Text

Attachment 1, Offsite Dose Calculation Manual, Revision 23
	ML101300371
Person / Time
Site:	Salem
Issue date:	08/06/2009
From:	Shelton J; Public Service Enterprise Group
To:	; Office of Nuclear Reactor Regulation
References
	LR-N10-0140
	Download: ML101300371 (158)
	v • d • e

Attachment 1 ODCM Revisions Salem ODCM Rev 23

OFFSITE DOSE CALCULATION MANUAL 1 . FOR

PSEG NUCLEAR LLC SALEM GENERATING STATION, Revision 23 Prepared By:

Jenny Shelton Date Reviewed by:

Sal~n Chemistry Manager Robert Bernard Date SQR 6 &/-061'2o 7 Is Reviewed by Date Accepted by:

0 PORC Chairman -'

Michael Gwirtz Date Meeting #:, -ýS2ooq -oQog Approved by:

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I Salem ODCM Rev. 23

Revision Summary

1. Revised the definitions 1.15 and 1.16 for MEMBER(S) OF THE PUBLIC to be consistent with I the definitions in 10CFR20 and 40CFRI90. A member of the public is defined in 10CFR20 as "any individual except when that individual is receiving an occup ational dose." A member of the public is defited in 40CFRI 90 as "any individual that can receive a radiation dose in the general environment, whether he may or may not also be exposed to radiation in an occupation associated with a nuclear fuel cycle. However, an individual is not considered a member of.the public during any period in which the individual is engaged in carrying out any operation which is part of a nuclear fuel cycle."

Justification:

The Salem ODCM definitions 1.15 and 1.16 for member of the public were revised to be consistent with the definitions in 10CFR2O, 40CFR190 and revision 2 of Regulatory Guide 4.1 Program or Monitoring Radioactivity in the Environment that was recently published in June 2009.

2009 RETS FASA 70096339- 0110. I

2. Revised the definition 1.29 for SITE BOUNDARY by adding the words "or property" after "land".

Justification:

This change is based on the new definitions for site boundary in revision 2 of Regulatory Guide 4.1 Program or Monitoring Radioactivity in the Environment that was recently published in June 2009.

This definition allows for the inclusion of "property" such as structures and transmission towers to

be part of the site boundary,, if they arelocated on the property line.

U') ~2009 RETS FASA 70096339-0 120.I

3. The following sentence was added to the end of footnote 2 in Table 3.12.1-1 "Sector 7 does not have a direct radiation monitoring station in the outer ring due to inaccessibility." Added an explanation for

<Z not having a TLD in the outer ring (5 - 11 kim) of sector 7 (SE) as requirediby Table 3.12.1-1 item 1.

(n Item 1states that "An outer ring of stations, one in each land-based 'meteorological sector in the 5 to 11-kI n

range from the-site (not bounded by or over water)". The footnote identifies that no direct radiation monitoring station are located in the outer ring in sector 8, but fails to mention'that no TLD is collectedin U') Sector 7. .

0 Justification:

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> This change was made to accurately reflect the fact that two sectors do not have direct radiation monitoring station located in the outer ring. Sector 7 is located in marsh land and no roads are available to provide access 0 to this sector location. NUREG 1301 Offsite Dose Calculation Manual Guidance'allows Table 3.12.1 Table Notation (2) third sentence states: "The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly." Therefore, no TLD location:in Sector 7 should be

> considered acceptable by the NRC since this sector location is surrounded by water.2009 REMP FASA; 0ay 0

70089372-0250.

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Salem ODCM Rev. 23 I* 4 Deleted the words "centerline of one reactor" in first sentence of the third paragraph section 6.9.1.7 since it was inaccurate. It was replaced with "midpoint of a line, between the center of the Salem units 1 & 2 containment domes". The same changewas made to Table 3.12.1-1 Table Notation 1 first sentence.

The words "from the centerline of one reactor" were replaced with "midpoint of a line between the center of the Salem units 1 & 2 containment domes.1' Justification:

This change was made to correct the location that was copied from NUREG 1301 Offsite Dose Calculation Manual Guidance, Table 3.12.1, Table Notation (1) namely, "centerline of one reactor."

This guidance is for a single reactor site. The reference location for Salem 1 and 2 is the midpoint of a line between the center of the Salem units 1 & 2 containinent domes. This reference point was used by PSEG drafting to determine the distance and direction to~each sampling/monitoring location (see Table E-l).

2009 REMP FASA 70089372-0240.

5. Appendix E Radiological Environmental Monitoring Program Table E- 1 REMP Sample Locations: Deleted the words "of vent" after the direction specified for each iStation Location.). Added some descriptions to Sample Stations 2S4, 4Si, 15S1, 16S1, I 1E2 and 2F3. Added the following alternate sampling locations for surface water locations on the Delaware River 1 Ala, l1C1a, 7Ela, and 1F2.

Justification: The reference point for Salem vent was ch.anged because it was inaccurate (see 4. above). The additional descriptions were added to some locations for clarification. The alternate surface water locations allow safer access to these locations on the river during the winter months. These alternate locations SI. are in the general vicinity of the primary location and wduld have the same distance and direction listed in Table E-1. 2009 REMP FASA 70089372-220

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6. Appendix E Radiological Environmental Monitoring Program: Add a footnote to the Station Location 1- which reads" *All distances and directions for the Station Locations are referenced to the midpoint between the two Salem units' containments. The WGS 84 coordinates for this site center point I F ~ location are: Latitude N 390 - 27' - 46.5" and Longitude W 750 - 32' - 10.6". This footnote more accurately explains the reference point that was used for determining distances and directions to the sample/monitoring locations (see 4 above).

VJustification:. The reference point for Salem 1 and 2 was changed because it was inaccurate (see 4. above).

z 2009 REMP FASA 70089372-0240

-- 7. Revise the location of the 5S1 air sampling station frown "1.0 miles E of vent, site access road" to "0.95 miles E, site access road". The movement of the air sampiling station was IAW DCP 80088766.

Z Justification: This change was implemented to provide a more reliable power source to the air sampling station. Air sampling station 5S1 has experiencing severial missed air samples per year as a result of unplanned power outages. This new feed has a back-up liesel generator which will supply power in the.

unlikely event of a loss of power. DCP 80088766 and Nbtifications 20426132 and 20426221 at

8. Added clarification to TABLE 3.3-12: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Table Notations Action 28, itemib and c, added the words "inside containment" Justification: This change was implemented to satisfy order 70088866. Which requested clarification as to Isampling 0

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what is meant by service water leakage on the containment fan coil unit and to clarify when increased should take place.

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I Salem ODCM Rev. 23 I

S 9. Revise Section 2.5.2: to add to the first sentence; "for the resident/dairy location" after the words "dose I assessment".

Justification: This clarification was made to this dose assessment method6logy to better align it with the derivation and justification of the simplified dose calculation described in Appendix D.

I 2009 RETS FASA 70096339-0140.

10. Revise Section 3.1: to align with the new definition of a "member of the public" (see 4. above). Also, I

identify the assumptions employed in current year dose calculation (for amember of the public inside the site boundary) to better align with the Requirements of section 6.9.1.8 Annual Radiological Effluent Release*

Report.

I Justification: This change was in response to a Program Improvement that was recommended in the 2009 RETS FASA 70096339-0150. To better align with the revised definition of "member of the public" and I requirements of section 6.9.1.8 Annual Radiological Effluent Release Report 2009 RETS FASA 70096339-0150.

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11. Editorial Change: Table 3.3-12, Action 28a, removed (Unit2).

Justification: Editorial change that corresponds to DCP 80059610, for the Salem Unit 1 (IR13) radiation I

monitors. Both units now have the same radiation monitors.

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Salem ODCM Rev. 23 5

INTRODUCTION TABLE OF CONTENTS

.............................................................................................. 9 PART I - RADIOLOGICAL EFFLUENT CONTROLS ............................... ..................... ............ 10 1.0 DEFIN ITION S ......................................................................................................................... 12 3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS.......................... 18 3/4.0 A PPLICABILITY................................................................................ I.................................. 18 3/4.3 INSTRUMENTATION ................................................. .............. 20 3/4.3.3.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ... 20 3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 26 3/4.11 RADIOACTIVE EFFLUENTS ......................................... 31 3/4.11.1 LIQUIDEFFLUENTS... . ...................................... 31 3/4.11.1.1 CONCENTRATION ................................................................................ ........... 31 3/4.11.1.22 D O SE .................. ...................................................................................................... 35 3/4.11.1.3 LIQUID RADWASTE TREATMENT ............................. 36 3/4.11.2 GASEOUS EFFLUENTS .........................................

. 37 3/4.11.2.1 DOSE RATE ............................................... 37 3/4.11.2.2 DOSE - NOBLE GASES ....................................... 40 3/4.11.2.3 DOSE - IODINE- 131, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM ............................................................ ............

. ......... 41 3/4.11.2.4 GASEOUS RADWASTE TREATMENT ............................ i...................... ;...... 42 3 4.11.4 ................................. .......................................................... ...... . .. 43 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING ...................... . 44 V) 3/4.12.1 MONITORING PROGRAM.......................... ............. 44 3 4.12.2 LAND USE CENSUS............................. .................. ..................... ..... 57 3/4.12.3 INTERLABORATORY COMPARISONPROGRAM ........................ 59 B A SE S .............................................................................................. ............................................... 60 z 3/4.3 INSTR UMENTATION ......... ..................................... 61 3/4.3.3.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION .... 61 3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 62 I <3/4.11 RADIOACTIVE EFFLUENTS..........4........................................64

) ~3/4.11.1 LIQUID EFFLUENTS..... .................................... 64 3/4.11.2 GASEOUS EFFLUENTS ..................................................................................... 65 o3/4.11.4 TOTAL DOSE ........... .............................................. 67 314.12 RADIOLOGICAL ENVIRONMENTAL MONITORING ....................... ..... 68 3/4.12.1 MONITORING PROGRAM ............. '" ...... .................. 68 3/4.12.2 LAND USE CENSUS........... .................. 68 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ........................................ 68

.t 5.0 DESIGN FEATURES ................................................. 70

> 5.1 SITE .......................................................................................................................................... 70 ay ,5.1.3 UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID 0 EFFLU EN TS ................................................................................. ................................... 70 6.0 ADMINISTRATIVE CONTROLS ........................................................... ............. 72

6.9.1.7 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATI'NG REPORT ......... 72 6.9.1.8 RADIOACTIVE EFFLUENT RELEASE REPORT ............................................... 72 n 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATM ENT SYSTEM S .......................................................................................... 74 Page 5 of 157 Uc

Salem ODCM Rev. 23 S PART II - CALCULATIONAL METHODOLOGIES ...................................................... ....... ..... 75 1.0 LIQU ID EFFLU EN TS ................................................................................. I........... ........ .76I6............

1.1 RadiationMonitoringInstrumentationand Controls.......................................................... 76 1.2 Liquid Effluent Monitor Setpoint Determination.2ý...... ............... ......... ................. :...... 76 1.2.1 Liquid Effluent Monitors (Radwaste, Steam Generator Blowdown, Chemical Waste Basin and Service W ater ...................................... ..........................

I 77 1.2.2 Conservative Default Values ....................................................................................... 78 1.3 Liquid Effluent ConcentrationLimits - 10 CFR :20 ............................................................... 79 1.4 Liquid Effluent Dose Calculation- 10 CFR 50. ... " ........................ ...................... 79 1.4.1 MEMBER OF THE PUBLIC Dose - Liquid Effluents. .......... ............ 79 1.4.2 Simplified Liquid Effluent Dose Calculation ..............

1.5 Secondary Side Radioactive Liquid Effluents and Dose CalculationsDuringPrimary to

. ...... 8....1....... I Secondary L eakage .......................................................................... ........................................ 81 1.6 Liquid Effl uent Dose Projections.......................................... ........... .......................... .. 82 I 2.0 GASEOUS EFFLUENTS ........ . ...... ..................................................................... 83 2.1 RadiationMonitoringInstrumentation and Controls........................... 83 2.2 Gaseous Effluent MonitorSetpoint Determination............................ .................................. 84 2.2.1 Containment and Plant Vent Monitor ...... ....... .............. ...... i................ 84 8,4.........................

2.2.2 Conservatiye Default Values ....................................... ............ 85 2.3 Gaseous Effluent InstantaneousDose Rate Calculations -10 CFR 20............................. 85 S 2.3.1 Site Boundary Dose Rate -Noble Gases".... .........................

2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates ..............

2.4 Noble Gas Effluent Dose Calculations- 10 CFR S*..............................

50 .............................

........................... 85

............. 87 87 U 2.4.1 UNRESTRICTED AREA Dose - Noble Gases .... ......................... 87 2.4.2 Simplified Dose Calculation for Noble Gases ................................ :............................. 88

< 2.5 Radioiodine and ParticulateDose Calculations- 10 CFR 50 ....... 89 2.5.1 UNRESTRICTED AREA Dose - Radioiodine and Particulates .... ................................. 89 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates ...... ........... 89

-2.6 Seqondary Side Radioactive Gaseous Effluents andDose Calculations............................... 90 D,2.7 G aseous Effl uent Dose Projection.................................. .................... .......... ........................ 92 9...

< 3.0 SPECIAL DOSE ANALYSES ........................................... 93 3.1 Doses Due To Activities Inside the SITE BOUNDARY ... , ............................. 93

z 3.2 Total dose to MEMBERS OF THE PUBLIC -.4.0 CFR 190............................................... 93 0 3.2.1 Effluent Dose Calculations ......................................................................................... 93

>,. 3.2.2 Dhiect Exposure Dose Determination ......... ............................. .............................. 94 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.............................. 94 o 4.1 Sampling Program............................................................ ............................................94 4.2 InterlaboratoryComparisonProgram.... *..... ............................... 94 i~iI Z-0En 0

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I iSalem ODCM Rev. 23 TABLES TABLE 1.1: OPERATIONAL M ODES ....................................................................................... 15 TABLE 1.2: FREQUENCY NOTATION .................................... 16 TABLE 3.3-12: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

............................................................................. ........... .................... ......... .................. ...... .......... 2 1l I TABLE 4.3-12: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

~ ~~SURVEILLANCE SUAN VE L E REQUIREMENTS RE U RE E TS .......

TABLE 3.3-13: RADIOACTIVE GASEOUS EFFLUENT MONITORING

. . .. . . . . 224 INSTRUM ENTATION ................................... ....................................................................

TABLE 4.3-13: RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ........................................... 29 TABLE. 4.11-1: RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

° °.. ......... .°.°......... °..... ... ............................. - ..... .... °... . ............ o.o °.......................... °....... ........... 32*

TABLE 4.11-2: RADIOACTIVE GASEOUS .-WASTE SAMPLING AND ANALYSIS PR O G R AM ................................................................................................................................... 38 TABLE 3.12.1-1: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM *..... 46 TABLE 3.12-2: REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES SA M EPL.......... ........ . .............................. .............. ................ ....... 53 TABLE 4.12-1: DETECTION CAPABILITIES [FOR ENVIRONMENTAL SAMPLE AN Table ALYSIS 1-1.1: (I), (2)

Parameters ...............

for Liquid Alarm Setpoint ...... ....................................................................

Determinations Unit 1. .............. 98 54 I Table 1-1.2: Parameters for Liquid Alarm Setpoint Determinations- Unit22............. ............ 99 TABLE 1-2: Site Related Ingestion Dose ComrAitment Factor, Ai. ........................ . .. . . . . . .. . . . . . . .. 100 Table 1-3: Bioaccumulation Factors ............ .... 102 Table 2-1: Dose Factors For Noble Gases 105 Gase Alarm s .......... D.t....nat........................................ 106

-i ~ Table ~~Table 2-2.1: Parameters for 2-2.2: for Gaseous Alarm Setpoint Determinations -- Unit 21 ..................

Set~oint Determinations ........... 107 106 Table 2-3: Controlling Locations, Pathw ays ...... ........................................................................ 108 Table 2-4: Pathway Dose Factors - Atmospheric Releases ........................ 109 D Table A.1: Calculation of Effective M PC - Unit! I ..................................................................... 124

< Table A-.2: Calculation of Effective MPC - Unit!2 ..................................................................... 125 Table B-1: Adult Dose Contributions - Fish and!Invertebrate Pathways - Unit 1 .................. 129 0" Table B-2: Adult Dose Contributions -. Fish and!Invertebrate Pathways - Unit 2 ...................... 130 Table C-i: Effective D ose Factors ...................... ....................................................................... 135 Z> Table D-l: Infant Dose Contributions ................ :...... I................................................................ 139.

ry TABLE E-1: REM P Sample Locations ...................................................................................... 142 I _Table F-l: M aximum Permissible Concentrations ..................................................................... 151 0

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I Salem ODCM Rev. 23 I S .I FIGURES FIGURE 5.1-3: AREA PLOT PLAN OF SITE..... ......................................................... 71 I Figure 1-1: Liquid Release Flowpath Unit 1 ................................ 95 Figure 1-2: Liquid Release Flowpath Unit 2."'.....................................................................

Figure 1-3: Liquid Radioactive W aste System ..................... ................ "..........................

96 97 I Figure 2-1: Salem Ventilation Exhaust Systems and Effluent Monitor Interfaces ........... 103 Figure 2-2: Gaseous Radioactive Waste Disposal System .......................

Figure 2-2: Gaseous Radioactive Waste Disposal System ......................

Figure E-1: ONSITE SAMPLING LOCATIONS ....... ...... ...........

103 104 148 I

Figure E-2: OFFSITE SAMPLING LOCATIIONS ............. ........................ ......... 149 I

APPENDICES APPENDIX A: EVALUATION OF DEFAULT PARAMETERS FOR LIQUID EFFLUENTS............................... 122 I APPENDIX B: TECHNICAL BASIS FOR SIMPLIFIED DOSE' CALCULATIONS - LIQUID EFFLUENTS . 127 APPENDIX APPENDIX C: TECHNICAL BASES FOR EFFECTIVE DOSE FACTORS - GASEOUS EFFLUENTS ................

D: TECHNICAL BASIS FOR SIPLIFIED DOSE-CALCULATION - GASEOUS EFFLUENTS .......

132 137 I APPENDIX E: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ...... ........... 141 APPENDIX F: MAXIMUM PERMISSIBLE CONCENTRATI6N (MPC) VALUES - LIQUID EFFLUENTS ........ 151 I

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1.10 The Limiting Conditions for Operation (LCOs) that were contained in theRadiological Effluent a: Technical Specifications were transferred to the OFFSITE DOSE CALCULATION MANUAL

_ (ODCM) and were renamed CONTROLS. This is to distinguish between those LCOs that were retained in the Technical Specifications and those LCOs or CONTROLS that were transferred to the ODCM.

i> DOSE EQUIVALENT 1-131 1.11 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram), which alone would produce the same thyroid dose as the quantity, and isotopic mixture of 1-131, 1-132, 1-w 133, 1-134, and 1-13 5 actually present. The thyroid dose conversion factors used for this calculation mO 0 shall be those listed in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Z Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and 0

no_ Ingestion".

