* N/A* 3.12E-2 9.52E-6 Iron-59 Ci N/A* N/A* N/D* N/D*

Zinc-65 Ci N/A* N/A* N/D* N/D*

Manganese-54 Ci N/A* N/A* N/D* N/D*

Chromium-51 Ci N/A* N/A* N/D* N/D*

Zirconium-Niobium-95 Ci N/A* N/A* N/D* N/D*

Molybdenum-99 Ci N/A* N/A* N/D* N/D*

Technetium-99m Ci N/A* N/A* N/D* N/D*

Barium-Lathanium-140 Ci N/A* N/A* N/D* N/D*

Cerium-141 Ci N/A* N/A* N/D* N/D*

Others- Iron-55 Ci N/A* N/A* 3.92E-4 N/D*

Antimony-1 25 Ci N/A* N/A* N/D* 4.56E-6 Unidentified Ci N/A* N/A* N/D* N/D*

Total for period (above) Ci N/A* N/A* 3.46E-2 3.14E-5 Xenon-133 Ci N/A* N/A* N/D* N/D*

Xenon-135 Ci N/A* N/A* N/D* N/D*

N/D*= Not Detected N/A*= Not Applicable 10

TABLE 3 Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report First Half, 2004 Solid Waste and Irradiated Fuel Shipments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel).

1. Type of Waste. Unit 6-Month Est. Total Error, Period
a. Spent resins, filter Cu. M. 13.8 sludges, etc. Ci. 7.85 E+01 +/- 25
b. Dry compressible waste, Cu. M. 19400 contaminated equipment, DAW, Ci. 1.84 E+01 +/- 25 cement.
c. Irradiated Hardware. Cu. M. 78.7 Ci. 3.33 E-01 +1-25
2. Estimate of major nuclide composition (by type of waste).
a. Co-60 2.39% 1.87E+00 Ni-63 78.27% 6.14E+01 Cs-1 37 17.15% 1.35E+01 Fe-55 1.78 1.40E+00
b. Co-60 24.25% 4.45E+00 Fe-55 6.34% 1.16E+00 Ni-63 41.65% 7.65E+00 Cs-1 37 24.60% 4.52E+00 Ce-144 1.21% 2.22E-01 Pu-241 0.31 5.69E-2 C. Co-60 10.23% 3.41 E-02 Fe-55 87.54% 2.92E-01 Ni-63 1.83% 6.1OE-3
3. Solid Waste Disposition Number of Shipments Mode of Destination Transportation 2 Trucking over Chem-Nuclear, Barnwell, highway S.C.

6 Trucking over Envirocare of Utah highway Clive, Utah 246 Rail Envirocare of Utah 11

Table 3 (Cont.)

B. Irradiated Fuel Shipments (Disposition): None Shipped.

Additional ODCM Appendix C requirements.

Solid Waste Class Volume (Cu. M.) Est. Activity Est. Total (Ci) Error A 1.95E+04 2.16E+01 +/- 25%

B 6.95E+00 7.56E+01 +I- 25%

C O.OOE+00 O.OOE+00 +/- 25%

Container Type Package Volume (Cu. M.)

Gondola Car Strong 68.0 Tight Container B-25 Steel Box Strong Tight 2.9 Container EL-142 Steel Liner Strong Tight 3.7 Container 12

TABLE 3 Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report Second Half, 2004 Solid Waste and Irradiated Fuel Shipments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel).

1. Type of Waste. Unit 6-Month Est. Total Error, %

Period b Dry compressible waste, Cu. M. 33400 contaminated equipment, Ci 1.07 E+01 +/- 25 DAW, cement.

2. Estimate of major nuclide composition (by type of waste).
b. Co-60 19.44% 1.74E+00 Ni-63 28.66% 2.56E+00 Cs-1 37 42.00% 3.76E+00 Fe-55 5.30 4.73E-01 H-3 1.89 1.69E-01 Cs-1 34 1.00 8.92E-02
3. Solid Waste Disposition Number of Shipments Mode of Destination Transportation 1 Trucking over Envirocare of Utah Highway Clive, Utah 424 Rail Envirocare of Utah Clive, Utah 13

TABLE 3 (Continued)

B. Irradiated Fuel Shipments (Disposition):

None Shipped.

Additional ODCM Appendix C requirements.

Solid Waste Class Volume (Cu. Est. Activity Est. Total M.) (c) Error A 3.34E+04 1.07E+01 +1-25%

B O.OOE+00 O.OOE+00 +1-25%

C O.OOE+00 O.OOE+00 +1-25%

Container Packaqe Volume (Cu. M.)

Gondola Car Strong Tight 68.0 Container B-25 Steel Box Strong Tight 2.9 Container 14

TABLE 4 Supplemental Information

1. Regulatory Limits Effluent Concentrations
a. Fission and activation gases 10 CFR 20; Appendix B, Table 2, Column 1
b. Iodines 10 CFR 20; Appendix B, Table 2, Column 1
c. Particulates, (with half lives greater than 8 days) 10 CFR 20; Appendix B, Table 2, Column 1
d. Liquid effluents: 10 CFR 20; Appendix B, Table 2, Column 2
e. Total noble gas concentration: 2.0 E-4 uCi/ml
2. Average Energy- Not Applicable
3. Measurements and Approximations of Radioactivity
a. Fission and Activation Gases Continuous Discharge- Primary Vent Stack and Fuel Building Exhaust Vent samples are analyzed monthly. Activity levels determined are assumed constant for the surveillance interval. The continuous Fuel Building Exhaust Vent monitor reading is used as a basis for increasing periodic sample frequency.

Batch Discharges- The waste gas hold-up drums were purged and removed from service in 1997 in preparation for decommissioning. With the permanent cessation of power operations and the removal of the nuclear fuel, containment purge operations are no longer required. Containment ventilation is directed to the Primary Vent Stack, and sampled as described above.

There are no gaseous effluent release paths associated with ISFSI Operations.

b. Iodines Iodine surveillance no longer applies due to the elapsed time since final plant shutdown from power operations.
c. Particulates Primary Vent Stack and Fuel Building Exhaust Vent particulate totals are taken from a minimum of weekly measurements of continuously collected in line particulate filters. The estimate total error for the particulate measurement has been increased to 35%. This estimated error is based on a detailed evaluation of sampling uncertainties with the particulate samplers. In the decommissioning, credit is not taken for HEPA filtration.

Without verification testing of the filters, it must be assumed that sample line plate-out may increase by up to a factor of three in the Primary Vent Stack and 3.8 in the Fuel Building Exhaust Vent. Detected particulate activity is reported in Tables 1A and 1B.

There are no gaseous effluent release paths associated with ISFSI Operations

d. Liquid Effluents There are no continuous discharges in the decommissioning mode.

