Text

Part 1 of 2, Three Mile Island Nuclear Station, Units 1 & 2, Combined 2001 Annual Radioactive Effluent Release Report
	ML021150025
Person / Time
Site:	Three Mile Island
Issue date:	04/17/2002
From:	George Gellrich; AmerGen Energy Co
To:	; Document Control Desk, NRC/FSME
References
	+kBR1SISP20060424, -RFPFR, 5928-02-20095, RG-1.021, Rev 1
	Download: ML021150025 (159)
	v • d • e

AmerGen Energy Company, LLC Telephone: 717 944-7621 AmerGen..,

An Exelon/BTitish Energy Company Three Mile Island Unit 1 Route 441 South, P.O. Box 480 Middletown, PA 17057 April 17, 2002 5928-02-20095 U. S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION UNIT 1 AND UNIT 2 OPERATING LICENSE NO. DPR-50 AND POSSESSION ONLY LICENSE NO. DPR 73 DOCKET NOS. 50-289 AND 50-320 COMBINED 2001 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT The Annual Radioactive Effluent Release Report required by TMI-1 Technical Specification 6.9.4.1, TMI-2 Technical Specifications 6.8.1.2, and 6.12, and the Off-Site Dose Calculation Manual Part 4, Section 2.1 is enclosed. contains a summary of the quantities of radioactive liquid and gaseous effluents released from the site as outlined in Reg. Guide 1.21, Rev. 1, with data summarized on a quarterly basis following the format of Appendix B thereof. contains information for each type of solid waste shipped offsite during the report period including the container volume, total curie quantity (specified as determined by measurement or estimate), principal radionuclides (specified as determined by measurement or estimate), type of waste, type of shipment and solidification agent(s). includes a summary of unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents made during the reporting period. describes any changes made during 2001 to the Process Control Program (PCP) documents or to the Offsite Dose Calculation Manual (ODCM) and a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Part 3, Section 8.2 of the ODCM. reports all instrumentation not returned to operable status within 30 days per the TMI ODCM Part 1, Sections 2.1.1.b and 2.1.2.b and Part 2, Section 2.1.2.b. is an annual summary of hourly meteorological data collected for 2001 in the form of joint frequency distribution of wind speed, wind direction and atmospheric stability.

$OCA

5928-02-20095 Page 2 of 2 is an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the respective unit during 2001. is an assessment of the radiation doses from the radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary during 2001. is an assessment of the radiation doses to the likely most exposed real individual from reactor releases and other nearby uranium fuel cycle sources including doses from primary effluent pathways and direct radiation for 2001. 0 is a summation of deviations from the sampling and analysis regime specified in the ODCM for TMI-1 and TMI-2. is a copy of the TMI Offsite Dose Calculation Manual (ODCM), revision 22, which was current as of December 31, 2001. There were two revisions made to the ODCM during 2001. Revision 21 was issued on January 19, 2001 and revision 22 was issued on May 4, 2001. is a copy of the procedure change request that modified the ODCM from revision 20 to 21. is a copy of the procedure change request that modified the ODCM from revision 21 to 22.

Please contact Adam Miller of TMI-1 Regulatory Assurance at 717-948-8128 if you have any questions concerning this report.

Sincerely, George H. Gellrich Plant Manager GHG/awm Attachments/Enclosures cc: Region 1 Administrator TMI Senior Resident Inspector TMI-1 Senior Project Manager TMI-2 Project Manager GPU Nuclear Cognizant Officer

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Summary of Radioactive Liquid and Gaseous Effluents Released from TMI during 2001

TABLE IA EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES TMI-1 UNITS 2001 11STQUARTER 1o2001 2001 2ND QUARTER 13RD QUARTER ,EST.TOTAL 1o2001 4TH QUARTER1 ERROR%

A. FISSION AND ACTIVATION GASES

1. TOTAL RELEASE Ci 2.8E-01 5.7E-01 1.1 E+0o 1.7E+00 25%

2. AVERAGE RELEASE RATE FOR PERIOD uCi/sec 3.5E-02 7.3E-02 1.4E-01 2.1 E-01

3. PERCENT OF TECH SPEC LIMIT % ....

B. IODINES

1. TOTAL IODINE 1-131 Ci 8.8E-07 7.3E-07 1.7E-07 1.IE-07 25%

2. AVERAGE RELEASE RATE FOR PERIOD uCi/sec 1.1 E-07 9.3E-08 2.1E-08 1.3E-08

3. PERCENT OF TECH SPEC LIMIT % ....

C. PARTICULATES

1. PARTICULATES WITH HALF-LIVES > 8 DAYS Ci 7.4E-06 <1.E-04 4.6E-08 1.2E-09 25

2. AVERAGE RELEASE RATE FOR PERIOD uCi/sec 9.5E-07 NA 5.8E-09 1.5E-10

3. PERCENT OF TECH SPEC LIMIT % ....

4. GROSS ALPHA RADIOACTIVITY Ci <1.E-11 <1.E-1I 2.2E-08 <1.E-11 D. TRITIUM

1. TOTAL RELEASE Ci 9.6E+00 2.4E+01 2.9E+01 7.7E+01 25%

2. AVERAGE RELEASE RATE FOR PERIOD uCi/sec 1.2E+00 3.1 E+00 3.6E+UU 9.7E+UU I -- I W I .

W W

3. PERCENT OF TECH SPEC LIMIT % I

% ODCM LIMITS: LISTED ON DOSE

SUMMARY

TABLE NOTE: ALL LESS THAN (<) VALUES ARE IN uCi/mI

TABLE 2A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES TMI-1 I 2001 UNITS lIST QUARTER 2001 12ND QUARTER 2001 1 I3RD QUARTERI4TH 2001 QUARTER [-EST. TOTALI I ERROR% %

A. FISSION AND ACTIVATION PRODUCTS

1. TOTAL RELEASES (NOT INCLUDING TRITIUM, GASES, ALPHA) Ci 4.1E-05 3.4E-03 6.1E-04 3.9E-03 25%I

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCi/ml 7.7E-12 5.OE-10 9.3E-11 5.OE-10

3. PERCENT OF APPLICABLE LIMIT %* * *

B. TRITIUM

1. TOTAL RELEASE Ci 1.OE+02 1.6E+02 2.OE+02 3.1E+01 25%

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCi/mI 1.8E-05 2.4E-05 3.OE-05 4.OE-06

3. PERCENT OF APPLICABLE LIMIT % * * *

C. DISSOLVED AND ENTRAINED GASES

1. TOTAL RELEASE Ci <1.E-04 1.2E-01 3.5E-06 <1.E-04 25%1

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCi/ml NA 2.E-08 5.E-13 NA

3. PERCENT OF APPLICABLE LIMIT %*....

D. GROSS ALPHA ACTIVITY

11. TOTAL RELEASE Ci <1.E-07 <1.E-07 <1.E-07 <1.E-07 25%

IE. VOLUME OF WASTE RELEASED (PRIOR TO DILUTION) liters 7.6E+06 1.2E+07 I .OE+07 9.2E+06 10%1 IF. VOLUME OF DILUTION WATER USED liters 5.3E+09 6.8E+09 6.6E+09 7.9E+09 100%1

% ODCM LIMITS: LISTED ON DOSE

SUMMARY

TABLE NOTE: ALL LESS THAN (<) VALUES ARE IN uCi/mI

TABLE 2B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT (2001)

LIQUID EFFLUENTS TMI-1 CONTINUOUS BATCH CONTINUOUS BATCH NUCLIDES RELEASED UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 QUARTER 3 QUARTER 4 CR 51 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5.E-07 <5.E-07 MN 54 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5.E-07 <5.E-07 FE 59 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5.E-07 <5.E-07 CO 58 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 1.8E-05 <5.E-07 <5.E-07 CO 60 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5.E-07 <5.E-07 ZN 65 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5.E-07 <5.E-07 SR 89 Ci <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 SR 90 Ci <5 E-08 <5 E-08 3.9E-07 2.4E-06 5.1E-07 1.4E-07 2.6E-06 1.1E-06 ZR 95 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 NB 95 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 MO 99 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 TC 99M Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 1 131 Ci <1 E-06 <1 E-06 <1 E-06 <1 E-06 <1 E-06 <1 E-06 <1 E-06 <1 E-06 CS 134 Ci <5 E-07 1.9E-04 <5 E-07 <5 E-07 <5 E-07 9.OE-05 <5 E-07 <5 E-07 CS 137 Ci 3.6E-05 3.2E-03 4.7E-06 1.8E-05 5.7E-05 3.8E-03 4.2E-05 2.5E-05 BA 140 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 LA 140 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 CE 141 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 FE 55 Ci <1 E-06 <1 E-06 <1 E-06 <1 E-06 <1 E-06 <1 E-06 5.1E-04 <1 E-06 TOTAL FOR PERIOD Ci 3.6E-05 3.4E-03 5.1E-06 2.0E-05 5.8E-05 3.9E-03 5.5E-04 2.6E-05 XE 133 Ci <1.E-04 <1.E-04 <1.E-04 1.2E-01 <1.E-04 <1.E-04 1.8E-06 <1.E-04 XE 135 Ci <1.E-04 <1.E-04 <1.E-04 <1.E-04 <1.E-04 <1.E-04 1.6E-06 <1.E-04 NOTE: ALL LESS THAN VALUES (<) ARE IN uCi/mI

TABLE 1C EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT (2001)

GASEOUS EFFLUENTS - GROUND LEVEL RELEASES TMI-1

- I I CONTINUOUS BATCH I CONTINUOUS I BATCH NUCLIDES RELEASED UNIT QUARTER I QUARTER 21 QUARTER 11 QUARTER 2 IQUARTER 3 IQUARTER 41QUARTER 3 1QUARTER 41

1. FISSION GASES AR 41 Ci <3 E-07 <3 E-07 1.2E-02 3.7E-02 <3 E-07 <3 E-07 1.2E-01 1.2E-01 KR 85M Ci <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 <5 E-08 3.OE-05 KR 85 Ci <2 E-05 <2 E-05 2.2E-01 3.9E-01 <2 E-05 <2 E-05 2.2E-02 5.3E-03 KR 87 Ci <1 E-07 <1 E-07 <1 E-07 <1 E-07 <1 E-07 <1 E-07 <1 E-07 <1 E-07 KR 88 Ci <2 E-07 <2 E-07 <2 E-07 <2 E-07 <2 E-07 <2 E-07 <2 E-07 <2 E-07 XE131M Ci <1E-6 <tE-6 <1E-6 <1E-6 <1E-6 <1E-6 <1E-6 8.8E-03 XE 133 Ci <2 E-07 <2 E-07 4.4E-02 1.5E-01 <2 E-07 <2 E-07 4.5E-01 1.5E+00 XE133M Ci <3 E-7 <3 E-7 <3 E-7 <3 E-7 <3 E-7 <3 E-7 <3 E-7 9.5E-03 XE 135M Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 XE 135 Ci <5 E-08 2.2E-04 <5 E-08 <5 E-08 5.5E-01 <5 E-08 7.4E-03 2.1E-05 XE 138 Ci <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 <5 E-07 TOTAL FOR PERIOD Ci NA 2.2E-04 2.8E-01 5.7E-01 5.5E-01 NA 5.9E-01 1.6E+00

2. IODINES 1 131 Ci 6.7E-07 3.7E-07 I2.1E-07 3.6E-07 I .7E-07 I1.1E-07 I<1 E-08I <1 E-08 I 133 Ci 5.9E-06 1.7E-06 5.8E-08 1.1E-07 6.2E-08 6.3E-06 <1 E-08 <1 E-08 ITOTAL FOR PERIOD I Ci I 6.5E-06 1 2.1E-06 2.7E-07 I 4.7E-07 I 2.3E-07 6.4E-06 I NA I NA

3. PARTICULATES CO588-EE8 i <1 E < E <<1 I E-1 I 1.2E-09 < E-08 <1

<E-08 CS 137 Ci 7.4E-06 <1 E-11 <1 E-08 <1 E-08 <1 E-11 <1E-11 4.6E-08 <1 E-08 NOTE: ALL LESS THAN VALUES (<) ARE IN uCi/mI

SUPPLEMENTAL INFORMATION FACILITY: TMI UNIT I LICENSE: DPR 50-289

1. REGULATORY LIMITS -- - REFER TO TMI OFFSITE DOSE CALCULATION MANUAL A. FISSION AND ACTIVATION GASES:

B. IODINES:

C. PARTICULATES, HALF-LIVES > 8 DAYS:

D. LIQUID EFFLUENTS:

2. MAXIMUM EFFLUENT CONCENTRATIONS -- - TEN TIMES 10 CFR 20, APPENDIX B TABLE 2 PROVIDE THE MAXIMUM EFFLUENT CONCENTRATIONS USED IN DETERMINING ALLOWABLE RELEASE RATES OR CONCENTRATIONS.

A. FISSION AND ACTIVATION GASES:

B. IODINES:

C. PARTICULATES, HALF-LIVES > 8 DAYS:

D. LIQUID EFFLUENTS:

3. AVERAGE ENERGY PROVIDE THE AVERAGE ENERGY (E-BAR) OF THE RADIONUCLIDE MIXTURE IN RELEASES OF FISSION AND ACTIVATION GASES, IF APPLICABLE E-BAR BETA = 2.95E-01 E-BAR GAMMA = 4.37E-01 E-BAR BETA AND GAMMA 7.32E-01

4. MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY PROVIDE THE METHODS USED TO MEASURE OR APPROXIMATE THE TOTAL RADIOACTIVITY IN EFFLUENTS AND THE METHODS USED TO DETERMINE RADIONUCLIDE COMPOSITION:

A. FISSION AND ACTIVATION GASES: HPGE SPECTROMETRY, LIQUID SCINTILLATION B. IODINES: HPGE SPECTROMETRY C. PARTICULATES HPGE SPECTROMETRY, GAS FLOW PROPORTIONAL, BETA SPECTROMETRY D. LIQUID EFFLUENTS: HPGE SPECTROMETRY, LIQUID SCINTILLATION

5. BATCH RELEASES PROVIDE THE FOLLOWING INFORMATION RELATING TO BATCH RELEASES OF RADIOACTIVITY MATERIALS IN LIQUID AND GASEOUS EFFLUENTS.

A. LIQUID (ALL TIMES IN MINUTES) [QUARTER 11QUARTER 2 QUARTER 31QUARTER 4

1. NUMBER OF BATCH RELEASES: 16 15 32 17

2. TOTAL TIME PERIOD FOR BATCH RELEASES: 4127 4082 8634 4512

3. MAXIMUM TIME PERIOD FOR A BATCH RELEASE: 565 347 332 290

4. AVERAGE TIME PERIOD FOR BATCH RELEASES: 257 272 269 265

5. MINIMUM TIME PERIOD FOR A BATCH RELEASE: 66 240 235 240

6. AVERAGE STREAM FLOW DURING PERIODS OF RELEASE OF EFFLUENT INTO A FLOWING STREAM: (CFM) I 2.2E+06 2.3E+06 4.2E+05 7.1EE+05 B. GASEOUS (ALL TIMES IN MINUTES)

1. NUMBER OF BATCH RELEASES: 3 3 4 18

2. TOTAL TIME PERIOD FOR BATCH RELEASES: 1536 2196 1916 85643

3. MAXIMUM TIME PERIOD FOR A BATCH RELEASE: 785 1260 1000 68571

4. AVERAGE TIME PERIOD FOR BATCH RELEASES: 512 732 479 4757

5. MINIMUM TIME PERIOD FOR A BATCH RELEASE: 24 106 6 15

6. ABNORMAL RELEASES A. LIQUID

1. NUMBER OF RELEASES: o -0

2. TOTAL ACTIVITY RELEASED: (CURIES) N/A N/A N/A N/A B. GASEOUS

1. NUMBER OF RELEASES: -0

2. TOTAL ACTIVITY RELEASED: (CURIES) N/A N/A N/A N/A

TABLE 1A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES TMI-2 I I 2001 2001 2001 2001 EST. TOTAL UNITS 1IST QUARTER I 2ND QUARTER 3RD QUARTER 14TH QUARTE ERROR %]

A. FISSION AND ACTIVATION GASES

1. TOTAL RELEASE Ci <LLD <LLD <LLD <LLD 25%

2. AVERAGE RELEASE RATE FOR PERIOD uCi/sec N/A N/A N/A N/A

3. PERCENT OF TECH SPEC LIMIT % ....

B. IODINES NOT APPLICABLE FOR TMI-2 C. PARTICULATES

1. PARTICULATES WITH HALF-LIVES > 8 DAYS Ci <LLD <LLD <LLD <LLD 25%

2. AVERAGE RELEASE RATE FOR PERIOD uCi/sec N/A <N/A <N/A <N/A

3. PERCENT OF TECH SPEC LIMIT % * * * *

4. GROSS ALPHA RADIOACTIVITY Ci <LLD <LLD <LLD <LLD D. TRITIUM

1. TOTAL RELEASE Ci 9.4E-01 5.OE-01 3.3E-01 9.9E-02 25%

2. AVERAGE RELEASE RA I FIO R PI-ODIU

3. PERCENT OF TECH SPEC LIMIT I

ui/sec

%

I 1 .2E-U1 "R

I O.'--UZ

-.- I 4r./tr--.U,, I I.zr---Uz I

BATCH RELEASES 0 0 0 0

% ODCM LIMITS: LISTED ON DOSE

SUMMARY

TABLE NOTE: ALL LESS THAN (<) VALUES ARE IN uCi/mI

TABLE 1C EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT GASEOUS EFFLUENTS-GROUND LEVEL RELEASES TMI-2 2001 I CONTINUOUS MODE I BATCH MODE I CONTINUOUS MODE BATCH MODE INUCLIDES RELEASED UNIT 1ST QUARTERI 2ND QUARTER 1ST QUARTERI ND QUARTE 13RD QUARTERI4TH QUARTERP3RD QUARTERI4TH QUARTE]

1. FISSION GASES KRYPTON-85 Ci <2 E-5 <2 E-5 <2 E-5 <2 E-5 <2 E-5 <2 E-5 <2 E-5 <2 E-5 KRYPTON-85M Ci <5E-8 <5E-8 <5E-8 <5E-8 <5E-8 <5E-8 <5E-8 <5E-8 KRYPTON-87 Ci <1 E-7 <1 E-7 <1 E-7 <1 E-7 <1 E-7 <1 E-7 <1 E-7 <1 E-7 KRYPTON-88 Ci <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 XENON-133 Ci <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 <2 E-7 XENON-1 35 Ci <5 E-8 <5 E-8 <5 E-8 <5 E-8 <5 E-8 <5 E-8 <5 E-8 <5 E-8 XENON-135M Ci <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 XENON-138 Ci <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 AR-41 Ci <3E-7 <3E-7 <3E-7 <3E-7 <3E-7 <3E-7 <3E-7 <3E-7 TOTAL FOR PERIOD Ci N/A N/A N/A N/A N/A N/A N/A N/A

2. IODINES NOT APPLICABLE TO TMI-2

3. PARTICULATES STRONTIUM-90 Ci <1 E-11 <1 E-11 N/A N/A <1 E-11 <1 E-11 N/A N/A COBALT 60 Ci <1 E-10 <1 E-10 N/A N/A <1 E-10 <1 E-10 N/A N/A ANTIMONY 125 Ci <1 E-10 <1 E-10 N/A N/A <1 E-10 <1 E-10 N/A N/A CESIUM-134 Ci <1 E-10 <1 E-10 N/A N/A <1 E-10 <1 E-10 N/A N/A CESIUM-137 Ci <1 E-10 <1 E-10 N/A N/A <1 E-10 <1 E-10 N/A N/A TOTAL FOR PERIOD Ci N/A N/A N/A N/A N/A N/A N/A N/A

4. TRITIUM ITRITIUM Ci 9.4E-01 I 5.OE-01 <1 E-6 <1 E-6 3.30E-01 9.89E-02 <1 E-6 <1 E-6 NOTE: ALL LESS THAN (<) VALUES ARE IN uCi/mI

TABLE 2A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES TMI-2 I I 2001 2001 2001 I 2001 EST. TOTAL UNITS 1ST QUARTER 12ND QUARTER I3RD QUARTER 4TH QUARTER ERROR% I A. FISSION AND ACTIVATION PRODUCTS

1. TOTAL RELEASES (NOT INCLUDING TRITIUM, GASES, ALPHA) Ci 3.4E-06 8.8E-06 5.4E-06 4.4E-06 25%1

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCi/mI 6.4E-13 1.3E-12 8.2E-13 5.6E-13

3. PERCENT OF APPLICABLE LIMIT %* * *

B. TRITIUM

1. TOTAL RELEASE Ci 3.OE-05 6.6E-04 2.9E-05 7.6E-05 25%1

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCi/mI 5,7E-12 9.7E-11 4.4E-12 9.6E-12

3. PERCENT OF APPLICABLE LIMIT %* * *

C. DISSOLVED AND ENTRAINED GASES

1. TOTAL RELEASE Ci <LLD <LLD <LLD <LLD 25%1

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCi/ml N/A N/A N/A N/A

3. PERCENT OF APPLICABLE LIMIT %* * *

D. GROSS ALPHA ACTIVITY

1. TOTAL RELEASE Ci <LLD <LLD <LLD I <LLD 25%1 E. VOLUME OF WASTE RELEASED (PRIOR TO DILUTION) liters 8.9E+02 1.4E+05 1.2E+03 1.1E+03 10%L F. VOLUME OF DILUTION WATER USED liters 5.3E+09 6.8E+09 6.6E+09 7.9E+09 107%

NUMBER OF BATCH RELEASES 1 3 1 1

% ODCM LIMITS: LISTED ON DOSE

SUMMARY

TABLE NOTE: ALL LESS THAN (<) VALUES ARE IN uCi/mI

TABLE 2B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT LIQUID EFFLUENTS TMI-2 2001 CONTINUOUS MODE BATCH MODE CONTINUOUS MODE BATCH MODE NUCLIDES RELEASED UNIT 1ST QUARTER 2ND QUARTER 1ST QUARTER 2ND QUARTER 3RD QUARTER 4TH QUARTER 3RD QUARTER 4TH QUARTER CO 60 Ci <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 SR 90 Ci <5 E-8 <5 E-8 2.8E-09 <5 E-8 <5 E-8 <5 E-8 6.8E-08 2.8E-08 SB 125 Ci <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 CS 134 Ci <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 <5 E-7 CS 137 Ci <5 E-7 <5 E-7 3.4E-06 8.8E-06 <5 E-7 <5 E-7 5.3E-06 4.4E-06 H-3 Ci <1 E-5 <1 E-5 3.OE-05 6.6E-04 <1 E-5 <1 E-5 2.9E-05 7.6E-05 TOTAL FOR PERIOD Ci NA NA 3.4E-05 6.7E-04 NA NA 3.4E-05 8.1E-05 NOTE: ALL LESS THAN VALUES (<) ARE IN uCi/mI

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-200095 Solid Waste Shipped Offsite during 2001

TABLE 3 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. Solid waste shipped off-site for burial or disposal (not irradiated fuel)

1. Type of waste UNIT 12 month EST. Total period Error %

a. Spent resins, filter sludges, m3 6.09m3 25%

Evaporator bottoms, etc. Ci 70.7 Ci

b. Dry compressible waste, m3 2128.9m 3 25%

contaminated equipment, etc. Ci 1.78Ci

c. Irradiated components, control mo N/A N/A rods, etc. Ci

d. Other (describe) : Solidified Liquid mo 29.45 m3 25%

Ci .0179Ci

2. Estimate of major nuclide composition (by type of waste)

a. Fe55 6.87%

Cs137 41.8%

Ni63 33.5%

Cs134 13.4%

b. Ni63 7.50%

Cs137 62.8%

Co58 22.9%

Sr90 1.66%

Cs134 3.35%

c. N/A %

d. Cs137 84.3 %

Sr90 .997%

Ni63 5.997%

Cs134 6.24%

3. Solid Waste Disposition Mode of Transportation Destination Number of Shipments See attached for this information B. Irradiated Fuel Shipments (Disposition)

Number of Shipments Mode of Transportation Destination N/A

WASTE SHIPPED AS FOLLOWS A.l.a One(l)- Stainless Steel reusable liner @ 215 ft3- Dewatered Resin A.1.b Thirty-four (34) Intermodal Containers @ 1040 ft ea. - Turbine metals Twelve (12) - Enviropak Bags @ 192 ft3 ea. - Turbine metals Three (3) - Enviropak Bags @ 5147 ft3 ea. - Turbine metals Three (3) - Enviropak Bags @ 1890 Wt3 ea. - Turbine metals Three (3) - Enviropak Bags @ 2770 ft3 ea. - Turbine metals One (1) Enviropak Bag @ 1470 ft3 - Turbine metals Fourteen (14) - steel boxes at 96 ft3 ea. - metals Twenty-two (22)- steel drums @ 7.5 ft3 ea. -building debris Four (4) - 20' cargo containers @ 1280 fM3 ea.- uncompacted DAW A-1-d One (1) - steel tanker @ 1040 ft3 - oil

A.3.a One Shipment Hittman Transport Cask Studsvik Process Ctr- Erwin,TN A.3.b Forty-four ship. MHF Logistics-Flatbed RACE- LLC Memphis ,TN Three Shipments Norfolk Southern-Rail RACE LLC Memphis,TN Three Shipments TSMT- Flatbed RACE LLC Memphis,TN.