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Salem ODCM Rev. 23 I O FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.
GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
MEMBER(S) OF THE PUBLIC 1.15 MEMBER(S) OF THE PUBLIC member of the public (10 CFR 20) - Means any individual except when that individual is receiving an occupational dose.
1.16 MEMBER(S) OF. THE PUBLIC (40 CFR 190) - Means any individual that can receive a radiation dose in the general environment, whether he may or may not also be exposed to radiation in an occupation associated with a nuclear fuel cycle. However, an individual is not considered a member of the public during any period in which the individual is engaged in carrying out any operation which is part of a nuclear fuel cycle.
OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculationi of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the I ~
conduct of the environmental radiological monitoring program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring'Programs required by Technical Specification Section 6.8.4 and (2) descriptions of the information that should be included I2 in the Annual Radiological Environmental Operating and the Radioactive Effluent Release Reports required by Technical Specification Sections 6.9.1.7 and 6.9.1.8, respectively.
D OPERABLE - OPERABILITY
< 1.18 A system, subsystem, train, component or device shall be OPERABLE orihave OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, normal or emergency electrical power source, cooling and seal water,
- lubrication or other auxiliary equipment that are required for the system, subsystem, train,
>. component or device to perform its specified safety function(s) are also capable of performing their related support function(s).
I OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination
> of core reactivity condition, power level and average reactor coolant temperature specified in Table I> 1.1.
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w PURGE - PURGING m 1.23 PURGE or PURGING shall be the controlled process of discharging air or gas from a W confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
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I Salem ODCM Rev. 23
. RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.
REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10CFR Part 50 or 10CFR 72.75.
SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land or property is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3.
SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel I sensor is exposed to either (a) an external.source of increased radioactivity, or (b) an internal source of radioactivity (keep-alive source), or-(c) an equivalent electronic source check.
THERMAL POWER 1.33 THERIAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
UNRESTRICTED AREA I 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to I radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential
" quarters or industrial, commercial, institutional, and/or recreational purposes.
VENTILATION EXHAUST TREATMENT SYSTEM 1.36 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine and radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).
Engineered Safet Feature (ESF) atmospheric cleanup systems are not considered to be Z VENTILATION EXHAUST TREATMENT SYSTEM components.
VENTING 1.37 VENTING shall be the controlled process of discharging air or gas from a confinement to z maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in ad: system names, does not imply a VENTING process.
0: WASTE GAS HOLDUP SYSTEM 1.41 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Z Coolant System and providing for delay or holdup for the purpose of reducing:the total radioactivity 0 prior to release to the environment.
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I I Salem ODCM Rev. 23 I. TABLE 1.1: OPERATIONAL MODES REACTIVITY AVERAGE COOLANT MODE CONDITION, Keff THERMAL POWER* TEMPERATURE
1. POWER OPERATION > 0.99 >5% > 350OF
2. STARTUP > 0.99 <5% > 350OF
3. HOT STANDBY < 0.99 0 > 350°F
4. HOT SHUTDOWN < 0.99 0 3500 F > Tavg
> 200OF
5. COLD SHUTDOWN < 0.99 0 < 200°F
6. REFUELING** <_0.95 0 < 140°F It.d
Excluding decay heat.
7 ** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
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I Salem ODCM Rev. 23 I
TABLE 1.2: FREQUENCY NOTATION I NOTATION FREQUENCY U S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . I w At least once per 7 days.
M At least once per.31 days.
I Q At least once per 92 days.
SA At least once per 6 months.
R At least once per 18 months.
S/U Prior to each reactor startup.
P Prior to each release.
N.A. Not applicable.
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Salem ODCM Rev. 23 i1 SECTIONS 3.0 AND 4.0 CONTROLS AND SURVEILLANCE REQUIREMENTS S
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I Salem ODCM Rev. 23 3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY CONTROLS 3.0.1 Compliance with the CONTROLS contained in the succeeding CONTROLS is required during I the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the CONTROL, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a CONTROL shall exist when the requirements of the CONTROLS and associated ACTION requirements are not met within the specified time intervals. If the CONTROL is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a CONTROL is not met except as provided in the associated ACTION requirements, within I one hour action shall be initiated to place the unifin a MODE in which the CONTROL does not apply by placing it, as applicable, in: I
1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , V)
S 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and
3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . I Where corrective measures are completed that permit operation under the ACTION requirements, the z ACTION may be taken in accordance with the specified time limits as measured from the time of failure
< to meet the CONTROL. Exceptions to these requirements are stated in the individual CONTROLS:.
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I< This CONTROL is not applicable in MODE 5 or 6.
(n 1 2" 3.0.4 Entry into an OPERATIONAL MODE or other specified condition:
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>y (a) shall not ACTION be made requires a shutdown conditions of the CONTROL are not met and the associated when the if.they W~e are not met within a specified time interval.
(b) may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time.
ry This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply
,O with ACTION requirements. Exceptions to these requirements are stated in the individual CONTROLS.
M 70 0~ ~I Wi~ I Page l8 of 157, CrC DI
Salem ODCM Rev. 23 APPLICABILITY I* SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual CONTROLS unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by CONTROL 4.0.2, shall constitute a failure to meet the OPERABILITY requirements for a CONTROL. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowed outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the I*
(I)
Surveillance Requirement(s) associated with the CONTROL has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to w OPERATIONAL MODES as required to comply with ACTION requirements.
0 Page 19 6f 157
Salem ODCM Rev. 23 S 3/4.3 INSTRUMENTATION i 3/ 40.3.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CONTROLS i 3.3.3.8 In accordance with Salem Units 1 and 2 Technical Specifications 6.8.4.g.1, the radioactive i liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of CONTROL 3.11.1.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).
APPLICABILITY: During all liquid releases via these pathways. i ACTION:
a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above CONTROL, without delay suspend the i release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative.
S b. With less than the minimum number of radioactive liquid effluent monitoring i Linstrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Exert z best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next radioactive effluent release report why the inoperability was not corrected in a timely manner.
z
c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
H--
(A 2£ SURVEILLANCE REQUIREMENTS 0
(A 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL
~CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies showni in Table 4.3-12.
LUj ry" I..t..
c1ý
~I n0 0~
(A LU ac PI Page 20 of 157 (A
03.
M M M M M M M Mm M M Mm M M M - MM USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES Salem ODCM Rev. 23 TABLE 3.3-12: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS T1IITR 1 ThATVIT OPERABLE ACTION
1. GROSS RADIOACTIVITY MONITORS PROVIDING AUTOMATIC TERMINATION OF RELEASE
a. Liquid Radwaste Effluent Line 1 (1R18, 2R18) 26
b. Steam Generator Blowdown Line 4 (1R19A-D, 2R19A-D) 27
2. GROSS RADIOACTIVITY MONITORS NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
a. Containment Fan Coolers - Service Water Line 2(Unit 1) (1R13A, B) 28 Discharge 2 (Unit 2) (2R13A, B)-
b. Chemical Waste Basin I (R37) 31
3. FLOW RATE MEASUREMENT DEVICES
a. Liquid Radwaste Effluent Line 1 (1FR1064, 2FR1064) 29
b. Steam Generator Blowdown Line 4 (1FA-3178, -3180, -3182, -3184, 29 2FA-3178, -3180, -3182, -3184).
Page 21 of 157
Salem ODCM Rev. 23 TABLE 3.3-12 (Continued)
TABLE NOTATION ACTION 26 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:
a. At least two independent samples, are analyzed in accordance with CONTROL 4.11.1.1.1, and
b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway. I ACTION 27 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement; .effluent releases via this pathway may continue provided grab samples are analyzed for principalgamma emitters, 1-131, and dissolved and entrained gases at the lower limits of detection required in 0DCM CONTROL Table 4.1 1-I.B, and the ODCM Surveillance Requirement 4.11.1.1.2 is performed:
a.
b.
At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT 1-131, or At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is I
"0 less than or equal to 0.01 microcuries/gram DOSE EQUIVALENT 1-131.
z "I
C~) ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that:
a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , local monitor readouts for the affected channels are D verified to be below their alarm setpoints, or
b. With a Service Water System leak (inside containment) on the Containment Fan Coil Unit associated with the inoperable monitor either:
z
_ 1. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are to be collected and analyzed for Wry principal gamma lower limits of detection 1-131, and emitters,specified dissolved in ODCM and entrained gases at the CONTROL Table 4.11 -1.B, and*
0 z_ the ODCM Surveillance Requirement 4.11.1.1.2 is performed, or
>- 2. Isolate the release pathway.
> c. With no identified service water leakage (inside containment) on the Containment ry Fan Coilgrab Unit associated with'the inoperable monitor, at-least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 0collect samples and analyze for principal gamma emitters, 1-131, and dissolved and entrained gases at the lower limits of detection specified in ODCM
_* CONTROL Table 4.11-1 .B, and the ODCM Surveillance Requirement 4.11.1..1.2 z0 is performed.
0~
(I)I a: Page 22 of 157
DI
Salem ODCM Rev. 23 1 0 TABLE 3.3-12 (Continued)
TABLE NOTATION ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, 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 performance curves may be used to estimate flow.
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 sampling is conducted in accordance with the following table:
Frequency Condition 1 per week During normal operation (all MODES) 1 per day During operation with an identified primary to secondary leak on either Salem Unit I* The samples shall be analyzed for principlal gamma emitters, 1-131, and dissolved and entrained gases at the lower limits qf detection specified in ODCM CONTROL Table 4.11-1..B, and the ODCM Surveillance Requirement 4.11.1.1.2 shall be performed.
<z
-J I.e 0*
Page 23 of 157
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES 00 Salem ODCM Rev. 23 TABLE 4.3-12: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST
1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE
a. Liquid Radwaste Effluent Line D P# R(3) Q(I)
b. Steam Generator Blowdown Line D M R(3) Q(1)
2. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
a. Containment Fan Coolers - Service Water Line D M R(3) Q(2)
Discharge
b. Chemical Waste Basin Line D M R(3) Q(5)
3. FLOW RATE MEASUREMENT DEVICES
a. Liquid Radwaste Effluent Line D(4) N.A. R N.A.
b. Steam Generator Blowdown Line D(4) N.A. R N.A.
Page 24 of 157 m - - - M Mm M M M-- M M M -MM
Salem ODCM Rev. 23 TABLE 4.3-12 (Continued)
TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates measured levels at br above the alarm/trip setpoint.
2. Circuit failure. (Loss of Power)
3. Control Room Instrument indicates a downscale failure.
i(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates measured levels at or above the alarm/trip setpoint.
2. Circuit failure. (Loss of Power) "
3. Control Room Instrument indicates a downscale failure.
4. Instrument controls not set in operate mode. (On instruments equipped with operate mode switches only {Unit 1})
w 0 (3) The initial CHANNEL CALIBRATION was performed using appropriate liquid or gaseous calibration sources obtained from reputable suppliers. The activity of the calibration sources
< 7- were (now reconfirmed using a multi-channel analyzer which was calibrated using one or more NBS NIST) standards.
Z (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release.
0o CHANNEL CHECK shall be made at least once. per. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.
o (5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates measured levels at or above the alarm/trip setpoint.
0 2. Circuit failure. (Loss of Power)
U- # The RM8's channels off-line channels which requires periodic decontamination. Any count rate ry indication above 10,000 cpm constitutes a SOURCE CHECK for compliance purposes.
0 U-V) Page 25 of 157
I Salem ODCM Rev. 23 3/4.3 INSTRUMENTATION I 3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT -MONITORING INSTRUMENTATION CONTROLS 3.3.3.9 In accordance with Salem Units I and 2 Technical Specifications 6.8.4.g.1, the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of CONTROL 3.11.2.1 are not exceeded. The alarm/trip setpoints of these 6hannels shall be determined in accordance with the ODCM.
APPLICABILITY: As shown in Table 3.3-13 I ACTION:
a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above CONTROL, without delay suspend the release of radioactive gaseous effluents monitored by the affected I channel or declare the channel inoperable or change the s.etpoint so it is acceptably conservative.
b. With less than the minimum number'of radioactive gaseous effluent monitoring (D instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
Exertbest efforts to return the instrument to OPERABLE status within 30 days (9 and, if unsuccessful, explain in the next radioactive effluent release report why the inoperability was not corrected in a timely manner.
0) c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-13.
0 I 0
cI
-. _1 cL LP 0A
I Page 26 of 157
m- M M M------ M- M- M- - M USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES Salem ODCM Rev. 23 TABLE 3.3-13: RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPI TC'AflTT 1TV APPY YPARYT TTV AAPTU) fCTTC'YNI
1. WASTE.GAS HOLDUP SYSTEM
a. Noble Gas Activity Monitor - Providing 1 (1R41A&D, 31 Alarm and Automatic Termination of Release 2R41A&D)
2. CONTAINMENT PURGE
a. Noble Gas Activity Monitor 1 (1R12A or 1R41A&D, ** 34 2R12A or 2R41A&D). #
3. CONTAINMENT PRESSURE - VACUUM RELIEF
a. Noble Gas Activity Monitor 1 (1R12A or 1R41A&D ** 37 2R12A or 2R41A & D) #
4. PLANT VENT HEADER
a. Nobl&Gas ActivitySYSTEM##
Mohitor
...'. 1 (1R4lA&D, 2R41A&D) ... ..* . .33 ..... .
b. Iodine Sampler 1 (1RME4, 5 or 1XT8911, 36 2RME4, 5 or 2XT8911)
c. Particulate Sampler 1 (1RME4, 5 or 1XT8911,
36 2RME4, 5 or 2XT8911)
d. Process Flow Rate Monitor (stack) I (1RM-1FA8603,
32 2RM-2FA8603)
e. Sampler Flow Rate Monitor 1 (IRM-1FA17079 or S1PAS-1FA6863Z,
32 2RM-2FA17079 or S2PAS-2FA6863Z)
The following process streams are routed to the plant vent where they are effectively monitored by the instruments described:
(a) Condenser Air Removal System (b) Auxiliary Building Ventilation System (c) Fuel Handling Building Ventilation System (d) Radwaste Area Ventilation System (e) Containment Purges & Pressure-Vacuum Relief Page 27 of 157
I Salem ODCM Rev. 23 I
S TABLE 3.3-13 (Continued)
TABLE NOTATION I
ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release: I
a. At least two independent samples of the tank's contents are analyzed, and
b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valving lineup; Otherwise, suspend release of radioactive effluents via this pathway.
I ACTION 32 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow I
rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab I
samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gaseous principal gamma emitters at the lower limits of detection required in ODCM CONTROL TABLE 4.11-2.A, B, or C within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, suspend release of radioactive effluents via this pathway.
I ACTION 34 -
With the number of channels OPERABLE- less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this I
pathway.
LU ACTION 36 - 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 I
4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months samples are continuously collected with auxiliary sampling equipment as required in 0-1 Table 4.11-2.
With the number of channels OPERABLE less than required by the Minimum Channels I
ACTION 37 -
c,-
LD OPERABLE requirement, Containment Pressure Reliefs may be performed provided that prior to initiating the release: I
a. At least two independent samples of containment are analyzed, and
b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations.
Otherwise, suspend release of radioactive ieffluents via this pathway.
I 0U-At all times, other than when the line is valved out and locked.
During Containment Purges OR Containment Pressure - Vacuum Relief I
U--
d-h 2:
0*
APPLICABILITY:
Modes 1-6, R4IA/D Monitors providing Alarm and Automatic Termination of Release, or Modes 1-5, R12A Monitor providing Alarm and Automatic Termination of Release, or I
0<L 0© Mode 6, 'R12AMonitor providing Alarm only (Automatic Termination of Release is not required).
0 LUi LUJ LU During Mode Undefined (Defueled) operation, containment purge is reclassified as a building ventilation process stream monitored by the PLANT VENT HEADER SYSTEM. I I
0 # During movement of irradiated fuel within containment with the Containment Equipment Hatch OPEN, only DJ D
R41A/D can be credited for MINIMUM CHANNEL OPERABLE.
During movement of irradiated fuel within containment with the Containment Equipment Hatch CLOSED, R41A/D or R12A may be credited for MINIMUM CHANNEL OPERABLE.
V)
I Page 28 of 157 I
- M Mm M---- M---- M M MM -
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES Salem ODCM Rev. 23 TABLE 4.3-13: RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURV1ELLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED
1. WASTE GAS HOLDUP SYSTEM
a. Noble Gas Activity Monitor - Providing P P R(3) Q(M)
Alarm and Automatic Termination of Release
2. CONTAINMENT PURGE AND PRESSURE - VACUUM RELIEF
a. Noble Gas Activity Monitor P P R(3) Q(M) **
3. PLANT VENT HEADER SYSTEM#
a. Noble Gas Activity Monitor D M R(3) Q(2) *
b. Iodine Sampler W N.A. N.A. N.A. *
c. Particulate Sampler W N.A. N.A. N.A. *
d. Process Flow Rate Monitor (stack)' D '-N.A. R' N.A.
e. Sampler Flow Rate Monitor W N.A. R N.A. *
The following process streams are routed to the plant vent where they are effectively monitored by the instruments described:
(a) Condenser Air Removal System (b) Auxiliary Building Ventilation System (c) Fuel Handling; Building Ventilation System (d) Radwaste Area Ventilation System -
(e) Containment Purges & Pressure-Vacuum Relief Page 29 of 157
Salem ODCM Rev.-23 TABLE 4.3-13 (Continued) I TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates measured levels above the alarm/trip setpoint. U
2. Circuit failure. (Loss of Power)
3. Control Room Instrument indicates a downscale failure. (Alarm Only)
I (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
1. Instrument indicates measured levels at or above the alarm/trip setpoint. -
2. Circuit failure. (Loss of Power)
3. Control Room Instrument indicates a downscale failure.
0 (3) The initial CHANNEL CALIBRATION was performed using appropriate liquid or gaseous C calibration sources obtained from reputable suppliers. The activity of the calibration sources were Z
< reconfirmed using a multi-channel analyzer which was calibrated using one or more NBS (now NIST) standards.
z
< At all times Dn
During Containment Purges OR Containment Pressure - Vacuum Relief
< Surveillance requirement -
Modes 1-6, R41A/D Monitors providing Alarm and Automatic Termination of Release Modes 1-5, R1 2A Monitors providing Alarm and Automatic Termination of Release 0 Mode 6, R1 2A Monitors providing ýAlarm only (Automatic Termination of Release
>n is not required).I
,wi During Mode Undefined Civ ventilation process stream(Defueled) monitored operation, containment purge is reclassified as a building by the PLANT VENT HEADER SYSTEM.
0
- During movement of irradiated fuel within containment with the Containment Equipment Hatch OPEN, only R4IA/D can be credited for MINIMUM CHANNEL OPERABLE.
> During movement of irradiated fuel within containment with the Containment Equipment Hatch 0: CLOSEDR41A/D or R12A may be credited for MINTMUM CHANNEL OPERABLE.
~I 0
0.
n P 0 kdPage 30 of 157
I O Salem ODCM Rev. 23 RADIOACTWE EFFLUENTS 3/4.11 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION I CONTROLS 13.11.I .1 In accordance with the Salem Units 1 and 2 Technical Specifications 6.8.4.g. 2 and 3, the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (See Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microcuries/ml.
APPLICABILITY: At all times.
ACTION:
With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS
°7 C" program 4.11.1.1.1in Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analyses Table 4.11 -1.
4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the ODCM to assure that Z the concentrations at the point of release are maintained within the limits of CONTROL 3.11.1.1.
Cr)
I-° 0
i, Page 31 of 157
I Salem ODCM Rev. 23 I TABLE 4.11-1: RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM I
Lower Limit Liquid Release Sampling Minimum Analysis Type of Activity of Detection (LLD)a I Type Frequency Frequency Analysis (ýICi/ml)
A. Batch Waste P P I Release Each Batch Each Batch Principal Gamma Tanksb Emittersc 5x 10-I 1-131 lx10"6 P M Dissolve and I One Batch-M Entrained Gases 1x10 P M (Gamma Emitters)
H-3 I
Each Batch Composited Gross Alpha lxi 04.
lx10.7 I
V)
S P Q Sr-89, Sr-90 5x10 8 I 2/
0Z Each Batch Composite C-)
Fe-55 ixl06 I
B. Continuous Principal Gamma z"
Releases'
1. Steam W
Grab Sample W Emittersc 5xl O-I Generator U) Blowdown I 1-131 1x10 6 w M M Dissolved and lx10"s I Grab Sample Entrained Gases 0
U-w U- W Grab Sample M
Composited H-3 1xl0"5 I MID 7 7-Gross Alpha 1x10 I
W Q Sr-89, Sr-90 CC" 7) 0 Grab Sample Composited Fe-55 5_l0" 1xl0. 6 I
U)
Lj I
Li U) Page 32 of 157
Salem ODCM Rev. 23 TABLE 4.11-1 (Continued)
TABLE NOTATION
a. The LLD is defined, for purposes of these CONTROLS 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 only 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 Sb E
V
2.22E6 e Y 9 exp(-,At)
Where:
LLD is the "a priori" lower limit of detection as defined above (as microcuries per unit mass or volume),
4.66 is the statistical factor from NUREG 1301 VSb is the standard deviation of the background counting rate or of the counting rate of a blank 0* sample as appropriate (as counts per minute),
r- E is the counting efficiency (as counts per disintegration),
<2 V is the sample size (in units of mass or volume),
2.22E6 is the number of disintegrations per minute per microcurie, V) Y is the fractional radiochemical yield (when applicable),
VX is the radioactive decay constant for the particular radionuclide, and w
I At for environmental samples is the elapsed time between sample collection (or end of the sample collection period) and time of counting.
Typical values of E, V, 4, and At should be used in the calculation.
o It should be recognized that the LLD is defined as an a prori(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.
I°w UK Page 33 of 157
I Salem ODCM Rev. 23 I
S TABLE 4.11-1 (Continued) I TABLE NOTATION I
b. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative I sampling.
c. The principal gamma emitters for which the LLD CONTROL applies exclusively are the I following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144*. This list does not mean that only these nuclides are to be detected and reported. Other peaks that aremeasurable and identifiable, together with the above nuclides, I shall also be identified and reported.
d. A composite sample is one in which the quantity of liquid sampled is proportional to the I quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.
I
e. A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a 0
w volume of a system that has an input flow during the continuous release.
I 0
I V)
V-I 0
z" 0
I The LLD for Ce-144 shall be 2x10 6 jLCi/ml.
0 I
I w
ry 0
U-I
-i Ld_
I W
0 n
I D
Page 34 of 157
Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1.2 DOSE CONTROLS 3.11.1.2 In accordance with Salem Units 1 and 2 Technical Specifications 6.8.4.g.4 and 5, the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-3) 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, and
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.
APPLICABILITY:At all times.
ACTION:
a. With the calculated dose from the release of radioacti*,e materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 (A days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the 0 cause(s) for exceeding the limit(s) and defines the corrective actions that have been W 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.
I b. The provisions of CONTROL 3.0.3 and 3.0.4 are not applicable.
IV SURVEILLANCE REQUIREMENTS
_0 4.11.1.2 Cumulative dose contributions from liquid effluents shall be determined in accordance with
> the ODCM at least once per 31 days.
U-0
_:D_ 0/
V) ~Page 35ofl157
I Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS I 3/4.11.1.3 LIQUID RADWASTE TREATMENT I CONTROLS 3.11.1.3 In accordance with the Salem Units 1 and 2 Technical Specifications 6.8.4.g.6,.the liquid I radwaste treatment system shall be used to reduce the radioactive materials liquid wastes prior to their discharge when the projected cumulative doses due to the liquid effluent from each reactor to UNRESTRICTED AREAS (see Figure 5.1-3) exceed 0.375 mrem to the total body or 1.25 mrem to any organ during any calendar quarter.
APPLICABILITY: At all times. I ACTION:
a. With the radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6,9.2, a Special Report that includes the following information:
1. Explanation of why liquid radwaste was being discharged without treatment, -
identification of any inoperable equipment or subsystems, and the reason for the Vinoperability.
0 C< 2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and
3. Summary description of action(s) taken to prevent a recurrence.
b. The provisions of CONTROL 3.0.3 and 3.0.4 are not applicable.
D I---
SURVEILLANCE REQUIREMENTS Cu> 4.11.1.3 Doses due to liquid releases shall be projected at least once per 31 days in accordance with the ODCM.
z 0,
7- MID 0
,I Page 36 of 157
Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE CONTROLS 3.11.2.1 In accordance with the Salem Units 1 and 2 Technical Specifications 6.8.4.g.3 and 7, the dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) 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 mrem/yr to the skin, and
b. For iodine- 131, for tritium, and for all radionuclides in particulate form-with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.
APPLICABILITY: At all times.