15

Each batch of potentially radioactive liquid is analyzed for gross alpha, tritium, dissolved gases, and gamma emitting isotopes prior to discharge.

Composite samples are made of liquid effluents for a quarterly analysis of Strontium-89, Strontium-90, and Iron-55.

There are no liquid effluent release paths associated with ISFSI Operations.

4. Batch Releases
a. Liquids
1. Number of Batch release: 91
2. Total time period for batch releases: 56914 minutes or 948 hours0.011 days <br />0.263 hours <br />0.00157 weeks <br />3.60714e-4 months <br /> & 34 minutes
3. Maximum time period for a batch release: 14352 minutes or 239 hours0.00277 days <br />0.0664 hours <br />3.95172e-4 weeks <br />9.09395e-5 months <br /> & 12 minutes
4. Average time period for batch releases: 625 minutes or 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> & 25 minutes
5. Minimum time period for a batch release: 35 minutes
6. Average stream flow during periods of release of effluents into a flowing stream:

N/A

7. Maximum gross release concentration (uCi/ml): 6.87E-4
b. Gaseous
1. Number of batch release: 0
2. Total time period for batch releases: Not Applicable
3. Maximum time period for a batch release: Not Applicable
4. Average time period for batch releases: Not Applicable
5. Minimum time period for a batch release: Not Applicable
6. Maximum gross release rate (uCi/sec): Not Applicable
5. Unplanned Releases
a. Liquid
1. Release- 3-8-04 While filling a liquid waste holding tank used to discharge processed waste water, a minor release of radioactivity from site occurred, which contained approximately 1.35 E-6 Curies of activity and resulted in a calculated exposure value of 9.0 E-1 1 mrem. This release was recognized as unscheduled and was reported as such to the NRC, and as required by State of Maine regulations.

Cause The cause of this release resulted from a 50gallon overflow of the tank that was in the process of being filled for discharge later in the week.

Corrective Actions Additional controls were established to ensure the tank does not reach a capacity that would allow an overflow condition. In addition, personnel were assigned to the fill point to observe the tank level while filling.

2. Release- 10-7-04 During routine sample analysis for Maine Yankee Outfall 006 (discharges directly to the mudflats), trace amounts of Cesium 137 (Cs-137) activity was Identified exiting the outfall piping.

The release was calculated as containing 3.18 E-6 Curies with an exposure value of 2.21 E-5 mrem. This release was recognized as unscheduled and was reported as such to the NRC, and as required by State of Maine regulations.

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Cause The activity was determined to be residual activity from Outfall -007(restricted area outfall) that contained residual activity from a rupture of a liquid waste discharge line. Sometime after the event, Outfall 007 was connected to Outfall 006 to allow remediation of the facilities Forebay once the liquid discharge path was rerouted to an alternate location specified in the Maine Yankee Offsite Dose Calculation Manual. A dam was created around the outfall 006 piping prior to remediation activities to prevent a release of activity from site. However, the dam failed and allowed the slightly contaminated water to be inadvertently discharged.

Corrective Action The end of the outfall was blocked to prevent a repeat occurrence and was removed later in the year.

b. Gaseous Release- 3/31/04 During the demolition of the Primary/Auxiliary Building (PAB), a trace amount of radioactivity was released to the environment. The calculated activity released, resulted in 4.7E-8 Curies with an exposure value of 4.26E-6 mrem. The release occurred at approximately 11am and ended approximately 15 minutes thereafter. This event was noticed as unscheduled and reported to the NRC, and as required by State of Maine regulations.

Cause The cause of this release resulted from a minor fire during thermal cutting activities of structural material during the demolition process for the structure of the PAB.

Corrective Actions Additional thermal cutting control measure were in place prior to the continuation of the thermal cutting activities.

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APPENDIX A Radioactive Effluent Monitoring Instrumentation Requirement: Radioactive effluent monitoring instrumentation channels are required to be operable in accordance with ODCM Sections 2.3.3 and 2.3.4. With less than the minimum number of channels operable and reasonable efforts to return the instrument(s) to operable status within 30 days of being unsuccessful, ODCM Sections 2.3.3 and 2.3.4 requires an explanation for the delay in correcting the inoperability in the next Annual Effluent Release Report.

Response: No radioactive effluent monitoring instrumentation was out of service for more than 30 consecutive days during the reporting period.

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APPENDIX B Liquid Radwaste Treatment System Requirement: With radioactive liquid waste being discharged without treatment, with estimated doses in excess of the limits in ODCM Section 2.1.5, a report must be submitted to the Commission in the Annual Effluent Release Report for the period.

Response: The requirements of ODCM Section 2.1.5 were met during this period and, therefore, no report is required.

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APPENDIX C Gaseous Radwaste Treatment System Requirement: With radioactive gaseous waste being discharged without treatment with doses in excess of the limits in ODCM Section 2.2.6, a report must be submitted to the Commission in the Annual Effluent Release Report for the period.

Response: The requirements of ODCM Section 2.2.6 were met during this period and, therefore, no report is required.

20

APPENDIX D Lower Limit of Detection for Radiological Analysis Requirement: ODCM Section 2.5 requires that when unusual circumstances result in LLD's Higher than required, the reasons shall be documented in the Annual Radioactive Effluent Release Report.

Response: All samples were counted in such a manner as to satisfy the specified priori lower limits of detection.

21

APPENDIX E Summary of Off-site Dose Calculation Manual Revisions Revision number: Change # 28 Date: 03/04 Summary; The purpose of this ODCM Change was to revise the internal setpoint calculation for the Liquid Effluent Release Monitor, RM-1664. The ODCM required the most limiting nuclide from 10 CFR 20, Table 2, Column 2 to be used in the calculation for the Effluent Concentration Limit (ECL).

The most limiting gamma radionuclide in Table 2, Column 2 of 10 CFR 20 is 1.0 E-6 uCi/mL, which is for Cesium 137 (Cs-1 37). Although the ECL for Cesium was appropriate for previous liquid discharges from Maine Yankee, it was not appropriate for the final discharge of the spent fuel pool water due to the elevated levels of Cobalt-60 (Co-60) after several iterations of clean-up and filtration efforts. Therefore a surrogate ECL was provided.

Revision number: Change # 29 Date: 07/04 Summary:

The purpose of this ODCM Change was to evaluate the need of the Spent Fuel Building / RCA Gaseous Waste Effluent Monitor (RM-19) and the Liquid Remote Monitoring System, RM-1664. In addition, the ODCM change determined the Radiological Environmental Monitoring Program requirements commen-surate with the remaining decommissioning activities and effluent pathways from the Maine Yankee site.