Four Shipments Hittman Transport-Flatbed Duratek-Oak Ridge, TN Two Shipments Kindrick Trucking - Flatbed ATG -Oak Ridge,TN.

A.3.d One Shipment Kindrick Trucking-Flatbed ATG- Oak Ridge,TN.

ALL SHIPMENT WERE TYPE A- LSA-II

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Summary of Unplanned Releases from the TMI Site During 2001 There were no unplanned releases to unrestricted areas from either the TMI-1 or TMI-2 site during 2001.

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Changes to the Process Control Program and the Offsite Dose Calculation Manual during 2001, And a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census

1. Changes to the Process Control Program Section 6.8.1 of the TMI-1 Technical Specifications requires that a procedure be established, implemented and maintained for the Process Control Program Implementation. TMI procedure 1104-281 (Waste Solidification Process Control Program) had been the document used to ensure that waste would be processed to meet waste stability requirements of 10 CFR 61.56 prior to disposal. Procedure 1104-281 was used in conjunction with operating procedures 11 04-28A(Evaporator Concentrate Processing) andl 104 -28C (Primary Resin and Precoat Processing) to ensure that radioactive wastes would be solidified or dewatered in compliance with 10 CFR Parts 20,61,71 ,State Regulations, burial ground requirements and other requirements that govern the disposal of radioactive waste. With TMI now part of a much larger organization, a goal of standardizing radioactive waste handling procedures was undertaken across the Exelon Fleet. The result has been a new Process Control Program for Radioactive Waste ,Procedure RW-AA-1 00, implemented at TMI on August 17, 2001. This procedure now looks the same at all of the Exelon operated plants. The advantage to TMI, is that now the boundaries and parameters for the preparation of procedures for processing ,sampling, analysis, packaging, storage and shipment of solid radwaste for compliance with local, state, federal, and burial site requirements are found in the same document. Previously these requirements were found throughout many documents, making it somewhat cumbersome to ensure documented compliance with the applicable regulations for a given waste stream. In conclusion, TMI is still applying the same process and bounding conditions for the preparation of radioactive waste for transportation and disposal as the previous version of the Process Control Program. The new procedure covers all radioactive waste generated at this station in one document versus being addressed over many procedures, thus making documented compliance with the federal, state, local, and burial site criteria much easier to monitor.

2. Changes to the Offsite Dose Calculation Manual during 2001 The Offsite Dose Calculation Manual (ODCM) was modified twice during 2001. These changes did not reduce the accuracy or reliability of dose calculations or setpoint determinations. The level of effluent controls required by 10 CFR 20.1301, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR 50 was not reduced and the accuracy or reliability of effluent, dose or setpoint calculations was not adversely impacted for the reasons stated below.

Revision 21 of the ODCM was issued on January 19, 2001. Revision 21 made the following changes to the ODCM:

"* Eliminated references to GPU Nuclear.

"* Changed the title of Director Operations and Maintenance to Plant Manager.

"* Clarifies the TMI-2 effluent flow rate as a monitoring device, since this recorder sums the flows from individual measuring devices. It also clarifies the available compensatory action if this monitoring device is out of service.

Clarifies that releases from the Waste Evaporator Condensate Storage Tanks (WECST) uses 20% of ODCM liquid concentration limits for tritium and 10% for other isotopes as an administrative means of ensuring compliance with ODCM Control 2.2.1.1. The limits specified in ODCM Control 2.2.1.1 are not affected.

Revision 22 of the ODCM was issued on May 4, 2001. Revision 22 made the following changes to the ODCM:

"* Eliminates the need to change the alarm setpoint for RM-L-6 each time a WECST is released.

"* Changes "National Bureau of Standards" to "National Institute of Standards and Technology."

"* Added typical flow rates for the gaseous release points in buildings located outside the "power-block."

3. A listing of new locations for dose calculations and/or environmental monitoring identified by the land use census Based on the results of the 2001 land use census, no changes to the radiological environmental monitoring program or the dose model are required.

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Instrumentation not returned to Operable status within 30 days during 2001 There was no instrumentation not returned to operable status within 30 days per the TMI ODCM Part 1, Sections 2.1.1 .b and 2.1.2.b and Part 2, Section 2.1.2.b during 2001.

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Annual Summary of Hourly Meteorological Data for 2001

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: A WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N S 15 35 34 1 0 0 85 NNE SSW 9 73 68 7 0 0 157 NE SW 21 45 22 2 0 0 90 ENE WSW 22 27 7 1 0 0 57 E W 24 34 20 10 1 0 89 ESE WNW 45 53 36 9 1 0 144 SE NW 68 120 81 35 5 0 309 SSE NNW 69 119 48 9 0 0 245 S N 13 29 27 0 0 0 69 SSW NNE 7 10 4 0 0 0 21 SW NE 5 5 0 0 0 0 10 WSW ENE 5 14 0 0 0 0 19 W E 9 16 8 0 0 0 33 WNW ESE 11 22 28 0 0 0 61 NW SE 19 18 18 2 0 0 57 NNW SSE 15 27 8 1 0 0 51 TOTAI, 357 647 409 77 7 0 1497

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: B WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N S 3 13 7 0 0 0 23 NNE SSW 4 12 11 3 0 0 30 NE SW 4 9 6 0 0 0 19 ENE WSW 8 1 2 1 0 0 12 E W 4 8 13 11 0 0 36 ESE WNW 7 11 21 17 2 0 58 SE NW 3 27 21 31 9 0 91 SSE NNW 8 14 8 3 2 0 35 S N 2 6 4 2 0 0 14 SSW NNE 3 3 1 0 0 0 7 SW NE 3 3 1 0 0 0 7 WSW ENE 5 4 0 0 0 0 9 W E 1 6 4 0 0 0 11 WNW ESE 4 9 6 1 0 0 20 NW SE 7 6 12 0 0 0 25 NNW SSE 2 3 4 0 0 0 9 TOTAL 68 135 121 69 13 0 406

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: C WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N S 1 6 3 2 0 0 12 NNE SSW 1 10 10 0 21 NE SW 0 4 3 0 7 ENE WSW 5 4 2 0 11 E W 1 5 8 4 18 ESE WNW 1 6 14 5 26 SE NW 3 13 8 12 41 SSE NNW 5 8 5 5 23 S N 4 3 1 0 8 SSW NNE 1 0 1 0 2 SW NE 1 4 0 0 5 WSW ENE 0 8 0 0 8 W E 2 7 6 0 15 WNW ESE 1 6 9 2 18 NW SE 1 2 7 1 11 NNW SSE 0 3 1 0 4 TOTAL 27 89 78 31 4 1 230

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: D WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N S 8 65 30 1 0 104 NNE SSW 14 82 47 4 0 147 NE SW 19 43 9 1 0 72 ENE WSW 20 25 13 1 0 59 E W 27 70 100 29 1 227 ESE WNW 39 73 121 83 9 325 SE NW 34 80 109 109 41 378 SSE NNW 37 52 53 24 17 183 S N 32 40 13 1 1 87 SSW NNE 27 30 3 0 0 60 SW NE 28 27 7 3 0 65 WSW ENE 28 52 7 4 0 91 W E 32 86 39 0 0 157 WNW ESE 37 67 85 10 0 199 NW SE 26 72 48 7 0 153 NNW SSE 22 54 17 0 0 93 TOTAL 430 918 701 277 69 5 2400

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: E WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N S 22 64 28 0 0 0 114 NNE SSW 27 86 20 3 1 0 137 NE SW 48 83 7 3 0 0 141 ENE WSW 47 79 14 1 0 0 141 E W 50 107 39 7 1 0 204 ESE WNW 45 88 75 18 2 0 228 SE NW 69 52 75 18 2 0 216 SSE NNW 58 91 19 5 0 0 173 S N 52 80 9 1 0 0 142 SSW NNE 36 22 2 0 0 0 60 SW NE 29 25 1 0 0 0 55 WSW ENE 39 27 2 0 0 0 68 W E 65 55 1 0 0 0 121 WNW ESE 42 42 15 0 0 0 99 NW SE 33 28 10 0 0 0 71 NNW SSE 32 44 1 0 0 0 77 TOTAL 694 973 318 56 6 0 2047

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: F WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL N S 38 9 0 47 NNE SSW 46 26 0 72 NE SW 60 25 0 86 ENE WSW 47 22 0 69 E W 57 22 0 81 ESE WNW 59 21 0 82 SE NW 58 8 0 69 SSE NNW 53 40 0 97 S N 32 31 0 63 SSW NNE 23 13 0 36 SW -NE 19 5 0 24 WSW ENE 22 3 0 25 W E 35 19 0 54 WNW ESE 56 7 0 63 NW SE 35 9 0 44 NNW SSE 36 5 0 43 TOTAL 676 265 10 4 0 0 955

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: G WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 26 6 0 0 0 0 32 N S 60 21 1 0 0 0 82 NNE SSW 44 9 1 0 0 0 54 NE SW 29 10 0 0 0 0 39 ENE WSW 26 5 1 0 0 0 32 E W 29 4 0 0 0 0 33 ESE WNW 32 9 3 0 0 0 44 SE NW 29 15 2 0 0 0 46 SSE NNW 21 7 0 0 0 0 28 S N 9 4 0 0 0 0 13 SSW NNE 14 0 0 0 0 0 14 SW NE 17 2 0 0 0 0 19 WSW ENE 27 12 0 0 0 0 39 W E 37 13 0 0 0 0 50 WNW ESE SE 43 5 0 0 0 0 48 NW 38 0 0 0 0 0 38 NNW SSE TOTAL 481 122 8 0 0 0 611

THREE MILE ISLAND METEROLOGICAL DATA 2001 JOINT FREQUENCY TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD OF RECORD: January 1, 2001 TO December 31, 2001 STABILITY CLASS: ALL WIND SPEED SECTOR WINDS TO FROM 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 113 198 102 4 0 0 417 N S NNE SSW 161 310 157 17 1 0 646 NE SW 196 218 49 6 0 0 469 178 168 38 4 0 0 388 ENE WSW E W 189 251 183 61 3 0 687 ESE WNW 225 256 268 133 14 0 896 SE NW 267 309 299 206 61 6 1148 259 339 137 48 19 0 802 SSE NNW S N 156 196 54 4 1 0 411 SSW NNE 106 82 11 0 0 0 199 SW NE 99 69 9 3 0 0 180 116 110 9 4 0 0 239 WSW ENE W E 171 201 58 0 0 0 430 WNW ESE 188 166 143 13 0 0 510 164 140 95 10 0 0 409 NW SE NNW SSE 145 136 33 1 0 0 315 TOTAL 2733 3149 1645 514 99 6 8146 Hours of Missing/Invalid Data: 614

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Assessment of Radiation Doses Due to Radioactive Liquid and Gaseous Effluents Released from TMI during 2001 TMI-1 The attached table presents the maximum hypothetical doses to an individual and the general population resulting from 2001 TMI-1 releases of gaseous and liquid effluents.

Provided below is a brief explanation of the table.

A. Liquid (Individual)

Calculations were performed on the four age groups and eight organs recommended in Regulatory Guide 1.109. The pathways considered for TMI-1 were the consumption of drinking water and fish and standing on the shoreline influenced by TMI-1 effluents. The latter two pathways are considered to be the primary recreational activities associated with the Susquehanna River in the vicinity of TMI. The "critical receptor" or Receptor I was that individual who 1) consumed Susquehanna River water from the nearest downstream drinking water supplier (Wrightsville Water Supply), 2) consumed fish residing in the vicinity of the TMI-1 liquid discharge outfall and 3) occupied an area of shoreline influenced by the TMI-1 liquid discharge.

For 2001, the calculated maximum whole body (or total body) dose from TMI-1 liquid effluents was 1.60E-1 mrem to an adult (line 1). The maximum organ dose was 2.41 E-1 mrem to the liver of a teen (line 2).

B. Gaseous (Individual)

There were six major pathways considered in the dose calculations for TMI-1 gaseous effluents. These were: (1) plume exposure (2) inhalation, consumption of; (3) cow milk, (4) vegetables and fruits, (5) meat, and (6) standing on contaminated ground. Real-time meteorology was used in all dose calculations for gaseous effluents.

Lines 3 and 4 present the maximum plume exposure at or beyond the site boundary. The notation of "air dose" is interpreted to mean that these doses are not to an individual, but are considered to be the maximum doses that would have occurred at or beyond the site boundary. The table presents the distance in meters to the location in the affected sector (compass point) where the theoretical maximum plume exposures occurred. The calculated maximum plume exposures were 9.85E-4 mrad and 1.26E-3 mrad for gamma and beta, respectively.

The maximum organ dose due to the release of iodines, particulates and tritium from TMI-1 in 2001 was 1.08E-2 mrem to the thyroid of an child residing 2150 meters from the site in the NNE sector (line 5). This dose again reflects the maximum exposed organ for the appropriate age group.

C. Liquid and Gaseous (Population)

Lines 6 - 9 present the person-rem doses resulting from 2001 TMI-1 liquid and gaseous effluents. These doses were summed over all pathways and the affected populations. The person-rem doses from liquid effluents were based upon the population encompassed within the region from the TMI-1 outfall extending down to the Chesapeake Bay (approximately 5,000,000 people). The person-rem doses from gaseous effluents were based upon the 1980 population and considered the population out to a distance of 50 miles around TMI (approximately 2,200,000 people).

Population doses were summed over all distances and sectors to give an aggregate dose.

The calculated maximum whole body dose to the affected population from TMI-1 liquid effluents was 1.03E+1 person-rem. The maximum critical organ population dose from liquid effluents was 1.04E+1 person-rem to the liver. TMI-1 gaseous effluents resulted in a whole body population dose of 4.99E-1 person-rem and a maximum critical organ population dose of 4.99E-1 person-rem to the liver, thyroid, kidney, lung and GI tract.

For 2001, TMI-1 liquid and gaseous effluents resulted in maximum hypothetical doses that were a small fraction of the quarterly and yearly ODCM dose limits.

TMI-1

SUMMARY

OF MAXIMUM INDIVIDUAL DOSES FOR TMI-1 FROM January 1, 2001 through December 31, 2001 Estimated Location % of Applicable Dose Age Dist Dir ODCM ODCM Dose Effluent Organ (mrem) Group (m) (to) Dose Limit Limit (mrem)

Quarter Annual Quarter Annual (1) Liquid Total Body 1.60E-1 Adult Receptor 1 1.07E+1 5.33E0 1.5 3 (2) Liquid Liver 2.41 E-1 Teen Receptor 1 4.82E0 2.41 E0 5 10 (3) Noble Air Dose 9.85E-4 610 NNE 1.97E-2 9.85E-3 5 10 Gas (gamma-mrad)

Air Dose 1.26E-3 610 NNE 1.26E-2 6.30E-3 10 20 (4) Noble (beta-mrad)

Gas (5) Iodine, Thyroid 1.08E-2 Child 2150 NNE 1.44E-1 7.20E-2 7.5 15 Tritium &

Particulates

SUMMARY

OF MAXIMUM POPULATION DOSES FOR TMI-1 FROM January 1, 2001 through December 31, 2001 Estimated Applicable Population Dose Effluent Organ (person-rem)

(6) Liquid Total Body 1.03E+1 (7) Liquid Liver 1.04E+1 (8) Gaseous Total Body 4.99E-1 (9) Gaseous Liver, Thyroid 4.99E-1 Kidney, Lung &

GI Tract

TMI-2 The attached table presents the maximum hypothetical doses to an individual and the general population resulting from 2001 TMI-2 releases of gaseous and liquid effluents.

Provided below is a brief explanation of the table.

A. Liquid (Individual)

Calculations were performed on the four age groups and eight organs recommended in Regulatory Guide 1.109. The pathways considered for TMI-2 were the consumption of drinking water and fish and standing on the shoreline influenced by TMI-2 effluents. The latter two pathways are considered to be the primary recreational activities associated with the Susquehanna River in the vicinity of TMI. The "critical receptor" or Receptor 1 was that individual who 1) consumed Susquehanna River water from the nearest downstream drinking water supplier (Wrightsville Water Supply), 2) consumed fish residing in the vicinity of the TMI-2 liquid discharge outfall and 3) occupied an area of shoreline influenced by the TMI-2 liquid discharge.

For 2001, the calculated maximum whole body (or total body) dose from TMI-2 liquid effluents was 4.74E-4 mrem to an adult (line 1). The maximum organ dose was 7.53E-4 mrem to the liver of a teen (line 2).

B. Gaseous (Individual)

There were six major pathways considered in the dose calculations for TMI-2 gaseous effluents. These were: (1) plume exposure (2) inhalation, consumption of; (3) cow milk, (4) vegetables and fruits, (5) meat, and (6) standing on contaminated ground. Real-time meteorology was used in all dose calculations for gaseous effluents.

Since there were no noble gases released from TMI-2 during 2001, the gamma and beta air doses (lines 3 and 4, respectively) were zero.

The maximum organ dose due to the release of particulates and tritium from TMI-2 in 2001 was 4.80E-5 mrem to the liver, total body, thyroid, kidney, lung and GI tract of a child residing 2000 meters from the site in the SE sector (line 5).

C. Liquid and Gaseous (Population)

Lines 6 - 9 present the person-rem doses resulting from 2001 TMI-2 liquid and gaseous effluents. These doses were summed over all pathways and the affected populations. The person-rem doses from liquid effluents were based upon the population encompassed within the region from the TMI-2 outfall extending down to the Chesapeake Bay (approximately 5,000,000 people). The person-rem doses from gaseous effluents were based upon the 1980 population and considered the population out to a

distance of 50 miles around TMI (approximately 2,200,000 people).

Population doses were summed over all distances and sectors to give an aggregate dose.

The calculated maximum whole body dose to the affected population from TMI-2 liquid effluents was 2.82E-4 person-rem. The maximum critical organ population dose from liquid effluents was 5.97E-4 person-rem to the bone. TMI-2 gaseous effluents resulted in a whole body population dose of 3.67E-3 person-rem and a maximum critical organ population dose of 3.67E-3 person-rem to the liver, thyroid, kidney, lung and GI tract.

For 2001, TMI-2 liquid and gaseous effluents resulted in maximum hypothetical doses that were a small fraction of the quarterly and yearly ODCM dose limits.

TMI-2

SUMMARY

OF MAXIMUM INDIVIDUAL DOSES FOR TMI-2 FROM January 1, 2001 through December 31, 2001 Location % of Estimated Age Dist Dir ODCM Dose ODCM Dose Effluent Applicable Organ Dose (mrem) Group m (to) Limit Limit (mrem)

Quarter Annual Quarter Annual (1) Liquid Total Body 4.74E-4 Adult Receptor 1 3.16E-2 1.58E-2 1.5 3 (2) Liquid Liver 7.53E-4 Teen Receptor 1 1.51 E-2 7.53E-3 5 10 (3) Noble Gas Air Dose 0 -- 0 0 5 10 (gamma-mrad)

(4) Noble Gas Air Dose 0 -- 0 0 10 20 (beta-mrad)

(5) Tritium & Liver, Total Body, 4.80E-5 Child 2000 SE 6.40E-4 3.20E-4 7.5 15 Particulate Thyroid, Kidney, Lung &

GI Tract

SUMMARY

OF MAXIMUM POPULATION DOSES FOR TMI-2 FROM January 1, 2001 through December 31, 2001 Estimated Applicable Population Dose Effluent Organ (person-rem)

(6) Liquid Total Body 2.82E-4 (7) Liquid Bone 5.97E-4 (8) Gaseous Total Body 3.67E-3 (9) Gaseous Liver, Thyroid, 3.67E-3 Kidney, Lung &

GI Tract

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Assessment of Radiation Doses from Liquid and Gaseous Effluents Releases to Members of the Public within the TMI Site Boundaries during 2001 The Offsite Dose Calculation Manual requires an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary during the reporting period.

The following are the assumptions made in this assessment:

1. A member of the public stays in the owner controlled area for 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months . The 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months is based upon shoreline recreation period given in Table E-5, of Regulatory Guide 1.109.

2. The individual is standing next to a radiologically controlled area, where the dose rate is 0.5 mR/hr. In areas where the dose rate is greater than 0.5 mR/hr, the area would be posted as a restricted area.

3. Liquid effluents are not a pathway to the individual on site.

4. The maximum airborne effluent per hour is characterized by Release G200110016, which is a RB purge. This effluent release had the highest concentration of radionuclides released into the air for year 2001. (Ar-41 1.1 E-6 ptCi/cc, Xe-1 33 3.72E-6 MCi/cc, and H-3 1.37E-6 pCi/cc)

5. Highest dispersion factor for gaseous effluents to personnel outside restricted area is 7.61E-5 sec/m3 for the 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months used. This is the value used in Final Safety Analysis Report section 2.5.4.2.1, Containment release to Yard intake. This intake is close to the protected area and is close to where the Reactor Building (Containment) would release.

The maximum total body dose to an individual is 34 mrem.

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Assessment of Radiation Dose to Most Likely Exposed Real Individual per 40 CFR 190 Dose calculations were performed to demonstrate compliance with 40 CFR 190 (ODCM Part IVSection 2.10). Gaseous and liquid effluents released from TMI-1 and TMI-2 in 2001 resulted in maximum individual doses (regardless of age group) of 0.02 mrem to the thyroid and 0.25 mrem to any other organ including the whole (total) body. The direct radiation component was determined using the highest quarterly fence-line exposure rate as measured by an environmental TLD, and subtracting from it, the lowest quarterly environmental TLD exposure rate.

Based on the maximum exposure rate of 6.1 mR/standard month, a person residing at the fence-line for 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months (shoreline exposure from Reg. Guide 1.109) received an exposure of 0.56 mR. Based on the lowest exposure rate of 3.5 mR/standard month and converting it by the same method yielded a background exposure of 0.32 mR.

Therefore, the net exposure from direct radiation from TMINS was 0.24 mR. Combining the direct radiation exposure (assumed to be equal to dose) with the maximum organ doses from liquid and gaseous releases, the maximum potential (total) doses were 0.26 mrem to the thyroid and 0.49 mrem to any other organ. Both doses were well below the limits specified in 40 CFR 190.

0 2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 Deviation from the ODCM Sampling and Analysis Regime during 2001 There was one deviation from the effluent sampling and analysis regime specified in the TMI Offsite Dose Calculation Manual during 2001.

The deviation was not obtaining a grab sample on a sump prior to discharging that sump. A condition report (TMI's corrective action system) was submitted as a result of the missed sample.

The maximum concentration value for the same type of water that was obtained during the month of release was used to account for the activity discharged. The dose and activity of this discharge is an insignificant value when compared to the plant's annual effluent.

2001 Annual Radioactive Effluent Releases Report for TMI 5928-02-20095 TMI Offsite Dose Calculation Manual, Revision 22 6610-PLN-4200.01 (Revision 22 was issued on May 4, 2001)

Number AmerGen TMI -Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 AppilicabilitylSope USAGE LEVEL Effective Date TM[ Division 3 05104101 This document is within QA plan scope X Yes No 50.59 Applicable X Yes No List of Effective Pages Revision Pag~e Revision Paae Revision Paae Revision 1 22 21 22 41 22 61 22 2 22 22 22 42 22 62. 22 3 22 23 22 43. 22 22 4 22 24 22 44 22 22 5 22 25 22 45 22 22 6 22 26 22 46 22 22 7 22 27 22 47 67 22 8 22 28 22 48 S68 22 9 22 29 22 49 22 10 22 30 22 so ~~71697 22 11 22 31 22 51 22 12 22 32 22 .52 72 22 13 22 33 22 53 .2 73 22 14 22 34 22 22 74 22 15 22 35 22 22 75 22 16 17 18 22 22 22 36 37 38 22 22 6k 22 22 22 76 77 78 22 22 22 19 20 22 22 39 40 22 69 22 22 79 80 22 22 K %i Procedure Owner is/ B. A. Parfift 03/06/01 Approver /si J. Telfer 04/06/01 1

Number TM] - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual- (ODCM) 22 List of Effective Pages Pane Revision Pagqe Revision Pace Revision Page Revision 81 22 121 22 161 22 82 22 122 22 162 22 83 22 123 22 163 22 84 22 124 22 164 22 85 22 .125 22 165 22 86 22 126 22 166 22 87 22 127 22 167 22 88 22 128 22 168 a

22 89 22 129 22 169 22 90 22 130 22 170 22 91 22 131 22 171 22 92 22 132 22 172 93 22 133" 22 173 94 22 134 22 174 95 22 135 22 175 96 22 136 22 176 97 22 137 22 177 98 22 138 22 22 99 22 139 22 22 100 22 140 22 22 101 22 141 22 22 102 22 142 22 22 103 22 143 22

.104 22 144 22 105 22 145.. V_ J 185 22 106 22 146 22

2U 0 184 186 22 107 22 147 187 22 108 22 148 188 22 109 22 149 22 189 22 110 22 22 190 22 111 22 22 191 22 112 22 22 192 22 113 22 22 193 22 114 22 22 194 22 115 22 22 195 22 116 22 22 196 22 117 22 157 22 118 22 158 22 119 7.2 159 22 120 M2 160 22 2

Number TMI - Unit 1 Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a supporting document of the Three Mile Island Nuclear Station (TMI) Unit I and Unit 2 PDMS Technical Specifications and implements TMI radiological effluent controls.