ACTION:
With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the
z above limit(s).
C.)
SURVEILLANCE REQUIREMENTS I "4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined continuously to be within the above limits in accordance with the ODCM.
>W 4.11.2.1.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with a: half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
rY 0
L-Page 37 of 157
I Salem ODCM Rev. 23 I S
TABLE 4.11-2: RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM I Minimum Lower Limit of Detection I
Gaseous Release Sampling Analysis Type of Activity (LLD)
Type Frequency Frequency Analysis (pCi/ml)
I A. Waste Gas P P Storage Tank Each Tank Grab Sample Each Tank Principal Gamma Emittersb lxlO-I B. Containment P P Principal Gamma 1x10.4 PURGE Each PURGE Grab Sample Each PURGE Emittersb I
C. Plant Vent Mc,'e H-3 Principal Gamma 1xl0"6 lxlO0 I
Grab Sample MN Emittersb H-3 Ix10-6 I D. All Release Continuousf Wg 1-131 lx 10-'2
(/A Uj 0
Types as Listed in A, Charcoal Sample I
B, and C z
(..
Above I H- Continuousf Wg Principal Gamma 1011 Emittersb C-)
Z Particulate Sample (1-131, Others)
I Continuousf M Gross Alpha 1x 10" L.,
U-"
Cz Composite Particulate I
Sample U---
I Continuousf Q Sr-89, Sr-90 1x10I1' I
0 Composite w Particulate 0
w z
cv" W
Continuousf Sample Noble Gas Noble Gasses 1xlO-6 I
Monitor Gross Beta or Gamma (A
(/3 D,
i Page 38 of 157 I
m
!Salem ODCM Rev. 23 TABLE 4.11-2 (Continued)
TABLE NOTATION
a. The LLD is defined in Table 4.11.1
b. The principal gamma emitters for which the LLD CONTROL applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m; Xe-135, Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
c. Sampling and analysis shall also be performed following shutdown, startup or a THERMAL POWER change that, within one hour, exceeds 15 percent of RATED THERMAL POWER unless:
1. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of three; and
2. The noble gas activity monitor shows that effluent activity has not increased by more V 'than a factor of three.
W< d. Tritium grab samples 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.
< e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from cV the spent fuel pool area whenever spent fuel is in the spent fuel pool.
m0 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 Z 3.11.2.1, 3.11.2.2 and 3.11.2.3.
g. Samples 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). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change that, within one hour, exceeds 15 percent of RATED THERMAL POWER and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 
> are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement o does not apply if (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 monitor shows
- that effluent activity has not increased by more than a factor of three.
IQ...
00, Page 39 of 157
I Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS I 3/4.11.2.2 DOSE - NOBLE GASES CONTROLS 3.11.2.2 In accordance with the Salem Units 1 and 2 Technical Specification 6.8.4.g.5 and 8, the air I dose due to noble gases released in gaseous effluents, from each reactor unit, from the site areas and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 1O-mrad for beta radiation and,
b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
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 Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for 0 exceeding the limit(s) and defines the corrective actions that have been taken to reduce
< the release and the proposed corrective actions to be taken to assure that subsequent z
b.
releases will be in compliance with the above limits.
The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
I (A
SURVEILLANCE REQUIREMENTS O 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year Cr)
> shall be determined in accordance with the ODCM at least once per 31 days.
0_I 0
0- Pae 0 f 5 Page I
Salem ODCM Rev. 23 I 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.2.3 DOSE - IODINE- 131. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM CONTROLS I 3.11.2.3 In accordance with the Salem Units 1 and 2 Technical Specification 6.8.4.g.5 and 9, the dose to a MEMBER OF THE PUBLIC from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
i a. During any calendar quarter: Less than or equal to 7.5 mre to any organ and,
b. During any calendar year: Less than or equal to 15 mrem to any organ.
APPLICABILITY: At all times.
ACTION:
a. With the calculated air dose from the release of iodine-131, tritium, and radionuclides in Lparticulate form with half-lives greater than 8 days, in gaseous effluents exceeding any
)of the above limits, prepare and submit to the Commission within 30 days, pursuant to z0 Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding
< the limit and defines the corrective actions that have been taken to reduce the release and o the proposed corrective actions to be taken to. assure that subsequent releases will be in z *compliance with the above limits.
b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
0 SURVEILLANCE REQUIREMEINTS Y"a 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year C for iodine- 131, tritium, and radionuclides in particulate form with half-lives greater than.8 days shall Z be determined in accordance with the ODCM at least once per 31 days..
U-132 n'-
0 V)
Page 41 of 157
Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS I 3/4.11.2.4 GASEOUS RADWASTE TREATMENT CONTROLS 3.11.2.4 In accordance with the Salem Units 1 and 2 Technical Specifications 6.8.4.g.6, the I GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases, -from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3), exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter. The VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous -waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed.1.875 mrem to any organ in any calendar quarter.
APPLICABILITY: At all times. I ACTION:
a. With gaseous wastebeing discharged without treatment and in excess of the above W.... limits, prepare and submit to the Commission within 30 days, pursuant to Technical 0 ,Specification 6.9.2, a Special Report that includes the following information:
1. Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.I
2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and
3. Summary description of action(s) taken to prevent a recurrence.
> b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable. I 0_z or" SURVEILLANCE REQUIREMENTS 4.11.2.4 Doses due to gaseous releases from the site shall be projected at least once per 31 days in o accordance with the ODCM.
0 LU0I PeVof V)Page 42 of 157
Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE CONTROLS 3.11.4. In accordance with Salem Units 1 and 2 Technical Specification s 6.8.4.g.11, the annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem).
APPLICABILITY: At all times ACTION:
a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of CONTROL 3.11.1.2a, 3.11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.11.2.3a, or 3.11.2.3b, calculations should be made including direct rdetermine Vr radiation contributions from the reactor units and from outside storage tanks to whether the limits of this CONTROL have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical U3. Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and
-- includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the z 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 Cr)
Dincludes 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) exceeds the above limits, and if the z .release condition resulting in violation of 40 CFR Part 190 or 10 CFR 72.104 has not 0 already been corrected, the Special Report shall include a request for a variance in
> accordance with the provisions of 40 CFR Part 190 and 10 CFR 72.104. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
U--
. b. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
ay SURVEILLANCE REQUIREMENTS 0
L"J 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in v, accordance with CONTROLS 4.11.1.2, 4.11.2.2, 4.11.2.3, and in accordance with the ODCM.
n 4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and from LU radwaste storage shall be determined in accordance with the ODCM.
Page 43 of 157
Salem ODCM Rev. 23 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1. MONITORING PROGRAM I CONTROLS 3.12.1. In accordance with Salem Units 1 and 2 Technical Specifications 6.8.4.h. 1, the radiologicalI environmental monitoring program shall be conducted as specified in Table 3.12-1.
APPLICABILITY: At all times.
ACTION: I
a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification 6.9.1.7, a description of the reasons for not conducting the program as :required and the plans for preventing a recurrence. I
b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and-defines the corrective actions to be taken to
- reduce radioactive effluents so that the potential annual dose* to aMEMBER OF THE n PUBLIC is less than the calendar year limits of CONTROLS 3.11.1.2, 3.11.2.2, and z 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the
1.0 o reporting level(1) reporting level (2)
When radionuclides other than those in Table 1.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose* to a
-Z- MEMBER OF THE PUBLIC from all radionuclides is 'equal to or greater than the
_ calendar year limits of CONTROLS 3.11.1.2, 3.11.2.2, and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; ry however, in such an event, the condition shall be reported and described in the Annual o Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7.
L_
II
_E *The methodology used to estimate the potential annual dose to a MEMBER OF THE PUBLIC o shall be indicated in this report.
VI) 122I DC Page 44 of 157
S..Salem ODCM Rev. 23 i -* 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM CONTROLS ACTION: (Cont'd)
c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program.
Pursuant to Technical Specification 6.9.1.8, identify the cause of the unavailability of samples and the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Report. Include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).
Is d. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
V) SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-I from the specific locations given in the table and figure(s) in the ODCM, and shall be
< analyzed pursuant to the requirements of Table 3.12-1, and the detection capabilities required by I Table 4.12-1.
V)
I-I 0
tLd U-0 U-_
Lý 0P V)Page 45 of 157
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES S a Salem ODCM Rev. 23 TABLE 3.12.1-1: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM
EXPOSURE PATHWAY NUMBER OF REPRESENTAIVE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLES AND SAMPLE COLLECTION OF ANALYSIS LOCATIONS (' FREQUENCY
1. DIRECT RADIATION (2) Forty-nine routine monitoring Quarterly Gamma dose quarterly stations with two or more dosimeters placed as follows:
An inner ring of stations one in each land based meteorological sector (not bounded by water) in the general area of the SITE BOUNDARY; An outer ring of stations, one in each land-based meteorological sector in the 5 to 1 1-km range from the site (not bounded by or over water); and The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.
The number, media, frequency, and location of samples may vary from site to site. This table presents an acceptable minimum program for a site at which each entry is applicable. Local site characteristics must be examined to determine if pathways not covered by this table may significantly contribute to an individual's dose and should be included in the sample program.
Page 46 of 157
- - - - - - - --- - --- /-/ - --
m -
=- - - - - - - - - - - - - - -
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES Salem ODCM Rev. 23 TABLE 3.12.1-1 (Cont'd)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY , NUMBER OF REPRESENTAIVE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE 'SAMPLES'AND SAMPLE COLLECTION' OF ANALYSIS LOCATIONS () FREQUENCY
2. AIRBORNE Radioiodine and Particulates Samples from 6 locations: Continuous sampler Radioiodine Canister I- 131 operation with sample analysis weekly.
collection weekly or more frequently if required by dust loading.
4 Samples - One sample from close Particulate Sampler Gross beta to the SITE BOUNDARY location radioactivity analysis following and three samples in different land filter change (3) .
based sectors of a high calculated annual average ground level D/Q One sample from the vicinity of a Gamma isotopic analysis(4) of community having a high calculated composites (by location) annual average ground- level D/Q; quarterly.
and One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction.
Page 47 of 157
USER RESPONSEBLE FOR VERtFYING REVtSION, STATUS AND CHANGES U 0 0 Salem ODCM Rev. 23 TABLE 3.12.1-1 (Cont'd)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF SAMPLING AND TYPE AND FREQUENCYOFANALYSIS PATHWAY REPRESENTAIVE SAMPLES COLLECTION AND/OR SAMPLE AND SAMPLE LOCATIONS () FREQUENCY
3. WATERBORNE
a. Surface(5 ) One sample upstream Grab sample monthly Gamma isotopic analysis(4) monthly. Composite One sample downstream for tritium analysis quarterly.
One sample outfall One sample cross-stream
b. Ground Samples from one or two sources Monthly Gamma isotopic analysis(4) monthly and tritium only if likely to be affected(7). analysis quarterly.
c. Drinking (10) One sample of the nearest water Composite sample I1131 analysis on each composite when the. dose supply affected by its discharge over two-week calculated for the consumption of the water is period'6 )when 1-131 greater than 1 mrem per year8 ). Composite for analysis is performed; gross beta and gamma isotopic analysis(4) monthly composite monthly Composite for tritium analysis quarterly otherwise.
d. Sediment One sample from downstream area Semiannually Gamma isotopic analysis(4) semiannually One sample from.cross-stream area One sample from outfall area One sample from upstream area One sample from a control location One sample from shoreline area One sample from Cooling Tower Blowdown Page48 of 157
M - -
1=11- M M M -M - - - - - /
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES Salem ODCM Rev. 23 TABLE 3.12.1-1 (Cont'dl RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTAIVE SAMPLING.AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLES AND SAMPLE COLLECTION OF-ANALYSIS LOCATIONS ") FREQUENCY
4. INGESTION
a. Milk Samples from milking animals in Semimonthly when animals Gamma isotopic (4) and 1-131 three locations within 5 km distance are on pasture, monthly at analysis semi-monthly when having the highest dose potential. If other time animals are on pasture; monthly there are none, then, one sample at other times from milking animals in each of three areas.between 5 to 8 km distant where doses are calculated to be greater than I mrem per yr(8)
One sample from milking animals at a control location 15 to 30 km distant.
b. Fish and Invertebrates One sample of each commercially Sample in season, or Gamma isotopic analysis(4) on and recreationally important species semiannually if they are not edible portions.
in vicinity of plant discharge area seasonal One sample of same species in area not influenced by plant discharge.
Page 49 of 157
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES S0 S 0 Salem ODCM Rev. 23 TABLE 3.12.1-1 (Cont'd)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBEROF SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE REPRESENTAIVE COLLECTION OF-ANALYSIS SAMPLES AND SAMPLE FREQUENCY LOCATIONS (1)
c. Food Products One sample of each principal At time of harvest (9) Gamma isotopic analysis (4) on class of food products from any edible portion.
area that is irrigated by water in which liquid plant wastes have been discharged Page 50 of 157
- - 1-- - - - 1-1-----1 1-i 1--
Salem ODCM Rev. 23 TABLE 3:* 12.1-1 (Continued)
TABLE NOTATION (1) Specific parameters of distance and direction sector from the midpoint of a line between the center of the Salem units 1 & 2 containment domes, and additional description where pertinent, shall be provided for each and every sample location in Table 3.12-1 in a table and figure(s) in the ODCM. Refer to NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Reg.
Guide 4.8 as amended by Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment, and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period..
All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological mI Environmental Monitoring Program given in the ODCM. Pursuant to CONTROL c6.9.1.8, submit in the next Radioactive Effluent Release Report documentation for a change in the ODCM including revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of o) samples for the pathway and justifying the selection. of the new location(s) for obtaining samples.
V(2) One or more instruments, such as pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters.
<) For the purposes of this table, a Dosimeter of Legal Record (DLR) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more 0 dosimeters. Film badges shall not be used for measuring direct radiation. The frequency 0of analysis or readout will depend upon the characteristics of the specific dosimetry system used and should be selected to obtain optimum dose information with minimal fading. No direct radiation monitoring stations are located in the inner ring sectors 8, 9, 12, 13 and 14 and the outer ring sector 8 as originally determined during plant licensing
_. and as permitted by Reg. Guide 4.8 as amended by The Branch Technical Position Revision 1, November 1979. Sector 7 does not have a direct radiation monitoring station in the outer ring due to inaccessibility.
0 (3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Ln or more after sampling to allow for radon and thoron daughter decay. If gross beta 01 activity in air particulate is greater than ten times the yearly mean of control samples, 0 - gamma isotopic analysis shall be performed on the individual samples.
ry P
Page 51 of 157
I Salem ODCM Rev. 23 (4) Gamma isotopic analysis means the identification and quantification of gamma-emitting I radionuclides that may be attributable to. the effluents from the facility.
TABLE 3.12.1-1 (Continued)
TABLE NOTATION (5) The "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall betaken in an area beyond but near the mixing zone. "Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence. Saltwater shall be sampled only when the receiving water is utilized for recreational activities.
(6) A composite sample is one which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short relative to the compositing period in order to assure obtaining a representative sample.
(7) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.
r< (8) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM. Additionally, 2 sample locations are monitored as management audit. Broad leaf vegetation may be obtained in lieu of milk collections.
(9) If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention, shall be paid to including samples of tuberous and root food products. The Delaware River at the
__ location of Salem and Hope Creek Nuclear Power Plants is a brackish water source. No ef)
>- irrigation of food products is performed using water-in the vicinity from which liquid plant wastes have been'discharged. However, 12 management audit food samples are 0 collected from various locations. "
=- II (10) No groundwater samples are required-as liquid effluents discharged from Salem and a:, Hope Creek Generating Stations do not directly affect this pathway. However for
>: management audit, one raw and one treated ground water sample from the nearest o unaffected water supply is required.
~I o-I U,,- ZI DPage 52 of 157
- M M Mm M- M-- M M- M M -M - M USER RESPONSIBLE FOR'VERIFYING REVISION, STATUS AND CHANGES Salem ODCM Rev. 23 TABLE 3.12-2: REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS Water Airborne Particulate Fish Milk Food Products Analysis (pCi/l or Gases (pCi/m3) (pCi/Kg, wet) (pCi/l) (pCi/Kg, wet)
H-3 3 x 104 Mn-54 Ix 103 3 x 10 4 Fe-59 4x10 2 x 104 Co-58 I x 103 . 3 x I0o4 Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104 Zr-Nb-95 4 x 102 1-131 20 0.9 3 1 x 102 Cs-134 30 10 1 xl03 60 1 x 103 Cs-137 50 20 2 x 103 70 2 x 103 Ba-La-140 2x 102 _ 3x 102 Page.53 of 157
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES W. S S Salem ODCM Rev. 23 TABLE 4.12-1: DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS')'(2)
LOWER LIMITS OF DETECTION (LLD)(3)
Water Airborne Particulate Fish Milk Food Products Sediment Analysis (pCi/i) or Gases (pCi/m3) (pCi/Kg, wet) (pCi/l) (pCi/Kg, wet) (pCi/Kg, dry) gross 4 1 x 10-2 beta H-3 3000 Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 Zn-65 30 260 Zr-Nb-95 15 1-131 10 7x 10 2 1 60 Cs-134 15 5x 10-2 130 15 60 150 Cs-137 18 6x 10-2 150 18 80 180 Bd-La-140 15 15 Page 54 of 157
Salem ODCM Rev. 23 TABLE 4.12-1 (Continued)
TABLE NOTATION (1) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7.
I (2) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.
I (3)The LLD is defined, for purposes of these CONTROLS 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 only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
03 I*LLD 460S E* V 0 2.22E6 , Y , exp(-AAt)
< Where:
LLD is the "a priori" lower limit of detection as defined above (as picocuries per unit mass or I<* volume),
03 4.66 is the statistical factor from NUREG 1301
<T Sb is the standard deviation of the background counting rate or of the countingrate of a blank Vsample as appropriate (as counts per minute),
0 E is the counting efficiency (as counts per disintegration),
> V is the sample size (in units of mass or volume),
tra 2.22 is the number of disintegrations per minute per picocurie, IZ Y is the fractional radiochemical yield (when applicable);
2:is the radioactive decay constant for the particular radionuclide, and X
> At for environmental samples is the elapsed time between sample collection (or end of the 0 sample collection period) and time of counting.
I Typical values of E, V, Y, and At should be used in the calculation.
I* Page 55 of' 157
i Salem ODCM Rev. 23 TABLE 4.12-1 (Continued) i TABLE NOTATION i 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. Analyses shall be performed in such a manner that the stated. LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7.
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Salem ODCM Rev. 23 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS I CONTROLS I 3.12.2. In accordance with the Salem Units 1 and 2 Technical Specifications 6.8.4.h.2, a land use census shall be conducted and shall identify within a distance of 8 km (5-miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden*
I of greater than 50 m 2 (500 ft2) producing broad leaf vegetation.
APPLICABILITY: At all times.
ACTION:
a. With a land use census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in CONTROL 4.11.2.3, identify the new location(s) in the next Radioactive Effluent Release Report, pursuant to CONTROL 6.9.1.8.
b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with CONTROL 3.12.1, add the C-r) *new location(s) to the radiological environmental monitoring program within 30 days.
The sampling location(s), excluding the control station location, having the lowest
< calculated dose or dose commitment(s), via the same exposure pathway, may be deleted Cn from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to CONTROL 6.9.1.8, identify the new location(s) in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and C/i table for the ODCM reflecting the new location(s).
I__ c. The provisions of CONTROLS 3.0.3 and 3.0.4 are not applicable.
S *Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Q in lieu of
> the garden census. CONTROLS for broadleaf vegetation sampling in Table 3:12-1.4c shall be followed, 0* including analysis of control samples.
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I Salem ODCM Rev. 23 I
3/4.12 RADIOLOGICAL ENVIRONMENTAL-MONITORING S3/4.12.2 LAND USE CENSUS (Cont'd) I SURVEILLANCE REQUIREMENTS I
4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, visual survey, aerial survey, or by consulting local agriculture authorities. The results of theiland use census shall be I included in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7.
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Salem ODCM Rev. 23 RADIOLOGICAL ENVIRONMENTAL MONITORING II 3/4.12.3 1NTERLABORATORY COMPARISON PROGRAM CONTROLS 3.12.3 In accordance with Salem Units I and 2 Technical Specifications 6.8.4.h.3, analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program.
APPLICABILITY: At all times.
ACTION:
a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7.
b. The provisions of CONTROLS 3.0.3 and 3.0.4. are not applicable.
SURVEILLANCE REQUIREMENTS I. 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.7.
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Salem ODCM Rev. 23 I
BASES FOR' SECTIONS 3.0 AND 4.0 CONTROLS AND LU S SURVEILLANCE REQUIREMENTS C-)
NOTE z
D The BASES contained in the succeeding pages summarize C) the reasons for the CONTROLS of Sections 3.0 and 4.0, 0
Cr) but are not considered a part of these CONTROLS.
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3/4.3 INSTRUMENTATION I* BASES 3/4.3.3.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The alarm/trip setpoints for these instruments shall be calculated and adjustedin accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A.to 10 CFR Part 50.
CROSS REFERENCE -TABLES 3.3-12 and 4.3-12 Unit 1:
T/S Table Item No. Instrument Description Acceptable RMS Channels 1a Liquid Radwaste Effluent Line Gross 1R18 Activity I I II I. (li lb Steam Generator Blowdown Line 1RI 9A, B, C, and D(')
Gross Activity ILW 2a Containment Fan Coolers Service 1R13A and B(')
C) -------- Water Line Discharge Gross Activity Unit 2:
T/S Table Item No. Instrument Description . Acceptable RMS Channels I-; z"0 la Liquid Radwaste Effluent Line Gross Activity 2R18 lb Steam Generator Blowdown Line 2R19A,B,C, Gross Activity and D(')
LU 2a Containment Fan Coolers - Service 2R13A and B(l) 14>-
Water Line Discharge Gross Activity ta_ 2b Chemical Waste Basin Line Gross R37 0-V-Activity z
LU (1) The channels listed are required to be operable to meet a single operable channel for the ODCM's "Minimum Channels Operable" requirement.
(,'y LIJa cV" Page 61 of 157
I Salem ODCM Rev. 23 I
3/4.3 INSTRUMENTATION BASES I 3/4.3.3.9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION I The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance I
with the procedures in the ODCM to. ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the I requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
CROSS REFERENCE - TABLES 3.3-13 and 4.3-13 I Unit 1:
I T/S Table Item No. Instrument Description Acceptable RMS Channels la Waste Gas Holdup System: 1R41A andD DX2) I Noble Gas Activity ,
2a Containment Purge and.
Pressure - Vacuum Relief IR12A I 0W Noble Gas Activity or 3a Plant Vent Header System 1R41A and D )(X2)
IR41A and D (1)2)
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I 00 Iodine Sampler (3) or Li r- ____'.__ _ 1XT8911 (iR45).
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IRME 4, 5 (1R41) or I 1XT8911 n
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(1) The channels listed are required to be operable to meet a single operable channel for the ODCM's I
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1R41D is the setpoint channel. 1R41A is the measurement channel.
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I Salem ODCM Rev. 23 3/4.3 INSTRUMENTATION I* BASES Unit 2:
T/S Table Instrument Description Acceptable RMS Item No. Channels la Waste Gas Holdup System Noble Gas Activity 2R41A and D(1X2) 2a Containment Purge and Pressure - Vacuum Relief 2R12A or 2R41A Noble Gas Activity and D(1)(
3a Plant Vent Header System Noble Gas Activity 2R41A and D(1 ) 2 )
3b Plant Vent Header System Iodine Sampler (3) RME 4, 5 (2R41) or 2XT89r11 (2R45) 3c Plant Vent Header System Particulate Sampler (3) 2RME 4, 5 (2R41) or 2XT8911 (2R45)