At the time of this ODCM change, all significant system components and piping were removed from the buildings and were shipped for burial. All fuel assemblies were loaded in the NAC-UMS Transportable Storag Canisters (TSC's) and stored at the Independent Spent Fuel Storage and the fuel pool was drained, decontaminated and in a stable condition (<1 0,000dpm/1 00cm 2 , which meets the building demolition criteria, when averaged over the surface area of the building). The waste generated from the fuel pool clean up activities was packaged and ready for shipment. The next stage of the project was to turn the buildings over for demolition and finally, Final Status Survey (FSS) in the restricted area yard.

22

ODCM Change 28 1.0 Purpose The purpose of this ODCM Change is to revise the internal setpoint calculation for the Liquid Effluent Release Monitor, RM-1664. Currently the ODCM requires the most limiting nuclide from 10 CFR 20, Table 2, Column 2 to be used in the calculation for the Effluent Concentration Limit (ECL). The most limiting gamma radionuclide in Table 2, Column 2 of 10 CFR 20 is 1.0 E-6 uCi/mL, which is for Cesium 137 (Cs-137). Although the ECL for Cesium was appropriate for previous liquid discharges from Maine Yankee, it is not appropriate for the final discharge of the spent fuel pool water due to the elevated levels of Cobalt-60 (Co-60) after several iterations of clean-up and filtration efforts. The waste treatment efforts are discussed in section 4 of this evaluation..

2.0 Requirements ODCM Section 2.1.3.1 states: "The concentration of radioactive material in liquid effluents released from the site to unrestricted areas shall be limited to not more than ten times the concentrations specified in 10 CFR, Part 20, Appendix B, Table 2 Column 2."

ODCM Section 2.3.3 states: "The radioactive liquid effluent monitoring instrumentation channels shown in Table 2.1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Section 2.1.3.1 are not exceed during periods of release of radioactive material through the pathways monitored." "The alarm/trip setpoints for these instruments are to ensure that the alarm/trip will occur prior to exceeding 10 times the limits of IOCFR Part 20 Table 2 Column 2".

ODCM Section 6.1 states: "Consistent with Section 2.1.3.1, the total allowable concentration of radioactivity for all releases entering the Back River at any given time shall be limited to a total Effluent Concentration Limit Ratio, ECL Ratio, (R) equal to or less than ten ..."

ODCM Section 6.1.1 states: "Internal monitor setpoints shall be established to monitor compliance with the release concentration limits specified in Section 2.1.3.1. Setpoints shall be calculated so as to alarm the monitor (and, if applicable, terminate the release) if the concentration in the discharge pathway may result in the concentration entering the Back River to exceed ten times the ECL for the most limiting isotope..." (The most limiting gamma emitting radionuclide (i) which potentially may be present in the release pathway (uCi/ml).)

Setpoint = ECL* [(D+Q)/Q)]

  • PF
  • RF
  • 10 D= Dilution Flow Q= Release Flow PF= Pathway Factor (0.9)

RF= Rad monitor Response Factor 1.59E8 cpm/uCi/ml

ODCM Change 28 3.0 Evaluation / Justification for ClIange The design function of the Liquid Waste Effluent Monitor (RM-1664) is to ensure concentrations greater than 10 times Table 2, Column 2 of 10 CFR 20 are not discharged from the Maine Yankee site. Currently, the ODCM internal setpoint calculation requires the most limiting nuclide in 10 CFR 20 Table 2, Column 2 to be used in the setpoint calculation. The most limiting nuclide in Table 2, Column 2 is Cs-137 at 1.0 E-6 uCi/mL. However, a review of the spent fuel pool water analysis results indicate a concentration for Cs-137 that is 1/3 of the effluent concentration limit prior to adding any dilution into the liquid discharge stream prior to discharging from site. However, Co-60 is almost 2 times the effluent concentration limit and is therefore the most limiting nuclide for the remaining spent fuel pool water to be released. As such, the expected instrument response and alarm/trip setpoints are calculated.

3.1 Instrument Expected Response ER= [E (Cu)i - (Cu)non-gamma]

  • RF Where; ER = The expected radiation monitor reponse (Counts per minute)

E(Cu)i = The sum of the concentrations of each of the radionuclides (i) as determined by the pre-release analysis.

(Cu) = Sum of the concentrations for non-gamma emitters RF= The radiation monitor response factor (sensitivity factor) as determined by the most recent monitor calibration (CPM/uCi/ml)

In the case of the spent fuel pool water, the expected response of the instrument is calculated as follows; Concentrations from the pre-dilution analysis in uCi/mL Tritium- 1.94 E-3 (beta emitter)

Co 5.14 E-5 (gamma emitter)

Cs-137- 3.41E-6 (gamma emitter)

ER = (1.99E 1.94 E-3)

  • 1.59E+8 ER= 5.48E-5
  • 1.59E+8 ER= 8714 counts

ODCM Change 28 The external setpoint is then calculated using the above results. The purpose of the external setpoint is to provide assurance that the pre-release analysis is representative of the release being made through the monitor and to alert the operator if a problem exist (i.e. concentrations approaching the effluent concentration limits.

3.2 External setpoint Calculation The external setpoint calculation is dependent of the ECL ratio derived from the liquid release permit.

Where; If R is < 5, then the following setpoint calculation shall be used; 2* ER + Background If R is > 5, then the following calculation shall be used;

[(I/R)

  • ER] + Background R= Is equal to the sum of the individual ECL ratio for each radionuclide present on the liquid release permit.

Assuming the total ECL ratio is < 5, the external setpoint is calculated as follows; 2

  • 8714 counts + 2500 counts Two times the expected monitor response, plus background (2500 counts per minute), establishes an external setpoint (requires observation of monitor response and operator action to terminate the release) of 19928 counts. As required by the ODCM and intent of the external setpoint, the limit must be a value less than the value established for the internal setpoint.

3.3 Internal Setpoint Calculation The purpose of the internal setpoint is to ensure concentrations in liquid effluents are not released in excess of 10 times Table 2, Column 2 of 10 CFR 20. The setpoint is calculated using instrument background, expected discharge flow rate, dilution flow rate, conservative pathway factor and instrument response factor in cpm/uCi/ml. In addition, the current limiting gamma emitting radionuclide used in the internal setpoint calculation is Cesium-137 for additional conservatism.

The internal setpoint is calculated using the expected dilution factor from the liquid release permit for the spent fuel pool discharge. The dilution factor (gallons of dilution + gallons of release / gallons of release) determines the required gallons of dilution needed to ensure post dilution concentrations are less than the effluent concentration limit (10 times Table 2, Column 2).