The ODCM contains the controls, bases, and surveillance requirements for liquid and gaseous radiological effluents.

In addition, the ODCM describes the methodology and parameters to be used in the calculation of off-site doses due to radioactive liquid and gaseous effluents. This document also describes the methodology used for calculation of the liquid and gaseous effluent monitoring instrumentation alarm/trip set points. Liquid and Gaseous Radwaste Treatment System configurations are also included.

The ODCM also is used to define the requirements for the TMI radiological environmental nr9Mting program (REMP) and contains a list and graphical description of the specific sample locations use EMP.

The ODCM is maintained at the Three Mile Island (TMI) site for use as a reference did training document of accepted methodologies and calculations. Changes in the calculation methods o ra ts will be incorporated into the ODCM to ensure the ODCM represents the present methodology in al le reas. Changes to the ODOM will be implemented in accordance with the TMI-1 and TMI-2 PDMS ech Td 2ecifications.

The ODCM follows the methodology and models suggested by NUREG-0133,\ Regulatory Guide 1.109, Revision 1 for calculation of off-site doses due to plant effluent releases. Simp -yingassumptions have been applied in this manual where applicable to provide a more workable docubr ent for implementation of the Radiological Effluent Controls requirements.

TMl implements the TMI Radiological Effluent Controls Pr iman egulatory Guide 1.21, Revision I (Annual Radioactive Effluent Release Report) requirements by A puterized system used to determine TMI effluent releases and to update cumulative effluent doses.

A,*

3

Number TMI- Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Reviston No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS PART I TMI-1 RADIOLOGICAL EFFLUENT CONTROLS Section Page 1.0 DEFINITIONS 15 Table 1-1, Frequency Notation 19 2.0 RADIOLOGICAL EFFLUENT CONTROLS AND BASES 21 2.1 Radioactive Effluent Instrumentation Q21 2.1.1 Radioactive Liquid Effluent Instrumentation 21 Table 2.1-1, Radioactive Liquid Effluent Instrumentation 22 2.1.2 Radioactive Gaseous Process and Effluent Mo In umentation 23 Table 2.1-2, Radioactive Gaseous Process and Effluent Monitoncg Instrumentation 24 2.2 Radiological Effluent Controls 30 2.2.1 Liquid Effluent Controls 30 2.2.2 Gaseous Effluent Control 33 2.2.3 Total Radioactive E nfn Is 39 3.0 SURVEILLANCES " 41 3.1 Radioactive Effluent ns en tion 41 3.1.1 Radioactiv id Effluent Instrumentation 41 Table 3.1-1, tive Liquid Effluent Monitoring Instrumentation anance Requirements 42 3.1.2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation 44 Table 3.1-2, Radioactive Gaseous Process and Effluent Monitoring Instrumentation 45 Surveillance Requirements 3.2 Radiological Effluents 49 3.2.1 Liquid Effluents 49 Table 3.2-1, Radioactive Liquid Waste Sampling and Analysis Program 50 4

Number TMI - Unit I Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Contd)

PART I TMI-1 RADIOLOGICAL EFFLUENT CONTROLS Section Pagze 3.2.2 Gaseous Effluents 53 Table 3.2-2, Radioactive Gaseous Waste Sampling and Analysis Program 55 3.2.3 Total Radioactive Effluents 59 4.0 PART I REFERENCES 60 Sl 5

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01.

Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Cont'd)

PART II TMI-2 RADIOLOGICAL EFFLUENT CONTROLS Section Pacie 1.0 DEFINITIONS 62 Table 1.1, Frequency Notation 64 2.0 CONTROLS AND BASES 65 2.1 RadioactiveRadioactive Effluent Instrumentation 2cQ 6565 2.1.1 Liquid Effluent Instrumentation 65 2.1.2 Radioactive Gaseous Process and Effluent Monitori tation 65 Table 2,1.2, Radioactive Gaseous Process and Effluent M riong strumentation 67 2.2 Radioactive Effluent Controls 68 2.2.1 Liquid Effluent Controls A 68 2.2.2 Gaseous Effluent Controls 71 2.2.3 Total Radioactive Effluen tr 76 3.0 SURVEILLANCES 78 3.1 Radioactive Effluent instr e ta 78 3.1.1 Radioacti L i uent Instrumentation 78 3.1.2 Radactiv eous Process and Effluents Monitoring Instrumentation 78 Table 3-1-* tive Gaseous Process and Effluent Monitoring Instrumentation 79 lance Requirements 3.2 Radiological Effluents 80 3.2.1 Liquid Effluents s0 Table 3.2-1, Radioactive Liquid Waste Sampling and Analysis Program 81 3.2.2 Gaseous Effluents 82 Table 3.2-2, Radioactive Gaseous Waste Sampling and Analysis Program 83 3.2.3 Total Radioactive Effluents 86 6

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Contd)

PART II TMI-2 RADIOLOGICAL EFFLUENT CONTROLS Section Eg7e 4.0 PART Il REFERENCES 87 03 7

Number TMI - Unit 1 Radiological Controls Procedure 661 O-PLN-4200.01 Ti91e Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLEOF CONTENTS (Cont'd)

PART III EFFLUENT DATA AND CALCULATIONAL METHODOLOGIES Section Paqe 1.0 LIQUID EFFLUENT MONITORS 89 1.1 TMI-1 and TMI-2 Liquid Radiation Monitor Set Points 89 1.2 TMI Liquid Release Points and Liquid Radiation Monitor Data 90 1.3 Control of Liquid Releases 91 2.0 LIQUID EFFLUENT DOSE ASSESSMENT 97 2.1 Liquid Effluents - 10 CFR 50 Appendix I 97 2.2 TMI Liquid Radwaste System Dose Calcs Once per Mon98 2.3 Alternative Dose Calculational Methodology 99 3.0 LIQUID EFFLUENT WASTE TREATMENT SYSTEM 104 3.1 TMI-1 Liquid Effluent Waste Treatment 104 3.2 Operability of TM I- Liquid Effluent tment System 105 3.3 TMI-2 Liquid Effluent Waste T rt Stern 105 4,0 GASEOUS EFFLUENT MONITO 108 4.1 TMI-1 Noble Gas Monit et oints 108 4.2 TMI-1 Particulate an a ioiodine Monitor Set Points 110 4.3 TMI-2 Gas ion Monitor Set Points 111 4.4 TMI-1 Gaseous ffluent Release Points and Gaseous Radiation Monitor Data 112 4.5 TMI-2 Gaseous Effluent Release Points and Gaseous Radiation Monitor Data 114 4.6 Control of Gaseous Effluent Releases 115 8

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Cont'd)

PART III EFFLUENT DATA AND CALCULATIONAL METHODOLOGIES Section Pace 5.0 GASEOUS EFFLUENT DOSE ASSESSMENT 129 5.1 Gaseous Effluents - Instantaneous Release Limits 129 5.1.1 Noble Gases 129 5..1.1 Total Body 129 5.1.1.2 Skin 130 5.1.2 lodines and Particulates 131 5.2 Gaseous Effluents -Ib CFR 50 Appendix I 132 5.2.1 Noble Gases 132 5.2.2 lodines and Particulates 133 5.3 Gaseous Radioactive System Dose ion nce per Month 135 5.4 Alternative Dose Calculational Me d g ios 136 6.0 GASEOUS EFFLUENT WASTE TREA T S TEM 158 6.1 Description of the TMI-1 G e'sadwaste Treatment System 158 6.2 Operability of the TMI-.s ous Radwaste Treatment System 158 7.0 EFFLUENT TOTAL DO SES'MENT 160 8.0 TMINS RADIOLO G*I ONMENTAL MONITORING PROGRAM (REMP) 161 8,1 Monitoring Pro ram Requirements 161 8.2 Land Use Census 163 8.3 Interlaboratory Comparison Program 165 9.0 PART HIlREFERENCES 182 9

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Cont'd)

PART III EFFLUENT DATA AND CALCULATIONAL METHODOLOGIES Section Page TABLES Table 1.1 TMI-1 Liquid Release Point and Liquid Radiation Monitor Data 93 Table 1.2 TMI-2 Sump. Capacities 94 Table 2.1 Liquid Dose Conversion Factors (DCF): DF6 ( 100 Table 2.2 Bioaccumulation Factors, BFI 103 Table 4.1 TMI-1 Gaseous Release Point & Gaseous Radiation Monitor 116 Table 4.2 TMI-2 Gaseous Release Point& Gaseous Radiation Mo Data 117 Table 4.3 Dose Factors for Noble Gases and Daughters 118 Table 4.4 Atmospheric Dispersion Factors for Three Mis nd - Station Vent 119 Table 4.5 Atmospheric Dispersion Factors for Thre Ila -Ground Release 120 Table 4.6 Dose Parameters for Radioiodines -0a, ctive Particulate In Gaseous Effluents 121 Table 5.2.1 Pathway Dose Factors, R, - InI aon 135 Table 5.2.2 Pathway Dose Factors, . l ahin136 Table 5.2.3 Pathway Dose Fact ,- T n, Inhalation 137 Table 5.2.4 Pathway Doseao - Adult, Inhalation 138 Table 5.3.1. Pathway D a rs, R, - All Age Groups, Ground Plane 139 Table 5.4.1 Pathway Dose actors, R, - Infant, Grass-Cow-Milk 140 Table 5.4.2 Pathway Dose Factors, R, - Child, Grass-Cow-Milk 141 Table 5.4.3 Pathway Dose Factors, R, - Teen, Grass-Cow-Milk 142 Table 5.4.4 Pathway Dose Factors, R, - Adult, Grass-Cow-Milk 143 Table 5.5.1 Pathway Dose Factors, R, - Infant, Grass-Goat-Milk 144 10

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 "Fife Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Cont'd)

PART III EFFLUENT DATA AND CALCULATIONAL METHODOLOGIES Section Pagle TABLES Table 5.5.2 Pathway Dose Factors, R, - Child, Grass-Goat-Milk 145 Table 5.5.3 Pathway Dose Factors, R, - Teen, Grass-Goat-Milk 146 Pathway Dose Factors, R, - Adult, Grass-Goat-Milk 147 Table 5.5.4 Pathway Dose Fattors, Ri - Infant, Grass-Cow-Meat 148 Table 5.6.1 Table 5.6.2 Pathway Dose Factors, R1 - Child, Grass-Cow-Meat 0, 149 Table 5.6.3 Pathway Dose Factors, R, - Teen, Grass-Cow-Meat 150 Table 5.6.4 Pathway Dose Factors, R1 - Adult, Grass-Cow-Meat 151 Table 5.7.1 Pathway Dose Factors, RI - Infant, Vegetation 1 152 Table 5.7.2 Pathway Dose Factors, R1 - Child, Veget A 153 Table 5.7.3 Pathway Dose Factors, R, - Teen, V ti 154 Table 5.7.4 Pathway Dose Factors, R - Ad et on 155 Table 8.1 Sample Collection and An *lirements 164 Table 8.2 Reporting Levels for a cti ty Concentrations in Environme ntal Samples 168 Table 8.3 Detection Cap eie rnvironmental Sample Analysis 169 Table 8.4 TMINS RE r Locations - Air Particulate and Air Iodine 171 Table 8.5 TMINS REMP S tion Locations - Direct Radiation (TLD) 171 Table 8.6 TMINS REMP Station Locations - Surface Water 173 Table 8.7 TMINS REMP Station Locations - Aquatic Sediment 173 Table 8.8 TMINS REMP Station Locations - Milk 174 Table 8.9 TMINS REMP Station Locations - Fish 174 Table 8.10 TMINS REMP Station Locations - Food Products 175 11

..-.-. L L Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 T-ite Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Cont'd)

PART Ill EFFLUENT DATA AND CALCULATIONAL METHODOLOGIES Section Pao~e TABLES MAP 8.1 Three Mile Island Nuclear Station Locations of Radiological Environmental 176 Monitoring Program Stations within 1 Mile of the Site MAP 8.2 Three Mile Island Nuclear Station Locations of Radiological Environmental 177 Monitoring Program Stations within 5 miles of the Site MAP 8.3 Three Mile Island Nuclear Station Locations of Radiological Environm Monitoring Program Stations Greater than 5 miles from the Site 4

C 178 FIGURES Figure 1.1 TM1-1 Liquid Effluent Pathways 95 Figure 1.2 TMI-2 Liquid Effluent Pathways 96 Figure 3.1 TMI-1 Liquid Radwaste 106 Figure 3.2 TMI-1 Liquid Waste Evaporators 107 Figure 4.1 TMI-i Gaseous Effluent Pathways 122 Figure 4,2 TMI-1 Auxiliary &Fuel Handl~engu 123 Figure 4.3 TMI-1 Reactor Building E t 1 124 TMI-1 Condenser 0 ga unt F Figure 4.4 athway 125 Figure 4.5 TMI-2 Gaseou fflu t iltration ýystern/Pathways 126 Figure 6.1 Waste Gas&s 157 12

Number "TMI - Unit 1 aafnnr _A.flfl 114 Irmauiuiuyicai Cnuiui rois- ceu r~uevv'-ruUW% aV Title Revision No.

OffMite Dose Calculation Manual (ODCM) 22 TABLE OF CONTENTS (Contd)

PART IV REPORTING REQUIREMENTS Section Page 1.0 TMI ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 182 2.0 TMI ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 183 3.0 PART IV REFERENCES 185 APPENDICES A. Pathway Dose Rate Parameter (Pt) 0 186 B. Inharation Pathway Dose Factor (R1) 187 C. Ground Plane Pathway Dose Factor (RI) 188 D. Grass-Cow-Milk Pathway Dose Factor (Ri) 189 E. Cow-Meat Pathway Dose Factor (R1) 191 F. Vegetation Pathway Dose Factor (Ri) 192 APPENDIX A- F REFERENCES 193 4._)

13

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.O1 "Mhr.

I m&l*

Revision No.

Offsite Dose Calculation Manual (ODCM) 22 PART TMI-1 RADJOLOCA UENT CONTROLS 14

Number TMI - Unit I Radioloaical Controls Procedure 6610-PLN-4200.01 Revision No.

Title Offsite Dose Calculation Manual (ODCM) 22 1.0 DEFINITIONS The following terms are defined for uniform interpretation of these controls And surveillances.

1.1 Reactor Operating Conditions

-1.1. Cold Shutdown The reactor is in the cpld shutdown condition when it is subcritical by at least one percent delta klk and Tavg is no more than 2000F. Pressure is defined by Technical Specification 3.1.2.

1.1.2 1.1.3 Hot Shutdown The reactor is in the hot shutdown condition when it is s delta k/k and Tavg is at or greater than 525°F.

Reactor Critical 4 by at least one percent The reactor is critical when the neutron Chain reactfol self-sustaining and Keff = 1.0.

1.1.4 Hot Standby ,1 The reactor is in the hot stanidhy cjn1diýhen all of the following conditions exist:

a. Tavg is greaterl n-r

b. The reacar cr

c. Indica &dutro in power on the power range channels is less than two power. Rated power is defined in Technical Specification n1.1.

1.1.5 Power' 4

'orlf in a power operating condition when the indicated neutron power is above toft*ted power as indicated on the power range channels. Rated power is eefiner Specification Definition 1.1.

in Tchnical 1.1.6 Refueling Shutdown The reactor is in the refueling shutdown condition when, even with all rods removed, the reactor would be subcritical by at least one percent delta k/k and the coolant temperature at the decay heat removal pump suction is no more than 140PF. Pressure is defined by Technical Specification 3.1.2. A refueling shutdown refers to a shutdown to replace or rearrange all or a portion of the fuel assemblies and/or control rods.

1.1.7 Refueling Operation An operation involving a change in core geometry by manipulation of fuel or control rods when the reactor vessel head is removed.

15

Number TMI - Unit I Title Radiological Controls Procedure 661 O-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 1.1.8 Refueling Interval Time between normal refuelings of the reactor. This is defined as once per 24 months.

1.119 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical.

1.1.10 Tave Tave is defined as the arithmetic average of the coolant temperar'skthe hot and cold legs of the loop with the greater number of reactor coolant pu operng, if such a distinction of loops can be made.

1.1,11 Heatup - Cooldown Mode The heatup-cooldown mode is the range of reac coo temperature greater than 2000F and less than 5250F. f 1.2 Operable A system, subsystem, train, component or de hall be OPERABLE or have OPERABILITY when it is capable of performing its specified func1n(s when aLl necessary attendant instrumentation, controls, electrical powc~ling1 seal water, lubrication or other auxiliary equipment that are required for the syste stem, train, component, or device to perform its function(s) are also capable of per Ithtli related support function(s).

1.3 Instrument Channel An instrument channel i combtion of sensor, wires, amplifiers, and output devices which are connected for the purpo asuring the value of a process variable for the purpose of observation, contra ndor rection. An instrument channel may be either analog or digital.

1.4 Instrumentatio urvei ce 1.4.1 Ch est A CHANNEL TEST shall bethe injection of a simulated signal into the channel as close to the sensor as practical to verify OPERABILITY, including alarm and/or trip functions.

1.4.2 Channel Check A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.

16

Number TMI - Unit I I Radiological Controls Procedure 6610-PLN4200.01 Title jRevision No.

Offsite Dose Calculation Manual (ODCM) 22 1.4.3 Source Check A SOURCE CHECK shall be the qualitative assessment of channel response when the.

channel sensor is exposed to a radioactive source.

1.4.4 Channel Calibration An instrument CHANNEL CALIBRATION is a test, and adjustment (if necessary), to establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values. Calibration shall encompass the entire channel, including euipment actuation, alarm, or trip and shall be deemed to include the channel test.

1.5 Dose Equivalent 1-131 The DOSE EQUIVALENT 1-131 shall be that concentration of 1-13 icrejcýrie/gram) which alone would produce the same thyroid dose as the quantity and isot re of 1-131, 1-132, 1-133, 1-134, and J-135 actually present The thyroid dose conversq n fa" rs used for this calculation shall be those Listed in Table IlI of TID 14844, "Calculation of D'ice Faors for Power and Test Reactor Sites". [Or in Table E-7 of NRC Regulatory Guide.1.1 0 evision 1, October 1977.1 1.6 Offsite Dose Calculation Manual (ODCM)

The OFFSITE DOSE CALCULATION MANUk Cj4) contains the methodology and parameters used in the calculationof offsite doses re lg fro dioactive gaseous and liquid effluent, in the calculation of gaseous and liquid efflue A g Alarm/Trip Setpoints, and in the conduct of the Radiological Environmental Monitori ,o . The ODCM also contains (1) the Radiological Effluent Controls, (2) the Radiologi En o fmentat Monitoring Program and (3) descriptions of the information that should be incl 04 the nnual Radiological Environmental Operating and Annual Radioactive Effluent ReleasK rts.

1.7 Gaseous Radwaste Tre e The GASEOUS RWAS REATMENT SYSTEM is the system designed and installed to reduce radioac~ve gaV s effluent by collecting primary coolant system off gases from the primary system and Qr6~ing f delay or holdup for the purpose of reducing the total radioactivity prior to release to t y ment.

1.8 Ventilation Exha st Treatment System A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluent by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodine or particulates from the gaseous exhaust system prior to the release to the environment.

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS.

1.9 Purge - Purging PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is required to purify the confinement.

17

Number TMI - Unit I Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 1.10 Venting VENTING is the controlled process of discharging air as gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is not provided. Vent used in system name does not imply a VENTING process.

1.11 Member(s) of the Public MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the GPU System, GPU contractors or vendors. Also excluded from this category are persons who enter the site t ce equipment or to make deliveries.

"1.12 Site Boundary The SITE BOUNDARY used as the basis for the limits on the seous effluents is as defined in Section 2.1.2.2 and shown on Figure 2.1-3 of thT . This boundary line includes portions of the Susquehanna River surface betw the t bank of the river and Three Mile Island and between Three Mile Island and Shelley Island>..

The SITE BOUNDARY used as the basis for the 'rits on the release of liquid effluents is as shown in Figure 1.1 in Part: I of this ODCM.

1.13 Frequency Notation The FREQUENCY NOTATION spe performance of Surveillance Requirements shall correspond to the intervals defined n Tab I-1. All Surveillance Requirements shall be performed within the specified time intervytla, ;ximum allowable extension not to exceed 25% of the surveillance interval. The 2 , ension applies to all frequency intervals with the exception of "F.

No extension is allowed f;*n ra 'esignated 'F."

18

Number TMI - Unit 1 Radioloaical Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 1-1 Frequency Notation Notation Frequency S Shiftly (once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months )

D Daily (once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

W Weekly (once per 7 days)

M Monthly (once per 31dO Q Quarterly (once per ,* ys)

S/A Semi-Annuall[ e 84 days)

R Refuelin ervalhce per 24 months)

P S/U Prior to each'jctor startup, if not done during the previous 7 da*

P , ompleted prior to each release N/A (NA) ,L NoP pplicable 6 E (.Once per 18 months F , " Not to exceed 24 months Bases Section 1.13 establishes the limit for hi e pecified time interval for Surveillance Requirements may be extended. It permits an allowable ensio the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operting c d ons that may not be suitable for conducting the surveillance; e.g.,

transient co nditions or other girpg su'eillance or maintenance activities. It also provides flexibility to accommodate the length of le for surveillances that are specified to be performed at least once each REFUELING INTERVAL. It is tended X that this provision be used repeatedly as a convenience to extend surveillance intervals beyond th specified for surveillances that are not performed once-each REFUELING INTERVAL. Likewise, it is not the intent that REFUELING INTERVAL surveillances be performed during power operation unless it is consistent with safe plant operation. The limitation of Section 1.13 is based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

19

6610-PLN-4200.01 Revision 22 FIGURE 1.1 Gaseous Effluent Release Points and Liquid Effluent Outfall Locations

.1 20

Number TMI - Unit I Radiological Controls Procedure 6610-PLN4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.0 RADIOLOGICAL EFFLUENT CONTROLS AND BASES 2.1 Radioactive Effluent Instrumentation 2.1.1 Radioactive Liquid Effluent Instrumentation CONTROL:

The radioactive liquid effluent monitoring instrumentation channels shown in Table 2.1-1 shall be OPERABLE with their alarmftrip setpoints set to ensure that the limits of Control 2.2.1.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the OFFS1TE DOSE CALCULATIrW1INUAL (ODCM).

APPLICABILITY: At all times

ACTION:

a. With a radioactive liquid effluent mo ' rin is umentation channel alarmitrip setpoint less conservative than req* iby th* above control, immediately suspend the release of radioactive liquidIfluent monitored by the affected channel or declare the channel inoperabte.

b.With less than the mil'm number of radioactive liquid effluent monitoring instrumentation channels ABLE, take the ACTION shown in Table 2.1-1. Exe f to return the instrumentation to OPERABLE status within 30 a , unsuccessful, explain in the next Annual Effluent Release Re inoperability was not corrected in a timely manner.

- ilN./k4, and RM-L6, operability is not required when barges are positively controled through the closure of L-V257.

For RM-L1 2 and associated IWTSIIWFS flow interlocks, operability is not required when discharges are positively

- controlled through the closure of IW-V72, 75 and IW-V280, 281.

For FT-146, operability is not required when discharges are positively controlled through the closure of WDL-V257, IW-V72, 75 and IW-V280, 281.

BASES The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluent during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding ten times the effluent concentrations of 10 CFR Part 20.

21

Number TMI- Unit I Title Radiological Controls Procedure j 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 2.1-1 Radioactive Liquid Effluent Instrumentation Minimum Channels Instrument Operable ACTION 1 Gross Radioactivity Monitors Providing Automatic Termination of Release a, Unit I Liquid Radwaste Effluent 120 Line (RM-L6) 1.* 20

b. IWTSIIWFS Discharge Line (RM-L12)

2. Flow Rate Measurement Devices

a. Unit I Liquid Radwaste Effluent (4

.4 Line (FT-84) 1- 21 b- Station Effluent Discharge . 4 1 21 (FT-146)

Tab oaQQ r ACTION 18 With the number of channels OPEFIIABL GtNe than required by the Minimum Channels OPERABLE requirement effluent releasesrq ntin e, provided that prior to initiating a release:

1. At least two in Sare analyzed in accordance with Surveillances 3.2.1.1.1 and1

2. At least ,te 6 qualified members of the Unit staff independently verify the rele rat is and verify the discharge valve lineup.

3. 1~ME ant Manager shall approve each release.

suspend release of radioactive effluents via this pathway-ACTION 20 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement effluent releases via this pathway may commence or continue provided that grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 1x10-7 microcuries/ml, prior to initiating a release and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during release.

ACTION 21 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement radioactive effluent releases via-this pathway may continue, provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

22

Number TMI - Unit 1 Radioloaical Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.1.2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation CONTROL:

The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 2.1-2 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Control 2.2.2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: As shown in Table 2.1-2.