IeVi)
(1) The channels listed are required to be operable to meet a single operable channel for the ODCM's "Minimum Channels Operable" requirement.
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(2) 2R41D is the setpoint channel. 2R41A is the measurement channel.
(3) Laboratory analysis of the sampler filters ensures that the limits of ODCM CONTROL 3.11.2.1 are not exceeded. Alarm/trip setpoints do not app ly to these passive components.
IV)
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I Salem ODCM Rev. 23 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS m 3/4.11. 1.1 CONCENTRATION The CONTROL is provided to ensure that the concentration of radioactive materials released in liquid waste effluents will be less than the concentration levels specified in 10 CFRPart 20, Appendix B Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the I
Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(a) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its I
MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection.(ICRP) Publication 2.
The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).
S3/4.11.1.2 DOSE V) This CONTROL is provided to implement the requirements of Sections II.A, III.A, and IV.A of L.
z Co Appendix I. 10 CFR Part 50.. The CONTROL implements the guides set forth in Section II.A of ACTION statements provide flexibility and at theofsame time IV.Athe required operating o*< Appendix implement I.theTheguides set forth in Section of Appendix I to assure that the releases radioactive material in liquid effluents will be kept "as low as is reasonably achievable." Also, for freshwater sites
< with drinking water supplies that can be potentially affected by plant operatiohis, there is reasonable assurance that the operation' of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculations in the V) ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Z" Appendix I be shown by calculational procedures based on models and data, such that the actual o exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual m r" release rates of radioactive materials in liquid effluents are consistent with the methodology provided in CD Regulatory Guide 1.109, "Calculation of Annual Doses to Man from RoutineReleases of Reactor z_ 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 w, Accidental and Routine Reactor Releases for the Purposes of Implementing Appendix I," April 1977.
o The CONTROL applies to the release of liquid effluents from each reactor at'the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among o__ the units sharing that system.
03 Lo ac Page 64 of 157
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Salem ODCM Rev. 23 I O RADIOACTIVE EFFLUENTS BASES I3/4.11.1.3 LIQUID RADWASTE TREATMENT The requirement:that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is ,
reasonably achievable". This 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 II.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system Were specified as a suitable fraction of the dose design objectives set forth the'Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.
3/4.11.2 GASEOUS EFFLUENTS'-
3/4.11.2.1 DOSE RATE This CONTROL is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part
20. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material Sdischarged in gaseous effluents will not result in the exposure of a MEMBER. OF THE PUBLIC either W) within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits I specified in Appendix B, Table tIof 10 CFR Part 20 [10 CFR Part 20.106(b)]. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the individual will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor I above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC with the appropriate occupancy factors shall be given in the ODCM. The specified release rate I
0 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 whole body and 3000 mrem/yr to the skin. These release rate limits. also restrict, at all 0 times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to z" less than or equal to 1500 mrem/year.
> This CONTROL applies to the release of gaseous effluents from all reactors at the site.
Iz 3/4.11.2.2 DOSE - NOBLE GASES
t. This CONTROL is provided to implement the requirements of Section II.B, III.A and IV.A of I Appendix I, 10 CFR Part 50. The CONTROL implements the guides set forth in Section II.B of a:ý Appendix I. The ACTION statements provide the required operating flexibility and at the same time 0
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 z Wguides of Appendix I be shown by calculational procedures based on models and data such that the n actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be*
Ua ry Page 65 of 157
!I Salem ODCM Rev. 23 m
I RADIOACTIVE EFFLUENTS BASES substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents 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 50, Appendix I,"
Revision I, 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 basedupon the historical average atmospheric conditions.. .
3/4.11.2.3 DOSE - IODINE-131, TRITIUM. AND RADIONUCLIDES IN PARTICULATE FORM.-
This CONTROL is provided to implement the requirements of Section II.C, III.A and LV.A of Appendix I, 10 CFR Part 50. The CONTROL are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the, guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in I gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods ,
specified in Surveillance Requirements implement the requirements in Section I1.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the. actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely. to be substantially Underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of ReactorZ.
z Effluents for the,Purpose of Evaluating Compliance with 10 CFR;50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.1.11, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1,. July 1977.
These equations also provide for determining, the actual dosebased upon the historical average 0 atmospheric conditions. The release rate controls for iodine- 131, tritium, and radionuclides in particulate form with half-life greater than 8 days are dependent-on the existing radionuclide pathways to'man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the m 0 development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) z 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.
3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM .
zm The requirement that. the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This CONTROL implements the requirements of 10s CFR Part 50.36a, 0
General Design Criterion 60 of Appendix A to 10CFR Part'50 and the design' objectives- given in Section I1.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate z
portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
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Salem ODCM Rev. 23 RADIOACTIVE EFFLUENTS
'BASES 3/4.11.4 TOTAL DOSE This CONTROL is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525 as well as the dose limitations specific to Independent Spent Fuel Storage Installation (ISFSI) operations in accordance with 10 CFR 72.104.
Over the long term, as more storage casks are placed on the ISFSI pads, it is expected that ISFSI operations will become the prominent contributor to the dose limits in this section. ISFSI dose contribution is in the form of direct radiation as no liquid or gas releases are expected to occur. The PSEG 10 CFR 72.212 Report prepared in accordance with 10 CFR 72 requirements assumes a certain array of casks exists on the pads. The dose contribution from this array of casks in combination with historical uranium fuel cycle operations prior to ISFSI operations was analyzed to be within the 40 CFR 190 and 10 CFR 72.104 limits. The CONTROL requires the preparation and submittal of a Special Report Whenever the calculated doses from plant including the ISFSI radioactive effluents 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 including outside storage tanks, etc. are kept small. The Special Report will describe a course of action that should result in the limitation of the I D annual dose to a MEMBER OF THE PUBLIC to withifi the 40 CFR Part 190 or 10 CFR 72.104 limits. For purposes of the Special Report, it may be. assumed that the dose commitment to the
, MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the Z exception that dose contributions from other nuclear fuel cycle facilities at the same. site or within a
- radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to
0) exceed the requirements of 40 CFR Part 190 or 10 CFR 724104, the Special Report with a request Z for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 or 10 CFR
<) 72.104 have not already been corrected), in accordance with the provisions of 40 CFR Part 190 or 10 CFR 72.104 and 10 CFR Part 20.405c, is considered to be a timely request Sand fulfills the requirements of 40 CFR Part 190 or 10 CFR 72.104 until NRC staff action is completed. The V) variance only relates to the limits of 40 CFR Part 190 or 10 CFR 72.104, and does not apply in any z" way to the other requirements for dose limitation of 10 CFR'Part 20, as addressed in CONTROLS o 3.11.1. and 3.11.2. An individual is not considered a MEMBER OF THE PUBLIC during any
-* period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
r-I/3 Page 67 of 157
Salem ODCM Rev. 23 RADIOACTIVE EFFLUENTS BASES 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING I 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this CONTROL provides I measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure-pathways. The initial specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience. I The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
0 3/4.12.2 LAND USE CENSUS This CONTROL is provided'to ensure that changesin the use of areas at and beyond the SITE 0* BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best information from the door-to-door
< survey, aerial survey or consulting with local agricultural authorities shall be used. This census satisfies V) the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50m 2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 I
2" kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To o determine this minimum garden size, the following assumptions were made: 1) 20% of the garden was
- used for growing broad leaf vegetation (i.e.,,similar to lettuce and cabbage), and 2) yield of 2 kg/mi. I CD 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM This requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in
> environmental sample matrices are performed as part of the quality assurance program for o environmental monitoring in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.
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I I Salem ODCM Rev. 23 ji ID I
I SECTION 5.0 I DESIGN FEATURES Ij C-)
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I Salem ODCM Rev. 23 I
S 5.0 DESIGN FEATURES I
5.1 SITE 5.1.3 UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS I
UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-3. (Provided FOR INFORMATION ONLY. Technical Specifications Section. 5.0 is controlling.)
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Salem ODCM Rev. 23 FIGURE 5.1-3: AREA PLOT PLAN OF SITE
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Salem ODCM Rev. 23 6.0 ADMINISTRATIVE CONTROLS 6.9.1.7 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.7 In accordance with Salem Units I and 2 Technical Specifications 6.9.1.7, The Annual Radiological Environmental Operating Report* covering the operation of the unit during the previous calendar year shall be-submitted prior to May 1 of each year.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies with operational controls (as appropriate), and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by CONTROL 3.12.2.. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all measurements taken during the period pursuant to the Table and Figures in the environmental radiation section of the ODCM; as well as summarized and tabulated results of locations specified in these analyses and measurements in the format of the table in Reg. Guide 4.8 as amended by Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as, possible in a supplementary report.
The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps, one covering sampling locations near the SITE
'" BOUNDARY and a second covering the more distant locations, all keyed to a table giving distances c~z and directions from the midpoint of a line between the centers of Salem units 1 & 2 containment domes; the results of licensee participation in the Interlaboratory Comparisorn Program, required by CONTROL 3.12.1; and discussion of all analyses in which the LLD required.-by Table 4.12-1 was not z achievable.
V)
D 6.9.1.8 RADIOACTIVE EFFLUENT RELEASE REPORT U) 6.9.1.8 In accordance with Salem Units I and'2 Technical Specifications. 6.9.1.8, The Annual z" Radiological Effluent Release Report* covering the operation of the unit'during the previous calendar V_ year shall be submitted prior to May 1 of each year. and in accordance with the requirements of
> 10CFR50.36a.
U The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21.
"Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive
> Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear:Power Plants," Revision a1 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.
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A single submittal may be made for a multiple unit station, The submittal should combine those z W sections that are common to all units at the station; however, for units with separate radwaste a- j systems, the submittal shall specify the releases of radioactive material from each unit.
V)I dc* Page 72 of 157
Salem ODCM Rev. 23 ii.The 6.9.1.8 RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and
-atmospheric stability. The report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. The report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The historical annual average meteorology or the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL.
The Radioactive Effluent Release Report shall identify those radiological environmental sample parameters and locations where it is not possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In addition, the cause of the unavailability of samples for the pathway and the new location(s) for obtaining replacement samples should be identified. The report should also include a revised figure(s) and table(s) for the ODCM reflecting the new location(s).
The Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel S cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection I V)
LU z
Standards for Nuclear Power Operation and 10 CFR 72.104, Criteria for Radioactive Materials in Effluents and Direct Radiation from an ISFSI or MRS. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.
The Radioactive Effluent Release Reports shall include the following information for each. class of V) solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:
~ a. Container volume, 0b. Total curie quantity (specify whether determined by measurement or estimate),
0_ *c. Principal radionuclides (specify whether determined by measurement or"estimate),
d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, CD evaporator bottoms),
e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and Id
f. Solidification agent or absorbent (e.g., cement, urea formaldehyde).
0 The Radioactive Effluent Release Report shall include a list of descriptions of unplanned releases from m S the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
ry Page 73 of 157
I Salem ODCM Rev. 23 6.9.1.8 RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) I The Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP), the OFFSITE DOSE CALCULATION MANUAL (ODCM), or radioactive waste systems. Also list new locations identified by the land use census pursuant to CONTROL 3.12.2. for dose calculations or environmental monitoring.
6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS 6.15.1 Licensee initiated major changes to the radioactive waste system (liquid, gaseous and solid):
1. Shall be reported to the Commission in the UFSAR for the periodin which the evaluation was reviewed by the Plant Operations Review Committee (PORC). The discussion of each change shall contain:
a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10CFR50.59;I
b. Sufficient detailed information to totally support the reason, for the change without benefit of additional or supplemental information;
c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
.D d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; D e. An evaluation of the change, which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ I,-.--
V) from those previously estimated in the license application and amendments
_" thereto;
> f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
g. An estimate of the exposure to plant operating personnel as a result of the
> change; and'
h. Documentation of the fact that the change was reviewed and found acceptable
'.' by the PORC.
zm 0
2. Shall become effective upon review and acceptance by the PORC.
I Page 74 of 157
Salem ODCM Rev. 23 I
PART II - CALCULATIONAL METHODOLOGIES 0
0 cn L-C/)
U-Page 75 of 157
I Salem ODCM Rev. 23 I
S1.0 LIQUID EFFLUENTS
I 1.1 Radiation Monitoring Instrumentation and Controls The liquid effluent monitoring instrumentation and controls at Salem for controlling and monitoring I normal radioactive material releases in accordance with the Salem Technical Specifications 6.8.4.g and ODCM CONTROLS are summarized as follows: i
1) Alarm (and Automatic Termination) - l-Rl8 (Unit 1) and 2-R18 (Unit'2) provide the alarm and automatic termination of liquid radioactive material releases as required by ODCM CONTROL 3.3.3.8.
l-R19 A, B, C, and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R19 A, B, C, and D provide this function for Unit 2.
2) Alarm (only) - The alarm functions for the Service Water System are provided by the radiation monitors on the Containment Fan Cooler discharges (1R 13 A and B for Unit 1 and I
2R 13 A and B for Unit 2).
Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored (with an alarm function) by R37 prior to release to the environment. I Liquid radioactive release flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 1-1 and 1-2 for Units 1 and 2, respectively. The
-)- Liquid Radioactive Waste System is presented in Figure 1-3.
<Z 1.2 Liquid Effluent Monitor Setpoint Determination 3D Per the requirements of ODCM CONTROL 3.3.3.8, alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of ODCM L CONTROL 3.11.1.1 are met (i.e., the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20, Appendix 0
B, Table II, Column 2, (Appendix F) for radionuclides and 2.OE-04 paCi/ml for dissolved or entrained
>j noble gases).
_z The following equation* must be satisfied to meet the liquid effluent restrictions:
C(F +f)(1)I f.
IWh
_.10 lmitofODM ONTOL3.1..1? iplmetin te 0I
=th efluntcocenraio 0 CFR 20 MPC (Appendix F) for the site, in /tCi/ml C . c the setpoint, in p.Ci/ml, of the radioactivity monitor measuring the radioactivity
'" concentration in the effluent line prior to dilution and subsequent release; the setpoint, Page 76 of 157
Salem ODCM Rev. 23 represents a value which, if exceeded, would result in concentrations exceeding the
limits of 10 CFR 20 (Appendix F) in the UNRESTRICTED AREA f the flow rate at the radiation monitor location, in volume per unit time, but in the same units as F, below F = the dilution water flow rate as measured prior to the release point, in volume per unit time
, [Note that if no dilution is provided, c < C. Also, note that when (F) is large compared to (f), then (F f)= F.]
.* Adapted from NUREG-0133 1.2.1 Liquid Effluent Monitors (Radwaste, Steam Generator Blowdown, Chemical Waste Basin and Service Water.
The setpoints for the liquid effluent monitors at the Salem Nuclear Generating Station are determined by the following equations:
SP* [MPCe*SEN*CW*CF*AF]+bkg(1.2)
S Owith:
Ci (gamma only)
-r ... MPCe= Ci (1.3)
(
heeMP 1 (gamma only)
< "Where:*
SP. = alarm setpoint corresponding to the maximum allowable release rate. (cpm)
V) MPCe = an effective MPC value for the mixture of gamma emitting radionuclides in the effluent z stream (gCi/ml) 0 Ci = the concentration of radionuclide i in the undiluted liquid effluents (ItCi/ml)
W> MPCi = the MPC value corresponding to radionuclide i from 10 CFR 20, Appendix B, Table II, Column 2 (Appendix F) (gtCi/ml)
SEN = the sensitivity-value to which the monitor is calibrated (cpm.per 4Ci/ml)
- CW = the circulating water flow rate (dilution water flow) at the time of release (gal/min)
RR = the liquid effluent release rate (gal/min),
> bkg = the background of the monitor (cpm) 0 CF = Correction factor to account for non-gamma emitting nuclides in setpoint calculations.
AF = an allocation factor applicable for steam generator blowdown me The radioactivity monitor setpoint equation (1.2) remains valid during outages when the circulating I
o water dilution is potentially at its lowest value. Reduction of the waste stream' flow (RR) may be necessary during these periods to meet the discharge criteria. However, in order to maximize the
.Page 77 of 157
Salem ODCM Rev. 23 available plant discharge dilution and thereby minimize the potential offsite doses, batch releases from either Unit-1 or Unit-2 may be routed to either the Unit- 1 or Unit-2 Circulating Water System discharge. Procedural restrictions prevent simultaneous batch releases from either a single unit or both units into a single Circulating Water System discharge.
I 1.2.2 Conservative Default Values Conservative alarm setpoints may be determined through the use of default parameters. Tables 1-1.1 and 1-1.2 summarize all current default values in use for Salem Unit-i and Uriit-2, respectively. They are based upon the following:
a) substitution of the effective MPC value with a default value of 6.05E-06 ýtCi/ml (Unit 1) and 4.81 E-06 ptCi/ml (Unit 2). (refer to Appendix A for justification);
b) for additional conservatism*, substitution of the 1-131 MPC value of3E-07 pCi/ml for the RI 9 Steam Generator Blowdown monitors, the. R-37 Chemical Waste Basin monitor and the R-13 Service Water monitors; I c) for conservatism, use of an allocation factor of 0.5 for the Steam Generator Blowdown monitors to limit consequences of potential simultaneous primary-to secondary leaks in two steam generators.** The allocation factor equals 1.0 for all liquid effluent setpoints; I
d) substitutions of the operational circulating water flow with the lowest flow, in gagmin;***
e) substitutions of the effluent release rate with the highest allowed rate, in gal/min; and, f) substitution of a Correction factor of 0.75 to account for non-gamma'emitting nuclides.
For batch liquid releases a fixed alarm setpoint is established for the 1, 2 R18 monitors and the release V) rate is controlled to ensure the inequality of equation 1.1 is maintained. With this approach, values 0" selected for the parameters in the setpoint calculation (e.g., Table 1-1.1 and Table 1-1.2) should be any
< reasonable values plusofbackground, set thattoprovide so as not a setpoint value reasonably above anticipated monitor response, yield spurious alarms. The release rate is controlled to ensure compliance I
with the requirements of ODCM CONTROL 3.3.3.8.
V)
Calculations, as performed by Engineering, to establish the actual fixed setpoints for use in the plant, I
F incorporate uncertainties and instrument drift. These factors will cause the actual installed instrument setpoint to be at a lower (conservative) value. However, for batch releases, when-the rate is controlled, I these uncertainties and drift should not be included in the evaluation, of acceptable release rate, since 0 this could cause a non-conservative correction, i.e., a higher allowable release rate. Therefore, for 1, 2
_ R1 8 monitors, the setpoint valueused for calculatingthe allowable release rate should be that value. l
,. prior to correction for uncertainty and drift. -
0:
zBased upon the potential for 1-131 to be present in the secondary and-service water systems,
_-, the use of the default effective MPC (MPCe) value as derived in AppendixA may be non-conservative for the 1, 2 R-19 SGBD monitors, the R-37 Chemical WasteBasin monitor and the R-13 Service Water monitors.
0
Setpoints using the Allocation Factor-of 0.5 become invalid if primary-to-secondary leaks are L
_J identified in more than two steam generators simultaneously.
0m ***The Containment Fan Coil Unit Discharge to Service Water Line is routed to the opposite o_.Unit's Circulating Water System discharge. Therefore, during periods when circulating aw Page 78 of 157
(/)
Salem ODCM Rev. 23
' water pumps are out of service, such as during refueling outages, the default setpoints of the other Unit's R13 radiation monitors are not valid.
1.3 Liquid Effluent Concentration Limits - 10 CFR 20 ODCM CONTROL 3.11.1.1 limits the concentration of radioactive material in liquid effluents (after dilution in the Circulating Water System) to less than the concentrations as specified in 10 CFR 20, Hare Appendix B, Table II, Column 2 (Appendix F) for radionuclides other than noble gases. Noble gases limited to a diluted concentration of 2.OE-04 ptCi/ml.
Release rates are controlled and radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded. However, in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of compliance with the concentration limits of ODCM CONTROL 3.11.1.1 may be performed using the following-equation:,
U ~ CW+/-RRi
MP (1Q.4)
Where:
Ci = actual concentration of radionuclide i as measured in the undiluted liquid effluent (pLCi/ml)
MPC = the MPC value corresponding to radionuclide i from 10 CFR 20, Appendix B, Table II, Column 2 (p.Ci/ml) [ODCM Appendix F]
= 2E-04 p.Ci/ml for dissolved or entrained noble gases I 03W RR = the actual liquid effluent release rate (gal/min)
) CW = the actual circulating water flow rate (dilution water flow) at the time of the release (gal/min) 0L 1.4 Liquid Effluent Dose Calculation - 10 CFR 50 I~ 1.4.1 MEMBER OF THE PUBLIC Dose - Liquid Effluents.
V)
Z" ODCM CONTROL 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC 0 from radioactive materials in liquid effluents from each unit of the Salem Nuclear Generating Station
> to:
uJ
- during any calendar quarter;
_ _< 1.5 mrem to total body per unit
< 5.0 mrem to any organ per unit
,-o - during any calendar year; IJ _<3.0 mrem to total body per unit
_j < 10.0 mrem to any organ per unit.
nP ryPage 79 of 157
I Salem ODCM Rev. 23 I
Per the surveillance requirements of ODCM CONTROL 4.11.1.2, the following calculational methods
shall be used for'determining the dose or dose commitment due to the liquid radioactive effluents from Salem: I D0 = 1.67E-02*VOL **(Ci* Aio" CW i (1.5)
I I
Where:
= dose or dose commitment to organ o (mrem). Total body dose can also be calculated I
using site-related total body dose commitment factor.
Aio = site-related ingestion dose commitment factor to the total body or any organ o for radionuclide i (mrem/hr per jtCi/ml)
I Ci = average concentration of radionuclide i, in undiluted liquid effluent representative of the VOL volume VOL (ptCi/ml).
= volume of liquid effluent released (gal) I CW = average circulating water discharge rate during release period (gal/min) 1.67E-02 = conversion factor (hr/min) I The site-related ingestion dose/dose commitment factors (Aio) are presented in Table 1-2 and have been V) derived in accordance with the requirements of NUREG-0133 by the equation:
I 03 S A o=1.14E +05* & 9T (1.6)
C)
Where: I z Ai. composite dose parameter for the total body or critical organ o of an adult for UI
=
=
radionuclide i, for the fish and invertebrate ingestion pathways (mrem/hr per piCi/ml) adult invertebrate consumption (5 kg/yr)