ODCM Change 28 Internal Setpoint Calculation lECL((D+Qi)/Q,)*PFI*RF*10l Background 2.50E+03 cpm ECL 1.QOE-06 Effluent Concentration Limit (ECL as specified in ODCM Section 2.1.3.1 of the most limiting gamma emitting l radionuclide (l) which potentially may be present in the _

_release _ pathway (uCi/ml). l l l l

- 6.50E+01 Minimum expected total Dilution Flowrate l l

__ 16.25 Flowrate per minute of the release (gallons) l_ l PF1 0.9 Pathway Factor as described in Section 6.1.1 of the OCDM RF 1.59E+08 Radiation monitor response, determined during the Primary Calibration New 7.16E+03 Calculated set point: __

New 9.66E+03 Calculated set point plus background As shown above, the calculation provides an internal setpoint of 9.660E+03 counts per minute, which is less than the calculated external setpoint. Therefore, additional dilution must be used until the internal setpoint is established at a value greater than the external and provides a comfortable margin that will ensure releases will not exceed effluent limit concentrations. The dilution factor required to achieve the above results calculates to a factor of 16 and would limit the liquid discharge to 4 gallons per minute or 27 days for the entire fuel pool batch release.

Internal Setpoint Calculation Using a Dilution Factor of 16 with Cs-137 as Limiting Radionuclide.

ECL 1.OOE-06 Effluent Concentration Limit (ECL as specified in ODCM Section 2.1.3.1 of the most limiting gamma emitting radionuclide (l) which potentially may be present in the release pathway (uCi/ml). I 1I1 I I I e I r D 16.50E+01 I Minimum expected total Dilution FlowrateII I II I I I1

ODCM Change 28 Q. 4.3 Flowrate per minute of the release (gallons) l II I I I PF, 0.9 Pathway Factor as described in Section 6.1.1 of the OCDM l

_ _ _ __ _ _I_ I I _ _

RF 1.59E+08 Radiation monitor response, determined during the Primary Calibration New 2.31 E+04 Calculated set point: I New 2.56E+04 Calculated set point plus background Using this same calculation and replacing the ECL with 3.0 E-6 uCi/mL (ECL for Co-60), the setpoint is calculated as follows; Background 2500 cpm ECL((D+Q,)/Q,)*PF1 RF*1 0 ECL 3.OOE-06 Effluent Concentration Limit (ECL as specified in ODCM

_ X_ _ TSection 2.1.3.1 of the most limiting gamma emitting radionuclide (I) which potentially may be present in the

==________ release pathway (uCi/mI). l l _

D 6.50E+01 Minimum expected total Dilution Flowrate Q_ 16.25 Flowrate per minute of the release (gallons)

PF1 RF 0.9 1.59E+08 Pathway Factor as described in Section 6.1.1 of the OCDM I I I I I.

Radiation monitor response, determined during the Primary Calibration New 2.15E+04 Calculated set point:

New 2.40E+04 Calculated set point plus

___ _ ___ ___ background __ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ __ _ _ _ _

ODCM Change 28 As shown above, the use of Cs-137 in the setpoint calculation unnecessarily limits the spent fuel pool discharge to 4 gallons per minute, when the after dilution concentration (using a dilution factor of 5) of this specific radionuclide is almost 20 times less than the effluent concentration limit.

As shown above (using Co-60 as the limiting nuclide), the calculation provides a setpoint of 24000 counts per minute and provides the same results (i.e. ensures liquid effluent concentrations do not exceed effluent concentration limits) . However, to maintain an increased level of conservatism in the calculation, a surrogate ECL will be used when deriving the internal setpoint for RM-1664. This is in addition to the conservatism that results from using the applied 0.9 pathway factor, which assumes 10% of the radionuclide concentration limit is already present in the receiving water. For example; Surrogate ECL ECL= "1/((l/ECLc6 0) + (cCs 137 /(Cco6eo*ECLcs 137 ))U Where; ECL = Surrogate effluent concentration used for RM-1664 internal setpoint calculation ECLc, 60= The effluent concentration limit for Co-60 as specified in Table 2, Column 2 of 10 CFR 20 Ccs-137 = The pre-dilution release concentration of Cs-137.

ECLcs. 13 7=The effluent concentration limit for Cs-1 37 as specified in Table 2, Column 2 of 10CFR20 Therefore ECL 1/((11 3.0 E-6) + (3.41 E-6 / (5.14E-5

  • 1.0 E-6))

= 1/(( 3.33E+05 + (3.41E-6/5.14E-11))

= 1/(3.33E+05 + 66342) 2.50E-6 ECL = 2.50E-6 uCi/mL The following table identifies the effluent concentration limits (ECL uCi/mL). The table also demonstrates that providing the ECL with the highest ECL ratio is used in the internal setpoint calculation, and the total ECL ratio is less than 10 (required by PMP 6.5.2 and ODCM),

then none of the gamma emitting nuclides would exceed the concentration limit of 10 times Table 2, Column 2 of 10 CFR 20.

ODCM Change 28 Isotope Before Dilution Dilution After ECL ECL Ratio Concentration Rate Dilution (uCi/ml)

(uCi/ml) Activity (uCi/ml)

Alpha 101 H-3 1.94E-03 10 1.94E-04 1.OOE-03 1.94E-01 Mn-54 10 3.00E-05 Co-57 10 6.00E-05 Co-60 1.10E-04 10 1.10E-05 3.00E-06 3.67E+00 Zn-65 10 5.OOE-06 Ag-110m 10 6.00E-06 Sn-113 10 3.OOE-05 Sb-1 25 10 O.OOE+O0 3.OOE-05 Cs-1 34 10 9.OOE-07 Cs-1 37 3.38E-05 10 3.38E-06 1.OOE-06 3.38E+00 Ce-1 44 10 3.00E-06 Other- Py 10 7.24E+00 4.0 Use of Liquid Radwaste Treatment System ODCM Section 2.1.5 states: "The Liquid Radwaste Treatment System shall be used in its designed modes of operation to reduce the radioactive materials in the liquid waste prior to its discharge when the estimated doses to the liquid effluent from the site, when averaged with all other liquid releases over the last 31 days, would exceed 0.06 mrem to the total body, or 0.2 mrem to any organ."

ODCM Figure 6.1 illustrates the Maine Yankee Liquid Radwaste System showing a pre-filter, a Duratek System or equivalent and a post filter prior to the batch waste release tank.