ACTION:

a. With a radioactive gaseous process or effluenreturn instrumentation channel alarm/trip setpoint less cons atinssansu,ired by the above control, immediately suspend the renleas rabive effluent monitored by the affected channel or declare thencn pn rabie.

b. With less than the mienimum n of active gaseous process or effluent monitoring instrumentation cha els OPERABLE. take the ACTiON shown in Table 2.1m-2. b t efforts to return the instrumentation to OPERABLE status withA dea~ys and, if unsuccessful, explain in the next Annual Effluent Relea why the inoperability was not corrected in a timely manner.~ L BASES The radioactive gas~dft Q instrumentation is provided to monitor and control, as applicable, the rel4b of radioactive materials in gaseous effluent during actual or potential relea. 're'armltrip setpoints for these instruments shall be calculated in accordance Ith ;!c arproved methods in the 00CM to provide reasonable assurance that the a ua I ~es are within the limits specified in 10 CER 20.1301.

The w ra ondenser offgas noble gas activity monitors also provide data for d riat! of steam generator primary to secondary leakage rate. Channel operability ra '*r ts are based on an ASLB Order No. LBP-84-47 dated October 31, 1984, and as ci in 20 NRC 1405 (1984).

23

6610-PLN-4200.01 Revision 22 Table 2.1-2 Radioa ctive Gaseous Process and Effluent Monitoring Instrumentation MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILI" FY ACTION I. Waste Gas Holdup System

a. Noble Gas Activity Monitor 1 25

b. Effluent System Flow Rate Iasur 1 26

2. Waste Gas Holdup System Explosive Gas

a. Hydrogen Monitor 2 ** 30

b. Oxygen Monitor 2 **

30

3. Containment Purge Monitoring System a.

b.

Noble Gas Activity Monitor (RM-A9)

Iodine Sampler (RM-A9)

1 27 31

c. Particulate Sampler (RM-A9) 31

d. Effluent System Flow Rate Measuring Device (FR-148) 26

e. Sampler Flow Rate Monitor I 26 0

24

6610-PLN-4200.01 Revision 22 Table 2.1-2 (Cont'd)

Radioactive Gaseous Process and Effluent Monitoring Instrumentation MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

4. Condenser Vent System

a. Low Range Noble Gas Act'fi nitor (RM-A5Lo and 2(1) # 32 Suitable Equivalent)

NOTE (1): For one of the channels, an operae ccha(neray be defined for purposes of this control and 3.1.2.1 only as a suitable equivalent monitoring system capable of being placed in ser-ceitAinpne hour. A suitable equivalent system shall include instrumentation with comparable sensitivity and response time to the RM-A5Lo rnoita'fng eLhapel. When the equivalent monitoring system is in service, indication will be continuously available to the operator, either through indiqat' $d alarm in the Control Room or through communication with a designated individual continuously observing local Indication.

4_

25

6610-PLN-4200.01 Revision 22 Table 2.1-2 (Cont'd)

Radioactive Gaseous Process and Effluent Monitoring Instrumentation MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILI rY ACTION

5. Auxiliary and Fuel Handling Building Ventilation System

a. Noble Gas Activity Monitor - or (RM-A4 and RM-A6) 1

27

b. Iodine Samples (RM-A8) or (R and RM-A6)

31

c. Particulate Sampler (RM-A8) or (RM4-A 6) 31

d. Effluent System Flow Rate Measuring Devicey l149 1

26 and FR-150)

e. Sampler Flow Rate Monitor 1 26

6. Fuel Handling Building ESF Air Treatment System

a. Noble Gas Activity Monitor (RM-A14 or Suitable Equivalent) "I A 27, 33 2)

b. Iodine Cartridge N/A1 31, 33

c. Particulate Filter 31, 33 NI':

d. Effluent System Flow (UR-1 104A/B) 26, 33 e, Sampler Flow Rate Monitor I 26, 33 NOTE 2: No instrumentation channel is provided, However, for determining operability, the equipmei must be installed and functional or the ACTION applies.

26

661 0-PLN-4200.01

.Revision 22 Table 2.1-2 (Cont'd)

Radioactive Gaseous Process and Effluent Monitoring Instrumentation MINIMUM CHANNELS ACTION INSTRUMENT OPERABLE APPLICABILII -Y

7. Chemical Cleaning Building Ventilation System

a. Noble Gas Activity Monitor C M-1-18) 27 31

b. Iodine Sampler (ALC RM-I)

I 31

c. Particulate Sampler (ALC RM-I,,-)

8, Waste Handling and Packaging Facility VeTh i Sy 31

a. Particulate Sampler (WHP-RIT-1) I

9. Respirator and Laundry Maintenance Facility Ventilation System

a. Particulate Sampler (RLM-RM-1) 31 NOTE 3: Channel only required when liquid radwaste is moved or processed 0r 27

Number TMI - Unit I Title Radiological Controls Procedure 661 O-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 2.1-2 Table Notation

At all times.

During waste gas holdup system operation.

Operability is not required when discharges are positively controlled through the closure of WDG-V47 and where RM-A8 (or RM-A4 and RM-A6), FT-149, and FT-1 50 are operable.

During Fuel Handling Building ESF Air Treatment System Operation.

At all times during containment purging.

At all times when condenser vacuum is established.

1. 1. During operation of the ventilation system.

ACTION 25 With the number of channels OPERABLE less than required by the Mi nels OPERABLE requirement, the contents of the tank may be released to the envirno rovided that prior to initiating the release:

1. At least two independent samples of the tank's co ten are nalyzed in accordance with Table 3.2-2, Item A, and

2. At least two technically qualified members of the Un' staff independently verify the release rate calculations and verify the ischarge valve.lineup.

3. The TM[ Plant Manager shall appr hrelease.

Otherwise, suspend release a ve effluent via this pathway.

ACTION 26 With the number of channels OPEBLNe thanrequired by the Minimum Channels OPERABLE requirement, effluent releases *pettpat ay may continue provided the flow rate is estimated at feast once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 27 With the number of cha es OPE BLE less than required by the Minimum Channels OPERABLE requirement, effluent rel s ia this pathway may continue provided grab samples are taken at least once per 12 urs an e initial samples are analyzed for gross activity (gamma scan) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the ch ne has been declared inoperable. If RM-A9 is declared inoperable, see also Technical S tion .5.1, Table 3-5.1, Item C.3.f.

ACTION 30 1. Wit umber of channels OPERABLE less than required by the Minimum Channels OPE BLE requirement, a grab sample shall be collected and analyzed for the inoperable gas channel(s) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With both channels inoperable, a grab sample shall be collected and analyzed for the inoperable gas channel(s):

(a) at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations.

(b) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations (e.g. Feed and Bleed).

28

Number TMI - Unit I Radioloaical Controls Procedure 6610-PLN-4200.01 Titte Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 2.1-2

2. If the inoperable gas channel(s) is not restored to service within 14 days, a special report shall be submitted to the Regional Administrator of the NRC Region I Office and a copy to the Director, Office of Inspection and Enforcement within 30 days of declaring the channel(s) inoperable. The report shall describe (a) the cause of the monitor inoperability, (b) action being taken to restore the instrument to service, and (c) action to be taken to prevent recurrence.

ACTION 31 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that within four hours after the channel has been declared inoperable, samples are continuously collec h auxiliary sampling equipment. - )

ACTION 32 With the number of channels OPERABLE less than required by the m Channels OPERABLE requirement, effluent releases via this pathway may continue for o 2 s, provided that one OPERABLE channel remains in service or is placed in service* * .or. After 28 days, or if one OPERABLE channel does not remain in service or ip no plad*l in service within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the provisions of Technical Specification 3.0.1 apply, *~ this ntrol were a Tech Spec Limiting Condition for Operation.

ACTION 33 With the number of channels OPERABLE les th~ required by the Minimum Channels OPERABLE requirement, either restore the inoperable char ne to OPERABLE status within 7 days, or prepare and submit a special report within 30 days od action(s) taken, the cause of the tf. slthe inoperability, and plans and schedule for tsystem to OPERABLE status.

29

Number TMI - Unit I Radiological Controls Procedure 661 O-PLN-4200.01

-rtie Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2 Radioactive Effluent Controls 2.2.1 Liquid Effluent Controls 2.2.1.1 Liquid Effluent Concentration CONTROL:

The concentration of radioactive material released at anytime from the unit to unrestricted areas shall be limited to ten'times the concentrations specified in 10 CFR Part 20.1001-20.2401, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained nobleifa-q. For dissolved or entrained noble gases, the concentration shall be to x 1iO. uCi/cc total activity.

APPLICABILITY: At all times ACTION: A With the concentration of radioactive ma aI released from the unit to unrestricted areas exceeding the above ffmits, immediately restore concentrations within theove limits BASES This control is pr4 r nsure that the concentration of radioactive materials rel n n1' id waste effluent from the unit to unrestricted areas will be lest n ten limes the concentration levels specified in 10 CFR Part 20.1001 .0Q41 pendix B, Table 2. This limitation provides additional assue wiat the levels of radioactive materials in bodies of water outside thAe -will result in exposures with (1) the Section II.A design objectives o*tgpe~dixI, 110 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the mis of 0 CFR Part 20.1301 to the population. The concentration limit for Q XnobIlases is based upon the assumption the Xe-1 35 is the controlling "iisotope

. and its MPC in air (submersion) was converted to an equivalent

ncentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

2.2.1.2 Liquid Effluent Dose CONTROL The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to the SITE BOUNDARY shall be limited:

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

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

30

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 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, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar quarters so that y'"q~ulative dose or dose commitment to any individual fr uch rleases during these four calendar quarters is wit m the total body and 10 mrem to any organ. Thi; I1Report shall also include (1) the result of radiol a es of the drinking water source, and (2) the radiologioaýýct on finished drinking water supplies with regard to te req e ents of 40 CFR 141, Safe Drinking Water Act.

BASES Thiscontrol and associ .action is provided to implement the requirements of Sections II.A, IIIA, Ann f Appendix 1, 10 CFR Part 50. The Control implements the geeet foh in Section ILA of Appendix 1. The ACTION statements proviC-ET-qjired operating flexibility and at the same time implement th e&*e t forth in Section IV.A of Appendix I to assure that the releases of r dioac v( material in liquid effluents will be kept "as low as is reason I""yle'. Also, for fresh water sites with drinking water suppli!1ich can be potentially affected by plant operations, there is re na ble urance that the operation of the facility will not result in io lideconcentrations in the finished drinking water that are in excess of th 'Teq,Jements of 10 CFR 20. The dose calcuJations in the ODCM

/\imp*l*ent The requirements in Section I[I.A. of Appendix I that conformance

. X(the guides of Appendix I is to be shown by calculational procedures Ilsed on models and data such that the actual exposure of a MEMBER OF "THEPUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,"Revision 1, October, 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April, 1977.

NUREG-01 33 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113.

31

Ii--L Number TMI - Unit 1 Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.1.3 Liquid Radwaste Treatment System CONTROL:

The appropriate portions of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the unit to unrestricted areas would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in any calendar month.

APPLICABILITY: At all times ACTION: hQa

a. With radioactive liquid waste bein i arged without treatment and in excess of the above limi per and submit to the NRC Region I Administrator withi s Special Report which includes the following i ~ma n:

1. Explanatidno liquid radwaste was being discharged witbdit treatment, identification of any inoperable equipment or subsystems, and the reason A inoperability,

2. Acti s) taken to restore the inoperable equipment

"\,*-4QOPERABLE status, and, Q % A summary description of action(s) taken to prevent a recurrence.

T r irement that the appropriate portions of this system be used, when "spe d, provides assurance that the releases of radioactive materials in id effluents will be kept as low as is reasonably achievable. This control

S plements the requirements of 10 CFR Part 50.36a, General Design "Criterion60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The intent of Section Il.D. is to reduce effluents to as low as is reasonably achievable in a cost effective manner. This control satisfies this intent by establishing a dose limit which is a small fraction (25%) of Section Il.A of Appendix 1,10 CFR Part 50 dose requirements. This margin, a factor of 4, constitutes a reasonable reduction.

32

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01

Title Revision No.

Offsite Dose Calculation Manual (ODCM) 1 22 2.2.1.4 Liquid Holdup Tanks CONTROL The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases..

a. Outside temporary tank APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive in any of the above listed tanks exceeding the aboveli W m tely suspend all additions of radioactive ma e ank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank content* wtne limit.

BASES Restricting the quantity of dioactive material contained in the specified tanks provides assurance tha in e event of an uncontrolled release of the tanks' contents, the resulting co t tions would be less than the limits of 10 CFR Part 20.1001-20-2*"T1, A endix B, Table 2, Column 2, at the nearest potable water su e nearest surface water supply in an unrestricted area.,

2.2.2 Gaseous Effluent C 0l CI 2.2.2.1 Ga .o E Int Dose Rate S dose rate due to radioactive materials released in gaseous effluent from site shall be limited to the following:

"a. For noble gases: less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mremfyr to the skin, and

b. For 1-131, 1-133, tritium and all radionuclides in particulate form with half lives greater than 8 days: less than or equal to 1500 mremlyr to any organ.

APPLICABILITY: At all times.

ACTION:

With the release rate(s) exceeding the above limits, immediately decrease the release rate to comply with the above limit(s).

33

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 T199e Revision No.

Offsite Dose Calculation Manual (ODCM) 22 BASES The control provides reasonable assurance that the annual dose at the SITE BOUNDARY from gaseous effluent from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas while providing sufficient operational flexibility in establishing effluent monitor setpoints.

These gaseous release rates provide reasonable assurance that radioactive material discharged in gaseous effluent will not result in the exposure of a MEMBER OF THE PUBLIC in an unrestricted area, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the values specified in Appendix B, Table 2 of 10 CFR Part 20. For MEMBERS OF THE PUBLIC who may at times be within the SIT N jt.UNDARY, the occupancy of the MEMBER OF THE PUBLIC will besufficinnty low to compensate for any increase in the atmospherin h I nctor above that for the exclusion area boundary. The specified r rate limits restrict, at all times, the corresponding gamma and beta se e above background to a MEMBER OF THE P3BLIC at or beyon E rOUNDARY to less than or equal to 500 mrem/year to the tota o or ess than or equal to 3000 mremlyear to the skin. These relea rate Ii lts also restrict, at all times, the corresponding thyroid dose rate above ,ground to a child via the inhalation pathway to less than or equalt 1500 mremlyear (NUREG 0133).

2.2.2.2 Gaseous Effluents Dose- ble Gases CONTROL: A The air dose 0 e gases released in gaseous effluents from the unit to areas at a bee nthe SITE BOUNDARY shall be limited to the following:

a. During any calendar quarter: less than or equal to 5 mrad for

'Nqi]mma radiation and less than or equal to 10 mrad for beta 4fadiaticn and,

b. During any calendar year: Jess than or equal to 10 mrad for Lgamma radiation and less than or equal to 20 mrad for beta

.'% radiation.

N,>' APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

34

Number STMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 BASES This control applies to the release of radioactive materials in gaseous effluents from TMi-1.

This control and associated action is provided to implement the requirements of Section IB, III.A and lV.A of Appendix 1,10 CFR Part 50. The Control implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance RequiremqMM*plement the requirements in Section llI.A of Appendix I that co olran~c with the guides of Appendix I be shown by calculational proced bksaw.dn models and data such that the actual exposure of a MEM* THE PUBLIC through the appropriate pathways is unlikely to be s tan underestimated. The dose calculation methodology and pararpste,. tallished in the ODCM for calculating the doses due to the actuj rel e es of radioactive noble gases in gaseous effluents are con~ ý nt wi he methodology provided in Regulatory Guide 1.109, "Calculation oj ual Doses to Man from Routine Release of Reactor Effluents for the Purlose of Evaluating Compliance with 10 CFR Part 50, Appendix " Revision 1, October 1977 and Regulatory Guide 1. 111, "Methods o stimating Atmospheric Transport and Dispersion of Gaseous Effluents I ne Releases from Light-Water Cooled Reactors," Revisjlo ,ýuly 1i*. The ODCM equations provided for determining the Ao and beyond the SITE BOUNDARY are based upon the histo,, ra e atmospheric conditions. NUREG-01 33 provides methods.for ose Ic consistent with Regulatory Guides 1.109 and 1.111.

2.2.2.3 Dos - di - 31, Iodine-133, Tritium, and Radionuclides In Particulate Form he dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133,

ium, and all radionuclides in particulate form with half lives greater than 8 d ays, in gaseous effluents released from the unit to areas at and beyond the SITE BOUNDARY shall be limited to the following:

a. During any calendar quarter. less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

35

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 "Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 ACTION:

With the calculated dose from the release of Iodine-1 31, lodine-133, Tritium, and radionuclides in particulate form with half lives greaterthan 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

BASES This control applies to the release of radioactivete Z gaseous effluents from TMI-1.

This control and associated action is prrement the requirements of Section Il.C, III.A and IV.A of Appe ix ,10 R Part 50. The Controls are the guides set forth in Section UE.f Appp dix I. The ACTION statement provides the required operating flexibilit 0d at the same time implements

.the guides set forth fn Section IV.A of A ~endix ]to assure that the releases of radioactive materials in 2aseous effluents will be kept "as low as is reasonably achievable." ODCM calculational methods specified in the surveillance requireme l1 ialment the requirements in Section IfI.A of Appendix I that co ance;th the guides of Appendix I be shown by calculational pro 4ed on models and data such that the actual exposure of a~Ml1 OF THE PUBLIC through appropriate pathways is unlikely to b subs?9n lly underestimated. The ODCM calculational methodo d..rameters for calculating the doses due to the actual reieasj'tes of-fe subject materials are consistent with the methodology pro~e in ýgulatory Guide 1.109, "Calculation of Annual Doses to Man "utin*Releases of Reactor Effluents for the Purpose of Evaluating lia ce with 10 CFR Part 50, Appendix I," Revision 1, October, 1977 and

,Reg* ry Guide 1.111, "Methods for Estimating Atmospheric Transport and is ersion of Gaseous Effluents in Routine Releases from

" h*t-Water-Cooled Reactors," Revision 1, July, 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for iodine-131, iodine-1 33, "tritiumand radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in areas at and beyond the SiTE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

36

Number TMI -Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) .22 2.2.2.4 Gaseous Radwaste Treatment System CONTROL The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in the gaseous waste prior to their discharge when the monthly projected gaseous effluent air doses due to untreated gaseous effluent releases from the unit would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM s lliused to reduce radioactive materials in gaseous waste prior to the otscharle when the monthly projected doses due to gaseous efflue e ea,,om the site would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the GASEOUS RADW TE TREATMENT SYSTEM and/or the VENTILAT J N EXHAUST TREATMENT SYSTEM inoperable for more thap amonth or with gaseous waste being discharged without tre a d in excess of the above limits, prepare and submitý"e NR egion I Administrator within 30 days, a SpeciKJ9pwhich includes the following information:

Identification of the inoperable equipment or subsystems and the reason for inoperability, Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. A summary description of action(s) taken to prevent a recu rrence.

'S 4N< BASES dyl The use of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that gaseous effluents are treated as appropriate prior to release to the environment. The appropriate portions of this system provide reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section 1I.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and I .C of Appendix 1,10 CFR Part 50, for gaseous effluents.

37

-I-Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.2.5 Explosive Gas Mixture CONTROL The concentration of oxygen in the Waste Gas Holdup System shall be limited to less than or equal to 2% by volume whenever the concentration of hydrogen in the Waste Gas Holdup System is greater than or equal to 4% by volume.

AVAILABILITY: At all times.

ACTION:

Whenever the concentration of hydrogen in the a Holdup System is greater than or equal to 4% by volume, and:

a. The concentration ofoxygen-te Gas Holdup System is greater than 2% by volu , b less han 4% by volume, without delay begin to reduce xygei¶concentration to within its ifmit.

b. The concentration of oxyge in the Waste Gas Holdup System is greater than oqequal to 4% by volume, immediately suspend additions Of te gas to the Waste Gas Holdup System and without de *a i to reduce the oxygen concentration to within its limn I

BASES:

Based ax4,$ntal data (Reference 1), lower limits of flammability for hydror.. 5% and for oxygen is 5% by volume. Therefore, if the copnn ati f either gas is kept below it lower limit, the other gas may be p(4sa&, n hlgher amounts without the danger of an explosive mixture.

M tai~ ng the concentrations of hydrogen and oxygen such that an explae mixture does not occur in the waste gas holdup system provides rance that the release of radioactive materials will be controlled in n fformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR 50.

REFERENCES (1) Bulletin 503, Bureau of Mines; Limits of Flammability of Gases and Vapors.

38

Number TMI - Unit I Radiological Controls Procedure 0o0'lU-r-LP.-4.UU.U7 "Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.2.6 Waste Gas Decay Tanks CONTROL:

The quantity of radioactivity contained in each waste gas decay tank shall be limited to less than or equal to 8800 curies noble gases (considered as Xe-133).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive matedi 1 any aste gas decay tank exceeding the above limit, im idt ,*Apend all additions of radioactive material to the tank ithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limi e ll BASES Restricting the quantity of radioactivity Nined in each waste gas decay tank provides assurance that in the eve of an uncontrolled release of the tanks contents, the resultinj total body exposure to a MEMBER OF THE PUBLIC at the nearest :xcljsion area boundary will not exceed 0.5 rem. This is consistent with Stan vew Plan 15.7.1, '"Waste Gas System Failure."

2.2.3 Total Radioactive Effluent Co4 7 2.2.3.1 Total Dose CONF<lna T e al 'calendar year) dose or dose commitment to any MEMBER OF T P LIC, due to releases of radioactivity and to radiation from uranium fuel e sources shall be limited to less than or equal to 25 mrem to the total t

or any organ except the thyroid, which shall be limited to less than or ual to 75 mrem.

"APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls 2.2.1.2.a, 2.2.1.2.b, 2.2.2.2.a, 2.2.2.2.b, 2.2.2.3.a, or, 2.2.2.3.b, calculations should be made including direct radiation contributions from the unit and from outside storage tanks to determine whether the above limits of Control 2.2.3.1 have been exceeded. If such is the case, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which 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. This Special Report, as defined in 39

-1L_

Nuenber TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 10 CFR Part 20.2203(b), shall include an analysis which estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceed the above limits, and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

BASES This control is provided to meet the dose Ii tion an 40 CFR Part 190 that have been incorporated into 10 CFR Pa .1*(d}. This control requires the preparation and submittal of a Sp ial po whenever the calculated doses from plant generated radioa eefflue'Ms and direct radiation exceed 25 mrem to the total body or any organ, r pt the thyroid, which shall be limited to less than or equal to 75 mrem. or sites containing up to 4 reactors, it is highly unfikelrthat the resultant dose to a MEMBER OF THE PUBLIC will exceed thejdT* limits of 40 CFR Part 190 if the individual reactors remain within tmt~ e dose design objectives of Appendix I, and if direct radiation dos W*om th feactor units and outside storage tanks are kept small. The S*cj ort will describe a course of action that should result in the llro&' rfofhe annual doseto a MEMBER OF THE PUBLIC to within the 40 FR 9190 limits. For.the purposes of the Special Report, it may be d t the dose commitment to the member of the public from other um cycle sources is negligible, with the exception that dose con bti o m other nuclear fuel cycle facilities at the same site or within a s of m must be considered. If the doseto any member of the public is* ed to exceed the requirements of 40 CFR Part 190, the Special SRet ith a request for a variance (provided the release conditions resulting "violationof 40 CFR Part 190 have not already been corrected), in

.0 rdance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 0..2203(b), is considered to be a timely request and fulfills the requirements of 40 CFR Part 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 10 CFR Part 20, as addressed in Controls 2.2.1.1 and 2.2.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which helshe is engaged in carrying out any operation that is part of the nuclear fuel cycle.

40

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.0 SURVEILLANCES 3.1 Radioactive Effluent Instrumentation 3.1.1 Radioactive Liquid Effluent Instrumentation SurveiUance Requirements 3.1.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, AND CHANNEL TEST operations during the MODES and at the frequencie n in Table 3.1-1.

A-Y 0*

41

6610-PLN-4200.01 Revision 22 Table 3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL SOURCE CHANNEL CHANNEL INSTRUMENT CHECK CHECK CALIBRATION TEST Radioactivity Monitors Providing Alarm and Automatic Isolation

a. Unit I Liquid Radwaste E nt ine (RM-L-6) D P R(2) b- IWTS/IWFS Discharge LinRM-L D P R(2) Q(1)

2. Flow Rate Monitors

a. Unit 1 Liquid Radwaste Effluent Lineýý4)^ D(3) N/A R Q

b. Station Effluent Discharge (FT-146) D(3) N/A R Q 0

42

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (DCM) 22 Table 3.1-1 Table Notation (1) The CHANNEL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the following condition exists:

1. Instrument indicates measured levels above the high alarm/trip setpoint. (Includes - circuit failure)

2. Instrument indicates a down scale failure. (Alarm function only.) (Includes - circuit failure)

3. Instrument controls moved from the operate mode (Alarm function only).

(2) The initial CHANNEL CALIBRATION for radioactivity measurement instrumen. Dn\Qj35e performed using one or more of the reference standards certified by the National Institute of rds and Technology or using standards that have been obtained from suppliers that participated r ment assurance activities with NIST. These standards should permit calibrating the sy r i intended range of energy and measurement range. For subsequent CHANNEL CALIBRATIO), s c hat have been related to the initial calibration should be used. (Operating plants may substituteviou established calibration procedures for this requirement)

(3) CHANNEL CHECK shall consist of verifying indication of ow during periods of release. CHANNEL CHECK shall be made at least once daily on any day on which co tinuous, periodic, or batch releases are made.