I BIi = bioaccumulation factor for radionuclide i in invertebrates from Table 1-3 (pCi/kg per pCifl)
I z0 UF = adult fish consumption (21 kg/yr) 0 (Y3 0
w BFj =
DFio bioaccumulation factor for radionuclide i in fish from Table 1-3 (pCi/kg per pCi/1) dose conversion factor for nuclide i for adults in pre-selected organ, o, from Table E- 1I I
of Regulatory Guide 1.109 (mrem/pCi)
(Y3 U--
1.14E+05 = conversion factor (pCi/*Ci
ml/kg per hr/yr) I The radionuclides included in the periodic dose assessment per the requirements of ODCM CONTROL 0
Z8 3/4.11.1.2 are those as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of ODCM CONTROL 3/4.11.1.1,.Table 4.11-1.
I m
r- Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of ODCM CONTROL I
Table 4.11-1.
L,J 03 I I
C-Page 80 of 157 0D
Salem ODCM Rev. 23 1.4.2 Simplified Liquid Effluent Dose Calculation.
In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following simplified dose calculation equation may be used for demonstrating compliance with the dose limits of ODCM CONTROL 3.11.1.2. (Refer to Appendix B for the derivation and justification for this simplified method.)
Total Body Dtb =1.21E + 03 *'VOL (1.7)
Dib-C CW Maximum Organ D m'ax =*E*2 52E + 04 *VOL Ci (1.8)
I Where: CW (
Ci = average concentration of radionuclide i, in undiluted. liquid effluent representative of the volume VOL (iLCi/ml)
VOL = volume of liquid effluent released (gal)
I S CW Dtb average circulating water discharge rate during release period (gal/min)
= conservatively evaluated total body dose (mrem)
Dmax = conservatively evaluated maximum organ dose (mrem)
Q 1.21 E+03 = conversion factor (hr/min) and the total bodydose conversion factor (Fe-59, total body r- -- 7.27E+04 mrem/hr per itCi/ml) 2.52E+04 = conversion factor (hr/min) and the conservative maximum organ dose conversion factor (Nb-95, GI-LLI -- 1.5 IE+06 mrem/hr per *tCi/ml)
D 1.5 Secondary Side Radioactive Liquid Effluents and Dose Calculations During Primary to I- Secondary Leakage H0 --
During periods of primary to secondary leakage (i.e., steam generator tube leaks), radioactive material will be transmitted from the primary system to the secondary system. The potential exists for the
, release of radioactive material to the off-site environment (Delaware River) via secondary system
d discharges. Potential releases are controlled/monitored by the Steam Generator Blowdown monitors I(R19)and the Chemical Waste Basin monitor (R37).
W However to ensure compliance with the regulatory limits on radioactive material releases, it may be
> desirable to account for potential releases from the secondary system during periods of primary to 0o secondary leakage. Any potentially significant releases -will be via the Chemical Waste Basin with the major source of activity being the Steam Generator Blowdown.
I o
S With identified radioactive material levels in the secondary system, appropriate samples should be collected and analyzed for the principal gamma emitting radionuclides. Based' on the identified Page 81 of 157
Salem ODCM Rev. 23 radioactive material levels and the volume of water discharged, the resulting environmental doses may be calculated based on equation (1.5).
Because the release rate from the secondary system is indirect (e.g., SG blowdown is normally routed to condenser where the condensate clean-up system will remove much of the radioactive material),
samples should be collected from the release point (i.e., Chemical Waste Basin) for quantifying the radioactive material releases. However, for conservatism and ease of controlling and quantifying all potential release paths, it is prudent to sample the SG blowdown and to assume all radioactive material is released directly to the environment via the Chemical Waste Basin. This approach while not exact is conservative and ensures timely analysis for regulatory compliance. Accounting for radioactive material retention of the condensate clean-up system ion exchange resins may be needed to more accurately account for actual releases.
In addition to the secondary releases described in this section, the Salem Ground Water Remediation System also can potentially discharge radioactive material to the Chemical Waste Basin. To ensure regulatory compliance, the releases are monitored by Radiation Monitor R-37. Samples are also collected, and analyzed for radionuclides. Based on the identified radioactive material levels and the volume of water discharged, the resulting environmental doses may be calculated based on equation (1.5).
1.6 Liquid Effluent Dose Projections S ODCM CONTROL 3.11.1.3 requires that the liquid radioactive waste processing system be used to reduce the radioactive material levels in the liquid waste prior to release when the quarterly projected cn doses exceed:
< - 0.375 mrem to the total body, or o - 1.25 mrem to any organ.
The applicable liquid waste processing system for maintaining radioactive material releases ALARA is the ion exchange system as delineated in Figure 1-3. Alternately, the waste evaporator as presented in the Salem FSAR has processing capabilities meeting the NRC ALARA design requirements and may be used in conjunction or in lieu of the ion exchange system for waste processing requirements in accordance with ODCM CONTROL 3.11.1.3. These processing requirements are applicable to each I
0 unit individually. Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit. I Dose projections are made at least once per 31 days by the following equations:
D tbp = D tb 9 (1.9)
I l l ,.
D ap=D
=px 91 (1.10)
_J Where:
00 Wm Dtbp = the total body dose projection for current calendar quarter (mrem)
P 2 Q .Page 82 of 157
(
Salem ODCM Rev. 23 Dtb the total body dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem)
I = the maximum organ dose projection for current calendar quarter (mrem)
Dmax = the maximum organ dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) d the number of days to date for current calendar quarter 91 the number of days in a calendar quarter 2.0 GASEOUS EFFLUENTS 2.1 Radiation Monitoring Instrumentation and Controls The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Technical Specifications 6.8.4.g and ODCM CONTROLS are summarized as follows:
1) Waste Gas Holdup System - The vent header gases are collected by the waste gas holdup system.
Gases may be recycled to provide cover gas for the CVCS hold-up tank or held in the waste gas tanks for decay prior to release. Waste gas decay tanks are batch released after sampling and analysis. The tanks are discharged via the Plant Vent. 1-R41D provides noble gas monitoring and automatic isolation I of waste gas decay tank releases for Unit-1. This function is provided by 2-R41D for Unit-2.
2) Containment Purge and Pressure/Vacuum Relief - containment purges and pressure/vacuum reliefs I1 are released to the atmosphere via the respective unit Plant Vent. Noble gas mronitoring and auto I isolation function are provided by 1-R41D for Unit-1 and 2-R41D for Unit-2. Additionally, in accordance with ODCM CONTROL 3.3.3.9, Table 3.3-13, 1-R12A and 2-RI2A may be used to CD provide the containment monitoring and automatic isolation function during purge and pressure/vacuum I~
(..)
reliefs (*)
I 3) Plant Vent -The Plant Vent for each respective unit receives discharges from the waste gas hold-up
< system, condenser evacuation system, containment purge and pressure/vacuum reliefs, and the 0 Auxiliary Building ventilation. Effluents are monitored by R41D, a flow through gross activity monitor I D
~ (for noble gas monitoring). Radioiodine and particulate sampling capabilities are provided by charcoal n cartridge and filter medium samplers. Additionally, back-up sampling capability for.radioiodine and particulates is provided at the I-R45 and 2-R45 sampling skids. Plant Vent flow rate is measured and o as a back-up may be determined'empirically as a function of fan operation (fan curves). Sampler flow
> rates are determined by flow rate instrumentation (e.g., venturi rotameter).
-v Gaseous radioactive effluent flow diagrams with the applicable, associated radiation monitoring Z instrumentation and controls are presented in Figures 2-1. A simplified diagram of the Gaseous u_ radioactive waste disposal system is provided in Figure 2-2.
r-
The R12A in Mode 6 provides containment monitoring and alarm functions Without automatic V) isolation P
~Page 83 of 157
I Salem ODCM Rev. 23 I
2.2 Gaseous Effluent Monitor Setpoint Determination 2.2.1 Containment and Plant Vent Monitor I Per the requirements of ODCM CONTROL 3.3.3.9, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of ODCM CONTROL 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of I
500 mremr/year to the total body or 3000 mrem/year to the skin.
Based on a grab sample analysis of the applicable release (i.e., grab sample of the Containment I
atmosphere, waste gas decay tank, or Plant Vent), the radiation monitoring alarm setpoints may be established by the following calculation method. Themeasured radionuclide concentrations and release rate are used to calculate the fraction of the allowable release rate, as limited by Specification:3.11.2.1, I
by the equation:
I FRA C + 02 *' '[4.72E
I VF*Z0079'_.. (2.1)
I FRA C [4.72 E+o02 *'Q VF* ONE (2.2)
= 'j ;Mf I
Where:
FRAC fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate I
z - annual average meteorological dispersion to the controlling site boundary location I (sec/i 3)
C/) VF Ci ventilation system flow rate for the applicable release poinrt and' monitor (ft3/min) concentration of noble gas radionuclide i as determined by radioianalysis of grab sample I
(P+/-Ci/cm 3) z 0.-
rK Ki = total body dose conversion factor for noble gas radionuclide i (mrem/yr per ýtCi/m 3 from Table 2-1)
I Li beta skin dose conversion factor for noble gas radionuclide i (irem/yr per ptCi/i 3 from z Mi
=
=
Table 2-1) gamma air dose conversion factor for noble gas radionuclide i (mrem/yr per .tCi/m 3 I from Table 2-1) w 0
1.1 500 =
= mremr skin dose per mrad gamma air dose (mrem/mrad) total body dose rate limit (mrem/yr)
I 3000 = skin dose rate limit (mrem/yr) z L-4.72 E+02 = conversion factor (cm 3/ft3
min/see) I 0
a--
Lial Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (R41D, and/or R12A) may be calculated by the equation: I I
D Page 84 of 157 I
Salem ODCM Rev. 23 iS
sp rCi*SEN1
+bkg (2.3)
H Where:
SP = alarm setpoint corresponding to the maximum allowable release rate (cpm)
SEN = monitor sensitivity (cpm per ýXCi/cm 3) bkg = background of the monitor (cpm)
AF = administrative allocation factor for the specific monitor and type release, which corresponds to the fraction of the total allowable release rate that is administratively allocated to the release.
The allocation factor (A.F) is an administrative control imposed to ensure that combined releases from Salem Units 1 and 2 and Hope Creek will not exceed the regulatory limits on release rate from the site I (i.e., the release rate limits of ODCM CONTROL 3.11.2.1). Normally, the combined AF value for Salem Units I and 2 is equal to 0.5 (0.25 per unit), with the remainder 0.5 allocated to Hope Creek.
Any increase in AF above 0.5 for the Salem Nuclear Generating Station will be coordinated with the Hope Creek Generating Station to ensure that the combined allocation factors for all units do not exceed 1.0.
I g 2.2.2 Conservative Default Values
) A conservative alarm setpoint can be established, in lieu of the individual radionuclide evaluation based Q on the grab sample analysis, to eliminate the potential of periodically having to adjust the setpoint to
.I reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default values presented in Table 2-2.1 and 2-2.2 for Units 1 and 2, respectively. These values are based upon:
-the maximum ventilation (or purge) flow rate; a radionuclide distributiona comprised of 95% Xe-133, 2% Xe-135, 1%Xe-133m, 1% Kr-88 and I1% -
Kr-85; and
- an administrative allocation factor of 0.25 to conservatively ensure that any simultaneous releases 0 from Salem Units 1 and 2 do not exceed the maximum allowable release rate. For this radionuclide I__. distribution, the alarm setpoint based on the total body dose rate is more restrictive than the
,.a corresponding setpoint based on the skin dose rate.
Z a) Adopted from ANSI N237-1976/ANS-18.1, Source Term Specifications, Table 6 h] 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations -10 CFR 20 2.3.1 Site Boundary Dose Rate - Noble Gases 0U-
,w, z
.
ODCM CONTROL 3.11.2. La limits the dose rate at the SITE BOUNDARY due to noble gas releases to <500 mrem/yr, total body and <3000 mrem/yr, skin. :Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded. Inbthe event any gaseous releases from on the station results in an alarm setpoint being exceeded, an evaluation of the SITE BOUNDARY dose
,., rate resulting from the release shall be performed using the following equations:
i*ry Page 85 of 157
Salem ODCM Rev. 23 I
S~ Q (2.4) and I Di=XQ - MKELM_ ;E:
~I (2.5)
Where:
Dtb = total body dose rate (mrem/yr)
D, = skin dose rate (mrem/yr)
Q = atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3)
Qi = average release rate of radionuclide i over the release period under evaluation (jtCi/sec)
Ki = total body dose conversion factor for noble gas radionuclide i (mrem/yr per gCi/mr3, from Table 2-1)I Li = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per pCi/m3 , from Table 2-1)
Mi = gamma air dose conversion factor for noble gas radionuclide i (mrad/yr per P.Ci/m 3, from n 1.1 =
Table 2-1) mrem skin dose per mrad gamma air dose (mrem/mrad) I As appropriate, simultaneous releases from Salem Units 1 and 2 and Hope Creek will be considered in evaluating compliance with the release rate limits of ODCM CONTROL 3.11 .2.1 a, following any
-C<
0 release exceeding the aboveprescribed alarm setpoints. I Monitor indications (readings) may be averaged over a time period not to exceed 15 minutes when determining noble gas release rate based on correlation of the monitor reading.and monitor sensitivity.
The 15-minute averaging is needed to allow for reasonable monitor' response to potentially changing radioactive material concentrations and to exclude potential electronic spikes in monitor readings that 0 may be unrelated to radioactive material releases. As identified, any electronic spiking monitor 2" responses may be excluded from the analysis.
NOTE: For administrative purposes, more conservative alarm setpoints than those as prescribed above may be imposed. However,. conditions exceeding these more limiting I
CD alarm setpoints do not necessarily indicate radioactive material release rates exceeding z the limits of ODCM CONTROL 3.11.2.1.a. Provided actual releases do not result in r_ radiation monitor indications exceeding alarm setpoint values based on the above uc> ofOcMriteria, ncoNTofurther ,.
analyses3.121aare required for demonstrating, compliance. with the limits zy 0
of ODCM CONTROL 3.11.2.1.a.
Lt, Actual meteorological conditions concurrent with the release period or the default, annual average
!1 dispersion parameters -j-as presented in Table 2-3 may be used for evaluating the gaseous effluent dose z rate.
0 LUI a:, Page 86 of 157 dL D I
Salem ODCM Rev. 23 2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates I
ODCM CONTROL 3.11.2.1.b limits the dose rate to <1500 mrem/yr to any organ for 1-131, tritium, and particulates with half-lives greater than 8 days. To demonstrate compliance with this limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (e.g,, nominally once per 7 days). The following equation shall be used for the dose rate evaluation:
ID 0 = (2.6)
Where:
Do = average organ dose rate over the sampling time period (mrem/yr)
VQ atmospheric dispersion to the controlling SITE BOUNDARY location for the inhalation pathway (sec/mr3 )
Rio. = dose parameter for radionuclide2-4 i (mrem/yr per PCi/m 3) and organ o for the child pathway from Table inhalation Qi = average release rate over the appropriate sampling period and analysis frequency for radionuclide i -- 1-131, tritium or other radionuclide in particulate form with half-life greater than 8 days (gtCi/sec)
By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for 1-131 can be determined. Based on the annual average meteorological dispersion (see Table 2-3) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid -- Rio = 1.62E+07 mrem/yr per pCi/i 3), the allowable release rate for 1-131 is 42 pCi/sec. Reducing this release rate by a factor of LW< 4 to account for potential dose contributions from other radioactive particulate material and other release points (e.g., Hope Creek), the corresponding release rate allocated to each of the Salem units is I~ 10.5 pCi/sec.
0 For a 7 day period, which is the nominal sampling and analysis frequency for 1-131, the cumulative I release is 6.3 Ci. Therefore, as long as the 1-131 releases in any 7 day period do not exceed 6.3 Cij no additional analyses are needed for verifying compliance with the ODCM CONTROL 3.11.2. .b limits on allowable release rate.
2.4 Noble Gas Effluent Dose Calculations - 10 CFR 50 I 2.4.1 UNRESTRICTED AREA Dose - Noble Gases
,_ ODCM CONTROL 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate I >
compliance with the quarterly dose limits of_<5 mrad, gamma-air and <10 mrad, beta-air and the calendar year limits <10 mrad, gamma-air and <20 mrad, beta-air. The limits are applicable separately to each unit and are not combined site limits. The following equations shall be used to calculate the gamma-air and beta-air doses:
Eo SDy =3.17E-08*Z M (2.7) iy Page 87 of 157
Salem ODCM Rev. 23 andI D6= 3.17E -08 A,*ZQ N (2.8)
Where:
Dr. = air dose due to gamma emissions for noble gas radionuclides (mrad)
D6 = air dose due to beta emissions for noble gas radionuclides (mrad)
= atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m3) 7Q Qi = cumulative release of noble gas radionuclide i over the period of interest ([LCi) where gLCi = (P.tCi/cc)*(cc released) or (pCi/sec)*(sec released)
Mi = air dose factor due to gamma emissions from noble gas radionuclide i N(mrad/yr per PCi/m 3, from Table 2-1) I Ni = air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per pCi/m3 , Table 2-1) 3.17E-08 = conversion factor (yr/sec) I 2.4.2 Simplified Dose Calculation for Noble Gases In lieu of the individual noble gas radionuclide dose assessment as presented above, the following simplified dose calculation equations may be used for verifying compliance with the dose limits of ODCM CONTROL 3.11.2.2. (Refer to Appendix C for the derivation and justification for this simplified method and for values of Meff, and Neff..)
LU3 3.17E-08 S-* *Meff*
Q2.9) i (2.9)
.0.50 0) and i--D,6 =- ~~~3.17E- 08 Nef *z~ (2.10) 0 z-0.50 Q *(10I ef*Q, 0.5 0
SWhere:
Meff = 5.3E+02, effective gamma-air dose factor (mrad/yr per p.Ci/m 3)
Neff = 1.1E+03, effective beta-air dose factor (mrad/yr per pCi/m 3 )
Qi = cumulative release for all noble gas radionuclides ( Ci), where-ptCi =(p.Ci/cc) * (cc released) or (p.iCi/sec) * (sec released) 00K 0.50 = conservatism factor to account for potential variability in the radionuclide distribution Actual meteorological conditions concurrent with the release period or the default, annual average zn dispersion parameters as presented in Table 2-3, may be used for the evaluation of the gamma-air and Z beta-air doses.
0-L I V) ry D/
Page 88 of 157
Salem ODCM Rev. 23 2.5 Radioiodine and Particulate Dose Calculations - 10 CFR 50 2.5.1 UNRESTRICTED AREA Dose - Radioiodine and Particulates In accordance with requirements of ODCM CONTROL 3.11.2.3, a periodic assessment shall be performed to evaluate compliance with.the quarterly dose limit of<7.5 mrem and calendar year limit
-<5mrem to any organ. The following equation shall be used to evaluate the maximum organ dose due to releases of 1-131, tritium and particulates with half-lives greater than 8 days:
Daop 3.17E-O8*W*SFp*" Y 2 Q" (2.11)
Where:
Daop dose or dose commitment via all pathways p and controlling age group a (as identified in Table 2-3) to organ o, including the total body (mrem)
W = atmospheric dispersion parameter to the controlling location(s) as identified in Table 2-3
'Y/Q
/ atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways (sec/m 3)
D/Q = atmospheric deposition for vegetation, milk and ground plane exposure pathways (m-')
Riop = dose factor for radionuclide i (mrem/yr per .tCi/m 3) or (m2 - mrerrlyr per .tCi/sec)
- .and organ o from Table 2-4 for each age group and the applicable pathway p as identified in Table 2-3. Values for Riop were derived in accordance with the methods described in NUREG-0133.
Q Q cumulative release over the period of interest for radionuclide i -- 1-131, tritium, or r radioactive material in particulate form with half-life greater than 8 days (ýiCi).
SFp = annual seasonal correction factor to account for the fraction of the year that the applicable exposure pathway does not exist.
1) For milk and vegetation exposure pathways:
A six month fresh vegetation and grazing season (Mayithrough October) 0.5
2) For inhalation and ground plane exposure pathways: =J1..0 For evaluating the maximum exposed individual, only the controlling pathways and age group as
_o identified in Table 2-3 need be evaluated for compliance with ODCM CONTROL 3.11.2.3.
wa 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates.
IZ_-_
In lieu of the individual radionuclide (1-131, tritium, and particulates) dose assessment for the resident/dairy location as presentedabove, the following simplified dose calculation equation may be used for verifying compliance with the dose limits of ODCM CONTROL 3.1 12.3 (refer to Appendix D ry for the derivation and justification of this simplified method).
0 Dmax=3.17E-08*W*SFp*R[-131*,Qi (2.12)
C/)
P o i~
La-or-ryPage 89 of 157
Salem ODCM Rev. 23 Where: I Dmax = maximum organ dose (mrem)
R1 -31 = 1-131 dose parameter for the thyroid for the identified controlling pathway
= 1.05E+12, infant thyroid dose parameter with the grass-cow-milk pathway W
mcontrllin
- mrem/yr per ItCi/sec)
= D/Q for radioiodineý 2.1E-10 1/m2 I
.Qi = cumulative release over the period of interest for radionuclide i.- 1-131, tritium, or radioactive material in particulate from with half life greater than 8 days (ýLCi)
The dose should be evaluated based on the predetermined controlling pathways as identified in Table 2-
3. If more limiting exposure pathways are determined to exist in the surrounding environment of Salem by the annual land-use census, Table 2-3 will be revised as specified in ODCM CONTROL 3.12.2.
2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations I During periods of primary to secondary leakage, minor levels of radioactive material may be released
via the secondary system to the atmosphere. Non-condensables (e.g., noble gases) will be I predominately released via the condenser evacuation system and will be monitored and quantified by the routine plant vent monitoring and sampling system and procedures (e.g., Ri 5 on condenser evacuation, R41D on plant vent, and the plant vent particulate and charcoal samplers).
SHowever, if the Steam Generator blowdown is routed directly to the Chemical Waste Basin (via the SG blowdown flash tank) instead of being recycled through the condenser, it may.:be desirable to account for the potential atmospheric releases of radioiodines and particulates from the flash tank vent (i.e.,
.. z releases due to moisture carry over). Since this pathway is not sampled or monitored, it is necessary to
< calculate potential releases.
Based on the guidance in NRC NUREG-0133, the releases of the radioiodines and particulates shall be calculated by the equation:
Qi=Ci*Rsgb*Fft* .(2.13) 0 Where:
'" Qi = the release rate of radionuclide,, i, from the steam generator flash tank vent (ýtCi/sec)
Ci = the concentration of radionuclide, i, in the secondary coolant water averaged over not more than one week (ýCi/ml)
Rsgb = the steam generator blowdown rate to the flash tank (ml/sec)
I Fft = the fraction of blowdown flashed in the tank determined from a heat balance taken around the flash tank at the applicable reactor power level 0 SQf'= the measured steam quality in the flash tank vent; or an assumed value of 0.85, based on NUREG-0017.
U Tritium releases via the steam flashing may also be quantified using the above equation with the o assumption of a steam quality (SQft,) equal to 0. Since the H-3 will be associated with the water w(/)
Uj I Page 90 of 157 V)
Salem ODCM Rev. 23
- molecules, it is not necessary to account for the moisture carry-over which is the transport media for the radioiodines and particulates.
Based on the design and operating conditions at Salem, the fraction of blowdown converted to steam (Fft) is approximately 0.48.. The equation simplifies to the following:
Qi 0.072*Ci*K gb (2.14)
For H-3, the simplified equation is:
Qi = 0.48 G* kgb (2.15)
Also during reactor shutdown operations with a radioactively contaminated secondary system, radioactive material may be released to the atmosphere via the atmospheric reliefs (PORV) and the safety reliefs on the main steam lines and via the steam driven auxiliary feed pump exhaust. The evaluation of the radioactive material concentration in the steam relative to that in the steam generator water is based on the guidance of NUREG-0017, Revision 1. The partitioning factors for the radioiodines is 0.01 and is 0.005 for all other particulate radioactive material. "Theresulting equation for quantifying releases via the atmospheric steam releases is:
QY 0=0.13.
C F7ii (2.16)
C/')
SWhere:
Qij = release rate of radionuclide i via pathway j,(QtCi/sec)
< Cij = concentration of radionuclide i, in pathway j,(ptCi/ml)
SFj = steam flow for release pathway j
- 400,000 lb/hr per PORV
< = 850,000 lb/hr per safety relief valve
= 62,500 lb/hr for auxiliary feed pump exhaust PFi = partitioning factor, ratio of concentration in steam to that in the water in the steam generator
- 0.01 for radioiodines
= 0.005 for all other particulates
= 1.0 for H-3 rc 0.13 = conversion factor - [(hr*ml) / (sec*lb)]
L= Any significant releases of noble gases via the atmospheric steam releases can be quantified in accordance with the calculation methods of the Salem Emergency Plan Implementation Procedure.
0 Alternately, the quantification of the release rate and cumulative releases may be based on secondary
w samples. The measured radionuclide concentration in the secondary system may be used for
_j quantifying the noble gases, radioiodine and particulate releases.
7 ryPage 91 of 157
I Salem ODCM Rev. 23 Note: The expected mode of operation would be to isolate the effected steam generator, thereby I reducing the potential releases during the shutdownrcooldown process. Use of the above calculation methods should consider actual operating conditions and release mechanisms.
The calculated quantities of radioactive materials may be used as inputs to the equation (2.11) or (2.12) to calculate offsite doses for demonstrating compliance with the Technical Specifications 6.8.4.g and the ODCM CONTROLS.
2.7 Gaseous Effluent Dose Projection ODCM CONTROL 3.11.2.4 requires that the GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected doses exceed one-half the annual design objective rate in any calendar quarter, i.e., exceeding:
- 0.625 mrad/quarter, gamma air;*
- 1.25 mrad/quarter, beta air; or
- 1.875 mrem/quarter, maximum, organ.
The applicable gaseous processing systems for maintaining radioactive material releases ALARA are the Auxiliary Building normal ventilation system (filtration systems # 1, 2 and 3) and the Waste Gas Decay Tanks as delineated in Figures 2-1 and 2-2. Dose projections are performed at least once per 31
days by the following equations:
(.)
~I Dflp= D *9 (2.18)
Dmaxp D max* (2.19)
I-) z" Where: I
> D = gamma air dose projection for current calendar quarter(mrad) tDy = gamma air dose to date for current calendar quarter as determined by Equation 2.7 or zo 2.9 (mrem)
>- Dpp = beta air dose projection for current calendar quarter (mrad)I Dp = beta air dose to date for current calendar quarter as determinied by Equation 2.8 or