Two releases have occurred since June 20, 2004. Both of these releases resulted in body and organ doses much lower those specified in the ODCM Section 2.1.5. When the estimated dose from proposed release from the PWST is averaged with these other two releases, the resulting average dose is less than those specified in ODCM Section 2.1.5 as shown below:

Dose Dose Date Source Quantity Cs-137 Co-60 Body Organ (Gallons) (uCi/ml) (uCi/ml) (mrem) (mrem) 6/25/04 Frac A 2,000 2.41E-7 5.86E-5 6.84E-5 7/12/04 Frac C 2,000 2.73E-7 4.59E-7 2.49E-4 3.76E-4 Pending -

PWST 155,000 3.41E-6 5.14E-5 1.65E-1 2.67E-1 Average: 5.5E-2 8.91E-2

ODCM Change 28 Nonetheless, since the pending PWST release itself is greater than the criteria specified in ODCM 2.1.5, it is prudent to show how the radwaste treatment of the PWST was equivalent to the Duratek System. Prior to the transfer of the water from the spent fuel pool to the PWST, the water was circulated through the in-pool demineralizer for multiple water volumes. During the transfer the water passed through two filters (0.45 micron and most of the water through 0.30 micron). Following placement of the water in the PWST, the water was processed through the Island Supply and Discharge System (ISADS), including the filters and demineralizer. This treatment resulted in the Cs-137 concentration being cut in half. (See EE-03-009 Engineering Evaluation for the ISADS). Following this treatment, polymer was added to the PWST as a flocculent and the water recirculated through filter for three water volumes. Therefore, the water being released has been process through the Maine Yankee Liquid Radwaste System in fulfilment of ODCM Section 2.1.5 Provided below is a time line of the radwaste treatment processes applied to the PWST water:

(to be provided by Steve Pratte)

TK-16 Recirc/Filtration 5-27-04

  • Recirc line in TK-16 re-routed to the top of the tank for better recirc.
  • Start transfer of SFP to TK-16 through FL-IOOOA (.45 micron filter) and FL-1000B (.3 micron filter).

6-1-04

  • Transfer SFP to TK- 16 through FL-I OOOA (.45 micron filter) and FL-IOOOB (.3 micron filter).
  • FL-1000B changed out at 150 mr. contact.

6-2-04

  • Transfer SFP to TK-16.

6-4-04

  • Transfer SFP to TK-16 6-5-04
  • Transfer SFP to TK-16.

6-7-04

  • Filter change FL-I000B at 400 mr. contact, 100 mr at a foot.
  • Transfer SFP to TK-16 and to stop at 3 inches on the floor.

6-8-04

  • Filter change FL-1000B at 800/200 mr/hr.
  • Start transfer SFP to TK-16.
  • Filter change FL-1000B at 500/150 mr/hr.
  • Filter change FL-I OOOB at 475/200 mr/hr.

6-9-04

  • Transfer SFP to TK-16 (Note: Mark said water was black).
  • TK-16 level now at 30 feet or 152,000 gallons.

ODCM Change 28 6-10-04

  • Transfer SFP to TK-16 (Sample FL-1000B Effluent Co-60 at 1.7E-4 and Cs-137 at 1.7E-5).
  • Filter change FL-1000B at 200/60 mr/hr. (Note: Pit down 8 inches).
  • New .3 micron filter installed in FL-IOOOB.

6-11-04

  • Start transfer SFP to TK-16.
  • Filter change FL-IOOOB due to dp dose at 20 mr/hr.
  • Filter change FL-I OOOB at 40/20 due to dp. (NOTE: went to 100 micron filter Bag).

6-17-04

  • Pump Coffer dam to SFP pit.
  • Recirc TK-16 through V-2 and V-4 demins on ISADS
  • Filter change FL-I000B 6-18-04
  • Recirc TK-16 through demin to lower CS-137.

6-21-04

  • Recirc TK-16 (Official Start, recirc time 51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br />.).
  • TK-16 cleanup isotopic: Co-60 at 3.36E-5, Cs-137 2.73E-7 6-23-04
  • Sample TK-16 for NRC, State and Plant.
  • TK-16 cleanup isotopic: Co-60 6.53E-5, Cs-137 6.53E-6 6-24-04
  • TK-16 samples off to NE Lab.

6-25-04

  • Transfer SFP to "A" Frac
  • Filter change FL-IOOOB at 50 mr/hr.
  • Filter change FL-IOOOB at 90 mr/hr.

(Removed .3 micron and installed 1.0 micron in FL-IOOOB) 6-26-04

  • Installed I micron in FL-I000B
  • Transfer SFP to "A" Frac.

6-28-04

  • Filter change FL-I OOOB at 400/80 mr/hr.
  • 1 micron filter still in FL-I1OOOB
  • Four inches in SFP pit.
  • Transfer Pit to "A" Frac.

6-29-04

  • Tried releasing TK-16.

6-30-04

  • TK-16 cleanup isotopic: Co-60 6.44E-5, Cs-137 3.42E-6 7- 1-04
  • Filter change FL-IOOOB at 25/10 mr/hr.

ODCM Change 28

  • Flushed lines on RMS
  • Recirc TK-16 through .3 micron filter FL-I000B.

7-6-04

  • TK-16 (S-U for discharge) iso: Co-60 6.22E-5, Cs-137 3.25E-6 7-9-04
  • Stretched and pumped five gallons of polymer to TK-16 7-10-04
  • Stretched and pumped eight liters of polymer to TK-16 and placed TK-16 on recirc.

7-12-04

  • TK-16 on recirc to mix polymer.
  • Filter change FL-I OOOB 45/15 7-13-04
  • TK-16 recirc re-sample Iso. Co-60 5.76E-5, Cs-137 3.45E-6
  • TK-16 on recirc through FL-IOOOB with .3 micron filter.

7-14-04

  • Recirc TK-16 through FL-IOOOB.
  • Filter change FL-IOOOB at 60/20 mr/hr.

7-15-04

  • Recirc TK-16 through FL-IOOOB
  • Start Release Recire for discharge sample 7-16-04
  • TK-16 Progress sample info: Co-60 5.78E-5, Cs-137 3.73E-6.
  • FL-IOOOB Filter change 80/20 mr/hr.
  • TK-16 on recirc through FL-IOOOB .3 micron filter 7-17-04
  • Filter change FL-I000B 100/20 mr/hr.
  • Recirc TK-16 through FL-1000B .3 micron filter.
  • Sample TK- 16 for release.

7-19-04

  • Tk-16 Release Iso. Co-60 5.14E-5, Cs-137 3.41E-6 5.0 Conclusion Setpoint calculation The change in the setpoint calculation continues to provide assurance that liquid discharges will not exceed effluent concentration limits of 10 times 10 CFR 20 Table 2, Coulmn2. This change will not adversely impact the accuracy or reliability of liquid effluent dose or setpoint calculations and will continue to maintain the level of radioactive effluent control required by 10 CFR 20, 40 CFR 190, 10 CFR 50.36a and Appendix I to 10 CFR 50 (3.2.1b).

ODCM Change 28 6.0 References

1. NUREG 0133 Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants
2. RM-1664 Primary Calibration Data
3. 10 CFR 20, 40 CFR 190, 10 CFR 50.36a and Appendix I to 10 CFR 50 (3.2.1b).
4. Maine Yankee Offsite Dose Calculation Manual

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements Purpose The purpose of this ODCM Change is to evaluate the need of the Spent Fuel Building / RCA Gaseous Waste Effluent Monitor (RM-19) and the Liquid Remote Monitoring System, RM-1664. In addition, this ODCM change will determine the Radiological Environmental Monitoring Program requirements commen-surate with the remaining decommissioning activities and effluent pathways from the Maine Yankee site.