0 43

Number TMI - Unit 1 Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No Offsite Dose Calculation Manual (ODCM) 22 3.1.2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation Surveillance Requirements 3.1.2.1 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations at the frequencies shown in Table 3.1-2.

0 0

44

6610-PLN-4200.01 Revision 22 Table 3.1-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL SOURCE CHANNEL CHANNEL INSTRUMENT CHECK CHECK CALIBRATION TEST APPLICABILITY

1. Waste Gas Holdup System

a. Noble Gas Activity Monitor ( -A7) P P E(3) Q(1)

b. Effluent System Flow R ; ur! Device (FT-123) P N/A E Q

2. Waste Gas Holdup System Explosive a onitoring System

a. Hydrogen au Monitor ,5? D N/A Q(4) M **

b. Oxygen Monitor D N/A Q(5) M

3. Containment Purge Vent System Q

a. Noble Gas Activity Monitor (RM-AQ) D P E(3) M(1)

b. Iodine Sampler (RM-A9) N/A N/A N/A

c. Particulate Sampler (RM-A9) N/A N/A N/A

d. Effluent System Flow Rate Measuring Device (FR-148) D E Q

e. Sampler Flow Rate Monitor D E N/A

4. Condenser Vent System

a. Noble Gas Activity Monitor (RM-A5 and Suitable D Q(2)

Equivalent - See Table 2.1-2, Item 4.a) 45

6610-PLN-4200.01 Revision 22 Table 3.1-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL SOURCE CHANNEL CHANNEL INSTRUMENT CHECK CHECK CALIBRATION TEST APPLICABILITY

5. Auxiliary and Fuel Handling Building Ventilation System

a. Noble Gas Activity Monitor(Iý-A8) or (RM-A4 and, D M E(3) Q(0)

RM-A6) N

b. Iodine Sampler (RM-A8) orXM-A4 RM-A6) W N/A NtA N/A

c. Particulate Sampler (RM-AB) or M-A44 .RM-A6 W N/A I)

N/A N/A

d. System Effluent Flow Rate Measure "ec N/A E Q (FR-1 49 and FR-1 50) D eP Sampler Flow Rate Monitor ( D N/A E N/A

6. Fuel Handling Building ESF Air Treatment System

a. Noble Gas Activity Monitor (RM-A14) M R(3) Q(2)

b. System Effluent Flow Rate *(UR-1104 A/B) NIA R Q 0

c. Sampler Flow Rate Measurement Device Q 0

46

6610-PLN-4200.01 Revision 22 Table 3.1-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL SOURCE CHANNEL CHANNEL INSTRUMENT CHECK CHECK CALIBRATION TEST APPLICABILITY

7. Chemical Cleaning Building Ventilation System

a. Noble Gas Activity Monitor (AA RIM-1-18) D M E(3) Q(2)

b. Iodine Sampler (ALC RM-I ,V W N/A N/A N/A

c. Particulate Sampler (ALC RM-I W N/A N/A N/A

8. Waste Handling and Packaging Facility Ventikloln;e)s

a. Particulate Sampler (WHIP-RIT-I) X/ D w SA W

9. Respirator and Laundry Maintenance Ventilation System

a. Particulate Sampler (RLM-RM-1) w SA W ruýý 0

47

I-Number TM! - Unit I "Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 3.1-2 Table Notation

At all times.

- During waste gas holdup system operation.

Operability is not required when discharges are positively controlled through the closure of WDG-V47, and where RM-A8 (or RM-A4 and RM-A6), FT-149, and FT-150 are operable.

During Fuel Handling Building ESF Air Treatment System Operation.

At all times during containment purging.

1. At all times when condenser vacuum is established.

1. 1. During operation of the ventilation system.

(1) The CHANNEL TEST shall also demonstrate that automatic isolation of this p a*N e Auxiliary and Fuel Handling Building Ventilation System, the supply ventilation is isolated ntrol room alarm annunciation occurs if the following condition exists:

I. Instrument indicates measured levels above the high alar ip (Includes circuit failure).

2. Instrument indicates a down scale failure (Alarm function Illudes circuit failure).

3. Instrument controls moved from the operate mod 4 (Alarm function only).

(2) The CHANNEL TEST shall also demonstrate that coretmor alarm annunciation occurs if any of the following conditions exist:

1 Instrument indicates measured ]eve ol t e alarm setpoint. (includes circuit failure)

2. Instrument indicates a down s ure ncludes circuit failure).

3. Instrument controls move r tf* erate mode.

(3) The initial CHANNEL CAB fr radioactivity measurement instrumentation shall be performed using one or more of the refere stand $ certified by the .National Institute of Standards and Technology or using standards that hye be o~tained from suppliers that participate in measurement assurance activities with NIST. These st~pr1ds sltuId permit calibrating the system over its intended range of energy and measurement rang . sequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shouldtj used. (Operating plants may substitute previously established calibration procedures for this requirement.)

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

I. One volume percent hydrogen, balance nitrogen, and

2. Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

1. One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

48

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2 Radiological Effluents 3.2.1 Liquid Effluents SURVEILLANCE REQUIREMENTS 3.2.1.1 Concentration 3.2.1.1.1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 3.2-1. The results of pre-release analyses shall be used with the calculational metyl*the ODCM to assure that the concentration at the p*r of ref ase is maintained within the limits of Control 2.2.1.1.

3.2.1.1.2 Post-release analysis of samplAo0 eited from batch releases shall be performed in acco able 3.2-1. The results of the previous post-releas na is s all be used with the calculational methods ' OD(ft4 to assure that the concentrations at the point blease were maintained within the limits of Control 2.2.1.1.

3.2.1.1.3 The radioa ivi concentration of liquids discharged from continuous7r'k e,,points shall be determined by collection and analy saml:* in accordance with Table 3.2-1. The results of thd hall-be used with the calculational methods of the sure that the concentration at the point of release is 3.2.1.2 Does within the limits of Control 2.2.1.1.

3.2.1 .2 Dose ain 33.1 umulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least once a month.

3;* quid Waste Treatment 3.2.1.3.1 Doses due to liquid releases shall be projected at least once a month, in accordance with the ODCM.

3.2.1.4 Liquid Holdup Tanks 3.2.1.4.1 The quantity of radioactive material contained in each of the tanks specified in Control 2.2.1.4 shall be determined to be within the limit by analyzing a representative sample of the tanks content weekly when radioactive materials are being added to the tank.

49

6610-PLN-4200,01 Revision 22 Table 3.2-1 Radioactive Liquid Waste Sampling and Analysis Program I

Sa8 Lower Limit of mpling Minimum Analysis -4 Type of Activity Detection (LLD)

Liquid Release Ty'pe Freaquency , Frequency Analysis (jiCi/mi) (Note a)

A.1 Batch Waste Release Tanks (Nco d) _I P , P H-3 Eacih Batch 1 x10° Each Batch Principal Gamma 5 x 10-7 Emitters (Note f) 1-131 1x iO-6 4

DiWsolved and 1 x 10-Entrained Gases (Gamma Emitters)

I P Q Gross alpha I x10"'

iBatch Composite (Note b)

Sr-89, Sr-90 8 6 x10" Fe-55

,e-5 ... .. xl... .. ,

A.2 Continuous Releases (Note e) " Principal Gamma Composite (Note c Emitters (Note I 5 x.10-7 (Note c) 1-131 1 x 1' Dissolved arid Entrained G~ ises (Gamma Emitters)

Continu *H-3 (Note 0]

ous 1 Xl10-7 ous Q -+Gross alpha 5lx 10"'

Continu (Note c) Composite *r9, Sr-90 I x t06 (Note c)

Fe-55 'Ix 10.

50

Number TMI - Unit I Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 3.2-1 Table Notation

a. The LLD is defined, for purposes of this surveillance, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD=

L4.66 Sb (j ExVx2.22 x 106 x Y x exp (-XAt) C Where:

LLD is the "a priori" lower limit of detection as defined abov(as urie per unit mass or volume),

Sb is the standard deviation of the background counting rate o of the counting rate-of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per integ n),

V is the sample size (in units of mass or v me,..

2.22 x 106 is the number of disn on per minute per microcurie, Y is the fractional radiochem I

I (when applicable),

X is the radioactive deant for the particular radionuclide, and At is the elapsed metween midpoint of sample collection and time of counting.

Typical valu , and At shall be used in the calculation.

It should be rec 'nized that the LLD is defined as an "a priorf' (before the fact) limit representing the capability of a measurement system and not as an "a posteriorP" (after the fact) limit for a particular measurement.

b. - A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged andin which the method of sampling employed results in a specimen which is representative of the liquids released.

c. To be representative of the quantities and concentrations of radioactive materials in liquid effluent, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

51

Ii-Number TMI - Unit I Tit~le T Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 3.2-1

d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and be thoroughly mixed, by a method described in the ODCM, to assure representative sampling.

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 the continuous release.

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.

This list does not mean that.only these nuclides are to be considered. Othtf `gma peaks that are identifiable, together with those of the above nuclides, shall also be analz and ported in the Annual Radioactive Effluent Release Report pursuant to TS 6.9.4.

<0 .

52

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2.2 Gaseous Effluents SURVEILLANCE REQUIREMENTS 3.2.2.1 Dose Rates 3.2.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of Control 2.2.2.-1a in accordance with the methods and procedures of the ODCM.

3.2.2.1.2 The dose rate of radioactive materials, other than noble gases, in gaseous effluents shall be determined tql59Xthin the limits of Control 2.2.2.1.b in accordance with .e. ods ind procedures of the ODCM by obtaining represen and performing analyses in accordance with the s§Jing and analysis program, specified in Table 3.2-2.

3.2.2.2 Dose, Noble Gas "

3.2.2.2.1 Cumulative dose contribut'l, /from noble gas effluents for the current calendar quarter an current calendar year shall be determined in accordance with the OFFSiTE DOSE CALCULAT MANUAL (ODCM) monthly.

3.2.2.3 Dose, Iodine-131, lvl 3 ritium, and Radionuclides In Particulate Form 3.2.2.3.1 a dose contributions from Iodine-131, Iodine-1 33, jritiurI a d radionuclides in particulate form with half lives athan 8 days for the current calendar quarter and current

< calendar year shall be determined in accordance with the FSITE DOSE CALCULATION MANUAL (ODCM) monthly.

3.2.2.4 G eo YWaste Treatment 1 2.4.1 Doses due to gaseous releases from the unit shall be projected monthly in accordance with the ODCM.

Explosive Gas Mixture 3.2.2.5.1 The concentrations of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the limits of Control 2.2.2.5 by monitoring the waste gases in the Waste Gas Holdup System with the hydrogen and oxygen monitors covered in Table 2.1-2 of Control 2.1.2.

53

II Number TMI- Unit 1 Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2.2.6 Waste Gas Decay Tank 3.2.2.6.1 The concentration of radioactivity contained in the vent header shall be determined weekly. If the concentration of the vent header exceeds 10.7 Ci/cc, daily samples shall be taken of each waste gas decay tank being added to, to determine if the tank(s) is less than or equal to 8800 Cl/tank.

54

6610-P LN-4200.01 Revision 22 Table 3.2-2 Radioactive Gaseous Waste Sampling and Analysis Program

'Min!mum ,Lower Limit of Sampling Analysis : Type of Activity Detection (LLD)

Gaseous Release Type Frequency Frequency Analysis , j.Cilml) (Note a)

Sp ;

P P Principal Gamma IX10 4 A. Waste Gas Gab Saml Each Tank , Emitters (Note g)

Decay Tank Grab Sample B. Containment P (Note b) Each H-3 1x 1 Purge Each Purge Grab Purge Principal Gamma 1x 10 Sample Emitters (Note g)

C. Auxiliary and M M(Notesc e) H-3 Fuel Handling Building M Principal Gamma x 10 Air Treatment System Emitters (Note g) mEmitters (Note g)

F. Fuel Handling Building A TaeSsm M (during H-3 ESF Air Treatment System ration) System Pdncipal Gamma I x t0 H. Renspriaor undryMaintenance FacilityiSeeGSection x 10Se AirTretme Sy)stem ntmp ofGrb Stabple of thisote of Emitters (Note g)of t table F. ChmialCeain Bilig irretmn stm , M (Notee h) , H-3 1 x 10-1 PupExas GrbSmle(oeh Principal Gamma I .,0 (Nt Grbhap) Emitters (Note g)' x10 F ChmclCennBuligAir Treatment System hi Mfal(Not

) ' H-3o thx tale , othsta 1f Gra Samplee H. Respirator and Laundry Maintenance Facility ; See Section I See Sectlio

See Section I See Section I Air Treatment System , of this table Iof this table , of this table of this table 55

6610-PLN-4200.01 Revision 22 Table 3.2-2 Radioactive Gaseous Waste Sampling and Analysis Program Minimum ,Lower Limit of Sampling , Analysis Type of Activity Detection (LLD)

Gaseous Release Type Frequency Frequency Analysis , (g.Ci/ml) (Note a)

All Release Types as Listed Above i B, C, D, F, G, Continuous W (Note d) and H (During System Operaon CoCharcoal (Note f) 1-131 1 x 1012 Sample Continuous W (Note d) Principal Gamma

,<'" ** (Note I

' Particulate Q I, Emitters (Note g)

(1-131, Others) I x 10"11 Continuous Composite G (Noe PrtculteGross Alpha (Note 0Particulate 1 x 10-11 Sample

. (* te t) Particulate0inuous

' Sr-89, Sr-90 1 x 10"1 i'*

',.t,* ..... Sample '

J. Condenser Vent Stack Continuous Iodine W(oNotuue d) ,. 1 10-12 Sampler (Note j) Charcoal' 1 S (N te k)u, Caroa

~SampleQ 56

Numl>er TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Tite Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Table 3.2-2 Table Notation

a. The LLD is defined, for purposes of this surveillance, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD 4.66Sb E x V x 2.22 x 106 x Y x exp (-4*At)

Where: ro_ a LLD is the "a priorn lower limit of detection as defined ab a urie per unit mass or volume),

Sb is the standard deviation of the background cajnting rate or of the counting rate of a blank sample as appropriate (as counts per minute*.*

E is the counting efficiency (as counts Itegon),

V is the sample size (in units of m 0 e),

2.22 x I 6 is the number of di per minute per microcurie, Y is the fractional radiocnical (when applicable),

X is the radioactiv" ecaytant for the particular radionuclide, and At is the elaps* time een midpoint of sample collection and time of counting.

Typical valu f_ Y and At shall be used in the calculation.

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.

b. Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

c. Tritium grab samples from the spent fuel pool area shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

57

SI I Number Radiological - Unit I Procedure TMIControls 6610O-PLN-4200.O'!

Tit e Revision No.

Offite Dose Calculation Manual (ODCM) 22 Table 3.2-2

d. Charcoal cartridges and particulate filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

e. Tritium grab samples shall be taken weekly from the spent fuel pool area whenever spent fuel is in the spent fuel pool.

f. The ratio of the sample flow rate to -the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls 2.2.2.1, 2.2.2.2, and 2.2.2.3.

g. The principal gamma emitters for which the LLD specification applies eX1 (Zvely re the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 Ags~Llmissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-137, Ce-I41 and C or particulate emissions.

This list does not mean that only these nuclides are to be consid d. r gamma peaks that are identifiable, together with those of the above nuclides, shall aI led and reported in the Annual Radioactive Effluent Release Report pursuant to T6.9.

h. Applicable only when condenser vacuum is established. San g and analysis shall also be performed following shutdown, startup, or a THERMAL POW change exceeding 15 percent of RATED THERMAL POWER within one hour uniss (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the pri- ry~oolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows 1]b~lt ent activity has not increased by more than a factor of 3. 7

j. Gross Alpha, Sr-89, and Sr-90 anao d t apply to the Fuel Handling Building ESF Air Treatment System.

j. If the Condenser Vent Stac Sinuous Iodine Sampler is unavailable, then alternate sampling equipment will be placedia srvi *ithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or a report will be prepared and submitted within 30 days from the time te pieple is found or made inoperable which identifies (a). the cause of the inoperability, (b) thcti taI n to restore representative sampling capability, (c) the action taken to prevent recurr , an quantification of the release via the pathway during the period and comparison to he Ii rescribed by Control 2.2.2.1.b.

k. Applicable condenser vacuum is established.

1. Applicable when liquid radwaste is moved or processed within the facility.

m. Iodine samples only required in the Chemical Cleaning Building when TMI-1 liquid radwaste is stored or processed in the facility.

58

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2.3 Total Radioactive Effluents 3.2.3.1 Dose Calculation 3.2.3.1.1 Cumulative annual dose contributions from liquid and gaseous effluents shall be determined in accordance with Surveillances 3.2.1.2.1, 3.2.2.2.1, and 3,2.2.3.1, including direct radiation contributions from the Unit and from outside storage tanks, and in accordance with the methodology contained in the ODCM.

OA-

<2i 59

LJ Number TMI - Unit I Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.0 PART IREFERENCES 4.1 Title 10, Code of Federal Regulations, "Energy" 4.2 Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routing Releases of Reactor Effluents forthe Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, "Revision 1, October 1977 4.3 TMI-1 Techniýal Specifications, attached to Facility Operating License No. DPR-50 4.4 TMI-1 FSAR A0 4>

60

0 0 -

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 PART II Definitions 1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout Part II of the ODCM.

PDMS 1.2 Post-Defueling Monitored Storage (PDMS) is that condition where TM I has been completed, the core debris removed from the reactor during the clea nperiod has been shipped off-site and the facility has been placed in a stable, safe and secu on .i ACTION 1.3 ACTION shall be those additional requirements specified as c ry statements to each control and shall be part of the controls.

OPERABLE - OPERABILITY 1.4 A system, subsystem, train, component o ice be OPERABLE or have OPERABILITY when it is capable of performing its specified fdI Implicit in this definition shall be the assumption that all necessary attendant instrumn , rontrols, norma] and emergency electrical power sources, cooling or seal water, lubrition r ether auxiliary equipment, that are required for the system, subsystem, train, com or vice to perform its function(s), are also capable of performing their related sup nction s).

(CHANNEL CALIBRATION 1.5 A CHANNEL CA ATLO all be the adjustment, as necessary, of the channel output such that it responds wit nece range and accuracy to known values of the parameter which the channel monitors. T' a*AN CALIBRATION shall encompass the entire channel including the sensor and alarm a t y functions, and shall include the CHANNEL FUNCTIONAL TEST. CHANNEL CALIBRATIO N~y be performed by any series of sequential, overlapping or total channel steps such that the en re channel is calibrated.

CHANNELCHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

62

Number TMI - Unit 1 Radioloaical Controls Procedure 6610-PLN-4200.01 Title IRevision No.

Offsite Dose Calculation Manual (ODCM) 22 CHANNEL FUNCTIONAL TEST 1.7 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. Bistable channels - the injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions-SOURCE CHECK 1.8 A SOURCE CHECK shall be the qualitative assessment of channel sensor is exposed to a radioactive source.

COMPOSITE SAMPLE 1.9 A COMPOSITE SAMPLE is a combination of individual s es o ned at regular intervals over a time period. Either the volume of each individual sample is pr*-tional to the flow rate discharge at the time of sampling or the number of equal volume samples i~proportional to the time period used to produce the composite. 4 GRAB SAMPLE 1.10 AC ,RAB SAMPLE is an individual in less than fifteen minutes.

BATCH RELEASE 1.11 A BATCH RELEASE is the CONTINUOUS RELEASE 1.12 A CONTINUOU EASVrs the discharge of fluid waste of a non-discrete volume, e.g., from a volume or sysm thaN/s an input flow during the CONTINUOUS RELEASE.

SITE B()UNDARY 1.13 'The SITE BOUNDARY used as the basis for the limits on the release of gaseous effluents is as defined in Section 2.1.2.2 and shown on Figure 2.1-3 of the TMI-1 FSAR. This boundary line includes portions of the Susquehanna River surface between the east bank of the river and Three Mile Island and between Three Mile Island and Shelley Island.

The SITE BOUNDARY used as the basis for the limits on the release of liquid effluents is as shown in Figure 1.1 in Part I of this ODCM.

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. All Surveillance Requirements shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.

63

ill Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 1.1 Frequency Notation NOTATION FREQUENCY S (Shiftly) At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D (Daily) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W (Weekly) At least once per 7 days.

M (Monthly) Atleast once pre

.

Q (Quarterly) At least on 492 days.

SA (Semi-Annually)

At le 184 days.

A (Annually) st 0 pr 12 months E Ale once per 18 months.

N.A. Not applicable.

P Completed prior to each release C

64

Number TMI - Unit 1 Radioloaical Controls Procedure 6610-PLN-4200.01 "Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.0 CONTROLS AND BASES 2.0.1 Controls and ACTION requirements shall be applicable during the conditions specified for each control.

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

2.0.3 In the event the Control and associated ACTION requirements cannot be satisfied because of circumstances in excess of those addressed in the Cp? initiate appropriate actions to rectify the problem to the extent possible under the ircmstal ces, and submit a special report to the Commission pursuant to TMI-2 PDM &richaYpecification (Tech. Spec.) Section 6.8.2 within 30 days unless otherwi ified.

2.1 Radioactive Effluent Instrumentation 2.1.1 Radioactive Liquid Effluent Instrumentation Radioactive Liquid Effluent Instrumentation is commnn between TMI-1 and TMI-2.

Controls, applicability, and actions are,¶pecified in ODCM Part 1, Control 2.1.1 2.1.2 Radioactive Gaseous Process and E Monitoring Instrumentation CONTROL; The radioactive gaseous racE and effluent monitoring instrumentation channels shown in Table 2.1-2 shall LE with their alarm/trip setpoints set to ensure that the limits of Control 2. 2. are not exceeded. The alarm/trip setpoints of these channels shall be deter in a ordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLI UTY:3As shown in Table 2.1-2.

a ! With a radioactive gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive effluent monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2.1-2. Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Effluent Release Report why the inoperability was not corrected in a timely manner.

65

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.O1 Trite Revision No.

Offsite Dose Calculation Manual (ODCM) 22 BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluent during actual or potential releases. The alarr/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to provide reasonable assurance that the annual releases are within the limits specified in 10 CFR 20.1301.

AD 411Y Is 66

6610-PLN-4200.01 Revision 22 Table 2.1-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION I. Containment Purge Monitoring System

a. Noble Gas Activity Moni -225) I NOTE 1 NOTE 2

b. Particulate Monitor (2HP-R-e55) *1 NOTE 1 NOTE 2 C. Effluent System Flow Rate Meas ina E '2AH-FR-5907 Pen 1) 1 NOTE 1 NOTE 3

2. Station Ventilation System 4

a. Noble Gas Activity Monitor (2HP-R-219) or 1 NOTE I NOTE 2

b. Particulate Monitor (2HP-R-219) or (2HP-R-21 1 NOTE I NOTE 2

c. Effluent System Flow Rate Monitoring Device I NOTE I NOTE 3 NOTES:

1. During operation of the monitored system.

2. With the number of ch.annels OPERABLE less than required by the Minimum Ch~a43ls progress. , ABLE requitement, secure Reactor Building Purge if in

3. With flow rate monitoring instrumentation out of service, flow rates from the Auxiliary, Fue a gnd Reactor Buildings may be summed individually or estimated using the maximum design flow for the exhaust fans in operation.

0 67

Number TMI - Unit I I Radiological Controls Procedure 6610-PLN-4200.01 "Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2 Radioactive Effluent Controls 2.2.1 Liquid Effluent Controls 2.2.1.1 Liquid Effluent Concentration CONTROL:

The concentration of radioactive material released at anytime from the unit to unrestricted areas shall be limited to ten times the concentrations specified in 10 CFR Part 20.1001-20.2401, Appendix B, Table 2, Column 2.

APPLICABILITY: Atalltimes

.ACTION:

With the concentration of radioactive m i a from the unit to unrestricted areas exceeding the abQb linim'* immediately restore concentrations within the above limi BASES This control is provided *sure that the concentration of radioactive materials released in liqui- e effluent from the unit to unrestricted areas will be less than teAIes theoncentration levels specified in 10 CFR Part 20.1001-20.2401,;ANlum B, Table 2. These Controls permit flexibility under unusuV"ittis, which may temporarily result in higher than normal releases, juttill*hwi ten times the concentrations, specified in 10 CFR20.

It is expfd ¥ using this flexibility under unusual conditions, and exerti! -ry effort to keep levels of radioactive material in liquid wastes as o,prac dble, the annual releases will not exceed a small fraction of the a nu verge concentrations specified in 10 CFR 20. As a result, this Control rovides reasonable assurance that the resulting annual exposure to an in idual in off-site areas will not exceed the design objectives of Section of Appendix I to 10 CFR Part 50, which were established as requirements r the cleanup of TMI-2 in the NRC's Statement of Policy of April 27, 1981.