> 2.10 (mrem) ay Dmaxp = maximum organ dose projection for current calendar quarter (mrem)
_J - Dmax = maximum organ dose to date for current calendar quarter as z determined by Equation 2.11 or 2.12 (morem) o d = number of days to date in current calendar quarter na-m 91 = number of days in a calendar quarter Cf).
Dj*
Salem ODCM Rev. 23 3.0 SPECIAL DOSE ANALYSES I
3.1 Doses Due To Activities Inside the SITE BOUNDARY In accordance with ODCM CONTROL.6.9.1.8, the Radioactive Effluent Release Report (RERR) shall include an assessment of radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.
The calculation methods as presented in Sections 2.4 and 2.5 may be used for determining the maximum potential dose to a MEMBER OF THE PUBLIC located inside the site boundary. For the purpose of this calculation, a MEMBER OF THE PUBLIC is an adult individual who is not subject to occupational exposure (i.e., an un-monitored site worker) performing duties within the site boundary, and who is exposed to radioactive material in gaseous effluent for 2,000 hours0 days 0 hours 0 weeks 0 months per year via the inhalation and ground plane exposure pathways. The values for the atmospheric dispersion coefficients at the point of interest inside the site boundary (e.g., 0.25 mile) shall be developed from the current year meteorological data.
3.2 Total dose to MEMBERS OF THE PUBLIC - 40 CFR 190 and 10 CFR 72.104 The Radioactive Effluent Release Report (RERR) shall also include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources (including dose contributions from effluents and direct radiation from on-site sources). For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Artificial Island, the sources of exposure need only consider the Salem Nuclear Generating Station and the Hope Creek Nuclear Generating Station which includes the Independent Spent Fuel Storage Installation (ISFSI): No other fuel cycle facilities contribute to the MEMBER OF THE PUBLIC i* dose for the Artificial Island vicinity.
r-C-) The dose contribution from the operation of Hope Creek Nuclear Generating Station will be estimated based on the methods as presented in the Hope Creek Offsite Dose Calculation Manual I(HCGS ODCM).
As. appropriate for demonstrating/evaluating compliance with the limits of ODCM CONTROL
< 3.11.4 (40 CFR 190), the results of the environmental monitoring program may be used for V providing data on actual measured levels of radioactive material in the actual pathways of exposure.
3.2.1 Effluent Dose Calculations w
tra For purposes of implementing the surveillance requirements of ODCM CONTROL 3/4.11.4 and the Z reporting requirements of 6.9.1.8 (RERR), dose calculations for the Salem Nuclear Generating
>- Station should be performed using the controlling pathways and locations of Table 2-3 and the ry calculation methods contained within this ODCM. If more limiting exposure pathways are
> determined to exist in the surrounding environment of Salem by the annual land-use census, Table 2-al*
0 3 will berevised as specified in ODCM CONTROL 3.12.2.
Average annual meteorological dispersion parameters or meteorological conditions concurrent with M the release period under evaluation may be used.
z Page 93 of 157
Salem ODCM Rev. 23 3.2.2 Direct Exposure Dose Determination.
Any potentially significant direct exposure contribution to off-site individual doses may be evaluated based on the results of the environmental measurements (e.g., DLR, ion chamber measurements) and/or by the use of a radiation transport and shielding calculation method.
Only during a non-typical condition will there exist any potential for significant on-site sources at Salem that would yield potentially significant off-site doses (i.e., in excess of 1 mrem per year to a MEMBER OF THE PUBLIC), that would require detailed evaluation for demonstrating compliance I with 40 CFR 190 or 10 CFR 72.104.
However, should a situation exist where the direct exposure contribution is potentially significant, on-site measurements, off-site measurements and/or calculation techniques will be used for determination of dose for assessing 40 CFR 190 or 10 CFR 72.104 compliance.
4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Program The operational phase of the Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of ODCM CONTROL 3.12. The objectives of the program are: I
- To determine whether any significant increases occur in the concentration of radionuclides in the critical pathways of exposure in the vicinity of Artificial Island; I
- To determine if the operation of the Salem Nuclear Generating Stations has resulted in any increase zc in the inventory of long lived radionuclides in the environment;
- To detect any changes in the ambient gamma radiation levels; and
- To verify that SNGS operations have no detrimental effects on the health and safety of the public or D on the environment.
V The sampling requirements (type of samples*, collection frequency and analysis) and sample locations are presented in Appendix E.
I S *NOTE: No public drinking water samples or irrigation water samples are required as these W" pathways are not directly effected by liquid effluents discharged from Salem Generating Station.
~I 4.2 Interlaboratory Comparison Program
> ODCM CONTROL 3.12.3 requires analyses be performed on radioactive material supplied as part of r" an Interlaboratory Comparison Program. Participation in an approved Interlaboratory Comparison 0
Program provides a check on the precision and accuracy of measurements of radioactive materials in
_J environmental samples.
zm A summary of the Interlaboratory Comparison Program results will be provided in the Annual a,-* Radiological Environmental Operating Report pursuant to ODCM CONTROL 6.9.1.7.
a* Page 94 of 157 VI)
m -m- m-M M n-M M M M M M M M MM USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES 0 Salem ODCl~ev. 23 FIGURE 1-1: LIQUID RELEASE FLOWPATH UNIT 1 I SERVICE WATER NITOR.K. ..
GROUND WATER TO NON-RAD (C) CONTAINMENT FAN COIL UNITS REMEDIATION 1 (Figure 1-2)
SYSTEM TO CIRC LATING WATER SYSTEM TO NON-RAD (B) R 13 MONITORS (Figure 1-2)
'4' To River Page 95 of 157
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES Salem ODCMev. 23 FIGURE 1-2: LIQUID RELEASE FLOWPATH UNIT 2 I?
n m m m m m m m m m m m m m m m m m m
- m M M nM M M M M M- M M M MM USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES 0 Salem ODCM41ev. 23 FIGuRE 1-3: LIQUID RADIOACTIVE WASTE SYSTEM Page 97 of 157
I Salem ODCM Rev. 23 I
Table 1-1.1" Parameters for Liauid Alarm Setnoint Determinations Unit 1 n Dee m nain Unit.
ia . lr ......
I Table ... 1:P m r tr o L Parameter Actual Default Comments MPCC Value Calculated Value 6.05E-06
Units iiCi/ml Calculated for each batch to be released.
I WPC 1-131 3.OE-07 N/A .Ci/ml 1-131"MPC conservatively used for SG blowdown and Service Water monitor setpoints.
I Ci Measured N/A JICi/ml Taken from gamma spectral analysis of liquid effluent.. I MPCi as determined N/A ýtCi/ml Taken from 10 CFR 20, Appendix B, Sensitivity as determined N/A cpm per Table II, Col 2 (Appendix F)..
Monitor sensitivities are controlled I
1-R18 Cti/ml under Public Service Blueprint 1-R19 (AB,C,D) l-R13 (A and B)
Document (PSBP) 315733 I
CW as determined 1.00E+05 gpm Circulating water system - single CW RR as determined gpm pump
Determined prior to release; release rate I
l-R18 120 ; can be adjusted for ODCM CONTROL compliance I
1-R19 250 Steam Generator blowdown rate per 1.00 E +05 Generator Service Water flow rate for I
0Z 1-R13 Containment fan coolers Setpoint I-R18 Calculated N/A cpm Monitor setpoints are controlled under Public Service Blueprint Document I
1-R19 (AB,C,D)** (PSBP) 315733 z*
V-)
D I-R13 (A and B)**
Correction Factor as determined_. 0.75 Unidess Default parameter to account for non-I (Non-Gamma) gamma emitting nuclides.
Allocation Factor 1-RI9 0.5 0.5 Unitless Conservatism factor to preclude exceeding MPC limit in the case of I
Q- simultaneous primaiy-to-secondary QA 0
H-leaks at both Salem Units I
Cii w
Refer to Appendix A for derivation
The MPC value of 1-131 (3E-07 .tCi/ml) has beenused for derivation of R19 Steam Generator Blowdown and R13 Service I
Water monitor setpoints as discussed in Section 1.2.2
__J w"
** During periods when Unit 2 Circulators are out of service, the CW flow for l-RI3 monitors is zero. See Section 1.2.2.
i 0C/)
al V)
U 0*
i i Page 98 of 157 i
Salem ODCM Rev. 23
i. Table 1-1.2: Parameters for Liquid Alarm Setpoint Determinations - Unit 2 Parameter Actual Default Units Value Value Comments MPC, Calculated 4.8 1E-06'* LCi/ml Calculated for each batch to be released.
MPC 1-131 3.0E-07 N/A 4Ci/mi 1-131 MPC conservatively used for SG bloýwdown, Service Water and Chemical Waste Basin monitor setpoints.
C, Measured N/A ACi/ml Taken from gamma spectral analysis of liquid effluent.
MPCi as determined N/A ptCi/ml Taken from 10 CFR 20, Appendix B, Table I1,Col. 2 (Appendix F)
Sensitivity as determined N/A cpm per. Monitor sensitivities are controlled 2-RI8 11Ci/ml under Public Service Blueprint Document 2R19(A,B,C,D) 315734 2-R13(A and B)
R37 CW as determined 1.0E+05 gpm Circulating Water System, single CW pump
RR as determined 120 gpm Determined prior to release; release rate
n. 2-R18 can be adjusted for ODCM CONTROL Compliance 250 2-119 Steam Generator Blowdown rate per "7-Generator 1.OE+05 2-R13 Circulating Water System, single CW 0
1200 Pump L(U R37 Chemical Waste Basin discharge V)
Setpoint Calculated N/A cpm Monitor setpoints are controlled under 2-R18 Public Service Blueprint Document i-) 2-R19(A,B,C,D)** (PSBP) 315734 2-R13(A and B)**
R37 **
LU 0
Correction Factor as determined 0.75 Unitless. Default parameter to account for non-(Non-Gamma) gamma emitting nuclides.
>- Allocation Factor 0.5 0.5 Unitless Conservatism factor to preclude Hi rý_ 2-R19 exceeding MPC limit in the case of simultaneous primary-to-secondary leaks at both Salem Units 0
Refer to Appendix A for derivation
The MPC value of 1-131 (3.OE-7 ýiCi/ml) has been used for derivation of the R13, R19 and R37 monitor setpoints as ry 1L.
discussed in Section 1.2.2 During periods when Unit 1 Circulators are out of service, the CW flow for 2-R13 monitors is zero. See Section 1.2.2.
(/3 Lii Page 99 of 157
I Salem ODCM Rev. 23 I
TABLE 1-2: Site Related Ingestion Dose Commitment Factor, Ai.
(Fish And Invertebrate Consumption)
(mrem/hr per .LCi/ml)
SI. Ip I g1N 1-3 - 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 C-14 3.45E+4 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 Na-14 .4.57E-1 2.90E-1 2.90E-1 2.90E-3 2.90E-1 2.90E-1 2.90E-Na-24 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 P-32 4.69E+6 2.91E+5 18E5 -- 5.27E+5 Cr-51 Mn-54 Mn-56 7.06E+3 1.78E+2 5.58E+O 1.35E+3 3.15E+1 3.34E+O 1.23E+O 2.10E+3 2.26E+2 7.40E+O 1.40E+3 2.16E+4 5.67E+3 I
Fe-55 5.11E+4 3.53E+4 8.23E+3 - 1.97E+4 2.03E+4 Fe-59 8.06E+4 1.90E+5 7.27E+4 - 5.30E+4 6.32E+5 Co-57 - 1.42E+2 2.36E+2 - 3.59E+3 Co-58 - 6.03E+2 1.35E+3 - - 1.22E+4 Co-60 - 1.73E+3 3.82E+3 - - 3.25E+4 Ni-63 4.96E+4 3.44E+3 1.67E+3 - - 7.18E+2 Ni-65 2.02E+2 2.62E+1 1.20E+1 - - 6.65E+2 Cu-64 - 2.14E+2 1.01E+2 - 5.40E+2 - 1.83E+4 Zn-65 1.61E+5 5.13E+5 2.32E+5 - 3.43E+5 - 3.23E+5 Zn-69 3.43E+2 6.56E+2 4.56E+1 - 4.26E+2 - 9.85E+1
, As-76 4.38E+2 1.16E+3 5.14E+3 3.42E+2 1.39E+3 3.58E+2 4.30E+4 Br-82 - -4.07E+O - 4.67E+O Br-83 - - 7.25E-2 - - 1.04E- I Co Br-84 - - 9.39E-2 - - 7.37E-7
< Br-85 - 3.86E-3 - -
o Rb-86 - 6.24E+2 2.91E+2 - - 1.23E+2 Rb-88 - 1.79E+O 9.49E-1 - - 2.47E- 11 z Rb-89 - 1.19E+O 8.34E-1 - - 6.89E-14 Sr-89 4.99E+3 - 1.43E+2 - - 8.OOE+2 D Sr-90 1.23E+5 - 3.01E+4 - - 3.55E+3 Sr-91 Sr-92 9.18E+l 3.48E+1 3.71E+O 1.51E+O 4.37E+2 6.90E+2 II Y-90 6.06E+O - 1.63E-1 - 6.42E+4 o Y-91m 5.73E-2 - 2.22E-3 - 1.68E-1
> Y-91 8.88E+1 - 2.37E+O 4.89E+4 Y-92 5.32E-1 , 1.56E-2 - - - 9.32E+3 Y-93 1.69E+O - 4.66E-2 - 5.35E+4 Z Zr-95 1.59E+1 5.11E+O 3.46E+O - 8.02E+O - 1.62E+4
" Zr-97 8.81E-1 1.78E-1 8.13E-2 2.68E-1 - 5.51E+4 Iy Nb-95 4.47E+2 2.49E+2 1.34E+2 2.46E+2 - 1.51E+6 w
> Nb-97 3.75E+O 9.49E-1 3.46E-1 -Ml1E+O - 3.50E+3 Mo-99 1.28E+2 2.43E+1 2.89E+2 - 2.96E+2
,. Tc-99m 1.30E-2 3.66E-2 4.66E-1 5.56E-1 1.79E-2 2.17E+1 Li Tc-101 1.33E-2 1.92E-2 1.88E-1 3.46E-1 9.81E-3 5.77E-14
__jJ 0
ry ~Page 100 of 157 DI
Salem ODCM Rev. 23 TABLE 1-2 (cont'd).
Site Related Ingestion Dose Commitment Factor, Aio i (Fish And Invertebrate Consumption)
(mrem/hr per pCi/mi)
Ru-103 1.07E+2 8.60E+1 4 - 4.07E+2 1.25E+4 Ru-105. 8.89E+- 3.51 .E+O 1.15E+2 5.44E+3 Ru-106 1.59E+3 - 2.01E+2 - 3.06El3 1.03E+5 I Te-103m Rh-1066- 1 - -.
Ag-110m 1.56E+3 1.45E+3 8.60E+2 -.-E 2.85E+3 - 5.91E+5 Sb-122 1.98E+1 4.55E-1 6.82E+ 3.06E-1 - 1.19E+1 7.5 1E+3 Sb-124 2.77E+2 5.23E+0 1.10E+2 6.71E-1 - 2.15E+2 7.86E+3 Sb-125 1.77E+2 1.98E+0 .4.21E-l 1.80E-1 - 1.36E+2 1.95E+3.
Sb-126 1.14E+2 2.31E+2 4.10E+l 6.96E-1 - 6.97E+1: 9.29E+3 Te-125m 2.17E+2 7.86E+1 2.91E+1 6.52E+1 8.82E+2 - 8.66E+2 Te-127m 5.48E+2 1.96E+2 6.68E+1 1.40E+2 2.23E+3 - 1.84E+3 Te-127 8.90E+1 3.20E+0 1.93E+O 6.60E+2 3.63E+1 - 7.03E+2 Te-1 297m 9.31.E+2 3E47E+2 1.47E+2 3.20E+2 3.89E+3 - 4.69E+3 Te-129 2.54E+O 9.55E-1 6.19E-1 1.95E+2 1.07E+1 - 1.92E-2 Te-131m 1.40E+2 6.85E+1 5.71E+1 1.08E+2 6.94E+2 - 6.80E+3 S i Te-131 1.59E+O 6.66E-1 5.03E-1I 1.31E- " 6.99E-iO 2.26E-1' S CTe-132 2.04E+2 1.32E+2 1.24E+2 1,46E+2 1,27E+3 7 6.24E+3
/) 1-130 3.96E+1 1.17E+2 4.61E+1 9.91E+3 1.82E+2 2 1.01E+2 C1-131 2.18E+2 3.12E+2 1.79E+2 1.02E+5 5.35E+2 -3 8.23E+1 1-132 1.06E+1 2.85E+1 9.96E+O 9.96E+2 4.54E+1 - 5.35E-5 133 7.45E+1 1.30E+2 3.95E+1 .90E+4 2.26E+2 - 1.136E+2 11-134 5.56E+3 1.51E+1 5.40E+ 2.62E+2
-72.40E+1 1.32E-2
< 1-135 2.32E+1 6.08E+1 2.24E+1 4.201E+3 9.75E+1 - 6.87E+1 Cs-134 6.84E+3 1.63E+4 1.33E+4 - 5.27E+3 1.75E+3 2.85E+2 Cs-136 7.16E+2 2.83E+3 2.04E+3 - 1.57E+3 2.16E+2 3.21 E+2 Cs-137 8.77E+3 1.20E+4 7.85E+3 - 4.07E+3 1.35E+3 2.32E+2 Cs-138 6.07E+0 2'.20E+ 5.94E+ - 8.81E+0 8.70E-1 5.12E-5 S
Ba139 Ba-140 7.85E+
1.64E+3 5.59E-3 2.06E+O 2.30E-1 1.08E+2
- 5.23E-3 7.02E-1 3.17E-3 1.39E+1
- 1.18E+0 3.38E+3
> Ba-141 3.891E-2 2*88E-3 1.29E-1 - 2.68E-3 1.63E-3 1.80E-9 ry Ba-142 1.72E+0 1.77E-3 1.08E-1 1.50E-3 1.00E-3 2.43E-18 I
Z La-14 La-,142 1.57E+O 8.06E-2 7.94E-1
-3.67E-2 -2.10E-1 9.13E-3 --2.68E+2 .5.83E+4
- Ce-141 3.43E+O 2.32E6+ 2.63E-1 !.08E+0 8.86E+3 Ce-143 6.04E-1 4.46E+2 4.94E-2 - 1.97E- I 1.67E+4
>Ce-144 1.79E+2 7.47E+1 9.59E+0 4.43E+1 6.04E+4 ay 0 Pr-143 5.79E+0 2.32E+0 2.87E-1P .34E+0 2.o4E+4 "Pr-144 1.90E-2 7.87E-3 9.64E-4 4. 44E-3 -2.73E-9 I" Nd-147 3.96E+0 4.58E+0 .4- 2.68E+0 2.20E+4
_Li LU W-487 9.16E+O 7.66E+O 2.68E+0 2.5 1E+3 zNp-239 3.53E-2 3.47E-3 1.91E-3 1.08E-2 7.11 E+2 0"
ý,t
a Page 10f* of 157
Salem ODCM Rev. 23 Table. 1-3: Bioaccumulation Factors (pCi/kg per pCi/liter)* I H 9.0E-01 9.3E-01 n C 1.8E+03 1.4E+03 CP 3.OE+03 3.OE+043 Cr 4.OE+02 2.0E+03 Mn Fe Co 5.5E+02 3.OE+03 1.0E+02
.40E+02 2.OE+04 1.0E+03 I
Ni 1.0E-02 2.5E+02 CU 6.7E+02 1.7E+03 Zn 2.OE+03 5.OE+04 As 3.3E+02 133E+02 Br, 1.5E-02 3.1E+00 Rb 8.3E+00 1.7E+01 Sr 2.OE+00 2.OE+01 Y 2.5E+01 L.OE+03 Zr 2;0OE+02 8.0E+01
Nb, 3.0E+04 1.0E+02'n Mo IL0E+01 1.0E+01 _
OTc* L0E+01 5.0E+01I Ru. 3.OE+00 1.0E+03 -
Rh L.OE+01 2.OE+03 o Ag,: 3.3E+03 3.3E+03 Sb 4.0E+01 5.4E+00
-- Te 1*0E+01 1.0E+02 r
I (L,)
< CS" Ba 4.OE+01 1.0E+01 A2.5E+01
,.0E+02 H- U3DLa 2.5E+01 .:.1.0E+03*
<Ce I2.5E+01 L.E+01 O-3.OE+O1
':6.0E+02 c*Pr. 2.5E+01 .L1.E+03 -- :
z*Nd. 2.5E+01 -l.0E+03 ...
SW 3.0E+01 3*.0E+01
_ Np 1.OE+01 1OE+01
-
Values in this table are taken from Regulatory Guide 1.109 except for phosphorus (fish) which is Z adapted from NUREG/CR- 1336 and silver, arsenic and antimony which are taken from UCRL 50564,
-_ Rev. 1, October 1972.
0 0~
ii w1 ry Page 102 of.157 D* Pae12oI5
Salem ODCM Rev. 23 FIGURE 2-1: SALEM VENTILATION EXHAUST SYSTEMS AND EFFLUENT MONITOR INTERFACES S Simplified One Line TO ATMOSPHERE tj1 u--
0 0
i-,
U.-
w)
U-f w
ry w
0._
0O ci-0 U-2 Page 103 of 157
Salem ODCM Rev. 23 FIGuRE 2-2: GASEOUS RADIOACTIVE WASTE DISPOSAL SYSTEM S Simplified One Line S
Vi)
Z C_)
D z
V-V/)
CA 0
z U-w w
0 V-7-
0 a_
01 c.
w ryPage 104 of 157 w
Salem ODCM Rev. 23 Table 2-1: Dose Factors For Noble Gases Total Body Skin Gamma Air Beta Air Dose Factor Dose Factor Dose Factor Dose Factor Ki Li Mi Ni Radionuclide LtCi/m3) LtCi/m3) iCi/m3) j.Ci/m3)
Kr-83m 7.56E-02 - 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.O1E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11-E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 0 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03 z
U-0 0
ry ~Page105 of 157
I Salem ODCM Rev. 23 I
p Table 2-2.1: Parameters for Gaseous Alarm Setpoint Determinations - Unit 1 I
Parameter Actual Value Default Value Units Comments I
X/Q calculated 2.2E-06 sec/m3 USNRC Salem Safety Evaluation, Sup 3 I
VF (Plant Vent) as measured or fan curves 1.30E+05 ft3/min Plant Vent - normal I
operation (Cont Purge) 3.50E+04 Containment Purge I
AF coordinated 0.25 N/A Administrative allocation factor with HCGS to ensure combined releases do not I
exceed release rate limit for Ci measured N/A UCi/cm 3 site.
Taken from gamma spectral I
nuclide specific N/A mrem/yr per3 analysis of gaseous effluent Values from Table 2-1 I
_tCi/m z Li nuclide specific N/A mrem/yr per
__Ci/m 3 Values from Table 2-1 I 0z Mi nuclide specific N/A mrem/yr per -Values from Table 2-1 I
L9J Sensitivities as determined N/A cpm, per Monitor sensitivities are z
1-1-R41 1-R12A gtCi/m 3 or cpm per controlled under Public.
Service Blueprint Dbcument I
Of gCi/cc (PSBP) 315733 00 03 z"
Setpoint I-R41D calculated N/A cpm or Monitor setpoints are controlled under Public I
I-R12A ** tXCi/sec Service Blueprint Document (A
03 (PSBP) 315733 I LU
Automatic Isolation function is applicable in all MODES except MODE 6 I
Ld
-j crý me I
D0 0a (A)
LUI I
LU.
.A Page 106 of 157 I
Salem ODCM Rev. 23
.C I, Table 2-2.2: Parameters for Gaseous Alarm Setpoint Determinations - Unit 2 Parameter Actual Default Units Value Value Comments X/Q Calculated 2.2E-6 sec/m 3 USNRC Salem Safety Evaluation, Sup 3 VF. as measured or Plant Vent fan curves 1.30E+05 ft3/min Plant Vent - normal operation' Cont. Purge 3.50E+04 Containment Purge AF Coordinated 0.25 N/A Administrative allocation factor with HCGS to ensure combined releases do not exceed release rate for site.
I C1 Measured N/A gtCi/cm 3 Taken from gamma spectral analysis of gaseous effluent K1 Nuclide N/A mrem/yr 3per Values from Table 2-1 specific p.Ci/m In L1 Nuclide N/A mrem/yr per PCi/m 3 Values from Table 2-1 0 Mt specific Nuclide N/A mrem/yr per 3
Values from Table 2-1 specific rCi/m Sensitivities as N/A cpm per Monitor sensitivities are I° I
.2-R41 2-R12A determined PtCi/m 3 or cpm per controlled under Public Service Blueprint Document (PSBP) pCi/cc 315734 7,i V) 0 Setpoint Calculated N/A cpm or Monitor setpoints are controlled 2-R41D pCi/sec under Public Service Blueprint
(-) 2-R12A** Document (PSBP) 315734
i. **Automatic Isolation function is applicable in all MODES. except MODE 6 II..
Z:
LLJ 0
I" V)
C7-t.
0 nv Page 107 of 157
Salem ODCM Rev. 23 I
I b Table 2-3: Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations
I Atmospheric Dispersion I
ODCM Location Pathway(s) Controlling X/Q D/Q CONTROL Age Group (sec/m3) (L/m2) I 3.11.2.1 a site boundary noble gases N/A 2.2E-06 N/A (0.83 mile, N) direct exposure I 3.11.2. lb site boundary Inhalation and child 2.2E-06 N/A (0.83 mile, N) ground plane I 3.11.2.2 site boundary gamma-air N/A 2.2E-06 N/A (0,83 mile, N) beta-air I 3.11.2.3 residence/dairy** milk, ground infant 5.4E-08 2.1E-10 (4.9 miles, W) plane and inhalation I S
The identified controlling locations, pathways and atmospheric dispersion are from the Safety I LU Evaluation Report, Supplement No. 3 for the Salem Nuclear Generating Station, Unit 2 (NUREG-0517, December 1978).
Location and distance are determined from the performance of the annual land use census as I
required by ODCM CONTROL 3.12.2.
H-
z 0
Lii I 0
z" V-0 I
0 I LUj 0
I zM I 0
LU1 I 0:
LUJ Page 108 of 157 I
LU 0D
Salem ODCM Rev. 23
1. Table 2-4: Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - ADULT (mrem/yr per 4Ci/m3)
H-1-3 1.26E+3 1.26E+3 I 1.26E+3 1.26E+3 1.26E+3 1.26E+3 C-14 1.82E+4 3.41E+3 3.41E+3 3.41E+3 3.41E+3 3.41E+3 3.41E+3 P-32 1.32E+6 7.71E+4 - - 8.64E+4 5.01E+4 Cr-51 - 5.95E+1 2.28E+1 1.44E+4 3.32E+3 1.OOE+2 Mn-54 3.96E+4 - 9.84E+3 1.40E+6 7.74E+4 6.30E+3 Fe-55 2.46E+4 1.70E+4 - 7.21E+4 6.03E+3 3.94E+3 Fe-59 1.18E+4 2.78E+4 - 1.02E+6 ,1.88E+5 1.06E+4 Co-57 6.92E+2 - 3.70E+5 3.14E+4 6.71E+2 Co-58 .1.58E+3 - 9.28E+5 1.06E+5 2.07E+3 Co-60 1.1 5E+4 - - 5.97E+6 2.85E+5 1.48E+4 Ni-63. 4.32E+5 3.14E+4 - - 1.78E+5 1.34E+4 1.45E+4 Zn-65 3.24E+4 1.03E+5 - 6.90E+4 8.64E+5 5.34E+4 4.66E+4 Rb-86 - 1.35E+5 - - - 1.66E+4 5.90E+4 Sr-89 3.04E+5 - 1.40E+6 3.50E+5 8.72E+3 Sr-90 9.92E+7 - 9.60E+6 7.22E+5 6.10E+6 Y-91 4.62E+5 - - 1.70E+6 3.85E+5 1.24E+4 Zr-95 1.07E+5 3.44E+4 - 5.42E+4 1.77E+6 1.50E+5 2.33E+4
u. Nb-95 Ru-103 1.41E+4 1.53E+3 7.82E+3 -
7.74E+3 5.83E+3 5.05E+5 5.05E+5 1.04E+5 1.1OE+5 4.21E+3 6.58E+2 Ru-106 6.91E+4 - 1.34E+5 9.36E+6 9.12E+5 8.72E+3 Ag-110m 1.08E+4 1.OOE+4 - 1.97E+4 4.63E+6 3.02E+5 5.94E+3 Sb-124 3.12E+4 5.89E+2 7.55E+1 - 2.48E+6 4.06E+5 1.24E+4 Sb-125 5.34E+4 5.95E+2 5.40E+1 - 1.74E+6 1.01E+5 1.26E+4 Te-125m 3.42E+3 1.58E+3 1.05E+3 1.24E+4 3.14E+5 7.06E+4 4.67E+2 C-) Te-127m 1.26E+-4 5.77E+3 3:29E+3 4.58E+4 9.60E+5 1.50E+5 1.57E+3 Te-129m 9.76E+3 4.67E+3 :3.44E+3 3.66E+4 1.16E+6 .3.83E+5 1.58E+3 1-131 2.52E+4 3.58E+4 1.19E+7 6..13E+4 - 6.28E+3 2.05E+4 1-132 1.16E+3 3.26E+3 1.14E+5 5.18E+3 4.06E+2 1.16E+3 1-133 8.64E+3 1.48E+4 2.15E+6 2.58E+4 - . 8.88E+3 4.52E+3 1-134 6.44E+2 1.73E+3 2.98E+4 2.75E+3 - 1.O1E+O 6.15E+2 1-135 2.68E+3 6.98E+3 4.48E+5 1.1 IE+4 - 5.25E+3 2.57E+3 Cs-134 3.73E+5 8.48E+5 2.87E+5 9.76E+4 1.04E+4 7.28E+5 0 Cs-136 3.90E+4 1.46E+5 - 8.56E+4 1.20E+4 1.17E+4 1.1OE+5 Cs-137 4.78E+5 6.21E+5 2.22E+5 7.52E+4 8.40E+3 4.28E+5 Ba-140 3.90E+4 4.90E+1 1.67E+1 1.27E+6 2.18E+5 2.57E+3 Ce-141 1.99E+4 1+/-j Ce-144 3.43E+6 1.35E+4 1.43E+6
- 6.26E+3 8.48E+5 3.62E+5 7.78E+6 1.20E+5 8.16E+5 1.53E+3 1.84E+5
>I Pr-143 Nd-147 9.36E+3 5.27E+3 3175E+3 6.10E+3 2.16E+3 3.56E+3 2.81E+5 2:21E+5 2.00E+5 1.73E+5 4.64E+2 3.65E+2 Page 109 of 157
I Salem ODCM Rev. 23 i
Table 2-4 (cont'd)
S Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - TEENAGER I (mrerm/yr per [tCi/m3)
I H-3 1.27E+3 I 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 I C-14 2.60E+4 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 P-32 Cr-51 1.89E+6 1.1OE+5 7.50E+1 3.07E+1 2.1OE+4 9.28E+4 3.OOE+3 7.16E+4 1.35E+2 I
Mn-54 Fe-55 3.34E+4 5.11E+4 2.38E+4 1.27E+4 1.98E+6 1.24E+5 6.68E+4 6.39E+3 8.40E+3 5.54E+3 I
Fe-59 1.59E+4 3.70E+4 - 1.53E+6 1.78E+5 1.43E+4 Co-57 Co-58 6.92E+2 2.07E+3 5.86E+5 1.34E+6 3.14E+4 9.52E+4 9.20E+2 2.78E+3 I
Co-60 - 1.51E+4 - - 8.72E+6 2.59E+5 1.98E+4 Ni-63 Zn-65 5.80E+5 3.86E+4 4.34E+4 1.34E+5 8.64E+4 3.07E+5 1.24E+6 1.42E+4 4.66E+4 1.98E+4 6.24E+4 I
Rb-8'6 1.90E+5 - 1-.77E+4 8.40E+4 Sr-89 Sr-90 4.34E+5 1.08E+8
-2.42E+6 1.65E+7 3.71E+5 7.65E+5 1.25E+4 6.68E+6 I
Y-91 6.61E+5 Zr-95 Nb-95 1.46E+5 1.86E+4 4.58E+4 1.03E+4 6.74E+4 1.OOE+4 2.94E+6 2.69E+6 7.51E+5 4.09E+5 1.49E+5 9.68E+4 1.77E+4 3.15E+4 5.66E+3 I
0 I
iV Ru-103 2.1OE+3 - -7.43E+3 7.83E+5 1.09E+5 8.96E+2 Ru-106 9.84E+4 - 1.90E+5 1.61E+7 9.60E+5 1.24E+4 C) -Ag-110m 1.38E+4 1.31E+4 2.50E+4 6.75E+6 2.73E+5 7.99E+3 Sb-124 Sb-125 4.30E+4 7.38E+4 7.94E+2 8.08E+2 9.76E+1 7.04E+1 3.85E+6 2.74E-6 3.98E+5 9.92E+4 1.68E+4 1.72E+4 I Te-125m 4.88E+3 2.24E+3 1.40E+3 - 5.36E+5 7.50E+4 6.67E+2 zD Te-127nm Te-129rn 1.80E+4 1.39E+4 8.16E+3 6.58E+3 4.38E+3 4.58E+3 6.54E+4 5.19E+4 1.66E+6 1.98E+6 1.59E+5 4.05E+5 2.18E+3 2.25E+3 I
0 1-131 3.54E+4 4.91E+4 1.46E+7 8.40E+4 - 6.49E+3 2.64E+4 tLd Li 1-132 1-133 1.59E+3 1.22E+4 4.38E+3 2.05E+4 1.51E+5 2.92E+6 6.92E+3 3.59E+4 1.27E+3 1.03E+4 1.58E+3 6.22E+3 I
1-134 8.88E+2 2.32E+3 3.95E+4 3.66E+3 - 2.04E+1 8.40E+2 z
0 U--
1-135 Cs-134 3.70E+3 5.02E+5 9.44E+3 1.13E+6 6.21E+5 1.49E+4 3.75E+5 1.46E+5 6.95E+3 9.76E+3 3.49E+3 5.49E+5 I
Cs-1.36 5.15E+4 1.94E+5 - 1.1OE+5 1.78E+4 1.09E+4 1.37E+5 Ci 0
Cs-137 Ba-140 6.70E+5 5.47E+4 8.48E+5 6.70E+1 -
- 3.04E+5 2.28E+1 1.21E+5 2.03E+6 8.48E+3 2.29E+5 3.11E+5 3.52E+3 I
Ce-i.41 2.84E+4 It90E+4 8.88E+3 6.14E+5 1.26E+5 2.17E+3 Q:ý 0a.
cn:
0J C) Ce-144 Pr-143 4.89E+6 1.34E+4 2.02E+6 5.31E+3
-1.21E+6 3.09E+3 1.34E+7 4{83E+5 8.64E+5 2.14E+5 2.62E+5 6.62E+2 I
0..
L'i Nd-147 Li L.i 7.86E+3 8.56E+3 5.02E+3 3172E+5 1.82E+5 5.13E+2 I
0O Page 110 of 1,57 I
Salem ODCM Rev. 23 Table 2-4 (cont'd)
I1 Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - CHILD (mrem/yr per pCi/m3)
H-3 - 1.12E+3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 C-14 3.59E+4 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 P-32 Cr-51 2.60E+6 1.14E+5 8.55E+1 7 2.43E+1 1.70E+4 4.22E+4 1.08E+3 9.88E+4 1.54E+2 Mn-54 - 4.29E+4 - 1.OOE+4 1.58E+6 .2.29E+4 9.51E+3 Fe-55 4.74E+4 2.52E+4 - 1.11E+5 2.87E+3 7.77E-+3 Fe-59 2.07E+4 3.34E+4 -- 1.27E+6 7.07E+4 1.67E+4 Co-57 - 9.03E+2 - 5.07E+5 1.32E+4 1.07E+3 Co-58 1.77E+3 - 1.11E+6 3.44E+4 3.16E+3 Co-60 - 1.3 1E+4 -- 7.07E+6 9.62E+4 2.26E+4 Ni-63 8.21E+5 4.63E+4 - 2.75E+5 6.33E+3 2.80E+4 Zn-65 4.26E+4 1.13E+5 - 7.14E+4 9.95E+5 1.63E+4 7.03E+4 Rb-86 - 1.98E+5 - 7.99E+3 1.14E+5 Sr-89 5.99E+5 - - - 2.16E+6 1.67E+5 1.72E+4 Sr-90 1.01E+8 - - - 1.48E+7 3.43E+5 6.44E+6 Y-91 9.14E+5 - - 2.63E+6 1.84E+5 2.44E+4 Zr-95 1.90E+5 4.18E+4 - 5.96E+4 2.23E+6 6.11E+4 3.70E+4 I* Nb-95 2.35E+4 9.18E+3 - 8.62E+3 6.14E+5 3.70E+4 6.55E+3 Ru-103 2.79E+3 - - 7.03E+3 6.62E+5 4.48E+4 1.07E+3 Ru-106 1.36E+5 - - 1.84E+5 1.43E+7 4.29E+5 1.69E+4 Ag-1lOr 1.69E+4 1.14E+4 - 2.12E+4 5.48E+6 1.OOE+5 9.14E+3 ci-) Sb-124 5.74E+4 7.40E+2 1.26E+2 - 3.24E+6 1.64E+5 2.OOE+4 Sb-125 9.84E+4 7.59E+2 9.10E+l - 2.32E+6 4.03E+4 2.07E+4 "I" Te-125m 6.73E+3 2.33E+3 1.92E+3 - 4.77E+5 3.38E+4 9.14E+2 0
w Te-127rn 2.49E+4 8.55E+3 6.07E+3 6.36E+4 1.48E+6 7.14E+4 3.02E+3
<J Te-129m 1.92E+4 6.85E+3 6.33E+3 5.03E+4 1.76E+6 1.82E+5 3.04E+3 1-131 4.81E+4 4.81E+4 1.62E+7 7.88E+4 - 2.84E+3 2.73E+4 1-132 2.12E+3 4.07E+3 1.94E+5 6.25E+3 - 3.22E+3 1.88E+3 1-133 1.66E+4 2.03E+4 3.85E+6 3.38E+4 - 5.48E+3 7.70E+3 1-134 1.17E+3 2.16E+3 5.07E+4 3.30E+3 - 9.55E+2 9.95E+2 V-1-135 4.92E+3 8.73E+3 7.92E+5 1.34E+4 - 4.44E+3 4.14E+3 Cs-134 6.51E+5 1.01E+6 3.30E+5 1.21E+5 3.85E+3 2.25E+5 Mz" Cs-136 6.51E+4 1.71E+5 - 9.55E+4 1.45E+4 4.18E+3 1.16E+5 Cs-137 9.07E+5 8.25E+5 - 2.82E+5 1.04E+5 3.62E+3 1.28E+5 V) Ba-140 7.40E+4 6.48E+1 - 2.11E+1 1.74E+6 1.02E+5 4.33E+3 0
Ce-141 3.92E+4 1.95E+4 - 8.55E+3 5.44E+5 5.66E+4 2.90E+3 Ce-144 6.77E+6 2.12E+6 - 1.17E+6 1.20E+7 3.89E+5 3.61E+5 Pr-143 1.85E+4 5.55E+3 3.OOE+3 4.33E+5 9.73E+4 9.14E+2 Nd-147 1.08E+4 8.73E+3 4.81E+3 3.28E+5 8.21E+4 6.81E+2 n*
Page 111 of 157
I Salem ODCM Rev. 23 I
Table 2-4 (cont'd)
S Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - INFANT I (mrem/yr per pCi/m3)
I H-3 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 C-14 P-32 2.65E+4 2.03E+6 5.31E+3 1.12E+5 5.31E+3 5.31E+3 5.31E+3 5.31E+3 1.61E+4 5.31E+3 7.74E+4 I Cr-51 5.75E+1 1.32E+l 1.28E+4 3.57E+2 8.95E+l Mn-54 Fe-55 1.97E+4 2.53E+4 1.17E+4 4.98E+3 1.OOE+6 8.69E+4 7.06E+3 1.09E+3 4.98E+3 3.33E+3 I
Fe-59 1.36E+4 2.35E+4 - 1.02E+6 2.48E+4 9.48E+3 Co-57 Co-58 6.51E+2 1.22E+3 -
3.79E+5 7.77E+5 4.86E+3 1.11E+4 6.41E+2 1.82E+3 I
Co-60 8.02E+3 - 4.51E+6 3.19E+4 1.18E+4 Ni-63 Zn-65 3.39E+5 1.93E+4 2.04E+4 6.26E+4
- 3.25E+4 2.09E+5 6.47E+5 2.42E+3 5.14E+4 1.16E+4 3.11E+4 I
Rb-86 - 1.90E+5 - 3.04E+3 8.82E+4 Sr-89 Sr-90 3.98E+5 4.09E+7 2.03E+6 1.12E+7 6.40E+4 1.31E+5 1.14E+4 2.59E+6 I
Y-91 5.88E+5 - - 2.45E+6 Zr-95 Nb-95 1.15E+5 1.57E+4 2.79E+4 6.43E+3 3.1iE+4 4.72E+3 1.75E+6 4.79E+5 7.03E+4 2.17E+4 I 27E+4 1.57E+4 2.03E+4 3.78E+3 I
0 Ru-103 Ru-106 Ag-1 IOm 2.02E+3 8.68E+4
- 4.24E+3 1.07E+5 5.52E+5 1.16E+7 1.61E+4 1.64E+5 6.79E+2 1.09E+4 I C) 9.98E+3 7.22E+3 - 1.09E+4 3.67E+6 3.30E+4 5.OOE+3 Sb-124 Sb-125 3.79E+4 5.17E+4 5.56E+2 4.77E+2 1.01E+2 6.23E+1 2.65E+6 1.64E+6 5.91E+4 1.47E+4 1.20E+4 1.09E+4 I 0 Te-125m 4.76E+3 1.99E+3 1.62E+3 - 4.47E+5 1.29E+4 6.58E+2 I--
r-y Te-127m Te-129m 1.67E+4 1.41E+4 6.90E+3 6.09E+3 4.87E+3 5.47E+3 3.75E+4 3.18E+4 1.31E+6 1.68E+6 2.73E+4 6.90E+4 2.07E+3 2.23E+3 I