Background

At this stage in the decommissioning project, all significant system components and piping are removed from the buildings and shipped for burial. All fuel assemblies are loaded in the NAC-UMS Transportable Storage Canisters (TSC's) and stored at the Independent Spent Fuel Storage and the fuel pool has been drained, decontaminated and in a stable condition (<10,000dpm/100cm 2, which meets the building demolition criteria, when averaged over the surface area of the building). The waste generated from the fuel pool clean up activities is packaged and ready for shipment. The next stage is to turn the building over for demolition and finally, perform Final Status Survey (FSS) in the restricted area yard.

Requirements Radioactive Effluents Controls Program Gaseous Effluents Section 2.2.5 of the ODCM specifies that the dose to a member of the public from Tritium and radioactive materials in particulate form with half lives greater than eight days in gaseous effluents released to areas at and beyond the site boundary shall be less than 7.5 mrem to any organ in a calendar quarter and less than 15 mrem to any organ in a calendar year.

Section 2.3.4 specifies that radioactive gaseous effluent monitoring instruments shown in Table 2.2 shall be operable to demonstrate that the limits in Section 2.2.3 are not exceeded during release of radioactive material via this pathway. Section 2.2.3 limits the dose rate (when averaged over one hour) due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary to the following;

a. For lodine-131, Iodine 133, tritium, and radioactive materials in particulate form with half-lives greater than eight days to less than or equal to 1500 mrem/yr to any organ.

The basis for these limits is to ensure that the dose rate at any time at the site boundary and beyond from gaseous effluents from all effluent release points will be within the annual dose limits of 10 CFR 20 while still providing operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration values in Appendix B.

Section 2.5.6 of the ODCM specifies that radioactive gaseous effluents from buildings serviced by the gaseous waste treatment system shall be monitored.

Section 2.2.6 of the ODCM specifies that the gaseous waste treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when estimated gaseous effluent air doses due to gaseous effluent releases from the site to areas at and beyond

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements the site boundary would exceed 0.2 mrad for gamma and 0.4 mrad for beta radiation over a period of 31 days.

This section also specifies that the gaseous waste treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when estimated doses due to gaseous effluent releases from the site to areas at and beyond the site boundary would exceed 0.3 mrem to any organ over 31 days.

Liquid Effluents As specified in section 2.1.2 of the Offsite Dose Calculation Manual (ODCM), the objective of liquid effluent controls is to establish conditions for the release to assure that doses to the public are within the limits specified in 10 CFR 20 and to ensure releases are ALARA and in accordance with 10 CFR 50 Appendix I. Inaddition, the controls ensure that the concentration being released islimited to ten times the concentrations specified in 10 CFR 20, Appendix B,Table 2 Column 2.

The dose or dose commitment to a member of the public from radioactive materials in liquid effluents being released from the site shall be limited to the following;

a. During any calendar quarter to 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 to less than or equal to 3 mrem to the total body, and less than or equal to 10 mrem to any organ.

Section 2.15 requires the use of the liquid radwaste treatment system in its design modes of operation to reduce radioactive materials in liquid waste prior to its discharge when the estimated doses due to the liquid effluent from the site, when averaged with all other liquid releases over the last 31 days, would exceed 0.06 mrem to the total body, or 0.2 mrem to any organ.

Radioactive liquid effluent monitoring instrumentation shall be operable with alarm trip setpoints to ensure the limits specified in section 2.1.3.1 are not exceeded.

Radiological Environmental Monitoring Program The Radiological Environmental Monitoring Program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1)representative measurements of radioactivity in the highest potential exposure pathways, and (2)verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2)conform to the guidance of Appendix I to 10 CFR 50, and (3)include the following; (Ref 5 & 6)

1) Monitoring, sampling, analysis and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.
2) A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of the census, AND
3) Participation in a Inter-laboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements matrices are performed as part of the Quality Assurance Program for environmental monitoring.

The monitoring program 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 measurement and modeling of the environmental exposure pathways. The Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. (Ref. 5) 10 CFR 50, Appendix I requires that the relationship between quantities of radioactive material released in effluents during normal operations, including anticipated operational occurrences , and resultant radiation doses to individuals from principal pathways of exposure be evaluated.

Evaluations / Conclusions Evaluation of Potential Gaseous Releases to The Environment.

At this stage in the project all buildings within the restricted area have either been demolished, surveyed to Final Status Survey criteria or have been remediated and are in the process of being turned over for demolition. The fuel pool however, is in the process of being remediated for meeting the building demolition criteria and is well below 5,000 dpm/100cm when averaged over the building interior. These low levels of activity would not result in an airborne release to the environment. However, the following calculation identifies the impact, if all of the activity was discharged in a single puff release. For conservatism, the 2

contamination level is assumed to be 10,000 dpm/1OOcm average with a ratio of 3000:1 beta- gamma to alpha activity.

Assumptions The Co-60 dose factor stated in the ODCM was used in the calculation since it yields the most conservative results for the nuclides available for release and is representative of the activity on the fuel pool surface.

The beta gamma to alpha nuclide ratio for the spent fuel pool is 3000:1.

This was taken from historical surveys for reactor coolant systems and items removed from the spent fuel pool.

2 The average contamination level of the spent fuel pool is 10,000 dpm/1 00cm .

It isassumed that 100% of activity isreleased in an instantaneous puff release.

Surface area of the Spent Fuel Pool is as follows; Floor= 37' x 41'= 1517 sq ft. = 1.41E+6 cm2 2 East & West Walls= 41'x 38'= 1558 x2 walls = 3116 sq ft. = 2.89E+6 cm 2 North & South Walls = 37' x 38' = 1406 sq ft. x2 = 2812 sq ft. = 2.61E+6 cm Calculation Dose to the critical organ, Dco, in mrem is:

This assumes the entire fuel pool surface is 10,000 dpm/100cm2 and released to the atmosphere.

(from beta-gamma sources-Dco=Qi^DFGico Assumes 100% Co-60)

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements Where Qi is the total activity of radionuclide I released during the period of interest, in Ci; and DFGico is the site specific Critical Organ Dose Factor for Co-60, in mrem/Ci (from Table 4.4 in the ODCM, this is 4.31 E+1 mrem/Ci)

Dco= 3.11 E-04 Ci

  • 4.31E+01 mrem/Ci Dco= 1.34 E-02 mrem Dose to the critical organ, Dco, in mrem is:

This assumes the entire fuel pool surface is 10,000 dpm/100cm2 and released to the atmosphere.