2,2.1.'V Liquid Effluent Dose "CONTROL The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to the SITE BOUNDARY shall be limited:

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

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

68

Number TMl - Unit I Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 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, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar quarters so that t ulative dose or dose commitment to any individual fro *uch rleases during these four calendar quarters is with' mlt*, the total body and 10 mrem to any organ. This I Report shall also include (1) the result of radiolo* l a I es of the drinking water source, and (2) the radiol ct n finished drinking water supplies with regard to th re **em nts of 40 CFR 141, Safe Drinking Water Act Y

BASES This Control requires th t eýdose to offsite personnel be limited to the design objectives of App Iof 10 CFR Part 50. This will assure the dose received by the pujJO inu DMS is equivalent to or less than that from a normal operating' T. - -e limits also assure that the environmental impacts are te ith those assessed in NUREG-0683, the TMI-2 Programmat Enviortfnental Impact Statement (PEIS). The ACTION stateme id e required flexibility under unusual conditions and at the same32 implement the guides set forth in Section IV.A of Appendix I to as e hat releases of radioactive material in liquid effluents will be kept "sI1,s i' reasonably achievable". The dose calculations in the ODCM irn$?em nt the requirements in Section II.A. of Appendix I that conformance withW guides of Appendix I is to be shown by calculational procedures 4 *ed on models and data such that the actual exposure of a MEMBER OF E PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October, 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,"April, 1977.

NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113.

69

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.1.3 Liquid Radwaste Treatment System CONTROL:

The appropriate portions of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the unit to unrestricted areas would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in any calendar month.

APPLICABILITY: At all times ACTION:

a. With radioactive liquid waste bei arged without treatment and in excess of the above lim' reqe and submit to the NRC Region I Administrator with i~.~ Special Report which includes the following in mat1 j 1.Explanation of liquid radwaste was being discharged with ut treatment, Identification of any inerable equipment or subsystems, and the reason noperability,

2. AActi s) taken to restore the inoperable equipment "N 'M-OPERABLEstatus, and, A summary description of action(s) taken to prevent a recurrence.

/ T e rement that the appropriate portions of this system (shared with

TM!- be used, when specified, provides assurance that the releases of A oactive materials in liquid effluents will be kept as low as is reasonably hievable. This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section 11.0 of Appendix I to 10 CFR Part 50. The intent of Section Il.D. is to reduce effluents to as low as is reasonably achievable in a cost effective manner. This control satisfies this intent by establishing a dose limit which is a small fraction (25%) of Section I1[A of Appendix 1,10 CFR Part 50 dose requirements. This margin, a factor of 4, constitutes a reasonable reduction.

70

Number TMI - Unit I Radioloq[cal Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.2 Gaseous Effluent Controls 2.2.2.1 Gaseous Effluent Dose Rate CONTROL:

The dose rate due to radioactive materials released in gaseous effluent from the site shall be limited to the following:

a. For noble gases: less than or equal to 500 mremlyr to the total body and less than or equal to 3000 mreml rto the skin, and

b. For tritium and all radionuclides in pa c late f with half lives greater than 8 days: less than or toN,6O mremfyr to any organ. '0 '

APPLICABILITY: At all times.

ACTION:

W~ith the release rate(s) yeeein li)ý mits, immediately decrease the release rate to comply wit he above limit(s).

BASES The control provi able assurance that the annual dose at the SITE BOUNDARY, a us effluent from all units on the site will be within the annual do elmits .f9i 0 CFR Part 20 for unrestricted areas. At the same time, th . a l permit flexibility under unusual conditions, which may temp rresult in higher than the design objective levels, but still within the do~IinTits *cified in 10 CFR 20 and within the design objectives of t-o 10 CFR 50. It is expected that using this flexibility under

, unsus~a4onditions, and by exerting every effort to keep levels of radioactive Nmat*'*al in gaseous wastes as low as practicable, the annual releases will not "1'b*ed a small fraction of the annual dose limits specified in 10 CFR 20 and

'S Sill not result in doses which exceed the design objectives of Appendix I to 10 CFR 50, which were endorsed as limits for the cleanup of TMI-2 by the NRC's Statement of Policy of April 27, 1981. These gaseous release rates provide reasonable assurance that radioactive material discharged in gaseous effluent will not result in the exposure of a MEMBER OF THE PUBLIC in an unrestricted area, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the values specified in Appendix B, Table 2 of 10 CFR Part 20. For MEMBERS OF THE PUBLIC who may at times bewithin the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.

71

_L L Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mremlyear to the skin. The absence of iodine ensures that the corresponding thyroid dose rate above background to an infant via the inhalation pathway is less than or equal to 1500 mrem/yr (NUREG 0133), thus there is no need to specify dose rate limits for these nuclides.

2.2.2.2 Gaseous Effluents Dose-Noble Gases CONTROL:

The air dose due to noble gases released in g o is Li'nts from the unit to areas at and beyond the SITE BOUNDARY be limited to the following:

a. During any calendar quart aror equal to 5 mrad for gamma radiation and le* thaer equal to 10 mrad for beta radiation and, Y

b. During any calendar year: ss than or equal to 10 mrad for "gammaradia n and less than or equal to 20 mrad for beta radiation.

AP PLICABILITY:W ,rntime ACTION:

a.. a calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above-limits, prepare and bmit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases wil] be in compliance with the above limits.

BASES This control applies to the release of radioactive materials in gaseous effluents from TMI-2.

This control and associated action is provided to implement the requirements of Section 11.3, III.A and IV.A of Appendix 1,10 CFR Part 50. The Control implements the guides set forth in Section I1.B of Appendix 1. The ACTION statements provide flexibility under unusual conditions and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that 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 72

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 the appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in 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, Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions. *M4EG-0133 provides methods for dose calculations consistent with Re ulitory Ouides 1.109 and 1.111.

2.2.2.3 Dose - iodine-131, Iodine-133, Tritium, an adi u ides In Particulate Form CONTROL:

The dose to a MEMBER OF THE PUB rom Tritium and all radionuclides in particulate form with half lives greate an 8 days, in gaseous effluents released from the unit to a,?as at and beyond the SITE BOUNDARY shall be limited to the following:,,,

a. During cal~ear quarter: less than or equal to 7.5 mrem to any 1usai

b. urn"g calendar year- less than or equal to 15 mrem to any CA -At all times.

A ,

the calculated dose from the release of Tritium and radionuclides in

,l<~ iculate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

73

Number TMI - Unit 1 Tie Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 BASES This control applies to the release of radioactive materials in gaseous effluents from TMI-2.

This control and associated action is provided to implement the requirements of Section 1i.C, IlI.A and IV.A of Appendix I, 10 CFR Part 50. The are the guides set forth in Section II.C of Appendix f. The ACTIONControls statement provides flexability during unusual conditions and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluepL&,ill be kept "as low as is reasonably achievable." The ODCM calculatior me ods specified in the surveillance requirements implement the requj*r nts i* Section Appendix I that conformance with the guides of lII.A of en e shown by calculational procedures based on models a l~uch that the actual exposure of a MEMBER OF THE PUBLIC t gh ropriate pathways is unlikely to be substantially underestJmat i*CM calculational methodology and parameters for calt atinJ doses due to the actual release rates of the subject mater, e con stent wfth the methodology provided in Regulatory Guide 1-109, "C11Ction of Annual Doses to Man from Routine Releases of Reactor EfflueEs for the Purpose of Evaluating Compliance with 10 CFR P rt 50, Appendix I," Revision 1, October, Regulatory Guide 1.111 " thods for Estimating Atmospheric Transport 1977 and and Dispersion of Gaseous E?%tts in Routine Releases from Light-Water-Coole rctors evision 1, July, 1977. These equations also provide for deterrn ctual doses based upon the historical average atmospheric c iod The release rate controls for iodine-131, iodine-133, tritium and r4 ionujd in particulate form with half lives greater than 8 days are depe ifrupo the existing radionuclide pathways to man, in areas at and be d thelE BOUNDARY. The pathways that were examined in the dev d9.e)+/-,f these calculations were: 1) individual inhalation of airborne r!a 5uclide, 2) deposition of radionuclides onto green leafy vegetation with a nt consumption by man, 3) deposition onto grassy areas where milk Sanim nd meat producing animals graze with consumption the milk and "eat y man, and 4) deposition on the ground with subsequentofexposure of The absence of iodines at the site eliminates the need to specify dose S]nits for these nuclides.

74

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Tftle Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.2.4 Ventilation Exhaust Treatment System CONTROL The VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE.

The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the monthly projected doses due to gaseous effluent releases from the site would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the VENTILATION EX#HAU 1 4 TMENT SYSTEM inoperable for more than a monfrjwl aseous waste being discharged without treatme *~~cess of the above limits, prepare and submit to R'e gion I Administrator within 30 R~

days, a Special Report hincl des the following information:

1. Identifiationof e inoperable equipment or s .~ystems and the reason for inoperability,

2. A ) taken to restore the inoperable equipment It RABLE status, and 3

/ summary description of action(s) taken to prevent a Km, } recurrence.

BASE T e f te VENTILATION EXHAUST TREATMENT SYSTEM ensures thatgasjo us effluents are treated as appropriate prior to release to the envirrrnent. The appropriate portions of this system provide reasonable rance that the releases of radioactive materials in gaseous effluents will "kept"aslow as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of "AppendixA to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to .10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections 1I.B and Il.C of Appendix 1,10 CFR Part 50, for gaseous effluents.

75

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.2.3 Total Radioactive Effluent Controls 2.2.3.1 Total Dose CONTROL The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ exceptthe thyroid, which shalJ be limited to less than or equal to 75 mrem.

APPLICABILITY: At all times.

ACTION: 44 With the calculated dose from the releas fmcive materials in liquid or gaseous effluents exceeding twice th2limit*f ontrols 2.2.1.2.a, 2.2.1.2.b, 2.2.2.2.a, 2.2.2.2.b, 2.2.2.3,a, or, 2. . .. b, ca ufations should be made including direct radiation contributions fr he unit and from outside storage tanks to determine whether the above li its of Control 2.2.3.1 have been exceeded. If such is the cre, prepare and submit to the NRC Region I Administrator within 30 ay a Special Report which defines the corrective action to be taken to reai;u sequent releases to prevent recurrence of exceeding the aboj its ar~ncludes the schedule for achieving conformance wit1 = limits. This Special Report, as defined in 10 CFR Part 1.2 3"* shall include an analysis which estimates the radiation exp sure ole) to a MEMBER OF THE PUBLIC from uranium fuel cycle so s, ing all effluent pathways and direct radiation, for the calenddl4Qar that includes the release(s) covered by this report. It shall also d b le9, of radiation and concentrations of radioactive material in olva an. the cause of the exposure levels or concentrations. If the es'ate dose(s) exceed the above limits, and if the release condition XS resu . in violation of 40 CFR 190 has not already been corrected, the jial Report shall include a request for a variance in accordance with the visions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

76

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation. Manual (ODCM) 22 BASES This control is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20.1301(d). This control requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and outs' rage tanks are kept small. The Special Report will describe a ou s* of aclon that should result in the limitation of the annual dose to a M151EJrHE PUBLIC to within the 40 CFR Part 190 limits. For the pur I of the Special Report, it may be assumed that the dose commitmen* th nber of the public from other uranium fuel cycle sources is neglo ih e exception that dose contributions from other nuclear fuel miclem 1cl at the same site or within a radius of 8 km must be considere N.f the N to any member of the public is estimated to exceed the requirement cf40 CFR Part 190, the Special Report with a request for a variance (pr rded the release conditions resulting in violation of 40 CFR Part 90 have not already been corrected), in accordance with the provisi ns of 40 CFR Part 190.11 and 10. CFR Part 20.2203(b), is conside ea timely request and fulfills the requirements of 40 CFR Part 190 1 NRQ taff action is completed. The variance only relates to the limit f R Part 190, and does not apply in any way to the other rqre ddpse limitation of 10 CFR Part 20, as addressed in Controls 2.2. .1 an 2.2.2.1. An individual is not considered a MEMBER OF THE PU urin any period in which hefshe is engaged in carrying out

<C) any o

on th is part of the nuclear fuel cycle.

77

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.0 SURVEILLANCES 3,0.1 Surveillance Requirements shall be applicable during the conditions specified for individual Controls unless otherwise stated in an individual Surveillance Requirement.

The Surveillance Requirements shall be performed to demonstrate compliance with the OPERABILITY requirements of the Control.

3.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.

3.0.3 Failure to perform a Surveillance Requirement within the time interval specified in Section 3.0.2 shall constitute non-compliance with OPERABILITY requiryfite for a Control. The time limits of the ACTION requirements are applicable at the tjov is i ntified that a Surveillance Requirement has not been performed. The A N'lOiements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit completion of the su i when the allowable outage time limits of the ACTION requirements are les an LQours. Surveillance Requirements do not have to be performed on inope* ujlUii ment.

3.1 Radioactive Effluent Instrumentation 3.1.1 Radioactive Liquid Effluent Instrumentation 0 Surveillance Requirements Radioactive Liquid Effluent InsJAi ntat! is common between TMI-1 and TMI-2.

Surveillances for this instrum* ]m - rwe specified in ODCM Part 1, Surveillance 3.1.1.

3.1.2 Radioactive Gaseous PrensnEffluent Monitoring Instrumentation SURVEILLANCE eft.IREMENTS 3.1.2.1 C ioactive gaseous process or effluent monitoring instrumentation Te hal be demonstrated OPERABLE by performance of the CHANNEL SSCH SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL T eoperations at the frequencies shown in Table 3.1-2.

78

66 10-PLN-4200.01 Re vision 22 Table 3.1-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK CALIBRATION TEST APPLICABILITY 1, Containment Purge Monitoring System

a. Noble Gas Activity Monitor H -225) D E M NOTE I

b. Particulate Sampler (2HP-fI45),>\ w N/A N/A NOTE I

2. Station Ventilation Monitoring System

a. Noble Gas Activity Monitor (2HP-R-2Y, D E M NOTE 1 (2HP-R-219A)

b. Particulate Sampler (2HP-R-219) and (21 w N/A N/A NOTE 1 1NOTES:

During operation of the monitored system.

-t

<2k>0 79

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2 Radioactive Effluents 3.2.1 Liquid Effluents SURVEILLANCE REQUIREMENTS 3.2.1.1 Concentration 3.2.1.1.1 The radioactivity content of each batch of radioactive liquid waste shall be determined by sampling and analysis in accordance with Table 3.2-1. The results of analyses shall be used with the calculational methods in the ODCM to aq"drat the concentration at the point of release i n~intaird within the limits of Control 2.2.1 1.

3.2.1:1.2 Analysis of samples composit e m h releases shall be performed in accordance w 4-. The results of the analysis shall be used w, .i.the*lcu ational methods in the ODCM to assure that t: ncee tions at the point of release were maintained within the y of Control 2.2.1.1.

3.2.1.1.3 The radioact[v. concentration of liquids discharged from continuousrl'se points shall be determined by collection and analysis fs § in accordance with Table 3.2-1:. The results of the a1 sis s lI be used with the calculational methods of the ODC ,,Jae that the concentration at the point of release is inilithin the limits of Control 2.2.1.1. ;

3.2.1.2 DoseC*C.

3. . *mutlative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at leastonce a month.

3.2. e Projections 3.2.1.3.1 Doses due to liquid releases shall be projected at least once a month, in accordance with the ODCM.

80

Number TMI - Unit I Radiological Controls Procedure J6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 3.2-1 Radioactive Liquid Waste Sampling and Analysis (4, 6)

A. Liquid Releases Sampling Frequency Type of Detectable Activity Analysis Concentration (3)

Jndividual Gamma 5E-7 ACi/ml (2)

Each Batch H-3 Q Gross Alpha I E-7 .Ci/ml Quarterly Composite (1) Sr-90 NOTES:

(1) A COMPOSITE SAMPLE is one in which the quantity of liquid sa ortional to the quantity of liquid waste discharged from the plant, (2) For certain mixtures of gamma emitters, it may not be pos'ible to measure radionuclides in concentrations near this sensitivity limit when other nuclides are presentsp the sample in much greater concentrations.

Under these circumstances, it will be more appropriatte.klculate the concentrations of such radionuclides using measured ratios with those radionuclides ware rot'nely Identified and measured.

(3) The detectability limits for radioactivity anally r sed on the technical feasibility and on the potential significance in the environment of the quan es r ead. For some nuclides, lower detection limits may. be readily achievable and when nuclides u d below the stated limits, they should also be reported.

(4) The results of these analyses sho d as the basis for recording and reporting the quantities of radioactive material released in efu ts during the sampling period. In estimating releases for a period when analyses were not , the average of the two adjacent data points spanning this period should be used. Such esti tes s be included in the effluent records and reports; however, they should be clearly identi tes, and the method used to obtain these data should be described.

(5) Deviations from theanalysis regime will be noted in the report specified in ODCM Part IV.

81

Number TMI - Unit l Radiological Controls Procedure 661 0-PLN-4200.D1 TMtle Revision No.

OfMsite Dose Calculation Manual (ODCM) 22 3.2.2 Gaseous Effluents SURVEILLANCE REQUIREMENTS 3.2.2.1 Dose Rates 3.2.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of Control 2.2.2,1.a in accordance with the methods and procedures of the ODCM.

3.2.2.1.2 The dose rate of radioactive materials, other than noble gases, in gaseous effluents shall be determined tqt'iN~hin the limits of Control 2.2.2.1 .b in accordance with ýetods4nd procedures of the ODCM by obtaining representaie sa~alp and performing analyses in accordance with the, g and analysis program, 3... Ds, specified in Table 3.2-2.

3.2.2.2 Dose, Noble Gas " - ,

3.2.2.2.1 Cumulative dose contributiorom noble gas effluents for the current calendar quarter and current calendar year shall be determined in.ccordance with the OFFSITE DOSE CALCULARL1 MANUAL (ODCM) monthly.

3.2.2.3 Dose, Ttibu m In Particulate Form 3.2.2.3.1 "(Pil!tib~dose contributions from Tritium and radionuclides in articu *te form with half lives greater than 8 days for the current Scb.ga~dr quarter and current calendar year shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL 0DCM) monthly.

3.2.24 V nE Exhaust Treatment

.* .2.4-1 Doses due to gaseous releases from the unit shall be projected monthly in accordance with the ODCM.

4`1ý 82

Number TML- Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 3.2-2 Radioactive Gaseous Waste Sampling and Analysis (3)

'"DETECTABLE SAMPLE SAMPLING TYPE OF CONCENTRATION(1)(

SAMPLE POINT TYPE FREQUENCY ACTIVITY ANALYSIS a)

Reactor Building Purge Releases P H-3 1 E-6 pCi/cc Gas Individual Each Purge Gamma Emitters 1E-4 p.Cicc (2)

Unit Exhaust Vent Release Points M Gas Monthly H-3

~~Individual A i.

1-5 i/Cicc

- ~lc 2 Gas Monthly Gamma Erer W Indlvi 1 0 IE-lQI.Ci/cc (2)

Weekly Gain a ___1E-10.__C______2 M

Monthly 1E-11 lCicc Particulates Composite M

Monthly'* Gross Alpha Emitters I E-11 pCi/cc Reactor Building Breather ,ndv. Gamma S u7y trs ((b) E-1 0 pCi/cc(2)

Emitters Particulates Semi -ally S-90 IE-11 rpCi/cc Alpha EmittersS*Gross I1E-1I ý.Ci/cc (1) The above detectability limits arb on technical feasibility and on the potential significance in the environment of the quantitie release For some nuclides, lower detection limits may be readily achievable and when nuclides are m red b ow the stated limits, they should also be reported.

(2) For ertain mixtures mamitters, it may be possible to measure radionuclides at levels near their sensitivity limits when e uclides are present in the sample at much higher levels, Under these circumstances, it will be re appropriate to calculate the levels of such radionuclides using observed ratios in the gaseous component in the reactor coolant for those radionuclides which are measurable.

(3) Deviations from the sampling and analysis regime will be noted in the report specified in ODCM Part IV.

83

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 3.2-2 Radioactive Gaseous Waste Sampling and Analysis Program Table Notation

a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% 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 sp E x V x 2.22 x10 6 x Y x exp (-7,At)

Where ' l LLD is the lower limit of detection as defined above (as p! 4 epe unit mass or volume).

Sb is the standard deviation of the background counting rate o rf the counting rate of a blank sample as appropriate (as counts per minute).

E is the counting efficiency (as counts per transf'i't)on),

V is the sample size (in units of mass orv 2.22 is the number of transformlo peroGinute per picocurie, Y is the fractional radiochemI li Id (when applicable),

X is the radioactive dent for the particular radionuclide, and r e At is the elapsed m etween midpoint of sample collection and time of counting (for plant effluents, not itronm I samples),

The value of s in the calculation of the LLD for a detection system shall be based on the actual obsergved d~rance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y, and At shall be used in the calculation. The background count rate is calculated from the background counts that are determined to be with +/- one FWHM (Full-Width-at-Half-Maximum) energy band about the energy of the gamma-ray peak used for the quantitative analysis for that radionuclide.

84

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 3.2-2

b. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn765, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

A-_

85

Number TMI - Unit 1 Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2.3 Total Radioactive Effluents 3.2.3.1 Dose Calculation 3.2.3.1.1 Cumulative annual dose contributions from liquid and gaseous effluents shall be determined in accordance with Surveillances 3.2.1.2.1, 3.2.2.2.1, and 3.2.2.3.1, including direct radiation contributions from the Unit and from outside storage tanks, and in accordance with the methodology contained in the ODCM.

Aýb 4 86

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.0 PART 11REFERENCES 4.1 NUREG-0683, "Final Programmatic Environmental Impact Statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979, accident Three Mile Island Nuclear Station, Unit 2," March 1981, and its supplements.

4.2 TMI-2 PDMS Technical Specifications, attached to Facility License No. DPR-73 4.3 Title 10, Code of Federal Regulations, "Energy" 4.4 "Statement of Policy Relative to the NRC Programmatic Environmental Impact Statement on the Cleanup of Three Mile Island Unit 2," dated April 27, 1981 4.5 Regulatory Guide 1.109, "Calculation of Annual Doses to Man from R eBses of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR P a t*Q.ppendix I,"Revision 1, October 1977 4.6 DOE/TIC-27601, Atmospheric Science and Power Red uc .

4.7 TMI-1 Technical Specifications, attached to Facility Operating rnse No. DPR-50 4.8 PDMS-SAR 87

Number TM1 - Unit 1 Radiological Controls Procedure Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 0

PART III EFFLUENT DATA AND M

ONAL METHODOLOGIES

Number TM! - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 1.0 LIQUID EFFLUENT MONITORS 1.1 TMI-1 and TMI-2 Liquid Radiation Monitor Set Points The liquid effluent off-line monitors are set such that the concentration(s) of radionuclides in the liquid effluents will not exceed ten times the concentrations specified in 10 CFR 20, Appendix B Table 2, Col 2. Table 1-1 lists the Liquid Effluent Release Points and their parameters; Figure 1.1 provides a Liquid Release Pathway Diagram.

To meet the above limit, the alarm/trip set points for liquid effluent monitors and flow measuring devices are set in accordance with the following equation:

1.1)

F _f < " *(eq where: "*

w =eten times the effluent concentration of 10 CFR 20 fo itml.

c = the set point, in I.Cilml, of the liquid effluent monitor mea ring the radioactivity concentration in the effluent line prior to dilution and releae. The set point is inversely proportional to the maximum volumetric flow of the effluent H e nd proportional to the minimal volumetric flow of the dilution stream plus the effluent stref'T_e alert set point value is set to ensure that advance warning occurs prior to exqS'g ant imits. The high alarm set point value is such that if it were exceeded, it would re*g centrations exceeding ten times the 10 CFR 20 concentrations for the unrestriIee f= flow set point as measure dq[*ation monitor location, in volume per unit time, but in the "sameunits as F below.('\

F = flow rate of diluti o* ,me sured prior to the release point, in volume per unit time.

The set point con tratio r reduced such that concentration contributions from multiple release points would n corn,*to

exceed ten times 10 CFR 20 concentrations. The set point concentratio¶ Tsnve d to set point scale units using appropriate radiation monitor calibration factors.

This section of e ODCM is implemented by the Radiation Monitor System Set Points procedure and, for batch releases, the Releasing Radioactive Liquid Waste procedure..

89

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 1.2 TMI Liquid Effluent Release Points and Liquid Radiation Monitor Data TMI-1 has two required liquid radiation monitors. These are RM-L6 and RM-L12. These liquid release point radiation monitors and sample points are shown in Table 1.1. (The TMI outfall radiation monitor, RM-L7, is also listed for information only.)

TMI-2 does not have any required liquid radiation monitors, but does utilize RM-L12, and RM-L7 for release of liquid waste.

1.2.1 RM-L6 RM-L6 is an off-line system, monitoring radioactive batch discharjgITgm the TMI-1 liquid radwaste system (see Figure 1,1). These batch releases are pled lid analyzed per site procedures prior to release. The release rate is based e reel~ one of two Waste Evaporator Condensate Storage Tanks (WECST) at a flo will add less than 10%,

often times the 10 CFR 20 concentrations [20% for H-3* ra uclide concentrations in the unrestricted area, including conservative default e r r-89, Sr-90, and Fe-55.