1-131 3.79E+4 4.44E+4 1.48E+7 5.18E+4 1.06E+3 1.96E+4 0.d 1-132 1-133 1.69E+3 1.32E+4 3.54E+3 1.92E+4 1.69E+5 3.56E+6 3.95E+5 2.24E+4 1.90E+3 2.61E+3 1.26E+3 5.60E+3 I
1-134 9.21E+2 1.88E+3 4.45E+4 2.09E+3 1.29E+3 6.65E+2 1-135 3.86E+3 7.60E+3 6.96E+5 847E+3 1.83E+3 2.77E+3 z
0 Cs-134 Cs-136 3.96E+5 4.83E+4 7.03E+5 1.35E+5 1.90E+5 5.64E+4 7.97E+4 1.18E+4 1.33E+3 1.43E+3 7.45E+4 5.29E+4 I
0 Cs-137 Ba-140 5.49E+5 5.60E+4 6.12E+5 5.60E+l 1.72E+5 1.34E+l 7.13E+4 1.60E+6 1.33E+3 3.84E+4 4.55E+4 2.90E+3 I
Ce-141 2.77E+4 1.67E+4 - 5.25E+3 5.17E+5 2.16E+4 1.99E+3
_.J Ce-144 Pr-143 3.19E+6 1.40E+4 1.21E+6 5.24E+3
- 5.38E+5 1.97E+3 9.84E+6 4.33E+5 1.48E+5 3.72E+4 1.76E+5 6.99E+2 I
D Nd-147 7.94E+3 8.13E+3 3.15E+3 3.22E+5 0
0l.
03 w.
- 3.12E+4 5.OOE+2 I
U-2 Page 112 of 157 0/ I
Salem ODCM Rev. 23 Table 2-4 (cont'd)
Pathway Dose Factors - Atmospheric Releases I, R(io), Grass7Cow-Milk Pathway Dose Factors - ADULT (mrem/yr per XtCi/m3) for H-3 and C-14 (m2
mrem/yr'per jtCi/sec) for others H-3 - 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 C-14 3.63E+5 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 P-32 1.71E+10 1.06E+9 - - - 1.92E+9 6.60E+8 Cr-51 - 11.71E+4 6.30E+3 3.80E+4 7.20E+6 2.86E+4 Mn-54 - 8.40E+6 - 2.50E+6 - 2.57E+7 1.60E+6 Fe-55 2.51E+7 1.73E+7 -I 9.67E+6 9.95E+6 4.04E+6 Fe-59 2.98E+7 7.OOE+7 - 1.95E+7 2.33E+8 2.68E+7 Co-57 - 1.28E+6 - 3.25E+7 2.13E+6 Co-58 .- 4.72E+6 - - - 9.57E+7 1.06E+7 Co-60 - 1.64E+7 - - - 3.08E+8 3.62E+7 Ni-63 6.73E+9 4.66E+8 - - - 9.73E+7 2.26E+8 Zn-65 1.37E+9 4.36E+9 -" 2.92E+9 - 2.75E+9 1.97E+9 Rb-86 - 2.59E+9 - - - 5.11E+8 1.21E+9 Sr-89 1.45E+9 .... 2.33E+8 4.16E+7 Sr-90 4.68E+10 - - - 1.35E+9 1.15E+10 Y-91 8.60E+3 - - I - - 4.73E+6 2.30E+2 I. Zr-95 Nb-95 Ru-103
- 9.46E+2 8.25E+4 1.02E+3 3.03E+2 4.59E+4 4.76E+2 4.54E+4 3.89E+3 9.62E+5 2.79E+8 1.19E+5 2.05E+2 2.47E+4 4.39E+2 Ru-106 2.04E+4 - - 3.94E+4 1.32E+6 2.58E+3 Cf) Ag-Ib0m 5.83E+7 5.39E+7 - 1.06E+8 2.20E+10 3.20E+7
.Sb-124 2.57E+7 4.86E+5 6.24E+4 - 2.OOE+7 7.31E+8 1.02E+7 Sb-125 2.04E+7 2.28E+5 2.08E+4 . 1.58E+7 2.25E+8 4.86E+6 C-)
Te-125m 1.63E+7 ý5.90E+6 4.90E+6 6.63E+7 6.50E+7 2.18E+6 Te-127m 4.58E+7 1.64E+7 1.17E+7 1.86E+8 1.54E+8 5.58E+6 Te-129m 6.04E+7 2.25E+7 2.08E+7 2.52E+8 3.04E+8 9,57E+6 1-131 2.96E+8 4.24E+8 1.39E+11 7.27E+8 1.12E+8 2.43E+8 1-132 1.64E-1 4.37E-1 1.53E+1 6.97E-1 8.22E-2 1.53E-1 1-133 3.97E+6 6.90E+6 1.01E+9 1.20E+7 6.20E+6 2.1OE+6 1-134- - - - -
1-135 1j39E+4 3.63E+4 2.40E+6 5.83E+4 4.1OE+4 1.34E+4 0 Cs-134 5.65E+9 1.34E+10 - 4.35E+9 1.44E+9 2.35E+8 1.10E+10 Cs-136 2.6,1E+8 1.03E+9 5.74E+8 7.87E+7 1.17E+8 7.42E+8 Cs-137. 7.38E+9 1.O1E+10 - 3.43E+9 1.14E+9 1.95E+8 6.61E+9 Ba-140 1269E+7 3.38E+4 -
l.15E+4 1.93E+4 5.54E+7 1.76E+6 Ce-141 4.84E+3 3.27E+3 - 1.52E+3 1.25E+7 3.71E+2 Ce-144 3.58E+5 l.50E+5 - 8.87E+4 1.21E+8 1.92E+4 Pr-143 1.59E+2 6.37E+l - 3.68E+1 6.96E+5 7.88E+O Nd-147 9.42E+1 1.09E+2 6.37E+1 5.23E+5 6.52E+O Page 113 of 157
Salem ODCM Rev. 23 i
Table 2-4 (cont'd) i p Pathway Dose Factors - Atmospheric Releases R(io), Grass-Cow-Milk Pathway Dose Factors - TEENAGER (mrem/yr per ptCi/m3) for H-3 and C- 14 (m2
mrem/yr per jiCi/sec) for others I
H-3 C-14 6.70E+5 9.94E+2 134E+5 9.94E+2 1.34E+5 9.94E+2 134E+5 I 9.94E+2 1.34E+5 9.94E+2 1.34E+5 9.94E+2 1.34E+5 I
P-32 3.15E+10 1.95E+9 - - - 2.65E+9 1.22E+9 Cr-51 Mn-54 1.40E+7 2.78E+4 1.1OE+4 4.17E+6 7.13E+4 8.40E+6 2.87E+7 5.OOE+4 2.78E+6 I
Fe-55 4.45E+7 3.16E+7 - 2.OOE+7 1.37E+7 7.36E+6 Fe-59 Co-57 5.20E+7 1.21E+8 2.25E+6 3.82E+7 2.87E+8 4.19E+7 4.68E+7 3.76E+6 I
Co-58 7.95E+6 - 1.1OE+8 1.83E+7 Co-60 Ni-63 1.18E+10 2.78E+7 8.35E+8 -
3.62E+8 1.33E+8 6.26E+7 4.01E+8 I
Zn-65 2.11E+9 7.31E+9 4.68E+9 - 3.1OE+9 3.41E+9 Rb-86 Sr-89 2.67E+9 4.73E+9 - - 7.OOE+8 3.18E+8 2.22E+9 7.66E+7 I
Sr-90 9.92E+7 9.60E+6 7.22E+5 6.1OE+6 Y-91 Zr-95 1.58E+4 1.65E+3 5.22E+2 7.67E+2 -
6.48E+6 1.20E+6 4.24E+2 3.59E+2 I
0 Nb-95 1.41E+5 7.80E+4 7.57E+4 - 3.34E8 4.30E+4 z Ru-103 Ru-106 1.81E+3 3.75E+4 -
6.40E+3 7.23E+4 1.52E+5
.1.80E+6 7.75E+2 4.73E+3 I
z Ag-'llnm 9.63E+7 9.11E+7 1.74E+8 2.56E+1O 5.54E+7 Sb-124 Sb-125 4.59E+7 3.65E+7 8.46E+5 3.99E+5 1.04E+5 3.49E+4
- 4.01E+7 3.21E+7 9.25E+8
-2.84E+8 1.79E+7 8.54E+6 I
Te-125m 3.00E+7 1.08E+7 8.39E+6 - 8.86E+7 4.02E+6 C/) Te,127m Te 129m 8.44E+7 1.11E+8 2.99E+7 4.1OE+7 2.01E+7 3.57E+7 3.42E+8 4.62E+8 -
2.1OE+8 4.15E+8 1.OOE+7 1.75E+7 I
Li 1-131 5.38E+8 7.53E+8 2.20E+11 I 1.30E+9 - 1.49E+8 4.04E+8 1-132 1-133 2.90E-1 7.24E+6 7.59E-1 1.23E+7 2.56E+1 1.72E+9 1.20E+O 2.15E+7 3.31E-i 9.30E+6 2.72E&1 3.75E+6 I
U-0 1-134 - - -
C/)
J-135 Cs434 2.47E+4 9.81E+9 6.35E+4.
2.31E+10 4.08E+6 1.OOE+5 7.34E+9 2.80E+9 7.03E+4 2.87E+8 2.35E+4 1.07E+10 I
Cs-136 4.45E+8 1.75E+9 - 9.53E+8 1.50E+8 1.41E+8 1.18E+9 LU_
c-J Cs-137 Ba-140 1.34E+10 4.85E+7 1.78E+l0 5.95E+4 6.06E+9 2.02E+4 2.35E+9 4.OOE+4 2,53E+8 7.49E+7 6.20E+9 3.13E+6 I
0 rv" Ce141 8.87E+3 1.35E+4 - 2.79E+3 -_ 1.69E+7 6.81E+2 0l (n.
Lu 0-v 0
LU Ce-144 Pr-143 6.58E+5 2.92E+2 2.72E+5 1.17E+2 1.63E+5 6.77E+ 1
- 1.66E+8 9.61E+5 3.54E+4 1.45E+1 I
Nd-147 1.81E+2 1.97E+2 - 1.16E+2 - 7.11E+5 1.18E+1 I
0_
Page 114 of 157 I
Salem ODCM Rev. 23 Table 2-4 (cont'd)
Pathway Dose Factors - Atmospheric Releases I* R(io), Grass-Cow-Milk Pathway Dose Factors - CHILD (mrem/yr per pCi/m3) for H-3 and C-14 (m2
mrem/yr per ptCi/sec) for others US .S _yS.d S _* S .5 H-3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 C-14 1.65E+6 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 P-32 7.77E+10 3.64E+9 - 2.15E+9 3.0OE+9 Cr-51 - 5.66E+4 1.55E+4 1.03E+5 5.41E+6 1.02E+5 Mn-54 2.09E+7 - 5.87E+6 - 1.76E+7 5.58E+6 Fe-55 1.12E+8 5.93E+7 - 3.35E+7 1.10E+7 1.84E+7 Fe-59, 1.20E+8 1.95E+8 - 5.65E+7 2.03E+8 9.71E+7 Co-57 3.84E+6 -3.14E+7 7.77E+6 Co-58 1.21E+7 - 7.08E+7 3.72E+7 Co-60 - 4.32E+7 - - 2.39E+8 1.27E+8 Ni-63 2.96E+10 1.59E+9 - 1.07E+8 1.01E+9 Zn-65 4.13E+9 1.1OE+10 - 6.94E+9 - 1.93E+9 6.85E+9 Rb-86 - 8.77E+9 - 5.64E+8 5.39E+9 Sr-89 6.62E+9 -... 2.56E+8 1.89E+8 Sr-90 1.12E+1l - 1.51E+9 2.83E+10 Y-91 3.91E+4 - - 5.21E+6 1.04E+3 I. Zr-95 Nb-95 3.84E+3 3.18E+5 8.45E+2 1.24E+5 1.21E+3 1.16E+5 8.81E+5 2.29E+8 7.52E+2 8.84E+4 Ru-103 4.29E+3 - 1.08E+4 - 1.11E+5 1.65E+3 Ru-106 9.24E+4 - 1.25E+5 - 1.44E+6 1.15E+4 Ag-ll0m 2.09E+8 1.41E+8 - 2.63E+8 1.68E+10 1.13E+8 Sb-124 1.09E+8 1.41E+8 2.40E+5 - 6.03E+7 6.79E+8 3.81E+7
<* Sb-125 8.70E+7 1.41E+6 8.06E+4 - 4.85E+7 2.08E+8 1.82E+7 C-) Te-125m 7.38E+7 2.OOE+7 2.07E+7 - 7.12E+7 9.84E+6 Te-127m 2.08E+8 5.60E+7 .4.97E+7 5.93E+8 - 1.68E+8 2.47E+7 Te-129m 2.72E+8 7.61E+7 8.78E+7 8.OOE+8 - 3.32E+8 4.23E+7 1-131 1.30E+9 1.31E+9 4.34E+11 2.15E+9 - 1.17E+8 7.46E+8 iv,) 1-132 6.86E-1 1.26E+0 5.85E+1 1.93E+0 - 1.48E+O 5.80E-1 I1> 1-133 1.76E+7 2.18E+7 4.04E+9 3.63E+7 - 8.77E+6 8.23E+6 1-134 .-..
V,-
1-135 5.84E+4 1.05E+5 9.30E+6 1.61E+5 - 8.E+4 4.97E+4 z
Cs-134 2.26E+10 3.71E+10 - 1.15E+10 4.13E+9 2.OOE+8 7.83E+9 Cs-136 1.OOE+9 2.76E+9 - 1.47E+9 2.19E+8 9.70E+7 1.79E+9 I, Cs-137 3.22E+10 3.09E+10 - 1.01E+10 3.62E+9 1.93E+8 4.55E+9 Ba440 1.17E+8 1.03E+5 - 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ce-141 2.19E+4 1.09E+4 - 4.78E+3 1.36E+7 1.62E+3 Ce-144 1.62E+6 5.09E+5 - 2.82E+5 1.33E+8 8.66E+4 0 Pr-143 7.23E+2 2.17E+2 - 1.17E+2 7.80E+5 3.59E+l Nd-147 4.45E+2 3.60E+2 - 1.98E+2 5.71E+5 2.79E+1 clY Page 115 of 157
I Salem ODCM Rev. 23 I
Table 2-4 (cont'd) a Pathway Dose Factors - Atmospheric Releases R(io), Grass-Cow-Milk Pathway Dose Factors - INFANT I (mrem/yr per ptCi/m3) for H-3 and C-14 (m2
mirenmyr per i.Ci/sec) for others I
H-3 C-14 3.23E+6 2.3 8E+3 6.89E+5 2.38E+3 6.89E+5 2.38E+3 6.89E+5 2.38E+3 6.89E+5 2.38E+3 6.89E+5 2.38E+3 6.89E+5 I P-32 1.60E+1 1 9.42E+9 - - 2.17E+9 6.21E+9 Cr-51 Mn-54 3.89E+7 1.05E+5 2.30E+4 8.63E+6 2.05E+5 4.71E+6 1.43E+7 1.61E+5 8.83E+6 I Fe-55 1.35E+8 8.72E+7 - 4.27E+7 1.11E+7 2.33E+7 Fe-59 Co-57 2.25E+8 3.93E+8 8.95E+6 1.16E+8 1.88E+8 3.05E+7 1.55E+8 1.46E+7 I Co-58 2.43E+7 - 6.05E+7 6.06E+7 Co-60 Ni-63 3.49E+/-10 8.81E+7 2.16E+9 2.1OE+8 1.07Et8 2.08E+8 1.21E+9 I Zn-65 5.55E+9 1.90E+10 - 9.23E+9 - 1.61E+10 8.78E+9 Rb-86 Sr-89 1.26E+10 2.22E+10 5.69E+8 2.59E+8 1.10E+1IO 3.61E+8 I
Sr-90 1.22E+11 - . - I 1.52E+9 3.10E+10 S
Y-91 Zr-95 7.33E+4 6.83E+3 1.66E+3 -
-6 1.79E+3 -
- 5.26E+6 8.28E+5 1.95E+3 1.18E+3 I
Nb-95 5.93E+5 2.44E+5 - 1.75E+5 - 2.06E+8 1.41E+5 I
Ln z Ru-103 8.69E+3 - - T1.81E+4 1 - 1.06E+5 2.91E+3 Ru-106 1.90E+5 - - 2.25E+5 - 1.44E+6 2.38E+4 Ag-110m 3.86E+8 2.82E+8 - 4.03E+8 - 1.46E+10 1.86E+8 0
Sb-124 Sb-125 2.09E+8 1.49E+8 3.08E+6 1.45E+6 5.56E+5 1.87E+5 1.31E+8 9.38E+7 6.46E+8 1.99E+8 6.49E+7 3.07E+7 I
0 Te-125m 1.51E+8 5.04E+7 5.07E+7 7.18E+7 2.04E+7 Lii Te-127m Te-129m 4.21E+8 5.59E+8 1.40E+8 1.92E+8 1.22E+8 2.15E+8 1.04E+9 1.40E+9 1.70E+8 3.34E+8 5.1OE+7 8.62E+7 I
1-131 2.72E+9 3.21E+9 1.05E+12 3.75E+9 .-1.15E+8 1.41E+9 z
-.y 1-132 1-133 1.42E+-O 3.72E+7 2.89E+O 5.41E+7 1.35E+2 9.84E+9 3.22E+O 6.36E+7 -
- 2.34E+O 9.16E+6 l+03E+O 1.58E+7 I
LU 1-134 - 1.01E-9 - - -
1-135 1.21E+5 2.41E+5 2.16E+7 2.69E+5 - 8.74E+4 8.80E+4 Cs-134 Cs-136 3.65E+10 1.96E+9 6.80E+10 5.77E+9 1.75E+10 2.30E+9 7.18E+9 4.70E+8 1.85E+8 8.76E+7 6;87E+9 2.15E+9 I
0*
A Cs-137 Ba-140 5.15E+10 2.41E+8 6.02E+10 2.41E+5 1.62E+10 5.73E+4 6.55E+9 1.48E+5 1.88E+8 5.92E+7 4.27E+9 1.24E+7 I
ry" Ce-141 4.33E+4 2.64E+4 - 8.15E+3 - 1.37E+7 3.11E+3 iI L) 0 a-LUt Ce-144 Pr-143 2.33E+6 1.49E+3 9.52E+5 5.59E+2 3.85E+5 2.08E+2 -
1-.33E+8 7.89E+5 1.30E+5 7.41E+1 I
0.* Nd-147 8.82E+2 9.06E+2 - 3.49E+2 - 5.74E+5 5.55E+1 LU 0l I
0 D/
Page 116 of 157 I
Salem ODCM Rev. 23 Table 2-4 (cont'd) 1I9 Pathway Dose Factors - Atmospheric Releases R(io), Vegetation Pathway Dose Factors - ADULT (mrem/yr per p.Ci/m3) for H-3 and C-14 (m2
mrem/yr per ptCi/sec) for others
-, Iý,~ -- u ,--i; ý . -.
H-3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 C-14 8.97E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 P-32 1.40E+9 8.73E+7 1.58E+8 5.42E+7 Cr-51 2.79E+4 1.03E+4 6.19E+4 - 1.17E+7 4.66E+4 Mn-54 - 3.11E+8 9.27E+7 9.54E+8 5.94E+7 Fe-55 2.09E+8 1.45E+8 8.06E+7 8.29E+7 3.37E+7 Fe-59 1.27E+8 2.99E+8 8.35E+7 9.96E+8 1.14E+8 Co-57 1.17E+7 2.97E+8 1.95E+7 Co-58 -_ 3.09E+7 - 6.26E+8 6.92E+7 Co-60 - 1.67E+8 - 3.14E+9 3.69E+8 Ni-63 1.04E+10 7.21E+8 -- - 1.50E+8 3.49E+8.
Zn-65 3.17E+8 1.01E+9 - 6.75E+8 - 6.36E+8 4.56E+8 Rb-86 2.19E+8 - - 4.32E+7 1.02E+8 Sr-89 9.96E+9 - - 1.60E+9 2.86E+8 Sr-90. 6.05E+11 - - - 1.75E+10 1.48E+10 Y-91 5.13E+6 .... 2.82E+9 1.37E+5 Zr-95 1.19E+6 3.81E+5 - 5.97E+5 - 1.21E+9 2.58E+5 Nb-95 1.42E+5 7.91E+4 - 7.81E+4 - 4.80E+8 4.25E+4 Ru-103 4.80E+6 - - 1.83E+7 - 5.61E+8 2.07E+6 Ru-106 1.93E+8 - - 3.72E+8 - 1.25E+10 2.44E+7 Ag-110m 1.06E+7 9.76E+6 - 1.92E+7 - 3.98E+9 5.80E+6 Sb-124 1.04E+8 1.96E+6 2.52E+5 8.08E+7 2.95E+9 4.11E+7 (1)
Sb-125 1.36E+8 1.52E+6 1.39E+5 - 1.05E+8 1.50E+9 3.25E+7 Te-125m 9.66E+7 3.50E+7 2.90E+7 3.93E+8 - 3.86E+8 1.29E+7 Te-127m 3.49E+8 1.25E+8 8.92E+7 1.42E+9 - 1.17E+9 4.26E+7 Te-129m 2.55E+8 9.50E+7 8.75E+7 1.06E+9 - 1:28E+9 4.03E+7 1-131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 - 3.05E+7 6.63E+7 1-132 5.74E+1 1.54E+2 5.38E+3 2.45E+2 , 2.89E+1 5.38E+1 1-133 2.12E+6 3.69E+6 5.42E+8 6.44E+6 - 3.31E+6 1.12E+6 1-134 1.06E-4 2.88E-4 5.OOE-3 4.59E-4 - 2.51E-7 1.03E-4 EM 0
1-135 4.08E+4 1.07E+5 7.04E+6 1.71E+5 - 1.21E+5 3.94E+4 Cs-134 4.66E+9 1.11E+10 - 3.59E+9 1.19E+9 1.94E+8 9.07E+9 Cs-136 4.20E+7 1.66E+8 - 9.24E+7 1.27E+7 1.89E+7 1.19E+8 LJ Cs-137 6.36E+9 8.70E+9 - 2.95E+9 9.81E+8 1.68E+8 5.70E+9 Ba-140 1.29E+8 1.62E+5 - 5.49E+4 9.25E+4 2.65E+8 8.43E+6 Ce-141 1.96E+5 1.33Ef5 - 6.17E+4 - 5.08E+8 1.51E+4 Ce-144 3.29E+7 1.38E+7 - 8.16E+6 - 1.11E+10 1.77E+6 0 Pr-143 6.34E+4 2.54E+4 - 1.47E+4 - 2.78E+8 3.14E+3 Nd-147 3.34E+4 3.86E+4 - 2.25E+4 - 1.85E+8 2.31E+3 Q:f Page 117 of 157
Salem ODCM Rev. 23 I
Table 2-4 (cont'd) I I Pathway Dose Factors - Atmospheric Releases R(io), Vegetation Pathway Dose Factors - TEENAGER (mren/yr per ýtCi/m3) for H-3 and C-14 (m2
mrem/yr per XCi/sec) for others I
H-3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 I
C-14 1.45E+6 2.91E+5 2.91E+5 2.91E+5 2.91E+5 2.91E+5 2.91E+5 P-32 Cr-51 1.61E+9 9.96E+7 3.44E+4 1.36E+4 8.85E+4 1.35E+8 1.04E+7 6.23E+7 6.20E+4 I
Mn-54 4.52E+8 - 1.35E+8 - 9.27E+8 8.97E+7 Fe-55 3.25E+8 2.31E+8 - 1.46E+8 9.98E+7 5.38E+7 Fe-59 Co-57 1.81E+8 4.22E+8 1.79E+7 1.33E+8 9.98E+8 3.34E+8 1.63E+8 3.00E+7 I
Co-58 4.38E+7 -- - 6.04E+8 1.01E+8 Co-60 2.49E+8 - - 3.24E+9 5.60E+8 Ni-63 1.61E+10 1.13E+9 - - 1.81E+8 5.45E+8 Zn-65 4.24E+8 1.47E+9 - 9.41E+8 - 6.23E+8 6.86E+8 Rb-86 Sr-89 1.51E+10 2.73E+8 -
4.05E+7 1.80E+9 1.28E+8 4.33E+8 I
Sr-90 7.51E+11 - - 2.I1E+10 1.85E+11 Y-91 Zr-95 7.87E+6 1.74E+6 5.49E+5
- 8.07E+5 3.23E+9 1.27E+9 2.11E+5 3.78E+5 I
Nb-95 1.92E+5 1.06E+5 - 1.03E+5 - 4.55E+8 5.86E+4 Ru-103 Ru-106 6.87E+6 3.09E+8 2.42E+7 5.97E+8 5.74E+8 1.48E+10 2.94E+6 3.90E+7 i
Ag-110m 1.52E+7 1.44E+7 - 2.74E+7 - 4.04E+9 8.74E+6 0
Sb-124 Sb-125 1.55E+8 2.14E+8 2.85E+6 2.34E+6 3.51E+5 2.04E+5 1.35E+8 1.88E+8 3.11E+9 1.66E+9 6.03E+7 5.OOE+7 I
Te-125m 1.48E+8 5.34E+7 4.14E+7 - 4.37E+8 1.98E+7 Ir Te-127m Te-129m 5.51E+8 3.67E+8 1.96E+8 1.36E+8 1.31E+8 1.18E+8 2.24E+9 1.54E+9 1.37E+9 1.38E+9 6.56E+7 5.81E+7 I
1-131 7.70E+7 1.08E+8 3.14E+10 1.85E+8 - 2.13E+7 5.79E+7 Mz 1-132 1-133 5.18E+1 1.97E+6 1.36E+2 3.34E+6 4.57E+3 4.66E+8 2.14E+2 5.86E+6 5.91E+1 2.53E+6 4.87E+1 1.02E+6 I
1-134 9.59E-5 2.54E-4 4.24E-3 4.01E-4 3.35E-6 9.13E-5 1-135 Cs-134 3.68E+4 7,09E+9 9.48E+4 1.67E+10 6.1OE+6 1.50E+5 5.30E+9 2.02E+9 1.05E+5 2.08E+8 3.52E+4 7.74E+9 I
Cs-136 4.29E+7 1.69E+8 - 9.19E+7 1.45E+7 1.36E+7 1.13E+8 Cs-137 L.O1E+l0 1.35E+10 - 4.59E+9 1.78E+9 1.92E+8 4.69E+9 Ba-140 Ce-141 1.38E+8 2.82E+5 1.69E+5 1.88E+5 5.75E+4 8.86E+4 1.14E+5 2.13E+8 5.38E+8 8.91E+6 2.16E+4 I
Ce-144 Pr-143 5.27E+7 7.12E+4 2.18E+7 2.84E+4 1.30E+7 1.65E+4 1.33E+10 2.34E+8 2.83E+6 3.55E+3 I
I Nd-147 3.63E+4 3.94E+4 - 2.32E+4 1.42E+8 2.36E+3 I
Page 118 of 157 I
Salem ODCM Rev. 23
-Table 2-4 (cont'd)
II Pathway Dose Factors - Atmospheric Releases R(io), Vegetation Pathway Dose Factors - CHILD (mremlyr per LCi/m3) for H-3 and C-14 (m2
mrem/yr per jLCi/sec) for others H-3 4.01E+3 4.01E+3 4.01E+3 4.01E+3 L4.01E+3 4.01E+3 C-14 3.50E+6 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 P-32 3.37E+9 1.58E+8 - - - [9.30E+7 1.30E+8 Cr-51 - 6.54E+4 1.79E+4 1.19E+5 6.25E+6 1.18E+5 Mn-54 - 6.61E+8 - 1.85E+8 - I 5.55E+8 1.76E+8 Fe-55 8.OOE+8 4.24E+8 - 2.40E+8 17.86E+7 1.31E+8 Fe-59 4.01E+8 6.49E+8 - 1.88E+8 6.76E+8 3.23E+8 Co-57 2.99E+7 - . 2.45E+8 6.04E+7 Co-58 6.47E+7 - 3.77E+8 1.98E+8 Co-60 - 3.78E+8 - -- 2.1OE+9 1.12E+9 Ni-63 3.95E+10 2.11E+9 .... i1.42E+8 1.34E+9 Zn-65 8.12E+8 2.16E+9 1.36E+9 - 3.80E+8 1.35E+9 Rb-86 - 4.52E+8 .... 2.91E+7 2.78E+8 Sr-89 3.59E+10 - 1.39E+9 1.03E+9 Sr-90 1.24E+12 -- 1.67E+10 3.15E+11 Y-91 1.87E+7 - - - 2.49E+9 5.01E+5 Zr-95 3.90E+6 8.58E+5 1.23E+6 - 8.95E+8 7.64E+5 I*)
Nb-95 4.1OE+5 1.59E+5 1.50E+5 2.95E+8 1.14E+5 Ru-103 1.55E+7 3.89E+7 3.99E+8 I 5.94E+6 Ru-106 7.45E+8 1.01E+9 1.16E+10 f9.30E+7 Ag-110m 3.22E+7 2.17E+7 4.05E+7 2.58E+9 I 1.74E+7 Sb-124 3.52E+8 4.57E+6 7.78E+5 - 1.96E+8 2.20E+9 1.23E+8 Sb-125 4.99E+8 3.85E+6 4.62E+5 - 2.78E+8 1.19E+9 1.05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 - 3.38E+8 4.67E+7 Te-127m 1.32E+9 3.56E+8 3.16E+8 3.77E+9 - 1.07E+9 1.57E+8 U- Te-129m 8.54E+8 2.39E+8 2.75E+8 2.51E+9 - 1.04E+9 1.33E+8 1-131 1.43E+8 1.44E+8 4.76E+10 2.36E+8 - 1.28E-7 &8.18E+7 1-132 9.20E+1 1.69E+2 7.84E+3 2.59E+2 - 1.99E+2 [7.77E+1 1-133 3.59E+6 4.44E+6 8.25E+8 7.40E+6 - 1.79E+6 1.68E+6 1-134 1.70E-4 3.16E-4 7.28E-3 4.84E-4 - 2.!OE-4 1.46E-4 1-135 6.54E+4 1.18E+5 1.04E+7 1.81E+5 - 8.98E+4 5.57E+4 0 Cs-134 1.60E+10 2.63E+10 8.14E+9 2.92E+9 1.42E+8 [5.54E+9 Cs-136 8.06E+7 2.22E+8 1..18E+8 1.76E+7 7.79E+6 '1.43E+8 Cs-137 2.39E+10 2.29E+10 7.46E+9 2.68E+9 1.43E+/-+8 3.38E+9 Ba-140 2.77E+8 2.43E+5 7.90E+4 1.45E+5 1.40E+8 1.62E+7 Ce-141 6.35E+5 3.26E+5 4.07E+8 4.84E+4
('1
-1.43E+5 -
Ce-144 1.27E+8 3.98E+7 .2.21E+7 - 1.04E+10 6.78E+6 Pr-143 1.48E+5 4.46E+4 2.41E+4 - 1.60E+8 7.37E+3 Nd-147 7.16E+4 5.80E+4 3.18E+4 - 9.18E+7 4.49E+3 Page 119 of 157
I Salem ODCM Rev. 23 I
Table 2-4 (cont'd)
S Pathway Dose Factors - Atmospheric Releases R(io), Ground Plane Pathway Dose Factors I
(m2
mrem/yr per .Ci/sec) ,1 Nuclide Any Organ I
H-3 C-14 P-32 Cr-51 4.68E+6 I
Mn-54 Fe-55 1.34E+9 I Fe-59 2.75E+8 Co-58 Co-60 3.82E+8 2.16E+10 I
Ni-63 Zn-65 Rb-86 7.45E+8 8.98E+6 I
Sr-89 Sr-90 Y-91 2.16E+4 1.08E+6 I
Zr-95 2.48E+8 Nb-95 1.36E+8 I
Ru- 103 1.09E+8 S
Ru-106 Ag-il0m 4.21E+8 3.47E+9 I 1.55E+6 0
Te-125m Te-127m Te-129m 9.17E+4 2.OOE+7 I
1-131 1.72E+7 Lu CD 1-132 1-133 1.24E+6 2.47E+6 I
1-134 4.49E+5 1-135 2.56E+6 I Cs-134 6.75E+9 z
0 Cs-136 Cs-137 Ba-140 1.49E+8 1.04E+10 2.05E+7 I
Ce- 141 Ce-144 1.36E+7 6.95E+7 I
Pr-143 CD Nd-147 8.40E+6 I
I Page 120 of 157 I
Salem ODCM Rev. 23 APPENDIX A Evaluation of Default Parameters for Liquid Effluents Lj S
V.)
(2 T
0 C-)
z" 0
o3 (A1 0
i, wL 0y Q:
La_
U-V~p w) nA V)
Page 121 of 157
I Salem ODCM Rev. 23 I
APPENDIX A: Evaluation of Default Parameters for Liquid Effluents A. Effective Maximum Permissible Concentration (MPCe) I In accordance with the requirements of ODCM CONTROL 3.3.3.8 the radioactive liquid effluent monitors shall be operable with alarm setpoints established to ensure that the concentration of radioactive material at the discharge point does not exceed the MPC value of 10 CFR 20, Appendix B, Table II, Column 2 (Appendix F). The determination of allowable radionuclide concentration and corresponding alarm setpoint is a function of the individual radionuclide distribution and corresponding MPC values.
In order to limit the need for routinely having to reestablish the alarm setpoints as a function of changing radionuclide distributions, a default alarm setpoint can be established. This default setpoint can be based on an evaluation of the radionuclide distribution of the liquid effluents from Salem and the effective MPC value for this distribution.
The effective MPC value for a radionuclide distribution iscalculated by the. equation:
MPe _,CI (gamma) mpi where: I z
0 MPCe = an effective MPC value for a mixture of gamma emitting radionuclides (ptCi/ml)
Ci = concentration of radionuclide i in the mixture o MPCi = the 10 CFR 20, Appendix B, Table II, Column 2 MPC value for radionuclide i (Appendix F) (itCi/ml) o The equation for determining the liquid effluent setpoints (Section 1.2.1, equation 1.2) is based on a multiplication of the effective MPC times the monitor sensitivity. Considering the average CO effective MPC value for the years 1993, 1994, and 1998, it is reasonable to select an MPCe value of 6.05E-06 pCi/ml for Unit 1 and 4.8 1E-06 jCi/ml for Unit 2 as typical of liquid radwaste 0 discharges.
° I W
I-t-
~I 0
©~I n
w0 I a:
V) " Page 122 of 157 I
Salem ODCM Rev. 23 B. Correction Factor I. The type of radiation detector used to monitor radioactive releases is not capable of detecting non-gamma emitting radionuclides such as H-3, Fe-55, and Sr-89, 90, as required by ODCM CONTROL 3.11.1.1. A conservative default safety factor can be determined to account for non-gamma emitting radionuclides. Non-gamma emitting radionuclides are analyzed at Salem station on a monthly basis from a composite sample of liquid releases.
Nuclide MPC (0Ci/ml) Activity (gCi/ml) Activity / MPC H-3 3E-3 5.2E-1 173.3 Fe-55 8E-4 2.5E-3 3.1 Sr-89 3E-6 2.OE-5 6.7 Sr-90 3E-7 7.2E-7 2.4 Total 185.5 The values in the table above represent the maximum reactor coolant values for non-gamma emitting nuclides in 1994 for Unit 1 and 2. Reactor coolant values were chosen to represent the maximum concentration of non-gamma emitting nuclides that could be released from Salem Station. The activity values in the table will be diluted by a minimum factor of 800 prior to release. The minimum dilution factor is obtained by using the minimum circulating water flowrate of 100,000 gpm and the maximum release rate of 120 gpm.
A conservative non-gamma factor for non-gamma emitting nuclides can be obtained using the highest Activity/MPC fraction and the minimum dilution factor as follows:
C-)
Non-Gamma Factor = 185.5 / 800 0.23 (Rounded up to 0.25)
U-)
Correction Factor = 1-0.25 0.75 C. Default setpoint determination:
Using the information and parameters described above a default setpoint can be calculated for Unit 1 and 2 liquid radwaste disposal process radiation monitors (RI 8).
Using these values to calculate the default RI 8 alarm setpoint value, results in a setpoint that:
0 1) Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and
2) Will provide for a liquid radwaste discharge rate (as evaluated for each batch release) that is compatible with plant operations (refer to Tables 1-1.1 and 1-1.2).
Page 123 of 157
I Salem ODCM Rev. 23 Table A-1: Calculation of Effective MPC - Unit 1 Activity Released (Ci)
Nuclide MPC* 1993 1994 1998
(.ci/ml) CURIES CURIES CURIES BE-7 2.OOE-03 8.88E-04 ND ND NA-24 3.OOE-05 6.68E-04 1.62E-04 1.OOE-04 CR-51 2.00E-03 5.38E-03 2.02E-03 ND MN-54 1.OOE-04 3.52E-02 1.37E-02 7.1-6E-04 MN-56 1.OOE-04 ND ND 0.00E+00 FE-59 5.OOE-05 4.76E-04 4.84E-03 ND CO-57 4.OOE-04 1.03E-02 3.1 OE-03 1.78E-05 CO-58 9.OOE-05 1.71E+00 6.47E-01 3.39E-02 CO-60 3.O0E-05 3.04E-01 1.OE-01 2.42E-02 ZR-95 6.OOE-05 3.29E-03 7.13E-04 ND NB-95 1.OOE-04 5.78E-03 1.28E-03 ND NB-97 9.OOE-04 1.27E-03 1.07E-03 4.90E-05 TC-99M 3.OOE-03 2.66E-04 ND ND SR-89 3.OOE-06 ND ND 7.32E-06 2.18E-04 ND I
SR-92 6.OOE-05 ND MO-99 4.OOE-05 1.76E-04 1.76E-04 ND AG-110m 3.OOE-05 1.19E-02 1.1OE-02 6.58E-05 l SN-113 8.OOE-05 7.88E-05 4.91E-05 ND SB-122 3.OOE-05 1.21E-03 5.35E-04 1.12E-03 SB-124 2.OOE-05 2.08E-02 1.75E-02 1.73E-02 SB-125 1.OOE-04 9.04E-02 8.23E-02 3.56E-02 SB-126 3.OOE-06 ND 6.18E-05 2.23E-04 Z 1-131 1-133
<I 3.00E-07 1.OOE-06 1.27E-01 2.16E-03 1.82E-02 1.88E-04 2.32E-03 8.32E-06 1-134 2.OOE-05 ND 3.63E-04 ND Z CE-141 9.OOE-05 ND 4.24E-05 ND
< CE-143 4.OOE-05 5.42E-05 ND ND l D CS-134 9.OOE-06 3.54E-01 6.46E-01 2.49E-02
< CS-136 6.OOE-05 3.61E-03 1.59E-03 ND CS-137 2.OOE-05 4.53E-01 8.54E-01 7.51E-02 z CS-138 3.OOE-06 4.15E-06 1.35E-04 ND 0 BA-140 2.OOE-05 ND 8.62E-05 ND Cr)
> LA-140 2.00E-05 2.12E-04 1.86E-04 ND a: RU-105 1.OOE-04 2.21E-04 1.35E-04 ND
© RU-106 1.OOE-05 ND 1.03E-03 ND Z ZN-65 1.OOE-04 6.72E-04 ND ND U--
Total Ci Gamma 3.14E+00 2.42E+00 2.16E-01
~I O MPCe (pCi/mi) 6.05E-06 1.28E-05 1.28E-05
_MPC value for unrestricted area from 10 CFR 20, Appendix B, Table II, Column 2.
m** ND - not detected c* ~I Page 124 of 157 wD LL3
Salem ODCM Rev. 23 I1 Table A-2: Calculation of Effective MPC - Unit 2 Activity Released (Ci)
Nuclide MPC* 1993 1994 1998 (A+/-Ci/M1) CURIES CURIES CURIES BE-7 2.OOE-03 1.59E-03 2.88E-04 ND NA-24 3.OOE-05 1.05E-03 5.77E-05 7.39E-05 CR-51 2.OOE-03 4.39E-03 1.55E-03 1.14E-04 MN-54 1.OOE-04 3.73E-02 1.37E-02 7.54E-04 MN-56 1.OOE-04 ND ND 4.66E-05 FE-59 5.OOE-05 4.83E-04 3.25E-03 ND CO-57 4.OOE-04 1.17E-02 3.24E-03 ND CO-58 9.OOE-05 1.75E+00 6.60E-01 4.52E-02 CO-60 3.OOE-05 3.47E-0 I 1.03E-01 2.12E-02 ZR-95 6.OOE-05 2.34E-03 3.22E-04 ND NB-95 1.OOE-04 3.97E-03 1.11E-03 ND NB-97 9.OOE-04 1.46E-03 1.1OE-03 4.22E-05 TC-99M 3.OOE-03 3.77E-04 ND 2.35E-06 SR-89 3.OOE-06 ND ND 2.71E-04 SR-92 6.OOE-05 ND 1.43E-05 ND MO-99 4.OOE-05 ND ND ND AG-1 lOm 3.OOE-05 1.03E-02 1.34E-02 ND SN- 113 8.OOE-05 7.45E-05 ND ND I< SB-122 SB- 124 3.OOE-05 2.OOE-05 1.20E-03 3.77E-02 ND 9.82E-03 6.37E-04 1.44E-02 SB-125 1.OOE-04 1.35E-01 6.03E-02 1.88E-02 SB-126 3.OOE-06 3.5 1E-04 ND 1.97E-04 1-131 3.OOE-07 1.87E-01 7.98E-03 3.14E-03 1-132 8.OOE-06 8.72E-05 ND 1.68E-04 1-134 2.OOE-05 2.39E-04 1.85E-04 ND U-CE-141 9.OOE-05 ND 2.87E-05 ND CE-143 4.OOE-05 ND ND ND 0'
CS-134 9.OOE-06 4.57E-01 6.44E-0 1 2.64E-02 V) CS-136 6.OOE-05 4.82E-03 1.51E-03 ND CS-137 ,2.OOE-05 5.70E-01 8.54E-01 7.97E-02 CS-138 3.OOE-06 ND ND 4,90E-05 BA-140 2.OOE-05 ND ND ND 0
LA-140 2.OOE-05 2.03E-03 1.11E-04 ND U- RU-105 1.OOE-04 4.07E-05 ND. ND
> ND, 4.38E-04 ND RU-106 1.OOE-05 1Y21 ND ND ZN-65 1.OOE-04 1.59E-04 0- W-187 6.OOE-05 ND 7.98E-05 ND L, Total Ci Gamma 3.57E+00 2.38E+00 2.31E-01 MPCe (PLCi/mnl) 4.81E-06 1.55E-05 1.12E-05
MPC value for unrestrictedarea from 10 CFR 20, Appendix B, Table II, Column 2.
zo.
t./3
ND = not detected 1,1 Page 125 of 157
Salem ODCM Rev. 23 APPENDIX B Technical Basis for Simplified Dose Calculations Liquid Radioactive Effluent, (A
Li z
(A.
z V) 0 0
0 (A
(A U--
O U--
U-
-4 0
U-(A_
DY Page 126 of 157
Salem ODCM Rev. 23 APPENDIX B: Technical Basis for Simplified Dose Calculations - Liquid Effluents The radioactive liquid effluents for the years 1993, 1994, and 1998 were evaluated to determine the dose contribution of the radionuclide distribution. These were the most recent years of full power operation for both Units. This'analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses, providing a simplified method of determining compliance with the
.dose limits of ODCM CONTROL 3.11.1.2.
For the radionuclide distribution of effluents from Salem, the controlling organ is typically the GI-LLI.
The calculated GI-LLL dose is predominately a function of the Fe-55, Co-58, Co-60, Fe-59 and Ag-1 10m releases. The radionuclides, Cs-134 and Cs-137 contribute the large majority of the calculated total body dose. The results of the evaluation for 1993, 1994, and 1998 are presented in
Table B-1 and Table B-2.
For purposes of simplifying the details of the dose calculational process, it is conservative to identify a controlling, dose significant radionuclide and limit the calculation process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e., cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
For the evaluation of the maximum organ dose, it is conservative to use the Nb-95 dose conversion I O factor (1.51 E+06 mrem/hr per pCi/ml, GI-LLI). By this approach, the maximum organ dose will be overestimated since this nuclide has the highest organ dose factor of all the radionuclides evaluated.
VI) jFor the total body calculation, the Fe-59 dose factor (2.32 E+05 mrem/hr per pCi/ml, total body) is the highest among the identified dominant nuclides. For evaluating compliance with the dose limits of ODCM CONTROL 3.11.1.2, the following simplified equations may be used:
< Total Body
< 1.67E-02*VOL 0
Z"
. Dtb ~CWi *AFe- 59,TB
Ci 0B.1)
"' Where:
Iz Dth = dose to the total body (mrem)
AFe-59.TB = 7.27E+04, total body ingestion dose conversion factor for Fe-59 (mrem/hr per ýtCi/ml)
VOL A = volume of liquid effluent released (gal)
I SCi C W
=
=
total concentration of all radionuclides (ptCi/ml) average circulating water discharge rate during release period(gal/min) 1.67E-02 = conversion factor (hr/min)
L°
__.1 w
ry Page 127 of 157 V)
Salem ODCM Rev. 23 I
Substituting the value for the Fe-59 total body dose conversion factor, the equation simplifies to: I
= 1.21E+03*VOL *,G CW (B.2)
I Maximum Organ I Dmax =-
1.67E-02*VOLAm9g,Gr-CVI.i
ZCj i (B.3) I Where: I Dmax =
ANb-95,GI-LLI =
maximum organ dose (mrem) 1.51E+06,.Gi-LLI ingestion dose.conversion factor for Nb-95 (mrem/hr per I pCi/ml)
Substituting the value for ANb-95,GI-LLI the equation simplifies to: I Dmax= 2.52E+O4*VOL ,
CW 7Ci (B.4) I 0
L2 7 Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from normal reactor releases is relatively negligible. The average annual tritium release I
Z 0
from each Salem Unit is approximately 350 curies. The calculated total body dose from such a release is 2.4E-03 mrem/yr via the fish and invertebrate ingestion pathways. This amounts to 0.08% of the design limit dose of 3 mrem/yr. Furthermore, the release of tritium is a function of operating time and I
0 0
(D.
power level and is essentially unrelated to radwaste system operation.
I I
z CL 0
H-I I
Li 0
0 I
c*j ILU 0!
Page 128 of 157 DI_
nM M-mn M-----M M M-- M -
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES 0 Salem ODCMt9'. 23 Table B-i: Adult Dose Contributions - Fish and Invertebrate Pathways - Unit 1 Nuclide Release (Ci) T.Body Dose Fraction GI-LLI Dose Fraction Liver Dose Fraction 1994 1993 1998 1994 1993 1998 1994 1993 1998 1994 1993 1998 Mn-54 1.32E-2 3.51E-2 7.16E-4 *
0.03 0.02 *
0.02
Fe-55 1.49E-1 6.40E-2 8.39E-2 0.07 0.04 0.37 0.12 0.03 0.52 0.19 0.14 0.67 Fe-59 4.84E-3 4.77E-4 N/D 0.02 *
0.12 0.01
0.03 0.01
Co-58 6.47E-1 1.71E+0 3.39E-2 0.05 0.18 0.02 0.31 0.51 0.13 0.01 0.07
Co-60 1.10E-1 3.04E-1 2.42E-2 0.02 0.09 0.05 0.14 0.24 0.24
0.03 0.01 Zn-65 NID 6.72E-4 N/D
0.01
0.01 *
0.02
Nb-95 1.28E-3 5.78E-3 N/D * * *
0.01 * * *
Ag-110m 1.1OE-2 1.19E-2 6.58E-5 *
0.26 0.17 0.01 * *