(from potential alpha sources Dco=Qi*DFGico -Assumes 3000:1 py to alpha ratio)

Where Qi is the total activity of alpha contamination released during the period of interest, in Ci; and DFGico is the site specific Critical Organ Dose Factor for radionuclide 1,in mrem/Ci (from Table 4.4 in the ODCM, this is 3.62E+4mremlCi).

Dco= 1.04 E-07 Ci

  • 3.62E+04 mrem/Ci Dco= 3.76 E-03 mrem Total Dose = 1.71 E-2 mrem The only potential for a release of particulate radioactive material is from the demolition of the remaining restricted area buildings and structures. TE-013-01(Rev 4 dated 4/22/03), Radiological Consequences of Hotside Building Demolition identifies a maximum dose to the public of 7.5 E-2 or 0.075 mrem. This potential dose from building demoltion combined with the drying of the spent fuel pool walls and subsequent release to the environment results in a maximum dose to the public of 0.0922 mrem. This is far less than any limits identified in the requirements section for gaseous releases.

10 CFR 20.1302(b)-i, specifies that compliance with the dose limits for members of the public may be demonstrated by measurement or calculation. The above calculation conservatively bounds the potential effluent release resulting from building demolition and the possibility of spent fuel pool walls drying out and releasing 100% of the loose surface activity to the atmosphere, and provide ample justification to abandon the Spent Fuel Building I RCA Gaseous Waste Effluent Monitor.

Evaluation of Potential Liquid Radioactive Releases to The Environment At this stage in the project, all significant radioactive systems, tanks, sumps and trenches have been drained and decontaminated with exception to the Primary Water Storage Tank, which contains water from the Spent Fuel Pool. The water from the spent fuel pool will need to be discharged using RM-1 664 to limit effluent concentrations to no more than 10 times Table 2, Column 2. Once the spent fuel pool water is released, the only potential source of radioactive water to be discharged from Maine Yankee will be groundwater which accumulates in excavations which are below the water table as the excavations are prepared for FSS activities.

The following table summarizes the activity of groundwater recently discharged to the Back River.

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements Table 2 reflects the effluent, environmental and Quality Assurance Program concentration limits for common radionuclides identified for liquids released.

NOTE The nuclides listed in Table 1 are the only nuclides that have been detected in groundwater releases from the Maine Yankee site. Liquid releases are however evaluated for the other nuclides shown in Table 2.6 of the Maine Yankee ODCM.

Table 1: EFFLUENT ACTIVITY CONCENTRATIONS FOR GROUDWATER RELEASES.

Permit # Nuclides identified Activity Concentration uCi/mL 4501 H-3 2.69E-5 Co-60 4.99E-8 Cs-1 37 9.23E-8 4503 H-3 1.34E-4 Co-60 8.42E-7 Cs-1 37 5.50E-7 4504 H-3 1.26E-4 Co-60 3.87E-7 Cs-1 37 5.66E-7 4505 H-3 5.67E-5 Co-60 5.79E-7 Cs-137 2.91 E-7 4506 H-3 2.94E-5 Co-60 3.42E-7 Cs-137 1.93E-7 4507 H-3 1.55E-8 4509 H-3 1.56E-5 Co-60 2.49E-7 Cs-1 37 3.80E-7 4510 H-3 1.31 E-5 Co-60 1.86E-7 Cs-1 37 2.44E-7 4511 H-3 2.78E-6 Co-60 5.16E-8 Cs-137 3.81 E-8 4512 H-3 7.76E-6 Co-60 4.16E-7 Cs-1 37 7.80E-7 Table 2: Effluent Concentrations Limits Nuclide Concentration Limit Quality Assurance Program Environmental (10 CFR 20) Concentration Limits Concentration Limits uci/cc uCi/cc uci/cc H-3 1.0 E-3 uCi/cc 1.OE-2 2.OE-6 Co-60 3.OE-6 uCi/cc 3.OE-5 1.5E-8 Cs-137 1.0E-6 uCi/cc 1.OE-5 1.8E-8 As demonstrated above, the radioactive concentrations in ground water are well below the effluent concentration limits specified in 10 CFR 20 and the Quality Assurance Program. The design function of RM-1 664 (liquid waste effluent monitor) is to ensure that radioactivity in liquid effluents does not exceed the concentration levels in 10 CFR 20 Table 2, Column 2 or in Maine Yankee's case, the limits specified in the Quality Assurance Program. Maine Yankee Quality Assurance Program limits concentrations to 10 times the values specified in Table 2, Column 2 of 10 CFR 20. Given the consistent concentrations of radionuclides in recent groundwater

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements discharges, it is not anticipated that the activity in future groundwater discharges would be any higher than what isshown in Table 1.Therefore, RM-1 664 is no longer necessary to perform its specific design function for future liquid discharges from the Maine Yankee site. If future concentrations in ground water are higher than 10 CFR 20 Table 2, Column 2 limits, then the water can easily be cleaned up to reduce the radioactivity in liquid effluents to concentrations less than those limited by 10 CFR 20.

Evaluation of Radioactive Environmental Monitoring / Sampling Requirements The monitoring program 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 measurement and modeling of the environmental exposure pathways. The Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1 1979. The initially specified monitoring program must be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. (Ref. 5)

The only radioactive release pathways remaining at the Maine Yankee site are liquid radioactive releases and a low potential for particulate releases resulting from the demolition of the restricted area buildings. At this stage in the project, all large components and highly radioactive waste has been shipped. Therefore, a direct dose component (TLD Dose) to the offsite dose contribution no longer exists from the plant.

Inhalation Pathway The only buildings within the restricted area that remain onsite are the Containment Building (which has been FSS'ed), the RCA Building and the Spent Fuel Pool, both of which have been remediated and are in the process of being turned over for demolition. These buildings will then be isolated from the ventilation system in accordance with the ODCM and site-specific procedures and will be the only source of a particulate release from the restricted area. TE-013-01 Rev 4 establishes a total dose of 0.075 mrem from particulates released via this pathway and defined the release to be a maximum distance of 200 ft. radius from the building being demolished. Since the REMP air particulate monitors are well beyond this maximum distance, the activity released will be monitored using the demolition air particulate monitors, which are required by the ODCM. The REMP air particulate monitors will then be removed from the Radioactive Environmental Monitoring Program, but will remain in place to monitor the demolition of the Containment Building, if Maine Yankee so desires.

Ingestion Pathway Maine Yankee has collected milk samples at Chewonki Foundation in Wiscasset and the Hanson Farm in Topsham as part of the REMP. Typically, the samples are analyzed for Sr-89, Sr-90 and 1-131 to detect activity released from the plant vs. monitoring for radionuclides resulting from the plate out in the environment from weapons testing. Since it is not possible for Maine Yankee to have a release of Iodine (lodine-131 was essentially removed 18 months after shutdown) and not Likely to have a release of Strontium, there is no longer a benefit in performing this sampling /

analysis and therefore is being removed from the ODCM. Additionally, the vegetation sampling will be removed along with the milk samples. Vegetation samples were only taken when milk samples were not available.