The release flow rate used is the most restrictiv two fIv rates calculated for each liquid batch release, per the approved plant proceduX Two Dilution Factors (DF) are calculated to ultimately calculate the batch release flow rate. These two DF's are calculated o sure each radionuclide released to the unrestricted area is less than 10 pe r-ft*fn times the 10CFR20 radionuclide concentrations, (20% for H-3), plao en~sre each liquid batch release boron concentration to the river will 0.7 ppm.

The maximum releasefl rate's en calculated by dividing the most restrictive (largest)

DF into 90 percent o r dilution flow rate of the Mechanical Draft Cooling Tower (MDCT). This cor ative flow rate is then multiplied by 0.9 for the allowable flow rate.

Calcaoft'he 10CFR20 concentration DF:

Z,* ') + (10% [20% for H-31 of ten times the IOCFR20 concentration)

Specific Activity of each Identified radionuclide Falculation of Boron DF:

DF 2 = Actual Tank Boron Concentration ÷ 0.7.

Maximum release flow rate calculation:

Max Flow = [(MDCT flow gpm

0.9) + (Most Restrictive DF)]

0.9 The dilution flow rate used is the current flow rate at the site. The minimum dilution flow rate is 5000 gpm per the TMI-1 FSAR. This ensures this batch release will meet the following equation.

F.(C 1IX1 ) + (CI-j2X,.3) < 0.1, (eq 1.2) s0

Number TMI - Unit 1 Radiological Controls.Procedure 661 0-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 where: Ci = diluted concentration of the Pi"radionuclide, other than H-3 X= Ten times the concentration for that radionuclide in the unrestricted area (10 CFR 20, App. B, Table 2, Col. 2). A value of 3E-3 .Cilml for dissolved and entrained noble gases shall be used, C] =diluted concentration of H-3 XH- =Ten times the concentration for H-3 in the restricted area (10 CFR 20, App. B, Table 2, Col. 2).

The set points for RM-L6 are based on the maximum release rate im), a minimum dilution flow (5000 gpm), and 25% of ten times the 10CFR2F c ntrat n for Cs-I 37, which is the most limiting radionuclide ata concentration of -5 1i 1. These inputs are used in Equation 1.1 to determine the RM-L-6 High All oint for all radionuclides being released, A high alarm on RM-L-6 will close vaiv L- - 7 and terminate any WECST releases to the environment

-1.2.2 RM-L212 RM-L12 is an off-line system, monitoring periodic conined releases from the Industrial Waste Treatment System/Industrial WaVte Filtration System (IWTS/IWFS). The inputto Iw'TSIIWFS originates in TMI-2 sumrs see Figures 1.1 and 1.2) and the TMI-1 Turbine Building sump (see Figure 1.1). Th oints are based on the maximum release rate from both IWTS and IWFS sim ousre see Figure 1.1) a minimum dilution flow rate, and 50% of ten times the 10 ntration for Cs-137, which is the most limiting radionuclide at a concent

pCi/ml. These inputs are used in equation 1.1 to determine the RM-L12 Hi h Ala t p*int for all radionucfides being released. A high alarm on RM-L12 wilS and IWFS release valves and trip release pumps to stop the release.

1.2.3 RM-L 1G RM-L10saN tector submerged in the TMI-I Turbine Building Sump. This deteo haen removed from service.

1.2.4 R RM-Lis not an ODCM required liquid radiation monitor. RM-L7 is an off-line system,.

monitoring the TMI outfall to the Susquehanna River (see Figures 1.1 and 1.2). This monitor is the final radiation monitor for TMI-I and TMI-2 normal liquid effluent releases.

1.3 Control of Liquid Releases TMI liquid effluent releases are controlled to less than ten times the I OCFR20 concentrations by limiting the percentage of this limit allowable from the two TMI liquid release points, RM-L6 and effluent sampling limit batch releases to less than or equal to 25% for all radionuclides, and RM-L12 and effluent sampling limit releases from TMI-I and TMF-2 to less than or equal to 50% for Cs-137.

These radiation monitor set points also include built in meter error factors to further ensure that TMI liquid effluent releases are less than ten times the 10CFR20 concentrations to the environment.

91

Number TMI - Unit I Titte I Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 The radioactivity content of each batch of radioactive liquid waste is determined prior to release by sampling and analysis in accordance with ODCM Part I Table 3.2-1 or ODCM Part 11,Table 3.2-1.

The results of analyses are used with the calculational methods in Section 1.1, to assure that the concentration at the point of release is maintained within the ODCM Part I Control 2.2.1.1, and ODCM Part II Control 2.2.1.1.

Post-release analysis of samples composited from batch releases are performed in accordance with ODCM Part I Table 3.2-1 or ODCM Part II Table 3.2-1. The results of the previous post-release analysis shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release were maintained within the ODCM Part I Control 2.2.1.1, and ODCM Part I1 Control 2.2.1.1.

The radioactivity concentration of liquids discharged from continuous re a poi 6sare determined by collection and analysis of samples in accordance with ODCM Part NaGe , or ODCM Part II Table 3.2-1. The results of the analysis are used with the calculati r.thods of the ODCM to assure that the concentration at the point of release is maintained inV ODCM Part I Control 2.2.1.1, and ODCM Part I1Control 2.2.1.1.

AJV 0s 92

Number TMI- Unit I Radioloqical Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 1.1 TMI Liquid Release Point and Liquid Radiation Monitor Data RELEASE LIQUID RADIATION LIQUID RELEASE TERMINATION MONITOR POINT (Maximum DISCHARGE FLOW INTERLOCK (DETECTOR) LOCATION Volume) RECORDER (YES/NO) VALVES RM-LG 281' Elevation WECST Batch FT- YES (Nal) TMI-1 Auxiliary Bldg Releases (8000 gal.) FT-84__ WDL-V257 RM-(7 South end of TMI-1 Station Discharge YES (Nal) TMI-1 and FT-146 WDL-V257 I, MDCT TM[-2, '*WDL-R-1311 YES IWTS/IWFS iW-V73, RM-L 12 IWFS Building NW Continuous IW-P16,17,18 (Nal) Corner Releases (300,000/

X;-3 80,000 gal.) 1 IW-V279, A IW-P29,30 WDL-R-1311 has beenflanged

____________________

S off as aTMI-2 liquid 1 t ____________________

WDL-R-1 311 has been. flanged off as a TMI-2 liquid litfall.

- RM-L7 is not an ODCM required liquid radiation or.

0 93

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 1.2 TMI-2 Sump Capacities Total Capacity Gallons Sump Gallons per Inch Turbine Building Sump 1346 22-43 Circulating Water Pump House Sump 572 10.59 Control Building Area Sump 718 9.96 Tendon Access Galley Sump 538 ( "* 9.96 Control to Service Building Sump 1346 \.,A2.43 Contaminated Drain Tank Room Sump 135 3.80 Chlorinator House Sump __-_______

Water Treatment Sump' 22.43 Air Intake Tunnel Normal Sump .ONO Air Intake Tunnel Emergency Sump 1_004Y 766.00 Condensate Polisher Sump* 2617 62.31 Sludge Collection Sump** '1106 26.33 Heater Drain Sump ' -

Solid Waste Staging Facility Sump 1476 24.00 Auxiliary Building Sump 10102 202.00 Decay Heat Vault Sump- 479 10.00 Building Spray Vault Sump 479 10.00

Condensate Polisher Sump is deac nd in PDMS condition.

- The Water Treatment and Slu Coll Sumps will be deactivated for PDMS.

94

6610-PLN-4200.01 Revision 22 Figure 1.1 Sampler C omposite Radiation Monitor TMI-1 Liquid Effluent Pathways M Radwaste Grab Disposal LUi Sample r

Component Cooling Mechanical Draft Cooling Tower Water TMI-1 TMI.1 95

Number TMI - Unit I Radiological Controfs Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 FIGURE 1.2 TMI-2 Liquid Effluent Pathways CONTROL CONTROL &

BUILDING SERVfCE SUMP AREA SUMP C-- COMPOSITE SAMPLER 96

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.0 LIQUID EFFLUENT DOSE ASSESSMENT 2.1 Liquid Effluents - 10 CFR 50 Appendix I The dose from liquid effluents results from the consumption of fish and drinking water. The location of the nearest potable water intake is PP&L Brunner Island Steam Electric Station located downstream of TMI. The use of the flow of the Susquehanna River as the dilution flow is justified based on the complete mixing in the river prior to the first potable water supply, adequately demonstrated by flume tracer die studies and additional liquid effluent release studies conducted using actual TMI-1 tritium releases. Other pathways contribute negligibly at Three Mile Island. The dose contribution from all radionuclides in liquid effluents released to the unrestricted area is calculated using the following expression:

Dose j (At)X (C!) X + AiiXFFDJX D (eq2.1) where:

Dose j = the cumulative dose commitment to the total or an organ, j, from the liquid effluents for the total time period, in mrem. 4 At = the length of the time period of actuall leases, over which C, and f are averaged for all liquid releases, in hours.

C,= the average concentration of nucli , i, in undiluted liquid effluent during time period For Fe-55, Sr-89, Sr- r to batch releases conservative concentration valu .i. be eld in the initial dose calculation based on similar past plant o9,*ins. LLD values are not used in dose calculations. i*

f= un *utdl waste flow, in gpm." "

FD = p di ion water flowrate during the period of release, in gpm FR = actual river flowrate during the period of release or average river flowrate for the month the release is occurring, in gpm.

DF = dilution factor as a result of mixing effects in the near field of the discharge structure of 0.2 (NUREG 0133) or taken to be 5 based on the inverse of 0,2.

AWij and AFij = the site-related ingestion dose commitment factor to the total body or any organ, j, for each identified principle gamma and beta emitter, in mrem/hr per tiCi/ml. AW is the factor for the water pathway and AF is the factor for the fish pathway.

97

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 Values for AWij are determined by the following equation:

AW11 = (1.14E5) x (Up) x (DF 1 ) (eq 2.2) where:

1.14E5 = (1.OES pCil/!Ci) x (1.0E3 mllkg) - (8760 hr/yr)

Uw= Water consumption rate for adult is 730 kglyr (Reg. Guide 1.109, Rev. 1).

DFij = ingestion dose conversion factor for radionuclide, i, for adults total body and for "worst case" organ, j, in mremfpCi, from Table 2.1 (Reg. Guide 1.109)"

Vwalues for AF11 are determined by the following equation:

AFij = (1.14E5) x (Ut) x (DFj) x (BFi) (eq 2.2.2) wthere:

1.14E5 = defined above Uf= adult fish consumption, assumed to b 21 kg/yr (Reg. Guide 1.109, Rev. 1).

DFij = ingestion dose conversion factor onuclide, i, for adult total body and for 'Worst case" organ, j, in mrernpCi,!? fable (Reg. Guide 1.109, Rev. 1).

BFi = Bioaccumulation factor f o lide, i, in fish, in pCilkg per pCi/L from Table 2.2 (Reg.

Guide 1.109, Rev. 1).

2.2 TMI Liquid Radwaste Syste 0 Ca cs Once/Month ODCM Part I Control 2.d1 .- d '*1-2 PDMS Tech Spec Section 6.7.4.a.6 requires that appropriate portions Ifti'Tiqutzd radwaste treatment system shall be used to reduce the radioactive materials in liquid rtes i to their discharge when the monthly projected doses due to the liquid effluent releas from a unit to unrestricted areas would exceed 0.06 mrem to the total body or 0.2 mrem toan rgart* any calendar month. The following calculational method is provided for performing o projection.

At least once per month, the total dose from all liquid releases for the quarter-to-date will be divided by the number of days into the quarter and multiplied by 31. Also, this dose projection shall include the estimated dose due to any anticipated unusual releases during the period for which the projection is made. If this projected dose exceeds 0.06 mrem total body or0.2 mrem any organ, appropriate portions of the Liquid Radwaste Treatment System, as defined in Section 3.1, shall be used to reduce radioactivity levels prior to release.

At the discretion of Radiological Engineering, time periods other than the current quarter-to-date may be used to project doses if the dose per day in the current quarter-to-date is not believed to be representative of the dose per day projected for the next month.

98

Number TMI - Unit 1 Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 2.3 Alternative Liquid Dose Calculational Methodology As an alternative, models in, or based upon, those presented in Regulatory Guide 1.109 (Rev. 1) may be used to make a comprehensive dose assessment. Default parameter values from Reg.

Guide 1.109 (Rev. 1) and/or actual site specific data are used where applicable.

As an alternative dose calculational methodology TMI calculates doses using SEEDS (simplified environmental effluent dosimetry system).

The onsite and SEEDS calculational models use actual liquid release data with actual monthly Susquehanna River flow data to assess the dispersion of effluents in the river.

0 4>;

99

Number TMI - Unit I Radiorogical Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 2.1 Liquid Dose Conversion Factors (DCF): DF 0 Page 1 of 3 Ingestion Dose Factors for Adults*

(MREM Per PCI Ingested).

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY -LUNG GI-LLI H 3 NO DATA 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1 1.05E-07 C 14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 .8- 5.68E-07

--

c-----

NA

24

.-. --. -- --.

1.70E-06

---- .-- . . ....

1.70E-06

.................

.....

1.70E-06

.................... . . . . . . . .

1.70E-06 1.70E-

.. -- --- *---

.70E-06

.......

------

1.70E-06 CR 51 NO DATA NO DATA 2.86E1-09 1.59E-09 5. - 3.53E-09 6.69E-07 MN 54 NO DATA 4.57E-06 8.72E-07 NO DATA tKENO DATA 1.40E.-05 MN 56 NO DATA 1.156-07 2.04E-08 NO DATA 14, -07 NO DATA 3.67E-06 FE 55 2.75E-06 1.90E-06 4.43E-07 NO ATA NO DATA 1.06E-06 1.09E-06 FE 59 4.34E-06 1.02E-05 3.91 E-06 NO TA NO DATA 2-85E-06 3.40E-05

--

CO

-- --- 58

-- ..--. .. ..o...

NO DATA

...... ...... ,*

...

7,452-07 o......... ...............

1.67E-06 0'{o0

..-- A

--- --- --

NO DATA

--- --- -- ...... I.. o...

NO DATA

........ ....

..

...

1.51 E-05

..

Co 60 NO DATA 2.14E-06 4.72E-06 NOtATA NO DATA NO DATA 4.02E-05 NI 4 3 6 63 1.30E-04 9.01E-06 . E- Q NO DATA NO DATA NO DATA 1.88E-06 NI 65 6.28E-07 4.8E-08 .KE-08 NO DATA NO DATA NO DATA 1-74E-06 CU 64- NO DATA 8.33E-08 .-910, NO DATA 2.10E-07 NO DATA 7.910E-06 ZN 95 4.84E-06 1,54E-05 96E-06 NO DATA 1.03E-05 NO DATA 9.70E-06 ZN 69 1.03E-08 1. 8 1.37E-09 NO DATA 1.28E-08 NO DATA 2.962-09 BR 83 NO DATA DA 4.02E-08 NO DATA NO DATA

---............................................................................................. NO DATA 5.79E-08 BR 84 NO DATA N 5.21 E-08 NO DATA NO DATA NO DATA 4.09E-13 S BR 85 NO DATA N DATA 2.14E-09 NO DATA NO DATA NO DATA LT2E-24 RB

...

86

- - ..- -... --.. - ..- -...- -..

NO ..DATA

..- -...- ... .. .. ...

2.11 IE-05 9.83E-06

.. .. ... -------------

.. ..

NO DATA

.-----------------------------------------

NO DATA NO DATA

...........................................................

4.16E-06 RB3 88 NO DATA 6.05E-08 3.212E-08 NO DATA NO DATA NO DAT A 8.36E- 19 RB 89 NO DATA 4.012E-08 2.82E-08 NO DATA NO DATA NO DATA 2.332-21 SIR 89 3.08E-04 NO DATA 8.84E-05 NO DATA NO DATA NO DATA 4.942-05 SIR 90 7.58E-03 NO DATA 1.86E-03 NO DATA NO DATA NO DATA 2,19E5-04 SR 91 5,67E-06 NO DATA 2.29E-07 NO DATA NO DATA NO DATA 2.70E-06 SIR 92 2.15E-06 NO DATA 9.30E-08 NO DATA NO DATA NO DATA 4.26E-05 Y 90 9.62E-09 NO DATA 2.58E-10 NO DATA NO DATA NO DATA 1.02E-04 100

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 2.1 Liquid Dose Conversion Factors (DCF): DFIj Page 2 of 3 Ingestion Dose Factors for Adults*

(MREM Per PCI Ingested)

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI Y 91M 9.09E-11 NO DATA 3.52E-12 NO DATA NO DATA NMTA 2.67E-1O Y 91 1.41 E-07 NO DATA 3.77E-09 NO DATA NO DATA NjJDA? 7.76E-05 Y 92 8.45E-10 NO DATA 2.47E-1l NO DATA NO DATA e0A.A 1.48E-05 Y 93 2.68E-09 NO DATA 7.40E-11 NO DATA NO D DATA .8-50E-05 ZR 95 3.04E-08 9.75E-09 6.60E-09 NO DATA 1. - 0 DATA 3.09E-05 ZR 97 1.68E-09 3.39E-10 1.55E-1O NO DATA .12E NO DATA 1.05E-04 NB 95 6.22E-09 3.46E-09 1.86E-09 NO DATA 3.4%-09 NO DATA 2.10E-05 MO 99 NO DATA 4.31 E-06 8.20E-07 NO QATA 9.76E-06 NO DATA 9.99E-06 TC 9SM 2.47E-10 6.98E-10 8.89E-09 NO AA 1.06E-08 3.42E-1 0 4.13E-07 So-oo--o -----------

TC 101 2..54E-10 3.66E-10 3.59E-09 0 D A 6.59E-09 1.87E-10 1.10E-21 RU 103 1.85E-07 NO DATA 7.97E-08 ATA 7.06E-07 NO DATA 2.16E-05 RU 105 1.54E-08 NO DATA 6.08E DATA 1.99E-07 NO DATA 9.42E-06 RU 106 2-75E-06 NO DATA ZE-0D NO DA " 5.31 E-06 NO DATA 1.78E-04 AG 110M 1-60E-07 IA8E-07 9 NODATA- 2.91E-07 NO DATA 6.04E-05 SB 125 1.79E-06 2.00E-08 26B47 1.82E-09 0.0 1.38E-06 1.97E-05 TE 125M 2.68E-06 9.71E--7

............. 3. 9E-07

-- .................. ..........................---...........................................---............

8.06E-07 1.09E-05 NO DATA 1.07E-05 TE 127M 6.77E-06 42E- 8.25E-07 1.73E-06 2.75E-05 NO DATA 2.271-05 TE 127 1.10E-07 t3XE-0 2.38E-08 8.1 5E-08. 4.48E-07 NO DATA 8.68E-06 TE 12gM 1.16E-05 __ .2906 4 1.82E-06 3.95E-06 4.80E-05 NO DATA 5.79E-05

........................... .oo~ o~ .................................................

,.............. ... ......... ',.----- ----------------------------

TE 129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 NO DATA 2.37E-08 TE 131M 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 NO DATA 8.40E-05 TE 131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 NO DATA 2.79E-09

-- - - - - - -- -................................. ............. ". - -..............................................................

TE 132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 NO DATA 7.71E-05 1 130 7.56E-07 2,23E-06 8.80E-07 1.89E-04 3.48E-06 NO DATA 1.92E-06 I 131 4.16E-06 5,95E-06 3.41E-06 1.95E-03 1.02E-05 NO DATA 1.57E-06 1 132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 NO DATA 1.02E-07 I 133 1.42E 2.47E-06 7.53E-07 3.63E-04 4.31 E-06 NO DATA 2.22E-06 1 134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 NO DATA 2.51 E-10 101

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 2.1 Liquid Dose Conversion Factors (DCF): DFu Page 3 of 3 Ingestion Dose Factors for Adults*

(MREM Per PCI Ingested)

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LL.

i 135 4-43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 N9;'9kTA 1.31E-06 CS 134 6.22E-05 1.48E-04 1.21 E-04 NO DATA 4.79E-05 I 59E-ld 2.59E-06 CS 136 6.51E-06 2.57E-05 1.85E-05 NO DATA 1.43E-05Z . E- 2.92E-06 CS 137 7.97E-05 1.09E-04 7.14E-05 NO DATA 3.70Ej "% 3E-05 2.11E-06 CS 138 5.52E-08 1.09E-07 5.40E-08 NO DATA 8. 7.91 E-09 4.65&-13 BA 139 9.70E-08 6.91 E-1 I 2.84E-09 NO DATA .46E-*j 3.92E-1 1 1.72E-07 BA 140 2.03E-05 2.55E-08 1.33E-06 NO DATA 8.I-09 1.46E-08 4.18E-05 BA 141 4.71E-08 3.56E-11 1.59E-09 NO 9ATA 3.31 E-11 2.02E-11 2.22E-17 BA 142 - 2.13E-08 2.IS E-11I 1,34E-09 NO0 ATIA 1.85E-11I 1.24E-1 I- 3.OOE-26 LA 140 2.50E-09 1.26E-09 3.33E-10 OD A NO DATA NO DATA 9.25E-05 LA 142 1.28E-10 5.82E-11 1.45E-11 ATA NO DATA NO DATA 4.25E*-07 CE 141 9.36E-09 6.33E-09 M.8EDAT 2.94E-09 NO DATA 2-42E-05

..................................... -----------------------------------------------------------------------

-- ------- I.......

CE 143 1.65E-09 1.22E-06 E-1 NO DATA 5.37E-10 NO DATA 4.56E-05 CE 144 4.88E-07 2.04E-07 .62 NO DATA 1.21 E-07 NO DATA 1.65E-04 PR 143 9.20E-09 3.69E-09 .56 10 NO DATA 2.13E-09 NO DATA 4.03E-05

*....-- 11111-------------

PR 144 3.01E-11 1.24 1 1.53E-12 NO DATA 7.05E-12 NO DATA 4.33E-18 ND 147 6.29E-09 A27E-& 4.35E-10 NO DATA 4.25E-09 NO DATA 3.49E-05 W 187 1.03E-07 4 8. E00 3.01E-08 NO DATA NO DATA NO DATA 2.82E-05 NP 239 1.19E-09 1{.7E-10 6.45E-11 NO DATA 3.65E-10 NO DATA 2.40E-05

Dose factors of internal exposure are for continuous intake over a one-year period and include the dose commitment over a 50-year period; from Reg. Guide 1.109 (Rev. 1)- Additional dose factors for nuclides not included in this table may be obtained from NUREG-0172.

102

Number TMI- Unit I Radiological Controls Procedure 661 O-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 2.2 Bioaccumulatlon Factors, BF, Bioaccumulation Factors to be Used in the Absence of Site-Specific Data*

(pCi/kg per pCi/liter)

Bioaccumulation factor values are taken from Reg. Guide 1.109 (Rev. 1), Table A-lj.

Sb bioaccumulation factor value is taken from EPRI NP-3840.

Ag bioaccumulation factor value is taken from Reg. Guide 1.109 (Rev. 0), Table A-8.

103

Number TM! -Unit 1 Title I Radiological Controls Procedure 6610-PLN-4200.01

.Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.0 TMI LIQUID EFFLUENT WASTE TREATMENT SYSTEMS 3.1 TMI-I Liquid Effluent Waste Treatment System 3.1.1 Description of the Liquid Radioactive Waste Treatment System (see Figure 3.1)

Reactor Coolant Train

a. Water Sources - (3) Reactor Coolant Bleed Tanks (RCBT)

- (1) Reactor Coolant Drain Tank (RCDT)

b. Liquid Processing - Reactor Coolant Waste Evaporator ure 3.2)

- Demineralizers prior to release qur

c. Liquid Effluent for Release - (2) Waste Evaporatoo ensa e Storage Tanks

- (WECST)

d. Dilution - Mechanical Draft Cooling Tower,(

- River Flow (2E7 gpm averagq Miscellaneous Waste Train Y

a. Water sources: - Auxiliary Bu ing Sump

- Reactor Ilg . Sump

- Misce eous . ste Storage Tank

- La ery eStorage Tank

- tr~rer Mixing Tank Neu IJ r Feed Tank

. sed recoat Tank

" *- ed Water Tank Tunnel Sump

. p at Exchanger Vault Sump

" -,endon Access Galley Sump

- Spent Fuel Pool Room Sump

- TMI-2 Miscellaneous Waste Holdup Tank b., Liq~uldrocessing - Miscellaneous Waste Evaporator, MWE (see Figure 3.2)

Li"- Demineralizers prior to release

c. Liquid Effluent for Release - (2) Waste Evaporator Condensate Storage Tanks

- (WECST)

d. Dilution - Mechanical Draft Cooling Tower (0-38k gpm)

- River Flow (2E7 gpm average) 104

Number TMI - Unit I Radiological Controls Procedure OU'IU-V-L1N14;Z.U.0.)

Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 3.2 Operability of the TMI-1 Liquid Effluent Waste Treatment System 3.2.1 The TMI-1 Liquid Waste Treatment System as described in Section 11 of the TMI-1 Final Safety Analysis Report is considered to be operable when one of each of the fo[lowing pieces of equipment is available to perform its intended function:

a) Miscellaneous Waste Evaporator (WDL-ZI B) or Reactor Coolant Evaporator (WDL-ZIA) b) Waste Evaporator Condensate Demineralizer (WDL-K3 A or B)

C) Waste Evaporator Condensate Storage Tank (WDL-T I11 )

d) Evaporator Condensate Pumps (WDL-P 14 A or B)i.