Sb-124 1.75E-2 2.58E-2 1.73E-2 * * * *
0.04 * *
Sb-125 8.23E-2 9.04E-2 3.56E-2 * * * *
0.02 * *
Cs-134 6.46E- 1 3.54E-1 2.49E-2 0.47 0.38 0.18 * *
0.38 0.37 0.09 Cs-137 8.54E-1 4.53E-1 7.51E-2 0.37 0.28 0.32 * *
0.37 0.35 0.20 Total 2.53E+0 3.21E+0 3.31E-1
Less than 0.01 N/D = not detected Page 129 of 157
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES e
Salem ODCMe. 23 Table B-2: Adult Dose Contributions - Fish and Invertebrate Pathways - Unit 2 Nuclide Release (Ci) T.Body Dose Fraction GI-LLI Dose Fraction Liver Dose Fraction 1994 1993 1998 1994 1993 1998 1994 1993 1998 1994 1993 1998 Mn-54 1.37E-2 3.73E-2 7.54E-4 * *
0.01 0.02 *
0.01
Fe-55 1.38E-1 6.61E-2 1.64E-2 0.06 0.04 0.10 0.10 0.03 0.18 0:18 0.12 0.27 Fe-59 3.25E-3 4.82E-4 N/D 0.01 *
0.08 0.01
0.02 *
Co-58 6.60E-1 1.75E+0 4.52E-2 0.05 0.16 0.04 0.29 0.51 0.29 0.01 0.06 0.01 Co-60 1.03E-1 3.47E-1 2.12E-2 0.02 0.09 0.06 0.12 0.27 0.37 0.01 0.03 0.02 Zn-65 N/D 1.59E-4 N/D * * * * * * * *
Nb-95 1.11E-3 3.97E-3 N/D * *
0.06 0.01 * * *
Ag-lOrn. -1.34E-2 1.03E-2 N/D *
0.31 0.14 * * *
Sb-124 9.82E-3 3.77E-2 1.44E-2 * * *
0.01 0.06 * *
Sb-125 6.03E-2 1.35E-1 .1.88E-2 * * *
0.01 0.02 * *
Cs-134 6.44E-1 4.58E-1 2.64E-2 0.48 0.41 0.26 0.01 *
0.39 0.40 0.20 Cs-137 8.54E-1 5.70E-1 7.97E-2 0.37 0.30 0.46 * *
0.38 0.36 0.45 Total 2.48E+0 3.65E+0 2.23E-1
Less than 0.01 N/D = not detected Page 130 of 157 M M-mmm- M- M- M M M M M -M MM
Salem ODCM Rev. 23 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent 0
FU C-)
z" ci-)
0 n
LU.
V) 0 Uv-Page 131 of 157
Salem ODCM Rev. 23 APPENDIX C: Technical Bases for Effective Dose Factors - Gaseous Effluents I 0 Overview I
The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose factors which are radionuclide specific.
I These effective factors, which can be based on typical radionuclide distributions of releases, can be applied to the total radioactivity released to approximate the dose in the environment (i.e., instead of having to perform individual radionuclide dose analyses only a single multiplication (Keff, Meff or Neff) I times the total quantity of radioactive material released would be needed).
This approach provides a reasonable estimate of the actual dose while eliminating the need for a I detailed calculational technique.
Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations:
I Keff AI (C.1)
I 0 Where: I LU I~ff= the effective total body dose factor due to gamma emissions from all noble gases released I Ki= the total body dose factor due to gamma emissions from each noble gas radionuclide i D
f released
= the fractional abundance of noble gas radionuclide i relative to the total noble gas I 0- activity 0
I 0
U--
e7~5~~ ~zi (C.2)
I iC-) Where:
V.)
0 U--
Z-LI (L + 1.1 M)eff = the effective skin dose factor due to beta and gamma emissions from all I
noble gases released
(_
0 A
(Li + 1.1 MO) = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released I
LUJ
z I LU_
0, Page 132 of 157
Salem ODCM Rev. 23 Meff Z AMC (C.3),
Where:
Meff the effective air dose factor due to gamma emissions from all noble gases released I
Mi the air dose factor due to gamma emissions from each noble gas radionuclide i released Neff.= A< (CA4)
Where:
SNe ff the effective air dose factor due to beta emissions from all noble gases released Ni =the air dose factor due to beta emissions from each noble gas radionuclide i released I .Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors. However, the noble gas releases from Salem have been maintained to such negligible quantities that the inherent variability in the data makes any meaningful evaluations difficult.
I O Therefore, in order to provide a reasonable basis for the derivation of the effective noble gas dose factors, the primary coolant source term from ANSI N237-1976/ANS-18.1, "Source Term Io
(
Specifications," has been used as representing a typical distribution. The effective dose factors as derived are presented in Table C-1.
Application
< To provide an additional degree of conservatism, a factor of 0.50 is introduced into the dose Cn calculational process when the effective dose transfer factor is used. This conservatism provides I< additional assurance that the evaluation of doses by the use of a single effective factor will not Usignificantly underestimate any actual doses in the environment.
0 For evaluating compliance with the dose limits of ODCM CONTROL 3.11.2.2, the following
> simplified equations may be used:
I*
ry,
-Y Dr=3.17E-08
XQ*Mff, *7 (C.5) 0.50 Q I> and 0
L-W
__ Dfl =3.17E-08,(6 *Nff*~ (C.6) 0.50 QNefi I _.9 Where:
Page 133 of 157
Salem ODCM Rev. 23 I
I 0 Dy = air dose due to gamma emissions for-the cumulative release of all noble gases (mrad)
Dp = air dose due to beta emissions for the cumulative release of all noble gases (mrad)
X/Q = atmospheric dispersion to the controlling site boundary (sec/m3) I Meff = 5.3E+02, effective gamma-air dose factor (mrad/yr per ýiCi/m3)
Neff = L.lE+03, effective beta-air dose factor (mrad/yr per pLCi/m3)
Qi = cumulative release for all noble gas radionuclides (ýtCi)
I 3.17E-08 = conversion factor (yr/sec) 0.50 = conservatism factor to account for the variability in the effluent data I Combining the constants, the dose calculational equations simplify to:
I Dr=3.5E-O5*/Q*,Qi (C.7) 1 I
and Dfi=7.OE-05,Q*ZQI (C.8)
I The effective dose factors are used on a very limited basis for the purpose of facilitating the timely I
assessment of radioactive effluent releases, particularly during periods of computer malfunction where (j3 0
a detailed dose assessment may be unavailable. I C
0 I
H-D I
Z I
z I
(I) o-0 I
0 0
I-J W
I IU-I Cu U-.
(V) 0y D
Page 134 of 157 I
Salem ODCM Rev. 23 Table C-1: Effective Dose Factors 1I Noble Gases - Total Body and Skin Total Body Effective Skin Effective Dose Factor Dose Factor Keff (L+ 1.1 M)eff Radionuclide fi0 (mrem/yr per ptCi/m 3) (mrem/yr per PCi/m 3)
Kr-85 Kr-88 0.01 1.4E+01 Xe-133m 0.01 1.5E+02 1.9E+02 0.01 2.5E+00 1.4E+01 Xe-133 Xe-135 0.95 2.8E+02 6.6E+02 0.02 3.6E+01 7.9E+01 Total 4.7E+02 9.6E+02 Noble Gases - Air Gamma Air Effective Beta Air Effective Dose Factor Dose Factor Is Meff Neff Radionuclide f0.0 (mrad/yr per pCi/m3) (mrad/yr per pLCi/m 3)
Kr-85 Kr-88 0.01 2.OE+01 Xe-133m 0.01 1.5E+02 2.9E+01 0.01 3.3E+00 1.5E+01 Xe-133 Xe-135 0.95 3.4E+02 1.OE+03 0.02 3.8E+01: 4.9E+01 V-)
0 D
Total 5.3E+02 1.1E+03 V"
iLi
Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1, "Source Term Specifications."
¢-y 0
U-LiJ Li w
c")
03 n,
Page 135 of 157
I Salem ODCM Rev. 23 I
.0 I
APPENDIX D I
Technical Basis for Simplified Dose Calculation I Gaseous Radioactive Effluent w
0 V-)
0 z
0 H--
z 0
0 w
0 Li_
z w
0 U-w o.
cn 0*
02 0n w-Page 136 of 157
Salem ODCM Rev. 23
APPENDIX D: Technical Basis for Simplified Dose Calculation - Gaseous Effluents The pathway dose factors for the controlling infant age group were evaluated to determine the controlling pathway, organ and radionuclide. This analysis was performed to provide asimpiified method for determining compliance with ODCM CONTROL 3.11.2.3 For the infant age group, the controlling pathway is the~grass-cow-milk (g/c/m) pathway. An infant receives a greater radiation dose from the g/c/m pathway than any other pathway. Of this g/c/m pathway, the maximum exposed organ including "thetotal body, is the thyroid, and the highest dose contributor is radionuclide I-131. The results for this evaltiation are presented in Table D-1.
For purposes of simplifying the details of the dose calculation process, it is conservative to identify a controlling, dose significant organ and radionuclide and limit the calculation process to the use of the dose conversion factor for the organ and radionuclide. Multiplication of the total release (i.e.
cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
For the evaluation of the dose commitment via a controlling pathway and-age group, it is conservative to use the infant, g/c/m, thyroid, 1-131 pathway dose factor (1.05E12 m2 mrem/yr per j.Ci/sec). By this approach, the maximum dose commitment will be overestimated since 1-131 has the highest pathway dose factor of all radionuclides evaluated.
E
For evaluating compliance with the dose limits of ODCM CONTROL 3.11.2.3, the following simplified equation may be used:
0 Z D. -=3.17E-08*W*R* -13I*zQ
<7 i Where:
IDx maximum organ dose (mrem)
W = atmospheric dispersion parameters to the controlling location(s) as identified in Table 3.2-4.
X/Q = atmospheric dispersion for inhalation pathway and H-3 dose contribution via other V) pathways (sec/m3)
D/Q = atmospheric deposition for vegetation, milk and ground plane exposure pathways (mi)
Vf) T gQn = cumulative release over the period, of interest for radioiodines and particulates 3.177E-8 = conversion factor (yr/sec) t, RI-131 = 1-131 dose parameter for the thyroid for the identified controlling pathway 1.05E-+12 (M2 mremlyr per .tCi/sec), infant thyroid dose parameter with the grass-cow-milk pathway controlling, uexposure and inhalation pathways need not be considered when the above simplified 0*
0 calculation thyroid dose. method is used because of the overall negligible contribution of these pathways to the total S_j 1I0 It is recognized that for some particulate radionuclides. (e.g., Co-60 and Cs-137), the ground epsr W pathway may represent a higher dose contribution than either the vegetation or milk pathway.
0*.
I0 Page 137 of 157 0D
I Salem ODCM Rev. 23 However, use of the I-131 thyroid dose parameter for all radionuclides will maximize the organ dose I
, calculation, especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway than 1-131 for the thyroid via the milk pathway (see Table D-1). I The dose should be evaluated based on. the predetermined controlling pathways as identified in Table 2-3. If more limiting pathways in the surrounding environment of Salem are identified by the annual land use census, Table 2-3 will be revised as specified in ODCM CONTROL 3.12.2.
I I
I I
0 I w
0 z I
.0 I
0 z0: I Z
I z
I U-U-
U-Q.:
.© 0
U-b_0 b.J Page 138 of 157 I
I TbeDl natDs otiuin I Salem ODCM Rev. 23 I, Table D-1: Infant Dose Contributions Fraction of Total Organ and Body Dose PATHWAYS I Target Organs Grass-Cow-Milk Ground Plane I Total Body 0.02 0.15 I Liver 0.23 0.14 Thyroid 0.59 0.15 Kidney 0.02 0.15 ID Lung 0.01 0.02 0 GI-LLI 0.02 0.15 77 Fraction of Dose Contribution by Pathway Pathway f r)
Grass-Cow-Milk 0.92 Ground Plane 0.08 z"
I.
Inhalation N/A U-t..t 0
c-j Page 139 of 157
Salem ODCM Rev. 23 0
APPENDIX E
.Radiological Environmental 'Monitoring Program Sample Type, Location and Analysis 0
0 V_)
0 Ld 0
zD 0
LJ 0
H-z 0
0 rý,,
U-LU 0
V-0_
Page 140 of 157 0q
Salem ODCM Rev. 23 1 0 APPENDIX E: Radiological Environmental Monitoring Program SAMPLE DESIGNATION Samples are identified by a three part code. The first two letters are the power station identification code, in this case "SA". The next three letters are for the media sampled.
AIO = Air Iodine IDM = Immersion Dose (DLR)
APT = Air.Particulates MLK = Milk ECH = Hard Shell Blue Crab PWR = Potable Water (Raw)
ESF = Edible Fish PWT = Potable Water (Treated)
ESS = SedimentSWA = Surface Water WWA = Well Water The last four symbols are a location code based on direction and distance from the site center point.
The midpoint of a line between the centers of Salem units 1 & 2 containment domes was used as the site center point. Of these, the first two represent each of the sixteen angular sectors of 22.5 degrees centered about the reactor site. Sector one is divided evenly by the north axis and other sectors are numbered in a clockwise direction; i.e., 2=NNE, 3=NE, 4=ENE, 5=E, 6=ESE, 7=SE, 8=SSE, 9=S, 10=SSW, 11=SW, 12=WSW, 13=W, 14=WNW, 15=NW and 16=NNW. The next digit is a letter which represents the radial distance from the plant:
I, S = On-site location E = 4-5 miles off-site A = 0-1 miles off-site F = 5-10 miles off-site B = 1-2 miles off-site G = 10-20 miles off-site C-)
C = 2-3 miles off-site H = > 20 miles off-site D 3-4 miles off-site The last number is the station numerical designation within each sector and zone; e.g., 1,2,3. For example; the designation SA-WWA-5D1 would indicate a sample in the SGS.and HCGS program (SA), consisting of well water (WWA), which had been collected in sector number 5, centered at 90' (due east) with, respect to the reactor site at a radial distance of 3 to 4 miles off-site, (therefore, radial distance D). The number 1 indicated that this is sampling station #1 in that particular sector.
SAMPLING LOCATIONS All sampling locations and specific information about the individual locations are given in Table E-1.
U-Maps E-1 and E-2 show the locations of sampling stations with respect to the site center.point.
0 U-U-
0 r© 0
V)
L, Cni n-0 D..
V)
Page 141 of 157
I Salem ODCM Rev. 23 TABLE E-1: REMP Sample Locations I
O A. Direct Radiation Monitoring Locations (IDM)
STATION CODE STATION LOCATION*
1S1 0.55 mi. N 2S2 0.4 mi. NNE 2S4 0.59 mi. NNE; N of equipment laydown area 3S1 0.58.mi. NE 4S1 0.60 mi ENE; site access road near intersection to TB-02 581 0.95 mi. E;site access road 6S2 0.21 mi. ESE; observation building 7S 1 0.12 mi. SE; station personnel gate' IoSI 0.14 mi. SSW; circ water bldg.
liS1 0.09 mi. SW; service water bldg.
15S1 0.57 mi. NW; near river and barge slip 16S1 0.54 mi. NNW; on road near fuel oil storage tank 4D2 3.7 mi. ENE; Alloway Creek Neck Road 5D1 3.5 mi. E; local farm IOD1 3.9 mi. SSW; Taylor's Bridge Spur 14D1 3.4 mi. WNW; Bay View, DE 15D1 3.8 mi. NW; Rt 9, Augustine Beach, DE O 2E1 4.4 mi. NNE; local farm 3El 4.1 mi. NE; local farm V 11 E2 5.0 mi. SW; Rt. 9 12EI 4.4 mi. WSW; Thomas Landing
< 13EI 4.2 mi. W; Diehl House Lab o 16EI 4.1 mi. NNW; Port Penn z
< WFI 5.8 mi. N; Fort Elfsborg 2F2 8.7 mi. NNE; Salem Substation 2F5 7.4,mi. NNE; Salem High School U' 2F6 7.3 mi. NNE; PSE&G Training Center Salem NJ 3F2 5.1 mi. NE; Hancocks Bridge, NJ Munc Bldg 2ý
_ 3F3 8.6 mi. NE; Quinton Township Elem. School NJ
>) 4F2 6.0 mi. ENE; Mays Lane, Harmersville, NJ U" 5F1 6.5 mi. E; Canton, NJ z
I-0 LU CO
_.J OI P 0F/i U')
Page 142 of 157
Salem ODCM Rev. 23 TABLE E-1 (Cont'd)
I
'0A.Direct Radiation Monitoring Locations (1DM) (Cont Id)
STATION CODE STATION LOCATION*
6F1 6.4 mi. ESE; Stow Neck Road 7F2 9.1 mi. SE; Bayside, NJ 9F1 5.3 mi. S; off Route #9, DE 10F2 5.8 mi. SSW; Rt. 9 1IF1 6.2 mi. SW; Taylors Bridge, DE 12F 1 9.4 mi. WSW; Townsend Elementary School, DE 13F2 6.5 mi. W; Odessa, DE 13F3 9.3 mi. W; Redding Middle School 13F4 9.8 mi. W; Middletown, DE 14F2 6.6 mi. WNW; Boyds Comer 15F3 5.4 mi. NW 16F2 8.1 mi. NNW; Delaware City Public School 1G3
19 mi. N; N. Church St. Wilmington, DE 3G1 17 mi. NE; local farm 1OGI 12 mi. SSW; Smyrna, DE
14G1 16G1 11.8 mi. WNW; Rte 286, Bethel Church Rd., DE 15 mi. NNW; Wilmington Airport 0 3H1 32 mi. NE; National Park, NJ C-)
I ~ B. Air Sampling Locations (AIO, APT)
O STATION CODE STATION LOCATION*
0 5S1 0.95 mi. E; site access road 5D1 3.5 mi. E; local farm 0 16EI 4.1 mi. NNW; Port Penn i1FI 5.8 mi. N; Fort Elfsborg 2F6 7.3 mi. NNE; PSE&G Training Center Salem, NJ 14G1 11.8 mi. WNW; Rte 286, Bethel Church Rd., DE Li w
D Page 143 of157
Salem ODCM Rev. 23 Table E-1 (Cont'd)"
C. Surface Water Locations (SWA) - Delaware River STATION CODE STATION LOCATION*
11A1 0.2 mi. SW; Salem Outfall Area I IlAla Alternate 0.15 SW location in plant barge slip area 12C1 2.5 mi. WSW; West bank of Delaware River 12C1a Alternate 3.7 mi.WSW at the tip of Augustine Beach Boat Ramp 7E1 4.5 mi. SE; River Bank 1.0 mi..W of Mad Horse Creek 7El a Alternate 8.87 mi SE at the end of Bayside Road 1F2 7.1 mi. N; midpoint of Delaware R.
16F 1 6.9 mi. NNW; C&D Canal D. Ground Water Locations (WWA)
STATION CODE STATION LOCATION*I 3E1 4.1 mi NE, local farm No groundwater samples are required as liquid effluents discharged from Hope Creek and I Salem Generating Stations do not directly affectthis pathway. However, this location (3E1) is being monitored as a management audit sample.
V) E. Drinking Water Locations (PWR, PWT)
STATION CODE STATION LOCATION*
C-) 2F3 8.0 mi NNE, City of Salem Water and Sewage Department
< No public drinking water samples or irrigation water samples are required as these pathways Do are not directly affected by liquid effluents discharged from Hope Creek and Salem Generating Stations. However, this location (2F3) is being monitored as a management audit sample.
H---
F. Water Sediment Locations (ESS) 0 _ _ _I 0 STATION CODE STATION LOCATION*
S1HAl 0.2 mi. SW; Salem outfall area Z 15A1 0.3 mi. NW; Hope Creek outfall area Uj,. 16A1 12C1 0.7 2.5 mi.
mi. NNW; WSW; South StormofDrain West bank outfall Delaware River 7E1 4.5 mi. SE; 1 mi West of Mad Horse Creek o 16F1 6.9 mi. NNW; C&D Canal U 6S2 0.2 mi. ESE; observation building m1I oU I UI (Y Page 144 of 157
Salem ODCM Rev. 23 Table E-1 (Cont'd)
E. G. Milk Sampling Locations (MLK)
STATION CODE STATION LOCATION*
2G3 12.0 mi. NNE, local farm 13E3 4.9 mi W, local farm 14F4 7.6 mi. WNW; local farm 3G1 17 mi. NE; local farm H. Fish and Invertebrate Locations (ESF, ECH)
STATION CODE STATION LOCATION*
11A1 0.2 mi. SW; Salem outfall area 12C1 2.5 mi. WSW; West bank of Delaware River 7E1 4.5 mi. SE; 1 mi West of Mad Horse Creek I. Food Product Locations STATION CODE STATION LOCATION*
The Delaware River at the location of Salem and Hope Creek Nuclear Power Plants is a brackish lie V) water source. No irrigation of food products is performed using water in the vicinity from which liquid plant wastes have been discharged. However, 12 management audit food samples are collected from various locations.
r *All distances and directions for the Station Locations are referenced to the midpoint between the z two Salem units' containments. The WGS 84 coordinates for this site center. point location are:
Latitude N 390 - 27' - 46.5" and Longitude W 75' - 32' - 10.6".
140 V)!
0L Page 145 of 157
Salem ODCM Rev. 23 I
SAMPLES COLLECTION AND ANALYSIS I
s SamIle Collection Method Analysis I Air Particulate Continuous low volume air sampler. Sample Gross Beta analysis on each weekly I collected every week sample. Gamma along with the filter change.
spectrometry shall be performed if I gross beta exceeds 10 times the yearly mean of the control I station value. Samples shall be analyzed 24 hrs or more after I collection to allow for radon and thorium daughter decay. Gamma I isotopic analysis on quarterly composites. I Vo 0
Air Iodine A TEDA impregnated charcoal cartridge is Iodine 131 analysis are performed on each weekly sample.
I (n
z connected to air
"-)
(_)
particulate air sampler and is collected weekly at filter change.
I z
co H--
Crab and Fish Two batch samples are sealed in a plastic Gamma isotopic analysis of edible I
H-z" 0
bag or jar and frozen semi-annually or when in season.
portion on collection.
I Sediment A sediment sample is taken semi-annually.
Gamma isotopic analysis I
(_9 7 semi-annually.
U-LJ Direct 2 DLR's will be Gamma dose quarterly.
I collected from each 0
U-location quarterly. I V) oL D_
I..1 0
I C-V..)
Page 146 of 157 I
Salem ODCM Rev. 23 SAMPLE COLLECTION AND ANALYSIS (Cont'd)
I0 Sample Collection Method Analysis Milk Sample of fresh milk Gamma isotopic is collected for each analysis and 1-131.
farm semi-monthly when analysis on each cows are in pasture, sample on collection.
monthly at other times.
Water Sample to be collected Gamma isotopic (Potable, monthly providing winter monthly H-3 on Surface) icing conditions allow. quarterly surface sample, monthly on ground water sample.
I*V)
LU 0
LuJ LuJ
>X Page 147 of 157 Ii
Salem ODCM Rev. 23 FIGURE E-1:. ONSITE SAMPLING LOCATIONS.
0 MAP B-1 ON-SITE SAMPLING LOCATIONS 1
S 13 5 z
0 V.)
0 H--
(/3 l:ý 0
0 co C2 0 9 W
U-V_
rrý c,- Page 148 of 157
Salem ODCM Rev. 23 FIGURE E-2: OFFSITE SAMPLING LOCATIONS IO A
F5 MW 16 NNW 1 AM) N. 2 NNE A13H? 1 3 H ITE CLAY./-,
C K HUNDqEO 0,_
EA PE IYN-N CASTUEe
.UNDRED 2G2 H SG NAL YA REFU ?Gr V 203 PENCADER UP HUNDRED VE 16F2 C!4 2F3 t
ZO IFF f NTON 1401 SAINT GEORGES JET 2
'-HqNDR2D
.0 .10 SFY 270 go-I, JF
-APPOQUINIMINK.-
HUNDRED 12EF LU rFFF GMNVAC14
_J to r2pr 7F2 C-)
D. BLACKBIRD HUNDRED r
nr
-cn Lu 0
roar 14i 1, c c L-, cu K60 CO t2 1bucK CREEK som a r-V is 100 NNATIONAL.
REFOO 9 sc-0 w,
Page 149 of 157
Salem ODCM Rev. 23 0
I APPENDIX F MAXIMUM PERMISSIBLE CONCENTRATIONS LIQUID EFFLUENTS I
I 0
Lu
-z-7' C-)
0z (I)
G,3 0
7-Ld 0
LU 0
-J (Y
LUJ Page 150 of 157 cy)
Salem ODCM Rev. 23 APPENDIX F: Maximum Permissible Concentration (MPC) Values - Liquid Effluents I* The following radionuclide concentrations were obtained from 10 CFR 20 Appendix B, Table I, Column 2 as revised January 1, 1991.
Table F-i: Maximum Permissible Concentrations Element Isotope Soluble Conc Insoluble Conc.
(p.Ci/rnl) (j.Ci/ml)
Actinium (89) Ac-227 2E-6 3E-4 Ac-228 9E-5 9E-5 Americium (95) Am-241 4E-6 3E-5 Am-242m 4E 9E-5 Am-242 1E-4 1E-4 Am-243 4E-6 3E-5 Am-244 5E-3 5E-3 Antimony (51) Sb- 122 3E-5 3E-5 Sb-124 2E-5 2E-5 Sb-125 1E-4 1E-4 Sb-126 3E-6 3E-6 Arsenic (33) As-73 5E-4 5E4 As-74 5E-5 5E-5 As-76 2E-5 2E-5 I, Astatine (85)
As-77 At-211 8E-5 2E-6 8E-5 7E-5 Barium (56) Ba-131 2E-4 2E-4
(-)
Ba-140 3E-5 2E-5 Berkelium (97) Bk-249 6E-4 6E-4 Bk-250 2E-4 2E-4 Beryllium (4) Be-7 2E-3 2E-3 Bismuth (83) Bi-206 4E-5 4E-5 V3 Bi-207 6E-5 6E-5 Bi-210 4E-5 4E-5 Bi-212 4E-4 4E-4 0
Bromine (35) Br-82 3E-4 4E-5 Br-83 3E-6 3E-6 ry Cadmium (48) Cd-109 2E-4 2E-4 Cd- 15m 3E-5 3E-5 Cd- 115 3E-5 4E-5 U-Calcium (20) Ca-45 9E-6 2E-4 Ca-47 5E-5 3E-5
> Californium (98) Cf-249 4E-6 2E-5 Lý. Cf-250 1E-5 3E-5 Cf-251 4&-6 3E-5 Ld Cf-252 7E-6 7E-6
_j Cf-253 1E-4 1E-4 Cf-254 1E-7 1E-7 Page 151 of 157
I Table F-1 (Continued)
Salem ODCM Rev. 23 I
Element Isotope Soluble. Conc.
_(Ci/ml)
Insoluble Conc.
(itCi/ml)
U Carbon (6) C- 14 8E-4 -------
Cerium (58) Ce-141 Ce-143 9E-5 4E-5 9E-5 4E-5 I
Cesium (55)
Ce-144 Cs-131 Cs-134m 1E-5 2E-3 6E-3 "1E-5 9E-4 1E-3 I
Cs-134 Cs-135 9E-6 1E-4 4E-5 2E-4 I Cs-136 9E-5 6E-5 Chlorine (17)
Cs-137 CI-36 2E-5 8E-5 4E-5 6E-5 I CI-38 4E-4 4E-4 Chromium (24)
Cobalt (27)
Cr-51 Co-57 2E-3 5E-4 2E-3 4E-4 I Co-58m 3E-3 2E-3 Co-58 Co-60 lE-4 5E-5 9E-5 3E-5 I
Copper (29) Cu-64 3E-4 2E-4 Curium (96) Cm-242 Cm-243 2E-5 5E-6 2E-5 2E-5 I
Cm-244 7E-6 3E-5 0 _
Cm-245 Cm-246 4E-6 4E-6 3E-5 3E-5 I
0 z
V)
Cm-247 Cm-248 Cm-249 4E-6 4E-7 2E-3 2E-5 1E-6 2E-3 I
V Dysprosium (66)
Einsteinium (99)
Dy-165 Dy-166 Es-253 4E-4 4E-5 2E-5 4E-4 4E-5 2E-5 I
2 0
Es-254m Es-254 2E-5 1E-5 2E-5 1E-5 I
__,_,_ Es-255 3E-5 3E-5 Z
0 Erbium (68) Er-169 Er-171 9E-5 IE-4 9E-5 1E-4 I
_ Europium (63) Eu-152 (9.2 hrs) 6E-5 6E-5 U, Eu- 152 (13 yrs)
Eu-154 8E-5 2E-5 8E-5 2E-5 I
o Eu-155 2E-4 2E-4 U-
_o Fermium (100) Fm-254 Fm-255 1E-4 3E-5 1E-4 3E-5 I
Fm-256 9E-7 9E-7 0
CL 0y w~t I
w-1 0./
Page 152 of 157 I
Salem ODCM Rev. 23 Table F-1 (Continued)
Element Isotope Soluble Conc. Insoluble Cone.
* *t~iml)(ýLCi/mnl)*
Fluorine (9) F-18 8E-4 5E-4 Gadolinium (64) Gd- 153 2E-4 2E-4 I Gallium (31)
Gd- 159 Ga-72 8E-5 4E-5 8E-5 4E-5 I Germanium (32)
Gold (79)
Ge-71 Au-196 2E-3 2E-4 2E-3 1E-4 Au-7198 5E-5 5E-5 Iio Hafnium (72)
Au- 199 Hf-181 2E-4 7E-5 2E-4 7E-5 I.0 Holmium (67) Ho-166 3E-5 3E-5 Hydrogen (3) H-3 3E-3 3E-3 Indium (49) In-113m 1E-3 1E-3 In-I 14m 2E-5 2E-5 In-115m 4E-4 4E-4 Io Iodine (53)
In- 115 1-125 9E-5 2E-7 9E-5 2E-4 1-126 3E-7 9E-5
__1-129 6E-8 2E-4 1-130 3E-6 3E-6 1-131 3E-7 6E-5 1-132 8E-6 2E-4 1-133 1E-6 4E-5 1-134 2E-5 6E-4 1-135 4E-6 7E-5 Iridium (77) Ir-190 2E-4 2E-4 I, Ir-192 4E-5 4E-5 Ir-194 3E-5 3E-5 Iron (26) Fe-55 SE-4 2E-3 I, Fe-59 6E-5 5E-5 Lanthanum (57) La-140 2E-5 2E-5 Lead (82) Pb-203 4E-4 4E-4 Pb-210 IE-7 2E-4 Pb-212 2E-5 2E-5 0 Lutetium (71) Lu-177 1E-4 IE-4 Manganese (25) Mn-52 3E-5 3E-5 Mn-54 1E-4 lE-4 Mn-56 IE-4 1E-4 Z-Mercury (80)_,, Hg-197m 2E-4 2E-4 Hg-197 31-4 5E-4 Hg-203 2E-5 1E-4 Molybdenum (42) Mo-99 2E-4 4E-5 w_
Page 153 of 157
I Table F-1 (Continued)
Salem ODCM Rev. 23 I
Element Isotope Soluble Conc.
(jiCi/ml)
Insoluble Conc.
(iLtCi/ml)
I Neodymium (60) Nd-144 Nd-147 7E-5 6E-5 8E-5 6E-5 I Nd-149 3E-4 3E-4 Neptunium (93) Np-237 3E-6 3E-5 Np-239 lE-4 1E-4 Nickel (28) Ni-59 2E-4 2E-3 Ni-63 3E-5 7E-4 Ni-65 1E-4 1E-4 Niobium (41) Nb-93m 4E-4 4E-4 Nb-95 1E-4 1E-4 Osmium (76)
Nb-97 Os-185 9E-4 7E-5 9E-4 7E-5 I
Os-191m Os-191 3E-3 2E-4 2E-3 2E-4 I
Os-193 6E-5 5E-5 Palladium (46) Pd-103 Pd- 109 3E-4 9E-5 3E-4 7E-5 I
Phosphorus (15) P-32 2E-5 2E-5 Platinum (78) Pt-191 1E-4 IE-4 'I Pt-193m 1E-3 IE-3 z
Pt-193 Pt- 197m Pt-197 9E-4 IE-3 1E-4 2E-3 9E-4 1E-4 I
0 Plutonium (94) Pu-238 z Pu-239 5E-6 5E-6 3E-5 3E-5 I
< Pu-240 5E-6 3E-5
__ _ Pu-241 Pu-242 2E-4 5E-6 1E-3 3E-5 I
__ Pu-243 3E-4 3E-4 0
V)
Polonium (84)
Potassium (19)
Po-21 0 K-42 7E-7 3E-4 3E-5 2E-5 'I
> Praseodymium(59) Pr-142 3E-5 3E-5 Promethium (61)
Pr-143 Pm-147 5E-5 2E-4 5E-5 2E-4 I
Pm-149 4E-5 4E-5 w, Protactinium(91) Pa-230 Pa-231 2E-4 9E-7 2E-4 2E-5 I
__Pa-233 1E-4 1E-4 wL
-j I
0 0-(I) wL I
0*
0*
wL Page 154 of 157 I
Salem ODCM Rev. 23 Table F-1 (Continued)
Element Isotope Soluble Conc'. Insoluble Conc.
(pCi/ml) (jICi/ml)
Radium (88) Ra-223 7E-7 4E-6 Ra-224 2E-6 5E-6 Ra-226 3E-8 3E-5 Ra-228 3E-8 3E-5 Rhenium (75) Re-183 6E-4 3E-4 Re- 186 9E-5 5E-5 Re-187 3E-3 2E-3 3E-5 10 Rhodium (45)
Re-188 Rh-103m 6E-5 IE-2 1E-2 Rh-105 IE-4 1E-4 Rubidium (37) Rb-86 7E-5 2E-5 Rb-87 1E-4 2E-4 Ruthenium (44) Ru-97 4E-4 3E-4 Ru-103 8E-5 8E-5 Ru-103m 3E-6 3E-6 I Ru-105 Ru-106 IE-4 1E-5 1E-4 1E-5 1 Samarium (62) Sm-147" Sm-151 6E-5 4E-4 7E-5 4E-4 Sm-153 8E-5 8E-5 Scandium (21) Sc-46 4E-5 4E-5 7, Sc-47 Sc-48 9E-5 3E-5 9E-5 3E-5 Selenium (34) Se-75 3E-4 3E-4 Silicon (14) Si-31 9E-4 2E-4 Silver (47) Ag-105 IE-4 1E-4 Ag-110m 3E-5 3E-5 10 Sodium (11)
Ag-1'1 Na-22 Na-24 4E-5 4E-5 2E-4 4E-5 3E-5 3E-5.
ef) Strontium (38) Sr-85m 7E-3 7E-3 n Sr-85 1E-4 2E-4 C-) Sr-89 3E-6 3E-5 Sr-90 3E-7 4E-5 r'Y Sr-91 7E-5 5E-5 Sr-92 7E-5 6E-5 Ii" pr- Sulfur (16)
Tantalum (73)
S-35 Ta-182 6E-5 4E-5 3E-4 4E-5 1-3 Page 155 of 157
I Table F-1 (Continued)
Salem ODCM Rev. 23 I Element Isotope Soluble Conc.
(yiCi/ml)
Insoluble Conc.
(yCi/ml)
I Technetium (43) Tc-96rn Tc-96 1E-2 lE-4 IE-2 5E-5 I Tc-97m 4E-4 2E-4 Tc-97 Tc-99m 2E-3 6E-3 8E-4 3E-3 I Tc-99 3E-4 2E-4 Tellurium (52) Te-125m Te-127m 2E-4 6E-5 lE-4 5E-5 I
Te-127 3E-4 2E-4 Te-129m Te-129 3E-5 8E-4 2E-5 8E-4 I
Te-131m 6E-5 4E-5 Terbium (65)
Te-132 Th-160 3E-5 4E-5 2E-5 4E-5 I
Thallium (81) TI-200 4E-4 2E-4 T1-201 TI-202 3E-4 1E-4 2E-4 7E-5 I
TI-204 1E-4 6E-5 Thorium (90) Th-227,.
Th-228 2E-5 2E-5 I 0 7E-6 1E-5 Cr LU Th-230 Th-231 2E-6 2E-4 3E-5 2E-4 I Th-232 2E-6 4E-5 Th-natural Th-234 2E-6.
2E-5 2E-5 2E-5 I 0 Thulium (69) Tm-170 5E-5 5E-5 V-)
Tin (50)
Tm-171 Sn-113 5E-4 9E-5 SE-4 8E-5 I Sn-124 2E-5 2E-5 Tungsten (74) W-181 W-185 4E-4 1E-4 3E-4 1E-4 I
0 W-187 7E-5 6E-5 LUJ Uranium (92) U-230 U-232 5E-6 3E-5 5E-6 3E-5 I
U-233 3E-5 3E-5 0
U-U-234 U-235 3E-5
.3E-5 3E-5 3E-5 I
me U-236 3E-5 3E-5 Fn 0
7* U-238 U-240 4E-5 3E-5 4E-5 3E-5 I
LLU U-natural 3E-5 3E-5
-J I
Vi)
D Page 156 of 157 I
Salem ODCM Rev. 23 Table F-I (Continued)
.Y. p Element Isotope Soluble Conc. Insoluble Conc.
I" (4Ci/ml) (jiCi/ml)
I Vanadium (23)
Ytterbium (70)
V-48 Yb-175 3E-5 1E-4 3E-5 1E-4 Yttrium Y-90 2E-5 2E-5 Y-91m 3E-3 3E-3 Y-91 3E-5 3E-5 Y-92 6E-5 6E-5 Y-93 3E-5 3E-5 Zinc (30) Zn-65 1E-4 2E-4 Zn-69m 7E-5 6E-5 Zn-69 2E-3 2E-3 I Zirconium (40) Zr-93 Zr-95 8E-4 6E-5 8E-4 6E-5 Zr-97 2E-5 2E-5 Any single radio- 3E-6 3E-6 nuclide not listed above with decay mode other than alpha emission or spontaneous fission and with radio -
active half-life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Any single radio- 3E-8 3E-8 nuclide not listed above, which decays by alpha emission or spontaneous fission.
Notes:
z"
1. If the identity of any radionuclide is not known, the limiting values for purposes of this table shall be: 3E-8 ýtCi/ml.
1-7
2. If the identity and concentration of each radionuclide are known, the limiting values should be 0 derived as follows: Determine, for each radionuclide in the mixture, the ratio between the quantity ILJ present in the mixture and the limit otherwise established in Appendix B for the specific
> radionuclide not in a mixture. The sum of such ratios for all the radionuclides in the mixture may not exceed "1" (i.e. "unity").
iio 1,
cký 0
12J_j V)i Page 157 of 157

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