As part of the ingestion pathway, Maine Yankee has collected fish, clams, mussels, lobster, and crabs due to liquid radioactive releases in the Back River. This sampling will continue for as long as radioactive material is released via this pathway.

Waterborne-Liquid Release Pathway

ODCM Change# 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements The sumps, trenches, cavity, spent fuel pool and tanks have been drained and released to the environment with exception to the PWST, which contains the water from the spent fuel pool. This source of water will be released in accordance with the ODCM, site discharge procedures and Maine State Law. The remaining source of water to be released is ground water to support Final Status Survey activities in the restricted area backyard. Estuary water samples will continue to be collected until such time the liquid release pathway no longer exists.

Historically, sediment samples have been taken (SE-16 & SE-18) around the old outfall area and Foxbird Island. Since the activity concentration is well defined at these locations, and liquid waste is no longer discharged and contributing to those areas, the samples do not provide value to the monitoring program and can be removed from the ODCM.

Direct Dose The direct dose component from plant-related fixed radiation sources to members of the public (the closest offsite area to the plant), as derived from 2003 TLD measurements was estimated to be 0.20 mrem. The neutron shield tank pieces were the largest contributor to this exposure. The estimated dose incorporated an occupancy time of 325 hours0.00376 days <br />0.0903 hours <br />5.373677e-4 weeks <br />1.236625e-4 months <br /> per year for worm diggers, as stated in the Maine Yankee ODCM. The receptor location used in the dose assessment was the center of the nearest portion of the mud flats exposed at low tide, approximately 150 meters from where the Primary Vent Stack once stood. It should be noted that most of the mud flat region in Bailey Cove that is used by the public issituated further away from this selected reference point.

As a result, actual exposures from direct radiation would be less than the above value. The following tables reflect current exposure rates for environmental TLD's and for the restricted area fence line for the 1St quarter of 2004. The exposure rates are likely to reduce even further with the remediation being performed throughout the year.

Table 1: Environmental TLD Exposure Rates-(1 st Quarter 2004)

TLD # Location Exposure Rate (uR/hr)

TL-01 Old Ferry Road 9.00 TL-02 Old Ferry Road (north) 10.44 TL-04 Westport Island-Rt. 144 7.30 TL-06 Route 144 & Greenleaf Rd. 9.37 TL-07 Westport Island-Rt. 144 8.90 TL-08 Near Maine Yankee Screen House 8.15 TL-09 Westport Island-Rt. 144 8.11 TL-10 Bailey Point 8.28 TL-1 1 Mason Station 9.56 TL-12 Westport Firehouse 8.42 TL-1 3 Foxbird Island 9.59 TL-14 Eaton Farm 7.51 TL-i5 Eaton Farm 7.99 TL- 6 Eaton Farm 8.30 TL-17 Eaton Farm Rd. 10.33 TL-18 Eaton Farm Rd 8.93 TL- 9 Eaton Farm Rd Lost TL-36 Boothbay Firehouse (Background Location) 8.58 TL-37 Bath Fire Station (Background Location) 11.59 TL-38 Dresden Substation (Background Location) 8.90 Table 2: Restricted Area Fence Line Dose Rates with Tissue Equivalent Micro-R Meter Area TLD Fence Exposure Rate Area TLD Fence Exposure Rate Location uR/hr Location uR/hr 18 North 25 45 Southwest 5 22 Northwest 10 46 South 8 21 West 10 47 South 6 43 West 6 59 Southeast 4 44 Southwest 6 24 South & East 6-10

ODCM Change # 29 Evaluation of the Station Effluent Monitors and Radiological Environmental Monitoring Requirements As demonstrated in Table 1,the exposure rate at the environmental TLD locations surrounding the site, are indistinguishable from background. The basis for environmental TLD's is to monitor exposure rates to member of the public to ensure the dose limits are in accordance with 40 CFR 190, where an individual at and beyond the site boundary is limited is limited to 25 mrem per year. Given the results of the 1st quarter TLD exposure rates, combined with the fact that the only remaining radioactivity onsite isresidual surface activity, it can be demonstrated by calculation that a member of the public cannot be exposed by direct radiation exposure from the Maine Yankee site and therefore, will not approach the 25 mrem per year limit. For example:

Using the highest exposure rate from the TLD's listed in Table 1 a member of the public would receive no more than 17 mrem. This assumes the individual has taken residence at the location for 8760 hrs per year and uses the highest exposure rate while subtracting the lowest exposure rate of the background TLD's.

TL-2 -TL-36 = net exposure rate x hours/hr =

10.44 uR/hr - 8.58 uR/hr = 1.86 uR/hr x 8760 hrs/year = 16.2 mrem/year.

As demonstrated in Table 2 (with exception to a section of the north fence), and using the equation above, the 25 mrem/year boundary could be established at the restricted area fence.

Therefore, the inner ring of TLD's can be removed from the environmental monitoring program.

Removing these TLD's from the ODCM will not adversely affect how direct dose calculations are performed since Appendix A of the ODCM uses the onsite area monitoring TLD's and the outer ring TLD's to calculate the direct dose component to the public. The outer ring TLD's are the control TLD's for the site. Specifically, exposure rates of area monitoring TLD's AM-31, 39 and 71 are used in comparison to the REMP control TLD's (TL-36,37 & 38). Exposure rates at the area monitoring TLD's are 8.91 uR/hr, 9.42 uR/hr and 8.41 uR/hr respectively. These TLD's adequately demonstrates dose limit compliance with 40 CFR 190.

Specific Changes;

-RM- 19 and the associated ventilation / monitoring system are being removed from the ODCM.

-Gaseous effluent dose factors are being removed from the ODCM

-RM-1664-Liquid waste Process Monitor- An exclusion was added to the ODCM to allow liquid to be discharged without the monitor when pre-dilution analysis is < Table 2, Column 2 of 10 CFR 20.

-The inner ring environmental TLD's are being removed from the ODCM

- Radioactive Environmental Monitoring Requirements associated with gaseous effluents are being removed from the ODCM.

- Sediment samples are being removed from the ODCM

- Deleted nuclides that have undergone 10 half-lives.

References

1. Regulatory Guide 1.109, Calculation of Annual Doses to Man From Routine Release of Reactor Effluents For The Purpose of Evaluating Compliance with 10 CFR Part 50 App. l.

2.10 CFR 50 3.10 CFR 20

4. Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Material in Liquid and Gaseous Effluents From Light-Water-Cooled Nuclear Power Plants.
5. NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors.
6. Maine Yankee Offsite Dose Calculation Manual (ODCM)
7. TE-013-01, Radiological Consequence of Hotside Building Demolition
8. Maine Yankee Quality Assurance Program Attachments None
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