3.2.2 TMI-1 Representative Sampling Prior to Discharge All liquid releases from the TMI-1 Liquid Waste T at t tem are made through the Waste Evaporator Condensate Storage Tanks. provirl thorough mixing and a representative sample, the contents of the tank are culated using one of the Waste Evaporator Condensate Transfer Pumps.

3.3 TMI-2 Liquid Effluent Waste Treatment Svste*

3.3.1 Description of the TMI-2 Liqui ctfWaste Treatment System The TMI-2 Liquid Radioa " W' t Treatment System has been out of service since the TMI-2 Accident in 1979. I-Li id Radioactive Waste is processed by the TMI-1 system described in 3.1 ror to release. In addition, TMI-2 releases water from various sumps a ks to e river (see Figures 1.1 and 1.2). These processes are governed by pjIt oc res that encompass proper sampling, sample analysis, and radiation mo .rdte~ niques.

105

Number TMI - Unit 1 Radioloaical Controls Procedure nm 6610-PL -4c2f Title Revision N0.

Offsite Dose Calculation Manual (ODCM) 22 FIGURE 3.1 TMI-1 Liquid Radwaste RC DRAIN MAKE UP DEMIN " WATER WATER STOR TANK RECLAIMED WATER SYSTEM 106

6610-PLN-4200.01 Revision 22 FIGURE 3.2 TMI-1 Liquid Waste Evaporators 107

Number TMI - Unit 1 Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.0 GASEOUS EFFLUENT MONITORS 4.1 TMI-1 Noble Gas Monitor Set Points The gaseous effluent monitor set points are established for each gaseous effluent radiation monitor to assure concentrations of radionuclides in gaseous effluents do not exceed the limits set forth in ODCM Part I Control 2.2.2.1. Table 4.1 lists Gaseous Effluent Release Points and their associated parameters; Figure 4.1 provides a Gaseous Effluent Release Pathway Diagram.

The set points are established to satisfy the more restrictive set point concentration in the following two equations:

500> (c)(F)(Kj)(Dv) (eq 4.1.1) and Qlc .

3000 > (cX)(Li + 1.1 M1)(Dv)(F) (eq 4.1.2) where:

c.= set point concentration based on Xe-l equivalent, in pCilcc F= gaseous effluent flowrate at th intorcc/se K = total body dose factor, in y r 4CI/m3 from Table 4.3 Dv = highest sector annub' e aseous atmospheric dispersion factor (X/Q) at or beyond the unrestricted ar oundary, in sec/m3 , from Table 4.4 for station vent releases and Table 4.5 for al i rr ses, (Condenser off gas, ESF FHB, and ground releases).

Maximum valýes p;ser y used are 7.17E-7 secfmn at sector NNE for station vent, and 1.16E-6 satcrrt sctors N and WNW for all other releases.

L,= skin ose due to beta emissions from radionuclide i, in mremlyr per uCi/m5 from T

T x M,= air dl ctor due to gamma emissions from radionuclide f, in mrad/yr per pCilm 3 from Table 4.3.

1.1 = mrem skin dose per mrad air dose.

500 = annual whole body dose rate limit for unrestricted areas, in mrem/yr.

3000 = annual skin dose rate limit for unrestricted areas, in mremiyr.

The set point concentration is further reduced such that the concentration contributions from multiple release points would not combine to exceed ODCM Control limits.

The set point concentration is converted to set point scale units on each radiation monitor using appropriate calibration factors.

108

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 This section of the ODCM is implemented by the Radiation Monitor System Set Points procedure and the procedure for Releasing Radioactive Gaseous Waste.

@ 0 109

Number TMI - Unit 1 Radiological Controls Procedure I 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 22 This section of the ODCM is implemented by the Radiation Monitor System Set Points procedure and the procedure for Releasing Radioactive Gaseous Waste.

0 AAIx.

109

Number TMI - Unit 1 T Radiological Controls Procedure 6610-PLN-4200.01 Revislon No.

Offsite Dose Calculation Manual (ODCM) 22 4.2 TMI-1 Particulate and Radioiodine Monitor Set Points Set points for monitors which detect radionuclides other than noble gases are also established to assure that concentrations of these radionuclides in gaseous effluents do not exceed the limits of ODCM Part I Control 2.2.2.1.

Set points are established so as to satisfy the following equations:

1500 > j (cj)(F)(Pj)(Dv) (eq 4.2) where:

cj = set point concentration based on 1-131 equivalent, in ACi/Vc = , )

F = gaseous effluent flow rate at the monitor, in ccfsec, P1 = pathway dose parameter, in mrermyr per ltCim3r tr 'ha ation pathway from Table 4.6. The dose factors are based on the al indi *dual organ and most restrictive age group (child) (NUREG-0133). >/

NOTE Appendix A contains Pj calcula l ethodology.

1500 = annual dose rate limit to any particulates and radioiodines and radionuclides (other than noble gases) fft'1*les I greater than eight days in mrem/yr.

Dv = highest sector annu e t aseousl dispersion factor (X/Q or DIQ) at or beyond the unrestricted area u ary from Table 4.4 for releases from the station vent and Table 4.5 for all other rel s used for the inhalation pathway. Maximum values of XJQ presently use ar 17E-7 sec/m3 for station vent, at sector SE, and 1. 16E-5 seclm3 for all other r ea a ctors N and WNW.

The set point cjncen ni is further reduced such that concentration contributions from multiple release poin Iv d n lcombine to exceed ODCM Control limits.

The set point c*entration is converted to set point scale units on each radiation monitor using appropriate calibration factors.

This section of the ODCM is implemented by the Radiation Monitor Systems Set Points procedure and the procedure for Releasing Radioactive Gaseous Waste.

110

.: *(

Number TMI - Unit I Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.3 TMI-2 Gaseous Radiation Monitor Set Points TMI-2 Gaseous Radiation Monitors have their set points described in TMI Plant Procedure 1101-2.1.

Figure 4.5 provides a gaseous effluent release pathway diagram. Table 4.2 provides TM[-2 Radiation Monitor Data.

These set points are set in accordance with the Controls delineated in Part [1of this ODCM.

ISV 111

Number TMI - Unit 1 Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Off'ite Dose Calculation Manual (ODCM) 22 4.4 TMI-I Gaseous Effluent Release Points and Gaseous Radiation Monitor Data TM[-1 has eleven (11) required effluent gaseous radiation monitors. These are RM-A4, RM-A5, RM-A15, RM-A6, RM-A7, RM-A8, RM-A9, RM-A14, ALC-RMI-18, WHP-RIT-1, and RLM-RM-1.

These gaseous release points, radiation monitors, and sample points are shown in Table 4.1.

4-4.1 RM-A4IRM-A6 Fuel Handling and Auxiliary Building Exhaust RM-A4 is the particulate, radioiodine and gaseous radiation monitor for the TMI-I Fuel Handling Building Ventilation (see Figures 4.1 and 4.2). RM-A6 is the particulate, radioiodine, and gaseous radiation monitor for the TMI-1 Auxiliary Building Ventilation (see Figures 4.1 and 4.2). High alarms on RM-A4 or RM-A6 nob-W channels will initiate shutdown of the related building ventilation air supply sy em. hese two radiation monitors concurrently will satisfy requirements for the Statio relse point in place of RM-A8.

4.4.2 RM-A8 Station Ventilation Exhaust RM-A8 is the particulate, radioiodine and gase4 radia monitor for the TMI-1 Station Ventilation (see Figures 4.1 and 4.2). This in plan ftient radiation monitor also has an associated sampling panel with sampling lines locat before the sample filters. High alarm on RM-A8 noble gas low channeV will initiate shutdown of the Station Ventilation air supply systems. (The Fuel Hand]ing] d Auxiliary Building Ventilation). This radiation monitor satisfies requirements for t.ation Vent release point in place of RM-A4 and RM-A6.

4.4-3 RM-A5/RM-AI 5 Condens.fýas Exhaust RM-A5 is the gaseo atio" monitor for the TMI-1 Condenser Off Gas exhaust (see Figures 4.1 and 4 M- is the back up gaseous radiation monitor for the TMI-1 Condenser Off a e t (see Figures 4.1 and 4.4). High alarms on RM-A5 low channel or R 15no gas channels will initiate the MAP-5 Radioiodine Processor Station. The o diation monitors together satisfy requirements for the Condenser Off Gas rel 8ep oih*

4.4.4 RM Was Gas Decay Tank Exhaust R 7i the gaseous radiation monitor for the TMI-1 Waste Gas Decay tanks (see Figur4.1 and 4.2). This in plant effluent radiation monitor also has an associated sampling panel. High alarm on RM-A7 noble gas channel will initiate shutdown of the Waste Gas Decay Tank release in progress. This radiation monitor satisfies requirements for batch gaseous releases to the Station Vent release point 4.4.5 RM-A9 Reactor Building Purge Exhaust RM-A9 is the particulate, radioiodine and gaseous radiation monitor for the TMI-1 Reactor Building Purge system (see Figures 4.1 and 4.3). This in plant effluent radiation monitor also has an associated sampling panel with sampling lines located before the sample filters. High alarm on RM-A9 noble gas low channel will initiate shutdown of the Reactor Building Purge System. This radiation monitor satisfies requirements for the Reactor Building Purge System release point.

112

Number TMI - Unit I Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.4.6 RM-A14 ESF FHB Ventilation System RM-A14 is the gaseous radiation monitor for the TMI-1 Emergency Safeguards Features (ESF) Fuel Handling Building Exhaust system (see Figures 4.1 and 4.2). This in plant effluent radiation monitor also has an associated sampling panel with sampling lines located before the sampler filters. High alarm on RM-A14 noble gas channel will initiate shutdown of the ESF Fuel Handling Building Exhaust System. This radiation monitor satisfies requirements for the ESF Fuel Handling Building Exhaust System release point.

4.4.7 ALC-RMI-18 Chemical Cleaning Facility (CCF) Ventilation Exhaust ALC-RMI-1 8 is an Victoreen particulate, radiolodine, and gaseo iation monitor for the Chemical Cleaning building exhaust. This monitor is locatei , in th Chemical Cleaning building on the ground floor, andoff of the has monitor. sample an associated Snpiing for is performed particulate activity 4.4.8 WHP-RIT-1 Waste Handlin and Packaging Facilit H aust WHP-RIT-1 is an EberlineAMS-3 particulate tion itorfor.theTMIWHPF. The monitor is located in the Mechanical Equipment R .n the WHPF. Sampling for particulate activity is performed off of the monitor, igh alarm will initiate shutdown of the ventilation air exhaust system.

4.4.9 RLM-RM-1 Respirator Cleaning aýtlndry Maintenance (RLM) Facility RLM-RM-1 is an Eberline A*- u rt radiation monitor for the TMI RLM Facility.

The monitor is located in ,l a ical Equipment Room in the RLM. Sampling for particulate activity is pe rm ofof the monitor.

113

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.5 TMI-2 Gaseous Effluent Release Points and Gaseous Radiation Monitor Data TMI-2 has three.(3) regulatory required gaseous effluent radiation monitors. These are HP-R-219, HP-R-219A and HP-R-225. These gaseous release points, radiation monitors, and sample points are shown in Table 4.2, and various gaseous effluent pathways are depicted in Figure 4.5.

4.5.1 HP-R-219 Station Ventilation Exhaust HP-R-219 is a Victoreen particulate and gaseous radiation monitor for the TMI-2 ventilation exhaust. This in-plant effluent radiation monitor is located in the TMI-2 Auxiliary Building 328 foot elevation and has an associated sample panel.

4.5.2 HP-R-219A Station Ventilation Exhaust HP-R-219A is a Victoreen particulate and gaseous radiati nitor the TMI-2 ventilation exhaust. This in-plant effluent radiation mo ior i ted in the TMI-2 Auxiliary Building 328 foot elevation, 4.5.3 HP-R-225 Reactor Building Purge Air Exhaust cta

--

HP-R-225 is a Victoreen particulate and gaseous ation monitor for the TMI-2 Reactor Building Purge Air Exhaust System. T is in-plant effluent radiation monitor is located in the TMI-2 Auxiliary Building 328'e on area.

0a 114

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 4.6 Control of Gaseous Effluent Releases TMI gaseous effluent combined releases are controlled (per ODCM Part I for TMI-I and ODCM Part II for TMI-2) by effluent samplingand radiation monitor set points. These measures assure that releases from the various vents do not combine to produce dose rates at the site boundary exceeding the most restrictive of 500 mrem per year to the total body or 3000 mrem per year to the skin, and 1500 mrem per year to the thyroid. This is done by restricting simultaneous releases and by limiting the dose rates that may be contributed by the various vents at any time. The various vent radiation monitor set points are each based on fractions of the above limits and do not exceed the above limits when summed together. These effluent radiation monitor set points are calculated using the methodology described in equations 4.1.1, or 4.1.2 and 4.2. The actual set points are then listed in TMI-1 Operations Procedure 1101-2.1.

The radioactive content of each batch of gaseous waste is determine N or ase by sampling and analyses in accordance with ODCM. Part I for TMI-1 and ODC a II for MI-2. The results of pre-release analyses are used with the calculational methods in ctio1-4l and 4.2 to assure that the dose rates at the site boundary are maintained below the Up NO M Part I for TMI-1 and ODCM Part II for TMI-2.

Post-release analyses of samples composited from batch an r'tinuous releases are performed in accordance with ODCM Part I for TMI-I and ODCM Part II forM 1-2. The results of the analyses 0

are used to assure that the dose rates at the site undary are maintained within the limits of ODCM Part I for TMI-1 and ODCM Part I1for TMI-2.

115

Number TMI - Unit 1 Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 4.1 TMI-1 Gaseous Release Point and Gaseous Radiation Monitor Data GASEOUS RELEASE RADIATION GASEOUS (F) TERMINATION MONITOR RELEASE FLOW INTERLOCK (DETECTOR) LOCATION POINT RECORDER (YES/NO) VALVES YES 306" Elevation Fuel Hand. AH-E-10 RM-A4 Auxiliary Bldg. Building Exhaust FR-149 AH-D-120 AH-C

..... AU-A - 22 306' Elevation Auxiliary E RM-A6 Auxiliary Bldg. Building Exhaust FR-1 50 *

1 YES DG-V47 RMA-819 Bldg. FR 4& AH-E-10 RM-A8 Near BWST Vent & FI Exhaust Starts MAP-5 Radioiodine Sampler YES d

322' Elevation Cond r Starts MAP-5 RM-A5 Second FloorTurbne

Off* u*Radiciodine sBdg. FR-1 113 Sarts MAP-S Turbine Bldg. a Sampler 322' Elevati on nser YES -S RM-A15 Second F as FR-1 113 Starts MAP Turbin g. _haust Radimiodine Sampler Waste Gas 6' EI v n Decay YES R0 xilia ldg. Tanks FR123 WDG-V47 (A,B,C)

YES Reactor AH-V-1A/B/C/D RMA-8/9 Bldg. Building FR-909/ WDG-5341535 Near BWST Purge FR-148 Starts MAP-5 Exhaust Radiiodine Sampler 331' Elevation ESF Fuel NO RM-A14 ESF FHB Handling FR-1104A/B Manual Outside Chem. Building Addition Bldg. Exhaust Actions ii6

Number TMI - Unit I Title Radiological Controls Procedure 6610-PLN-4200.01 Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 4.1 TMI-1 Gaseous Release Point and Gaseous Radiation Monitor Data RELEASE GASEOUS TERMINATION RADIATION INTERLOCK MONITOR GASEOUS (YES/NO)

(DETECTOR) LOCATION RELEASE POINT VALVES Chemical CCB Exhaust ALC-RMI-18 Cleaning Bldg. System NONE 304' Elevation - (Typical flow rate is 10,000 cfm)

WHPF WHPF.Exhaust WHP-RIT-1 Mechanical s (Typical flow rate Equipment Room yi flow rate W p o is 7,500 cfm) i RLM Exhaust RLM-RM-I RLM-Mechanical System Equipment Room (Typical flow rate Y NONE is 00 cfm TABLE , a TMI-2 Gaseous Release Point VQ.se Radiation Monitor Data

" Rai da' RELEASE GASEOUS TERMINATION RADIATION GASEOUS INTERLOCK MONITOR RELEASE (YES/NO)

(DETECTOR) C POINT VALVES El . ation Station HP-R-219 Aux4 Vent NONE 9K Btlding Exhaust

,728' Elevation Station Vent HP- Auxiliary Exhaust NONE Building Reha ustor Bl HP-R-225 328' Elevation Reactor Bldg Auxiliary Buildin Auxliay BildngDuct Purge"A" Exhaust NONE 147

Number TMI - Unit I Radiological Controls Procedure 6610-PLN-4200.01 Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 4.3 Dose Factors for Noble Gases and Daughters*

Gamma Beta Total Body Skin Dose Gamma Air Dose Factor(a) Factor(b) Dose Factor Beta Air K* LI ; MI Dose Factor (mremlyr per (mremlyr per (mradlyr per Ni Radionuclide I!Ci/m 3) I . Ci/m3 ) I .CiCVm) (mradfyr per Ci/m 3)

Kr-83m 7.56E-02** -- 1.93E+01 /,,88E+02 Kr-85m 1.17E+03 1.46E+03 123E+03 1E+03 Kr-85 1.61E+01 1.34E+03 1.72E+014 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+-t., 1.03E+04 Kr-88 1.47E+04 2.37E+03 1. +04 2.93E+03 Kr-89 1.66E÷04 1.01 E+04 1.73E+X 1.06E+04 Kr-90 1.56E+04 7.29E+03 o 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 Is, 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94 0 3.27E+02 1.48E+03 Xe-133 2,94E+02 (0 6'N0 3.53E+02. 1.05E+03 Xe-135m 3.12E+03 7. 02 3.36E+03 7.39E+02 Xe-1 35 1.81 E+03,a0 .86E+03 1.92E+03 2.46E+03 Xe-137 1.42-03:ý 1.22E+04 1.51E+03 1.27E+04 Xe-I 38 83E 4.13E+03 9.21 E+03 4.75E+03 Ar-41 +3 2.69E+03 9.30E+03 3.28E+03 Dose factors are for immersion exposure in uniform semi-infinite cloud of noble gas radionuclides. that may be detected ingaseous effluents. Dose factor values are taken from Regulatory Guide 1.109 (Rev. 1), Table B-I.

7.56E-02 = 7.56 x 102.

(a) Total body dose factor for gamma penetration depth of 5 cm into the body.

(b) Skin dose factor at a tissue depth or tissue density thickness of 7 mg/cm 2.

i18

Number TMI - Unit I Radioloaical Controls Procedure-Title Revision No.

Offsite Dose Calculation Manual (ODCM) 22 TABLE 4.4 Atmospheric Dispersion Factors for Three Mile Island

"*STATION VENT DISTANCE

"*SECTOR AVERAGE XIQ ON SECIM) (IN METERS) SEASON - ANNUAL SECTOR 610 2413 4022 5631 7240 12067 24135 40225 56315 72405 N 1.18E-07 5.32E-07 2.95E-07 1.93E-07 1.39E-07 5.52E-08 1.91E-08 5.02E-09 1.88E-09 1.09E-09 NNE 1.70E-07 7.17E-07 3.45E-07 2.OOE-07 1.39E-07 5.58E-08 1.705-08 4.77E-09 1.98E-09 9.695-10 NE 1.12E-07 1.75E-07 3.26E-07 1.86E-07 12..1E-07 5.005-06 1.67E-08 4.67E-09 1,85E-09 9.93E-10 ENE 1.09E-07 2.13E-07 2.67E-07 1.53E-07 1.05E-07 4.31E-08 1.42E-08 4.425:-09 1.59E-09 8.64E-10 E 2.31 E-07 1.71E-07 1.52E-07 1.49E-07 1.06E-07 4.63E-08 1.52E-08 5-19-09 .48E-09 1.50E-09 ESE 3.50E-07 2.12E-07 2.50E-07 1.48E-07 9.48E-08 3.98E-08 1.502-08 , 9 k.92E-09 1.93E-09 SE 4.19E-07 3.79E-07 2.53E-07 1.55E-07 1.11E-07 4,82E-08 1.81E-. E-09 3.30&-09 2.22E-09 SSE 2.90E-07 3.62E-07 2.55E-07 1.49E-07 1.11E-07 5.02E-08 1.9800k 6. -09 2.94E-09 1.70E-09 S 1.87E-07 6.47E-08 2.16E-07 1.30E-07 8.65E-08 4.09E-08 I1 .96E-09 1.99E-09 1.04E-09 SSW 6.13E-08 4.16E-08 1.56E-07 1.03E-07 6.81E-08 2.72E-08 .7452 3.01E-09 1.50E-09 8.23E-10 SW 5.76E-08 1.14E-07 1.705-07 1.052-07 6.93E-08 2.51 E-08 9.' 09 2.722-09 1.33E-09 8.33E-10 WSW 8.52E-08 3.75E-07 2.14E-07 1.26E-07 7.74E-08 3.08E-08 1.E-08 3.28E-09 1.39E-09 9.69E-10 W 1.15E-07 5.80E-07 2.88E-07 1.63E-07 1.18E-07 f.23E-08 1.72E-08 5.08E-09 1.98E-09 1.25E-09 WNW 1.41E-07 6.28E-07 3.30E-07 2.19E-07 1.482-(L .. 68E-08 1.95E-08 6.32E-09 2.16E-09 1.34E-09 NW 1.42E-07 5,67E-07 3.17E-07 1.93E-07 1.3 7 5. VE-08 2.06E-08 5.90E-09 2.70E-09 1.45E-09 NNW 1.00E-07 5.77E-07 3.18E-07 1.802-07 ýN 5.202-08 1.77E-08 4.82E-09 2.01E-09 1.22E-09

" STATION VENT D SECTOR AVERAGE D/Q (IN M2) _,__ IN SEASON - ANNUAL

_TERS)

SECTOR 610 2413 4022 ^1 -7240 12067 24135 40225 56315 72405 N 2.51E-09 8.72E-10 4.84E-1 2..82- 2.50E-10 8.57E-11 2.51E-11" 4.98E-12 1.572-12 7.84E-13 NNE 3.89E-09 1.98E-09 9.542-V_. 2V-10 3.38E-10 1.10E-10 2.89E-1 1 6.06E-12 2.10E-12 8.89E-13 NE 2.58E-09 6.70E-10 9.OE-10 4 2E-10 2.97E-10 1.042-10 2.87E-11 6.01E-12 1.992E-12 9.23E-13 ENE 2.15E-09 5.85E-10 9.5410, 3.065-10 2.05E-10 8.30E-11 2.32E-11 5.412-12 1,6325-12 7.64E-13 E 5.54E-09 1.23E- .172A 4.592-10 3.63E-10 1.34E-10 3.66E-11 9.44E-12 3.77E-12 1.97E-12 ESE 9.17E-09 2.05 - 1. E-09 8.66E-10 5.11E-10 1.82E-10 5.77E-11 1.72E-11 7,072-12 4.07E-12 SE 1.22E-08 2.882-09 .84E-09 1.02E-09 6.85E-10 2.60E-10 8.30E-11 2.34E-11 9.42E-12 5.51E-12 SSE 7.50E-09 1.622-09 1.08E-09 5.89E-10 4.49E-10 1.872E-10 6.16E-11 1.612-11 5.67E-12 2.832-12 S 3.86E-09 6.53E-10 6.27E-10 3.59E-10 2.32E-10 1.06E-10 3.05E-11 8.10E-12 2.73E-12 1.23E-12 SSW 1.13E-09 2.94E-10 4.192-10 2-53E-10 1.56E-10 5.382-11 1.68E-11 3.91E-12 1.642-12 7.842-13 Sw 1.19E-09 3.84E-10 4.96E-10 2.80E-10 1.70E-10 5.242-11 1.65E-11 3.62E-12 1.49E-12 8.122-13 WSW 1.77E-09 8.312-10 6.492-10 3.50E-10 1.99E-10 6.73E-11 1.89E-11 4.58E-12 1.632-12 9.90E-13 W 2.412-09 1.29E-09 6.81E-10 3.65E-10 2.96E-10 1.12E-10 3.11E-11 6.90E-12 2.26E-12 1.25E-12 WNW 3.202-09 1.39E-09 7,73E-10 5.91E-10 3.66E-10 1.19E-10 3.43E-11 8.36E-12 2.39E-12 1.29E-12 NW 3.25E-09 1.23E-09 7.39E-10 4.22E-10 2.77E-10 1.14E-10 7.282-1i1 7.612-12 2.92E-12 1.36E-12 NNW 1.982-09 9.882-10 5.712-10 3.055-10 2.23E-10 8.212-11 2.41"E-11 4.93E-12 1.722-12 9.03E-13 DATA FROM 1/1/78 THROUGH 12/31/86 USED IN CALCULATIONS 119

ML021150025

Contents

Text

SUBJECT:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu

ML021150025

Text

SUBJECT:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu

Search